WorldWideScience

Sample records for gas cooled core

  1. Core catcher cooling for a gas-cooled fast breeder

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.; Schretzmann, K.

    1976-01-01

    Water, molten salts, and liquid metals are under discussion as coolants for the core catcher of a gas-cooled fast breeder. The authors state that there is still no technically mature method of cooling a core melt. However, the investigations carried out so far suggest that there is a solution to this problem. (RW/AK) [de

  2. 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

  3. 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

  4. The first high resolution image of coronal gas in a starbursting cool core cluster

    Science.gov (United States)

    Johnson, Sean

    2017-08-01

    Galaxy clusters represent a unique laboratory for directly observing gas cooling and feedback due to their high masses and correspondingly high gas densities and temperatures. Cooling of X-ray gas observed in 1/3 of clusters, known as cool-core clusters, should fuel star formation at prodigious rates, but such high levels of star formation are rarely observed. Feedback from active galactic nuclei (AGN) is a leading explanation for the lack of star formation in most cool clusters, and AGN power is sufficient to offset gas cooling on average. Nevertheless, some cool core clusters exhibit massive starbursts indicating that our understanding of cooling and feedback is incomplete. Observations of 10^5 K coronal gas in cool core clusters through OVI emission offers a sensitive means of testing our understanding of cooling and feedback because OVI emission is a dominant coolant and sensitive tracer of shocked gas. Recently, Hayes et al. 2016 demonstrated that synthetic narrow-band imaging of OVI emission is possible through subtraction of long-pass filters with the ACS+SBC for targets at z=0.23-0.29. Here, we propose to use this exciting new technique to directly image coronal OVI emitting gas at high resolution in Abell 1835, a prototypical starbursting cool-core cluster at z=0.252. Abell 1835 hosts a strong cooling core, massive starburst, radio AGN, and at z=0.252, it offers a unique opportunity to directly image OVI at hi-res in the UV with ACS+SBC. With just 15 orbits of ACS+SBC imaging, the proposed observations will complete the existing rich multi-wavelength dataset available for Abell 1835 to provide new insights into cooling and feedback in clusters.

  5. Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    1978-09-01

    The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure

  6. Aseismic study of high temperature gas-cooled reactor core with block-type fuel, 3

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1985-01-01

    A two-dimensional horizontal seismic experiment with single axis and simultaneous two-axes excitations was performed to obtain the core seismic design data on the block-type high temperature gas-cooled reactor. Effects of excitation directions and core side support stiffness on characteristics of core displacements and reaction forces of support were revealed. The values of the side reaction forces are the largest in the excitation of flat-to-flat of hexagonal block. Preload from the core periphery to the core center are effective to decrease core displacements and side reaction forces. (author)

  7. A design method to isothermalize the core of high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Takano, M.; Sawa, K.

    1987-01-01

    A practical design method is developed to isothermalize the core of block-type high-temperature gas-cooled reactors (HTGRs). Isothermalization plays an important role in increasing the design margin on fuel temperature. In this method, the fuel enrichment and the size and boron content of the burnable poison rod are determined over the core blockwise so that the axially exponential and radially flat power distribution are kept from the beginning to the end of core life. The method enables conventional HTGRs to raise the outlet gas temperature without increasing the maximum fuel temperature

  8. 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.

  9. Seismic response of high temperature gas-cooled reactor core with block-type fuel, (2)

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1980-01-01

    For the aseismic design of a high temperature gas-cooled reactor (HTGR) with block-type fuel, it is necessary to predict the motion and force of core columns and blocks. To reveal column vibration characteristics in three-dimensional space and impact response, column vibration tests were carried out with a scale model of a one-region section (seven columns) of the HTGR core. The results are as follows: (1) the column has a soft spring characteristic based on stacked blocks connected with loose pins, (2) the column has whirling phenomena, (3) the compression spring force simulating the gas pressure has the effect of raising the column resonance frequency, and (4) the vibration behavior of the stacked block column and impact response of the surrounding columns show agreement between experiment and analysis. (author)

  10. A fast converging CFD model for thermal hydraulic analysis of gas cooled reactor cores

    International Nuclear Information System (INIS)

    Chen, Gary; Anghaie, Samim

    1999-01-01

    A computational fluid dynamics (CFD) approach to the solution of Navier-Stokes equations for the thermal and flow fields of gas cooled reactor cores is presented. An implicit-explicit MacCormack method based on finite volume discretization scheme, in conjunction with the Gauss-Seidel line iteration procedure is utilized to solve axisymmetric, thin-layer Navier-Stokes equations. This numerical method requires only the inversion of block bidiagonal systems rather than block tridiagonal systems, thus yielding savings in computer time and storage requirements. A two-layer algebraic eddy viscosity turbulence model is used in this study. The effects of turbulence are simulated in terms of the eddy viscosity coefficient, which is calculated for an inner and an outer region separately. An enthalpy-rebalancing scheme is implemented to allow the convergence solutions to be obtained with the application of a wall heat flux. The detailed computational analysis developed in this work is used to evaluate many different Nusselt number equations, property corrections, and axial distance corrections. The calculation based on this CFD model is compared with other published results. The good agreement indicates the usefulness of the presented model for the prediction of flow and temperature distributions for gas cooled reactor cores. (author)

  11. Emergency core cooling system

    International Nuclear Information System (INIS)

    Arai, Kenji; Oikawa, Hirohide.

    1990-01-01

    The device according to this invention can ensure cooling water required for emerency core cooling upon emergence such as abnormally, for example, loss of coolant accident, without using dynamic equipments such as a centrifugal pump or large-scaled tank. The device comprises a pressure accumulation tank containing a high pressure nitrogen gas and cooling water inside, a condensate storage tank, a pressure suppression pool and a jet stream pump. In this device there are disposed a pipeline for guiding cooling water in the pressure accumulation tank as a jetting water to a jetting stream pump, a pipeline for guiding cooling water stored in the condensate storage tank and the pressure suppression pool as pumped water to the jetting pump and, further, a pipeline for guiding the discharged water from the jet stream pump which is a mixed stream of pumped water and jetting water into the reactor pressure vessel. In this constitution, a sufficient amount of water ranging from relatively high pressure to low pressure can be supplied into the reactor pressure vessel, without increasing the size of the pressure accumulation tank. (I.S.)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  14. Draft of standard for graphite core components in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shibata, Taiju; Sawa, Kazuhiro; Eto, Motokuni; Kunimoto, Eiji; Shiozawa, Shusaku; Oku, Tatsuo; Maruyama, Tadashi

    2010-01-01

    For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)

  15. Gas cooled reactors

    International Nuclear Information System (INIS)

    Kojima, Masayuki.

    1985-01-01

    Purpose: To enable direct cooling of reactor cores thereby improving the cooling efficiency upon accidents. Constitution: A plurality sets of heat exchange pipe groups are disposed around the reactor core, which are connected by way of communication pipes with a feedwater recycling device comprising gas/liquid separation device, recycling pump, feedwater pump and emergency water tank. Upon occurrence of loss of primary coolants accidents, the heat exchange pipe groups directly absorb the heat from the reactor core through radiation and convection. Although the water in the heat exchange pipe groups are boiled to evaporate if the forcive circulation is interrupted by the loss of electric power source, water in the emergency tank is supplied due to the head to the heat exchange pipe groups to continue the cooling. Furthermore, since the heat exchange pipe groups surround the entire circumference of the reactor core, cooling is carried out uniformly without resulting deformation or stresses due to the thermal imbalance. (Sekiya, K.)

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

    International Nuclear Information System (INIS)

    El Ganaoui, K.

    2006-09-01

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

  17. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  18. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations

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

    International Nuclear Information System (INIS)

    Sui Zhe; Sun Jun; Ma Yuanle; Zhang Ruipeng

    2013-01-01

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

  20. VALIDATION OF NUMERICAL METHODS TO CALCULATE BYPASS FLOW IN A PRISMATIC GAS-COOLED REACTOR CORE

    Directory of Open Access Journals (Sweden)

    NAM-IL TAK

    2013-11-01

    Full Text Available For thermo-fluid and safety analyses of a High Temperature Gas-cooled Reactor (HTGR, intensive efforts are in progress in the developments of the GAMMA+ code of Korea Atomic Energy Research Institute (KAERI and the AGREE code of the University of Michigan (U of M. One of the important requirements for GAMMA+ and AGREE is an accurate modeling capability of a bypass flow in a prismatic core. Recently, a series of air experiments were performed at Seoul National University (SNU in order to understand bypass flow behavior and generate an experimental database for the validation of computer codes. The main objective of the present work is to validate the GAMMA+ and AGREE codes using the experimental data published by SNU. The numerical results of the two codes were compared with the measured data. A good agreement was found between the calculations and the measurement. It was concluded that GAMMA+ and AGREE can reliably simulate the bypass flow behavior in a prismatic core.

  1. Study on the seismic verification test program on the experimental multi-purpose high-temperature gas cooled reactor core

    International Nuclear Information System (INIS)

    Taketani, K.; Aochi, T.; Yasuno, T.; Ikushima, T.; Shiraki, K.; Honma, T.; Kawamura, N.

    1978-01-01

    The paper describes a program of experimental research necessary for qualitative and quantitative determination of vibration characteristics and aseismic safety on structure of reactor core in the multipurpose high temperature gas-cooled experimental reactor (VHTR Experimental Reactor) by the Japan Atomic Energy Research Institute

  2. 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.)

  3. Core design studies on various forms of coolants and fuel materials. 2. Studies on liquid heavy metal and gas cooled cores, small cores and evaluation of 4-type cores

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Sakashita, Yoshiyuki; Naganuma, Masayuki; Takaki, Naoyuki; Mizuno, Tomoyasu; Ikegami, Tetsuo

    2001-01-01

    Alternative concepts to sodium cooled fast reactors, such as heavy metal liquid cooled reactors and gas cooled fast reactors were studied in Phase-1 of the feasibility studies, aiming at simplification of the system, high thermal efficiency and enhancing safety. Fuel and core specifications and nuclear characteristics were surveyed to meet the targets for commercialization of fast reactor cycle. Nuclear characteristics of small fast reactor cores were also surveyed from the perspective of the possibility of multi-purpose use and dispersed power stations. The key points of the design study for each concept in Phase-2 were summarized from the aspect of the screening of the candidates for FR commercialization. (author)

  4. 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

  5. MORECA: A computer code for simulating modular high-temperature gas-cooled reactor core heatup accidents

    International Nuclear Information System (INIS)

    Ball, S.J.

    1991-10-01

    The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to the unavailable. Simulations of long-term loss-of-forced-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the US Department of Energy reference design of a standard MHTGR

  6. Emergency core cooling system

    International Nuclear Information System (INIS)

    Kato, Ken.

    1989-01-01

    In PWR type reactors, a cooling water spray portion of emergency core cooling pipelines incorporated into pipelines on high temperature side is protruded to the inside of an upper plenum. Upon rupture of primary pipelines, pressure in a pressure vessel is abruptly reduced to generate a great amount of steams in the reactor core, which are discharged at a high flow rate into the primary pipelines on high temperature side. However, since the inside of the upper plenum has a larger area and the steam flow is slow, as compared with that of the pipelines on the high temperature side, ECCS water can surely be supplied into the reactor core to promote the re-flooding of the reactor core and effectively cool the reactor. Since the nuclear reactor can effectively be cooled to enable the promotion of pressure reduction and effective supply of coolants during the period of pressure reduction upon LOCA, the capacity of the pressure accumulation vessel can be decreased. Further, the re-flooding time for the reactor is shortened to provide an effect contributing to the improvement of the safety and the reduction of the cost. (N.H.)

  7. Emergency core cooling device

    International Nuclear Information System (INIS)

    Suzaki, Kiyoshi; Inoue, Akihiro.

    1979-01-01

    Purpose: To improve core cooling effect by making the operation region for a plurality of water injection pumps more broader. Constitution: An emergency reactor core cooling device actuated upon failure of recycling pipe ways is adapted to be fed with cooling water through a thermal sleeve by way of a plurality of water injection pump from pool water in a condensate storage tank and a pressure suppression chamber as water feed source. Exhaust pipes and suction pipes of each of the pumps are connected by way of switching valves and the valves are switched so that the pumps are set to a series operation if the pressure in the pressure vessel is high and the pumps are set to a parallel operation if the pressure in the pressure vessel is low. (Furukawa, Y.)

  8. Safety analysis of RSG-GAS Silicide core using one line cooling system

    International Nuclear Information System (INIS)

    Endiah-Puji-Hastuti

    2003-01-01

    In the frame of minimizing the operation-cost, operation mode using one line cooling system is being evaluated. Maximum reactor has been determined and to continuing this program, steady state and transient analysis were done. The analysis was done by means of a core thermal hydraulic code, COOLOD-N, and PARET. The codes solves core thermal hydraulic equation at steady state conditions and transient, respectively. By using silicide core data and coast down flow rate as the input, thermal hydraulics parameters such as fuel cladding and fuel meat temperatures as well as safety margin against flow instability were calculated. Imposing the safety criteria to the results of steady state and transient analysis, maximum permissible power for this operation was obtained as much as 17.1 MW

  9. Fuel cycles and advanced core designs for the Gas-Cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Simon, R.H.; Hamilton, C.J.; Hunter, R.S.

    1982-01-01

    Studies indicate that a 1200 MW(e) Gas-Cooled Fast Breeder Reactor could achieve compound system doubling times of under ten years when using advanced oxide or carbide fuels. In addition, when thorium is used in the breeding blankets, enough U-233 can be generated in each GCFR to supply several advanced converter reactors with fissionable material and this symbiotic relationship could provide energy for the world for centuries. (author)

  10. Emergency core cooling system

    International Nuclear Information System (INIS)

    Ando, Masaki.

    1987-01-01

    Purpose: To actuate an automatic pressure down system (ADS) and a low pressure emergency core cooling system (ECCS) upon water level reduction of a nuclear reactor other than loss of coolant accidents (LOCA). Constitution: ADS in a BWR type reactor is disposed for reducing the pressure in a reactor container thereby enabling coolant injection from a low pressure ECCS upon LOCA. That is, ADS has been actuated by AND signal for a reactor water level low signal and a dry well pressure high signal. In the present invention, ADS can be actuated further also by AND signal of the reactor water level low signal, the high pressure ECCS and not-operation signal of reactor isolation cooling system. In such an emergency core cooling system thus constituted, ADS operates in the same manner as usual upon LOCA and, further, ADS is operated also upon loss of feedwater accident in the reactor pressure vessel in the case where there is a necessity for actuating the low pressure ECCS, although other high pressure ECCS and reactor isolation cooling system are not operated. Accordingly, it is possible to improve the reliability upon reactor core accident and mitigate the operator burden. (Horiuchi, T.)

  11. Core cooling systems

    International Nuclear Information System (INIS)

    Hoeppner, G.

    1980-01-01

    The reactor cooling system transports the heat liberated in the reactor core to the component - heat exchanger, steam generator or turbine - where the energy is removed. This basic task can be performed with a variety of coolants circulating in appropriately designed cooling systems. The choice of any one system is governed by principles of economics and natural policies, the design is determined by the laws of nuclear physics, thermal-hydraulics and by the requirement of reliability and public safety. PWR- and BWR- reactors today generate the bulk of nuclear energy. Their primary cooling systems are discussed under the following aspects: 1. General design, nuclear physics constraints, energy transfer, hydraulics, thermodynamics. 2. Design and performance under conditions of steady state and mild transients; control systems. 3. Design and performance under conditions of severe transients and loss of coolant accidents; safety systems. (orig./RW)

  12. 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.)

  13. Scaling analysis of the coupled heat transfer process in the high-temperature gas-cooled reactor core

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1986-08-01

    The differential equations representing the coupled heat transfer from the solid nuclear core components to the helium in the coolant channels are scaled in terms of representative quantities. This scaling process identifies the relative importance of the various terms of the coupled differential equations. The relative importance of these terms is then used to simplify the numerical solution of the coupled heat transfer for two bounding cases of full-power operation and depressurization from full-system operating pressure for the Fort St. Vrain High-Temperature Gas-Cooled Reactor. This analysis rigorously justifies the simplified system of equations used in the nuclear safety analysis effort at Oak Ridge National Laboratory

  14. Gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    1974-01-01

    The invention aims at simplying gas-cooled nuclear reactors. For the cooling gas, the reactor is provided with a main circulation system comprising one or several energy conversion main groups such as gas turbines, and an auxiliary circulation system comprising at least one steam-generating boiler heated by the gas after its passage through the reactor core and adapted to feed a steam turbine with motive steam. The invention can be applied to reactors the main groups of which are direct-cycle gas turbines [fr

  15. Emergency core cooling system

    International Nuclear Information System (INIS)

    Abe, Nobuaki.

    1993-01-01

    A reactor comprises a static emergency reactor core cooling system having an automatic depressurization system and a gravitationally dropping type water injection system and a container cooling system by an isolation condenser. A depressurization pipeline of the automatic depressurization system connected to a reactor pressure vessel branches in the midway. The branched depressurizing pipelines are extended into an upper dry well and a lower dry well, in which depressurization valves are disposed at the top end portions of the pipelines respectively. If loss-of-coolant accidents should occur, the depressurization valve of the automatic depressurization system is actuated by lowering of water level in the pressure vessel. This causes nitrogen gases in the upper and the lower dry wells to transfer together with discharged steams effectively to a suppression pool passing through a bent tube. Accordingly, the gravitationally dropping type water injection system can be actuated faster. Further, subsequent cooling for the reactor vessel can be ensured sufficiently by the isolation condenser. (I.N.)

  16. 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.)

  17. 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.)

  18. Emergency core cooling systems

    International Nuclear Information System (INIS)

    Kubokoya, Takashi; Okataku, Yasukuni.

    1984-01-01

    Purpose: To maintain the fuel soundness upon loss of primary coolant accidents in a pressure tube type nuclear reactor by injecting cooling heavy water at an early stage, to suppress the temperature of fuel cans at a lower level. Constitution: When a thermometer detects the temperature rise and a pressure gauge detects that the pressure for the primary coolants is reduced slightly from that in the normal operation upon loss of coolant accidents in the vicinity of the primary coolant circuit, heavy water is caused to flow in the heavy water feed pipeway by a controller. This enables to inject the heavy water into the reactor core in a short time upon loss of the primary coolant accidents to suppress the temperature rise in the fuel can thereby maintain the fuel soundness. (Moriyama, K.)

  19. Casting core for a cooling arrangement for a gas turbine component

    Science.gov (United States)

    Lee, Ching-Pang; Heneveld, Benjamin E

    2015-01-20

    A ceramic casting core, including: a plurality of rows (162, 166, 168) of gaps (164), each gap (164) defining an airfoil shape; interstitial core material (172) that defines and separates adjacent gaps (164) in each row (162, 166, 168); and connecting core material (178) that connects adjacent rows (170, 174, 176) of interstitial core material (172). Ends of interstitial core material (172) in one row (170, 174, 176) align with ends of interstitial core material (172) in an adjacent row (170, 174, 176) to form a plurality of continuous and serpentine shaped structures each including interstitial core material (172) from at least two adjacent rows (170, 174, 176) and connecting core material (178).

  20. Gas-cooled reactors

    International Nuclear Information System (INIS)

    Vakilian, M.

    1977-05-01

    The present study is the second part of a general survey of Gas Cooled Reactors (GCRs). In this part, the course of development, overall performance and present development status of High Temperature Gas Cooled Reactors (HTCRs) and advances of HTGR systems are reviewed. (author)

  1. Special power supply and control system for the gas-cooled fast reactor-core flow test loop

    International Nuclear Information System (INIS)

    Hudson, T.L.

    1981-09-01

    The test bundle in the Gas-Cooled Fast Reactor-Core Flow Test Loop (GCFR-CFTL) requires a source of electrical power that can be controlled accurately and reliably over a wide range of steady-state and transient power levels and skewed power distributions to simulate GCFR operating conditions. Both ac and dc power systems were studied, and only those employing silicon-controlled rectifiers (SCRs) could meet the requirements. This report summarizes the studies, tests, evaluations, and development work leading to the selection. it also presents the design, procurement, testing, and evaluation of the first 500-kVa LMPL supply. The results show that the LMPL can control 60-Hz sine wave power from 200 W to 500 kVA

  2. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Majumdar, Saurindranath; Srinivasan, Makuteswara

    2013-01-01

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite

  3. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Majumdar, Saurindranath [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

    2013-07-15

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.

  4. Emergency core cooling system

    International Nuclear Information System (INIS)

    Sato, Akira; Kobayashi, Masahide.

    1983-01-01

    Purpose: To enable a stable operation of an emergency core cooling system by preventing the system from the automatic stopping at an abnormally high level of the reactor water during its operation. Constitution: A pump flow rate signal and a reactor water level signal are used and, when the reactor water level is increased to a predetermined level, the pump flow rate is controlled by the reactor water level signal instead of the flow rate signal. Specifically, when the reactor water level is gradually increased by the water injection from the pump and exceeds a setting signal for the water level, the water level deviation signal acts as a demand signal for the decrease in the flow rate of the pump and the output signal from the water level controller is also decreased depending on the control constant. At a certain point, the output signal from the water level controller becomes smaller than the output signal from the flow rate controller. Thus, the output signal from the water level controller is outputted as the output signal for the lower level preference device. In this way, the reactor water level and the pump flow rate can be controlled within a range not exceeding the predetermined pump flow rate. (Horiuchi, T.)

  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. 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.)

  7. Fuel Summary for Peach Bottom Unit 1 High-Temperature Gas-Cooled Reactor Cores 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Karel I. Kingrey

    2003-04-01

    This fuel summary report contains background and summary information for the Peach Bottom Unit 1, High-Temperature, Gas-Cooled Reactor Cores 1 and 2. This report contains detailed information about the fuel in the two cores, the Peach Bottom Unit 1 operating history, nuclear parameters, physical and chemical characteristics, and shipping and storage canister related data. The data in this document have been compiled from a large number of sources and are not qualified beyond the qualification of the source documents. This report is intended to provide an overview of the existing data pertaining to spent fuel management and point to pertinent reference source documents. For design applications, the original source documentation must be used. While all referenced sources are available as records or controlled documents at the Idaho National Engineering and Environmental Laboratory (INEEL), some of the sources were marked as informal or draft reports. This is noted where applicable. In some instances, source documents are not consistent. Where they are known, this document identifies those instances and provides clarification where possible. However, as stated above, this document has not been independently qualified and such clarifications are only included for information purposes. Some of the information in this summary is available in multiple source documents. An effort has been made to clearly identify at least one record document as the source for the information included in this report.

  8. Thermohydraulics in a high-temperature gas-cooled reactor primary loop during early phases of unrestricted core-heatup accidents

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Colman, J.; Hsu, C.J.

    1983-01-01

    In High Temperature Gas Cooled Reactor (HTGR) siting considerations, the Unrestricted Core Heatup Accidents (UCHA) are considered as accidents of highest consequence, corresponding to core meltdown accidents in light water reactors. Initiation of such accidents can be, for instance, due to station blackout, resulting in scram and loss of all main loop forced circulation, with none of the core auxiliary cooling system loops being started. The result is a slow but continuing core heatup, extending over days. During the initial phases of such UCHA scenarios, the primary loop remains pressurized, with the system pressure slowly increasing until the relief valve setpoint is reached. The major objectives of the work described here were to determine times to depressurization as well as approximate loop component temperatures up to depressurization

  9. Rhapsody-G simulations I: the cool cores, hot gas and stellar content of massive galaxy clusters

    International Nuclear Information System (INIS)

    Hahn, Oliver; Martizzi, Davide; Wu, Hao-Yi

    2017-01-01

    We present the rhapsody-g suite of cosmological hydrodynamic zoom simulations of 10 massive galaxy clusters at the M vir ~10 15 M ⊙ scale. These simulations include cooling and subresolution models for star formation and stellar and supermassive black hole feedback. The sample is selected to capture the whole gamut of assembly histories that produce clusters of similar final mass. We present an overview of the successes and shortcomings of such simulations in reproducing both the stellar properties of galaxies as well as properties of the hot plasma in clusters. In our simulations, a long-lived cool-core/non-cool-core dichotomy arises naturally, and the emergence of non-cool cores is related to low angular momentum major mergers. Nevertheless, the cool-core clusters exhibit a low central entropy compared to observations, which cannot be alleviated by thermal active galactic nuclei feedback. For cluster scaling relations, we find that the simulations match well the M 500 –Y 500 scaling of Planck Sunyaev–Zeldovich clusters but deviate somewhat from the observed X-ray luminosity and temperature scaling relations in the sense of being slightly too bright and too cool at fixed mass, respectively. Stars are produced at an efficiency consistent with abundance-matching constraints and central galaxies have star formation rates consistent with recent observations. In conclusion, while our simulations thus match various key properties remarkably well, we conclude that the shortcomings strongly suggest an important role for non-thermal processes (through feedback or otherwise) or thermal conduction in shaping the intracluster medium.

  10. Gas-cooled reactors

    International Nuclear Information System (INIS)

    Schulten, R.; Trauger, D.B.

    1976-01-01

    Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

  11. Emergency cooling system for a gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Cook, R.K.; Burylo, P.S.

    1975-01-01

    The site of the gas-cooled reactor with direct-circuit gas turbine is preferably the sea coast. An emergency cooling system with safety valve and emergency feed-water addition is designed which affects at least a part of the reactor core coolant after leaving the core. The emergency cooling system includes a water emergency cooling circuit with heat exchanger for the core coolant. The safety valve releases water or steam from the emergency coolant circuit when a certain temperature is exceeded; this is, however, replaced by the emergency feed-water. If the gas turbine exhibits a high and low pressure turbine stage, which are flowed through by coolant one behind another, a part of the coolant can be removed in front of each part turbine by two valves and be added to the haet exchanger. (RW/LH) [de

  12. Thermal response of core and central-cavity components of a high-temperature gas-cooled reactor in the absence of forced convection coolant flow

    International Nuclear Information System (INIS)

    Whaley, R.L.; Sanders, J.P.

    1976-09-01

    A means of determining the thermal responses of the core and the components of a high-temperature gas-cooled reactor after loss of forced coolant flow is discussed. A computer program, using a finite-difference technique, is presented together with a solution of the confined natural convection. The results obtained are reasonable and demonstrate that the computer program adequately represents the confined natural convection

  13. Specialists' meeting on gas-cooled reactor core and high temperature instrumentation, Windermere, UK, 15-17 June 1982. Summary report

    International Nuclear Information System (INIS)

    1982-09-01

    The Specialists' Meeting on ''Gas-Cooled Reactor Core and High Temperature Instrumentation'' was held at the Beech Hill Hotel, Windermere in England on June 15-17 1982. The meeting was sponsored by the IAEA on the recommendation of the International Working Group on Gas Cooled Reactors and was hosted by the Windscale Nuclear Power Development Laboratories of the UKAEA. The meeting was attended by 43 participants from Belgium, France, Federal Republic of Germany, Japan, United Kingdom of Great Britain and Northern Ireland and the United States of America. The objective of the meeting was to provide a forum, both formal and informal, for the exchange and discussion of technical information relating to instrumentation being used or under development for the measurement of core parameters, neutron flux, temperature, coolant flow etc. in gas cooled reactors. The technical part of the meeting was divided into five subject sessions: (A) Temperature Measurement (B) Neutron Detection Instrumentation (C) HTR Instrumentation - General (D) Gas Analysis and Failed Fuel Detection (E) Coolant Mass Flow and Leak Detection. A total of twenty-five papers were presented by the participants on behalf of their organizations during the meeting. A programme of the meeting and list of participants are given in appendices to this report

  14. Thermohydraulics in a high-temperature gas-cooled reactor prestressed-concrete reactor vessel during unrestricted core-heatup accidents

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Colman, J.; Araj, K.

    1983-01-01

    The hypothetical accident considered for siting considerations in High Temperature Gas-Cooled Reactors (HTGR) is the so called Unrestricted Core Heatup Accident (UCHA), in which all forced circulation is lost at initiation, and none of the auxillary cooling loops can be started. The result is a gradual slow core heatup, extending over days. Whether the liner cooling system (LCS) operates during this time is of crucial importance. If it does not, the resulting concrete decomposition of the prestressed concrete reactor vessel (PCRV) will ultimately cause containment building (CB) failure after about 6 to 10 days. The primary objective of the work described here was to establish for such accident conditions the core temperatures and approximate fuel failure rates, to check for potential thermal barrier failures, and to follow the PCRV concrete temperatures, as well as PCRV gas releases from concrete decomposition. The work was done for the General Atomic Corporation Base Line Zero reactor of 2240 MW(t). Most results apply at least qualitatively also to other large HTGR steam cycle designs

  15. A Computational Fluid Dynamic and Heat Transfer Model for Gaseous Core and Gas Cooled Space Power and Propulsion Reactors

    Science.gov (United States)

    Anghaie, S.; Chen, G.

    1996-01-01

    A computational model based on the axisymmetric, thin-layer Navier-Stokes equations is developed to predict the convective, radiation and conductive heat transfer in high temperature space nuclear reactors. An implicit-explicit, finite volume, MacCormack method in conjunction with the Gauss-Seidel line iteration procedure is utilized to solve the thermal and fluid governing equations. Simulation of coolant and propellant flows in these reactors involves the subsonic and supersonic flows of hydrogen, helium and uranium tetrafluoride under variable boundary conditions. An enthalpy-rebalancing scheme is developed and implemented to enhance and accelerate the rate of convergence when a wall heat flux boundary condition is used. The model also incorporated the Baldwin and Lomax two-layer algebraic turbulence scheme for the calculation of the turbulent kinetic energy and eddy diffusivity of energy. The Rosseland diffusion approximation is used to simulate the radiative energy transfer in the optically thick environment of gas core reactors. The computational model is benchmarked with experimental data on flow separation angle and drag force acting on a suspended sphere in a cylindrical tube. The heat transfer is validated by comparing the computed results with the standard heat transfer correlations predictions. The model is used to simulate flow and heat transfer under a variety of design conditions. The effect of internal heat generation on the heat transfer in the gas core reactors is examined for a variety of power densities, 100 W/cc, 500 W/cc and 1000 W/cc. The maximum temperature, corresponding with the heat generation rates, are 2150 K, 2750 K and 3550 K, respectively. This analysis shows that the maximum temperature is strongly dependent on the value of heat generation rate. It also indicates that a heat generation rate higher than 1000 W/cc is necessary to maintain the gas temperature at about 3500 K, which is typical design temperature required to achieve high

  16. Finite element modeling of fluid/thermal/structural interaction for a gas-cooled fast reactor core

    International Nuclear Information System (INIS)

    Bennett, J.G.; Ju, F.D.

    1980-01-01

    Two nonlinear finite element formulations for application to a series of experiments in the Gas-Cooled Fast Reactor (GCFR) development program are described. An efficient beam column element for moderately large deformations is combined with a finite element developed for an engineering description of a convecting fluid. Typical results from both elements are illustrated. A combined application for a problem typical of the GCFR loss-of-coolant experiments is illustrated. These problems are not the usual fluid structural interaction problems in that the inertia coupling is negligible while the thermal coupling is very important

  17. Warming rays in cluster cool cores

    Science.gov (United States)

    Colafrancesco, S.; Marchegiani, P.

    2008-06-01

    Context: Cosmic rays are confined in the atmospheres of galaxy clusters and, therefore, they can play a crucial role in the heating of their cool cores. Aims: We discuss here the thermal and non-thermal features of a model of cosmic ray heating of cluster cores that can provide a solution to the cooling-flow problems. To this aim, we generalize a model originally proposed by Colafrancesco, Dar & DeRujula (2004) and we show that our model predicts specific correlations between the thermal and non-thermal properties of galaxy clusters and enables various observational tests. Methods: The model reproduces the observed temperature distribution in clusters by using an energy balance condition in which the X-ray energy emitted by clusters is supplied, in a quasi-steady state, by the hadronic cosmic rays, which act as “warming rays” (WRs). The temperature profile of the intracluster (IC) gas is strictly correlated with the pressure distribution of the WRs and, consequently, with the non-thermal emission (radio, hard X-ray and gamma-ray) induced by the interaction of the WRs with the IC gas and the IC magnetic field. Results: The temperature distribution of the IC gas in both cool-core and non cool-core clusters is successfully predicted from the measured IC plasma density distribution. Under this contraint, the WR model is also able to reproduce the thermal and non-thermal pressure distribution in clusters, as well as their radial entropy distribution, as shown by the analysis of three clusters studied in detail: Perseus, A2199 and Hydra. The WR model provides other observable features of galaxy clusters: a correlation of the pressure ratio (WRs to thermal IC gas) with the inner cluster temperature (P_WR/P_th) ˜ (kT_inner)-2/3, a correlation of the gamma-ray luminosity with the inner cluster temperature Lγ ˜ (kT_inner)4/3, a substantial number of cool-core clusters observable with the GLAST-LAT experiment, a surface brightness of radio halos in cool-core clusters

  18. Magnitude and reactivity consequences of accidental moisture ingress into the Modular High-Temperature Gas-Cooled Reactor core

    International Nuclear Information System (INIS)

    Smith, O.L.

    1992-01-01

    Accidental admission of moisture into the primary system of a Modular High-Temperature Gas-Cooled Reactor (MHTGR) has been identified in US Department of Energy-sponsored studies as an important safety concern. The work described here develops an analytical methodology to quantify the pressure and reactivity consequences of steam-generator tube rupture and other moistureingress-related incidents. Important neutronic and thermohydraulic processes are coupled with reactivity feedback and safety and control system responses. Rate and magnitude of steam buildup are found to be dominated by major system features such as break size in comparison with safety valve capacity and reliability, while being less sensitive to factors such as heat transfer coefficients. The results indicate that ingress transients progress at a slower pace than previously predicted by bounding analyses, with milder power overshoots and more time for operator or automatic corrective actions

  19. Magnitude and reactivity consequences of moisture ingress into the modular High-Temperature Gas-Cooled Reactor core

    International Nuclear Information System (INIS)

    Smith, O.L.

    1992-12-01

    Inadvertent admission of moisture into the primary system of a modular high-temperature gas-cooled reactor has been identified in US Department of Energy-sponsored studies as an important safety concern. The work described here develops an analytical methodology to quantify the pressure and reactivity consequences of steam-generator tube rupture and other moisture-ingress-related incidents. Important neutronic and thermohydraulic processes are coupled with reactivity feedback and safety and control system responses. The rate and magnitude of steam buildup are found to be dominated by major system features such as break size compared with safety valve capacity and reliability and less sensitive to factors such as heat transfer coefficients. The results indicate that ingress transients progress at a slower pace than previously predicted by bounding analyses, with milder power overshoots and more time for operator or automatic corrective actions

  20. Experiments on graphite block gaps connected with leak flow in bottom-core structure of experimental very high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Kikuchi, Kenji; Futakawa, Masatoshi; Takizuka, Takakazu; Kaburaki, Hideo; Sanokawa, Konomo

    1984-01-01

    In order to minimize the leak flow rate of an experimental VHTR (a multi-purpose very high-temperature gas-cooled reactor), the graphite blocks are tightened to reduce the gap distance between blocks by core restrainers surrounded outside of the fixed reflectors of the bottom-core structure and seal elements are placed in the gaps. By using a 1/2.75-scale model of the bottom-core structure, the experiments on the following items have been carried out: a relationship between core restraint force and block gap, a relationship between core restraint force and inclined angle of the model, leak flow characteristics of seal elements etc. The conclusions derived from the experiments are as follows: (1) Core restraint force is significantly effective for decreasing the gap distance between hot plenum blocks, but ineffective for the gap between hot plenum block and fixed reflector. (2) Graphite seal element reduces the leak flow rate from the top surface of hot plenum block into plenum region to one-third. (author)

  1. 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

  2. CORTAP: a coupled neutron kinetics-heat transfer digital computer program for the dynamic simulation of the high temperature gas cooled reactor core

    International Nuclear Information System (INIS)

    Cleveland, J.C.

    1977-01-01

    CORTAP (Core Transient Analysis Program) was developed to predict the dynamic behavior of the High Temperature Gas Cooled Reactor (HTGR) core under normal operational transients and postulated accident conditions. CORTAP is used both as a stand-alone component simulation and as part of the HTGR nuclear steam supply (NSS) system simulation code ORTAP. The core thermal neutronic response is determined by solving the heat transfer equations for the fuel, moderator and coolant in an average powered region of the reactor core. The space independent neutron kinetics equations are coupled to the heat transfer equations through a rapidly converging iterative technique. The code has the capability to determine conservative fuel, moderator, and coolant temperatures in the ''hot'' fuel region. For transients involving a reactor trip, the core heat generation rate is determined from an expression for decay heat following a scram. Nonlinear effects introduced by temperature dependent fuel, moderator, and coolant properties are included in the model. CORTAP predictions will be compared with dynamic test results obtained from the Fort St. Vrain reactor owned by Public Service of Colorado, and, based on these comparisons, appropriate improvements will be made in CORTAP

  3. Full Core Criticality Modeling of Gas-Cooled Fast Reactor Using the SCALE6.0 and MCNP5 Code Packages

    International Nuclear Information System (INIS)

    Matijevic, M.; Jecmenica, R.; Pevec, D.; Trontl, K.

    2012-01-01

    The Gas-Cooled Fast Reactor (GFR) is one of the reactor concepts selected by the Generation IV International Forum (GIF) for the next generation of innovative nuclear energy systems. It was selected among a group of more than 100 prototypes and his commercial availability is expected by 2030. GFR has common goals of the rest GIF advanced reactor types: economy, safety, proliferation resistance, availability and sustainability. Several GFR fuel design concepts such as plates, rod pins and pebbles are currently being investigated in order to meet the high temperature constraints characteristic for a GFR working enviroment. In the previous study we have compared the fuel depletion results for heterogeneous GFR fuel assembly (FA), obtained with TRITON6 sequence of SCALE6.0 code system, with the MCNPX-CINDER90 and TRIPOLI-4-D codes. Present work is a continuation of neutronic criticality analysis of heterogeneous FA and full core configurations of a GFR concept using 3-D Monte Carlo codes KENO-VI/SCALE6.0 and MCNP5. The FA is based on a hexagonal mesh of fuel rods (uranium and plutonium carbide fuel, silicon carbide clad, helium gas coolant) with axial reflector thickness being varied for the purpose of optimization. Three reflector materials were analysed: zirconium carbide (ZrC), silicon carbide (SiC) and natural uranium. ZrC has been selected as a reflector material, having the best contribution to the neutron economy and to the reactivity of the core. The core safety parameters were also analysed: a negative temperature coefficient of reactivity was verified for the heavy metal fuel and coolant density loss. Criticality calculations of different FA active heights were performed and the reflector thickness was also adjusted. Finally, GFR full core criticality calculations using different active fuel rod heights and fixed ZrC reflector height were done to find the optimal height of the core. The Shannon entropy of the GFR core fission distribution was proved to be

  4. Gas cooled leads

    International Nuclear Information System (INIS)

    Shutt, R.P.; Rehak, M.L.; Hornik, K.E.

    1993-01-01

    The intent of this paper is to cover as completely as possible and in sufficient detail the topics relevant to lead design. The first part identifies the problems associated with lead design, states the mathematical formulation, and shows the results of numerical and analytical solutions. The second part presents the results of a parametric study whose object is to determine the best choice for cooling method, material, and geometry. These findings axe applied in a third part to the design of high-current leads whose end temperatures are determined from the surrounding equipment. It is found that cooling method or improved heat transfer are not critical once good heat exchange is established. The range 5 5 but extends over a large of values. Mass flow needed to prevent thermal runaway varies linearly with current above a given threshold. Below that value, the mass flow is constant with current. Transient analysis shows no evidence of hysteresis. If cooling is interrupted, the mass flow needed to restore the lead to its initially cooled state grows exponentially with the time that the lead was left without cooling

  5. Gas hydrate cool storage system

    Science.gov (United States)

    Ternes, M.P.; Kedl, R.J.

    1984-09-12

    The invention presented relates to the development of a process utilizing a gas hydrate as a cool storage medium for alleviating electric load demands during peak usage periods. Several objectives of the invention are mentioned concerning the formation of the gas hydrate as storage material in a thermal energy storage system within a heat pump cycle system. The gas hydrate was formed using a refrigerant in water and an example with R-12 refrigerant is included. (BCS)

  6. Measurement of reactivity worths of burnable poison rods in enriched uranium graphite-moderated core simulated to high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Akino, Fujiyoshi; Takeuchi, Motoyoshi; Kitadate, Kenji; Yoshifuji, Hisashi; Kaneko, Yoshihiko

    1980-11-01

    As the core design for the Experimental Very High Temperature Gas Cooled Reactor progresses, evaluation of design precision has become increasingly important. For a high precision design, it is required to have adequate group constants based on accurate nuclear data, as well as calculation methods properly describing the physical behavior of neutrons. We, therefore, assembled a simulation core for VHTR, SHE-14, using a graphite-moderated 20%-enriched uranium Semi-Homogeneous Experimental Critical Facility (SHE), and obtained useful experimental data in evaluating the design precision. The VHTR is designed to accommodate burnable poison and control rods for reactivity compensation. Accordingly, the experimental burnable poison rods which are similar to those to be used in the experimental reactor were prepared, and their reactivity values were measured in the SHE-14 core. One to three rods of the above experimental burnable poison rods were inserted into the central column of the SHE-14 core, and the reactivity values were measured by the period and fuel rod substitution method. The results of the measurements have clearly shown that due to the self-shielding effect of B 4 C particles the reactivity value decreases with increasing particle diameter. For the particle diameter, the reactivity value is found to increase linearly with the logarithm of boron content. The measured values and those calculated are found to agree with each other within 5%. These results indicate that the reactivity of the burnable poison rod can be estimated fairly accurately by taking into account the self-shielding effect of B 4 C particles and the heterogeneity of the lattice cell. (author)

  7. Gas cooled HTR

    International Nuclear Information System (INIS)

    Schweiger, F.

    1985-01-01

    In the He-cooled, graphite-moderated HTR with spherical fuel elements, the steam generator is fixed outside the pressure vessel. The heat exchangers are above the reactor level. The hot gases stream from the reactor bottom over the heat exchanger, through an annular space around the heat exchanger and through feed lines in the side reflector of the reactor back to its top part. This way, in case of shutdown there is a supplementary natural draught that helps the inner natural circulation (chimney draught effect). (orig./PW)

  8. OSCIL: one-dimensional spring-mass system simulator for seismic analysis of high temperature gas cooled reactor core

    International Nuclear Information System (INIS)

    Lasker, L.

    1976-01-01

    OSCIL is a program to predict the effects of seismic input on a HTGR core. The present model is a one-dimensional array of blocks with appropriate spring constants, inter-elemental and ground damping, and clearances. It can be used more generally for systems of moving masses separated by nonlinear springs and dampers

  9. OSCIL: one-dimensional spring-mass system simulator for seismic analysis of high temperature gas cooled reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Lasker, L. (ed.)

    1976-01-01

    OSCIL is a program to predict the effects of seismic input on a HTGR core. The present model is a one-dimensional array of blocks with appropriate spring constants, inter-elemental and ground damping, and clearances. It can be used more generally for systems of moving masses separated by nonlinear springs and dampers.

  10. Gas cooled traction drive inverter

    Science.gov (United States)

    Chinthavali, Madhu Sudhan

    2013-10-08

    The present invention provides a modular circuit card configuration for distributing heat among a plurality of circuit cards. Each circuit card includes a housing adapted to dissipate heat in response to gas flow over the housing. In one aspect, a gas-cooled inverter includes a plurality of inverter circuit cards, and a plurality of circuit card housings, each of which encloses one of the plurality of inverter cards.

  11. Development of C/C composite for the core component of the high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. Y.; Kim, W. J.; Ryu, W. S.; Jang, J. H

    2005-01-15

    This report reviewed a state of the art on development of C/C composite for the core components for VHTR and described the followings items. The fabrication methods of C/C composites. Summary on the JAERI report (JAERI-Res 2002-026) on the process screening test for the selection of a proper C/C composite material. Review of the proceedings presented at the GEN-IV VHTR material PMB meeting. A status of the domestic commercial C/C composite. The published property data and the characteristics of the commercial C/C composite.

  12. Development of C/C composite for the core component of the high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Park, J. Y.; Kim, W. J.; Ryu, W. S.; Jang, J. H.

    2005-01-01

    This report reviewed a state of the art on development of C/C composite for the core components for VHTR and described the followings items. The fabrication methods of C/C composites. Summary on the JAERI report (JAERI-Res 2002-026) on the process screening test for the selection of a proper C/C composite material. Review of the proceedings presented at the GEN-IV VHTR material PMB meeting. A status of the domestic commercial C/C composite. The published property data and the characteristics of the commercial C/C composite

  13. Reference core design Mark-III of the experimental multi-purpose, high-temperature, gas-cooled reactor

    International Nuclear Information System (INIS)

    Shindo, Ryuiti; Watanabe, Takashi; Ishiguro, Okikazu; Kuroki, Syuzi

    1977-10-01

    The reactivity control system is one of the important items in reactor design, but it is much restricted by structural design of fuel element and pressure vessel in the experimental multi-purpose, high-temperature reactor. Preceding the first conceptual design of the reactor, therefore, the reactivity control system composed of control rod, burnable poison and reserve shutdown system in Mark-II design was re-studied, and several improvements were indicated. (1) The diameter of control rods must be as large as possible because it is impossible to increase the number of control rods. (2) The accuracy in estimation of the reactivity to be compensated with control rods is important because of the mutual interference of pair control rods with the twin configuration in a fuel element. (3) The improvement of core performance in burnup is accompanied by the reduction of design margin for control rods. (4) Increase of the reactivity to be compensated with the burnable poison leads to increase of the core reactivity recovery with burnup, and the assertion of the decrease for recovery of reactivity leads to increase of the temperature dependency of reactivity compensated with control rods. (5) Reduction of reactivity to be compensated with control rods is thus limited by cancellation of the effects in the reactivity recovery and the reactivity temperature dependency. (6) The reserve shutdown system can be designed with margin under the condition of excluding the reactivity of burnup from that to be compensated. (auth.)

  14. Stress relaxation and creep of high-temperature gas-cooled reactor core support ceramic materials: a literature search

    International Nuclear Information System (INIS)

    Selle, J.E.; Tennery, V.J.

    1980-05-01

    Creep and stress relaxation in structural ceramics are important properties to the high-temperature design and safety analysis of the core support structure of the HTGR. The ability of the support structure to function for the lifetime of the reactor is directly related to the allowable creep strain and the ability of the structure to withstand thermal transients. The thermal-mechanical response of the core support pads to steady-state stresses and potential thermal transients depends on variables, including the ability of the ceramics to undergo some stress relaxation in relatively short times. Creep and stress relaxation phenomena in structural ceramics of interest were examined. Of the materials considered (fused silica, alumina, silicon nitride, and silicon carbide), alumina has been more extensively investigated in creep. Activation energies reported varied between 482 and 837 kJ/mole, and consequently, variations in the assigned mechanisms were noted. Nabarro-Herring creep is considered as the primary creep mechanism and no definite grain size dependence has been identified. Results for silicon nitride are in better agreement with reported activation energies. No creep data were found for fused silica or silicon carbide and no stress relaxation data were found for any of the candidate materials. While creep and stress relaxation are similar and it is theoretically possible to derive the value of one property when the other is known, no explicit demonstrated relationship exists between the two. For a given structural ceramic material, both properties must be experimentally determined to obtain the information necessary for use in high-temperature design and safety analyses

  15. High temperature gas cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hosegood, S.B.; Lockett, G.E.

    1975-01-01

    For high-temperature gas cooled reactors it is considered advantageous to design the core so that the moderator blocks can be removed and replaced by some means of standpipes normally situated in the top of the reactor vessel. An arrangement is here described to facilitate these operations. The blocks have end faces shaped as irregular hexagons with three long sides of equal length and three short sides also of equal length, one short side being located between each pair of adjacent long sides, and the long sides being inclined towards one another at 60 0 . The block defines a number of coolant channels located parallel to its sides. Application of the arrangement to a high temperature gas-cooled reactor with refuelling standpipes is described. The standpipes are located in the top of the reactor vessel above the tops of the columns and are disposed coaxially above the hexagonal channels, with diameters that allow the passage of the blocks. (U.K.)

  16. Updating of ASME Nuclear Code Case N-201 to Accommodate the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development

    International Nuclear Information System (INIS)

    Basol, Mit; Kielb, John F.; MuHooly, John F.; Smit, Kobus

    2007-01-01

    On September 29, 2005, ASME Standards Technology, LLC (ASME ST-LLC) executed a multi-year, cooperative agreement with the United States DOE for the Generation IV Reactor Materials project. The project's objective is to update and expand appropriate materials, construction, and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. Task 4 was embarked upon in recognition of the large quantity of ongoing reactor designs utilizing high temperature technology. Since Code Case N-201 had not seen a significant revision (except for a minor revision in September, 2006 to change the SA-336 forging reference for 304SS and 316SS to SA-965 in Tables 1.2(a) and 1.2(b), and some minor editorial changes) since December 1994, identifying recommended updates to support the current high temperature Core Support Structure (CSS) designs and potential new designs was important. As anticipated, the Task 4 effort identified a number of Code Case N-201 issues. Items requiring further consideration range from addressing apparent inconsistencies in definitions and certain material properties between CC-N-201 and Subsection NH, to inclusion of additional materials to provide the designer more flexibility of design. Task 4 developed a design parameter survey that requested input from the CSS designers of ongoing high temperature gas cooled reactor metallic core support designs. The responses to the survey provided Task 4 valuable input to identify the design operating parameters and future needs of the CSS designers. Types of materials, metal temperature, time of exposure, design pressure, design life, and fluence levels were included in the Task 4 survey responses. The results of the survey are included in this report. This research proves that additional work must be done to update Code Case N-201. Task 4 activities provide the framework for the Code Case N-201 update and future work to provide input on materials. Candidate

  17. Emergency cooling method and system for gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Kumpf, H.

    1982-01-01

    For emergency cooling of gas-cooled fast breeder reactors (GSB), which have a core consisting of a fission zone and a breeding zone, water is sprayed out of nozzles on to the core from above in the case of an incident. The water which is not treated with boron is taken out of a reservoir in the form of a storage tank in such a maximum quantity that the cooling water gathering in the space below the core rises at most up to the lower edge of the fission zone. (orig./GL) [de

  18. 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.)

  19. AGN Heating in Simulated Cool-core Clusters

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yuan; Ruszkowski, Mateusz [Department of Astronomy, University of Michigan, 1085 S. University Avenue, Ann Arbor, MI 48109 (United States); Bryan, Greg L., E-mail: yuanlium@umich.edu [Department of Astronomy, Columbia University, Pupin Physics Laboratories, New York, NY 10027 (United States)

    2017-10-01

    We analyze heating and cooling processes in an idealized simulation of a cool-core cluster, where momentum-driven AGN feedback balances radiative cooling in a time-averaged sense. We find that, on average, energy dissipation via shock waves is almost an order of magnitude higher than via turbulence. Most of the shock waves in the simulation are very weak shocks with Mach numbers smaller than 1.5, but the stronger shocks, although rare, dissipate energy more effectively. We find that shock dissipation is a steep function of radius, with most of the energy dissipated within 30 kpc, more spatially concentrated than radiative cooling loss. However, adiabatic processes and mixing (of post-shock materials and the surrounding gas) are able to redistribute the heat throughout the core. A considerable fraction of the AGN energy also escapes the core region. The cluster goes through cycles of AGN outbursts accompanied by periods of enhanced precipitation and star formation, over gigayear timescales. The cluster core is under-heated at the end of each cycle, but over-heated at the peak of the AGN outburst. During the heating-dominant phase, turbulent dissipation alone is often able to balance radiative cooling at every radius but, when this is occurs, shock waves inevitably dissipate even more energy. Our simulation explains why some clusters, such as Abell 2029, are cooling dominated, while in some other clusters, such as Perseus, various heating mechanisms including shock heating, turbulent dissipation and bubble mixing can all individually balance cooling, and together, over-heat the core.

  20. Impingement jet cooling in gas turbines

    CERN Document Server

    Amano, R S

    2014-01-01

    Due to the requirement for enhanced cooling technologies on modern gas turbine engines, advanced research and development has had to take place in field of thermal engineering. Impingement jet cooling is one of the most effective in terms of cooling, manufacturability and cost. This is the first to book to focus on impingement cooling alone.

  1. Prevention and investigations of core degradation in case of beyond design accidents of the 2400 MWTH gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Bertrand, F.; Gatin, V.; Bentivoglio, F.; Gueneau, C.

    2011-01-01

    The present paper deals with studies carried out to assess the ability of the core of the Gas Fast Reactor (GFR) to withstand beyond design accidents. The work presented here is aimed at simulating the behaviour of this core by using analytical models whose input parameters are calculated with the CATHARE2 code. Among possible severe accident initiators, the Unprotected Loss Of Coolant Accident (ULOCA of 3 Inches diameter) is investigated in detail in the paper with CATHARE2. Additionally, a simplified pessimistic assessment of the effect of a postulated power excursion that could result from the failure of prevention provisions is presented. (author)

  2. R + D work on gas-cooled breeder development

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.; Jacobs, G.; Meyer, L.; Rehme, K.; Schumacher, G.; Wilhelm, D.

    1978-01-01

    The development work for the gas-cooled breeder in the Karlsruhe Nuclear Research Center may be assigned to two different groups: a) Investigations on fuel elements. b) Studies concerning the safety of gas-cooled fast breeder reactors. To the first group there belongs the work related to the: - heat transfer between fuel elements and coolant gas, - influence of increased content of water vapor in helium or the fuel rods. The second group concerns: - establishing a computer code for transient calculations in the primary and secondary circuit of a gas-cooled fast breeder reactor, - steam reactivity coefficients, - the core destruction phase of hypothetical accidents, - the core-catcher using borax. (orig./RW) [de

  3. Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

    Science.gov (United States)

    Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

    2017-07-01

    Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (Dpower is 300 MWth. First calculation, we calculate survey parameter for UN-PuN fuel with fissile contain from Plutonium waste LWR for each geometry. The minimum power density is around 72 Watt/cc, and maximum power density 114 Watt/cc. After we calculate with various geometry core, when we use the balance geometry, the k-eff value flattest and more stable than the others.

  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. Natural gas cooling: Part of the solution

    International Nuclear Information System (INIS)

    Jones, D.R.

    1992-01-01

    This paper reviews and compares the efficiencies and performance of a number of gas cooling systems with a comparable electric cooling system. The results show that gas cooling systems compare favorably with the electric equivalents, offering a new dimension to air conditioning and refrigeration systems. The paper goes on to compare the air quality benefits of natural gas to coal or oil-burning fuel systems which are used to generate the electricity for the electric cooling systems. Finally, the paper discusses the regulatory bias that the author feels exists towards the use of natural gas and the need for modification in the existing regulations to provide a 'level-playing field' for the gas cooling industry

  6. Scram device for gas-cooled reactor

    International Nuclear Information System (INIS)

    Murakami, Atsushi; Takahashi, Suehiro.

    1989-01-01

    A scram device for gas-cooled reactors has a hopper disposed below a stand pipe standing upright passing through a reactor container and electromagnets disposed therein. It further comprises neutron absorbing steel balls maintained between the electromagnets and the hopper upon energization of the electromagnets. Upon emergency reactor shutdown, energization for the electromagnets is interrupted to drop the neutron absorption stainless steel balls into the reactor core. It is an object of the present invention to keep the mechanical strength of the electromagnets in a high temperature gas atmosphere and not to reduce the insulation performance. That is, coils for the electromagnets are constituted with a small oxide-insulated metal sheath cable (MI cable). As the feature of the MI cable, it can maintain the mechanical strength even when exposed to high temperature gas coolant and the insulation performance thereof does not reduce by virture of its gas sealing property. Accordingly, a scram device of stable reliability can be obtained. (K.M.)

  7. Heating up the gas cooling market

    International Nuclear Information System (INIS)

    Watt, G.

    2001-01-01

    Gas cooling is an exciting technology with a potentially bright future. It comprises the production of cooling (and heating) in buildings and industry, by substituting environmentally-friendlier natural gas or LPG over predominantly coal-fired electricity in air conditioning equipment. There are currently four established technologies using gas to provide cooling energy or conditioned air. These are: absorption, both direct gas-fired and utilising hot water or steam; gas engine driven vapour compression (GED); cogeneration, with absorption cooling driven by recovered heat; and desiccant systems. The emergence of gas cooling technologies has been, and remains, one of evolution rather than revolution. However, further development of the technology has had a revolutionary effect on the performance, reliability and consumer acceptability of gas cooling products. Developments from world-renowned manufacturers such as York, Hitachi, Robur and Thermax have produced a range of absorption equipment variously offering: the use of 100 percent environmentally-friendly refrigerants, with zero global warming potential; the ideal utilisation of waste heat from cogeneration systems; a reduction in electrical distribution and stand-by generation capacity; long product life expectancy; far less noise and vibration; performance efficiency maintained down to about 20 percent of load capacity; and highly automated and low-cost maintenance. It is expected that hybrid systems, that is a mixture of gas and electric cooling technologies, will dominate the future market, reflecting the uncertainty in the electricity market and the prospects of stable future gas prices

  8. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    Markoczy, G.; Hudina, M.; Richmond, R.; Wydler, P.; Stratton, R.W.; Burgsmueller, P.

    1980-03-01

    The research and development work in the field of core thermal-hydraulics, steam generator research and development, experimental and analytical physics and carbide fuel development carried out 1979 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  9. 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

  10. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    International Nuclear Information System (INIS)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm 2 , 1000 0 C cladding temperature, and (2) 40 h at 40 W/cm 2 , 1200 0 C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370 0 C

  11. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

  12. Processing of coke oven gas. Primary gas cooling

    Energy Technology Data Exchange (ETDEWEB)

    Ullrich, H [Otto (C.) und Co. G.m.b.H., Bochum (Germany, F.R.)

    1976-11-01

    The primary cooler is an indispensable part of all byproduct processing plants. Its purpose is to cool the raw gas from the coke oven battery and to remove the accompanying water vapor. The greater part of the cooling capacity is utilized for the condensation of water vapor and only a small capacity is needed for the gas cooling. Impurities in the gas, like naphthalene, tar and solid particles, necessitate a special design in view of the inclination to dirt accumulation. Standard types of direct and indirect primary gas coolers are described, with a discussion of their advantages and disadvantages.

  13. Gas cooled fast reactor research in Europe

    International Nuclear Information System (INIS)

    Stainsby, Richard; Peers, Karen; Mitchell, Colin; Poette, Christian; Mikityuk, Konstantin; Somers, Joe

    2011-01-01

    Research on the gas-cooled fast reactor system is directed towards fulfilling the ambitious long term goals of Generation IV (Gen IV), i.e., to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. In common with other fast reactors, gas-cooled fast reactors (GFRs) have exceptional potential as sustainable energy sources, for both the utilisation of fissile material and minimisation of nuclear waste through transmutation of minor actinides. The primary goal of GFR research is to develop the system primarily to be a reliable and economic electricity generator, with good safety and sustainability characteristics. However, for the longer term, GFR retains the potential for hydrogen production and other process heat applications facilitated through a high core outlet temperature which, in this case, is not limited by the characteristics of the coolant. In this respect, GFR can inherit the non-electricity applications of the thermal HTRs in a sustainable manner in a future in which natural uranium becomes scarce. GFR research within Europe is performed directly by those states who have signed the 'System Arrangement' document within the Generation IV International Forum (the GIF), specifically France and Switzerland and Euratom. Importantly, Euratom provides a route by which researchers in other European states, and other non-European affiliates, can contribute to the work of the GIF, even when these states are not signatories to the GFR System Arrangement in their own right. This paper is written from the perspective of Euratom's involvement in research on the GFR system, starting with the 5th Framework Programme (FP5) GCFR project in 2000, through the FP6 project between 2005 and 2009 and looking ahead to the proposed activities within the current 7th Framework Programme (FP7). The evolution of the GFR concept from the 1960s onwards is discussed briefly, followed by the current perceived role, objectives and progress with

  14. A Massive, Cooling-Flow-Induced Starburst in the Core of a Highly Luminous Galaxy Cluster

    Science.gov (United States)

    McDonald, M.; Bayliss, M.; Benson, B. A.; Foley, R. J.; Ruel, J.; Sullivan, P.; Veilleux, S.; Aird, K. A.; Ashby, M. L. N.; Bautz, M.; hide

    2012-01-01

    In the cores of some galaxy clusters the hot intracluster plasma is dense enough that it should cool radiatively in the cluster s lifetime, leading to continuous "cooling flows" of gas sinking towards the cluster center, yet no such cooling flow has been observed. The low observed star formation rates and cool gas masses for these "cool core" clusters suggest that much of the cooling must be offset by astrophysical feedback to prevent the formation of a runaway cooling flow. Here we report X-ray, optical, and infrared observations of the galaxy cluster SPT-CLJ2344-4243 at z = 0.596. These observations reveal an exceptionally luminous (L(sub 2-10 keV) = 8.2 10(exp 45) erg/s) galaxy cluster which hosts an extremely strong cooling flow (M(sub cool) = 3820 +/- 530 Stellar Mass/yr). Further, the central galaxy in this cluster appears to be experiencing a massive starburst (740 +/- 160 Stellar Mass/ yr), which suggests that the feedback source responsible for preventing runaway cooling in nearby cool core clusters may not yet be fully established in SPT-CLJ2344-4243. This large star formation rate implies that a significant fraction of the stars in the central galaxy of this cluster may form via accretion of the intracluster medium, rather than the current picture of central galaxies assembling entirely via mergers.

  15. Effect of heat source shape on the thermal field in the pebble bed core of High Temperature Gas-cooled Reactor (HTGR)

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Leisheng; Lee, Jaeyoung [Handong Global University, Pohang (Korea, Republic of)

    2015-10-15

    In this study, in order to minimize the error brought by non-uniform heat flux, the spherical heaters are employed as heat source; subsequently, thermal field and heat transfer characteristics of the pebbles are investigated. The thermal field of the pebble surface in PBR is measured with heat source in different shapes. The HTGR design concept exhibits excellent safety features due to the low power density and the large amount of graphite present in the core which gives a large thermal inertia in an accident such as loss of coolant. However, the possible appearance of hot spots in the pebble bed cores of HTGR may affect the integrity of the pebbles, which has drawn the attention of many scientists to investigate the thermal field and to predict the maximum temperature locations in the pebbles using CFD method, Lee et.al has also done some experimental work on measuring the surface temperature of the pebbles as well as visualizing flow patterns of the coolant gas, and it was found that the temperature near the contacting points between pebbles was not higher than the flow stagnation points due to the higher thermal conductivity of the pebble. Certain error of temperature measurement might occur because of not very uniform heat flux in the pebbles since heater in cylindrical shape was utilized as heat source in previous experiment. More uniform heat flux and more complicated thermal profile are found in the result obtained using spherical heaters. The result shows that the temperature in contact point is higher than that in the top point, which is different from the previous results. The complex thermal phenomena observed in the lower-half side-sphere can be explained by the flow pattern near the surface.

  16. Hot gas path component cooling system

    Science.gov (United States)

    Lacy, Benjamin Paul; Bunker, Ronald Scott; Itzel, Gary Michael

    2014-02-18

    A cooling system for a hot gas path component is disclosed. The cooling system may include a component layer and a cover layer. The component layer may include a first inner surface and a second outer surface. The second outer surface may define a plurality of channels. The component layer may further define a plurality of passages extending generally between the first inner surface and the second outer surface. Each of the plurality of channels may be fluidly connected to at least one of the plurality of passages. The cover layer may be situated adjacent the second outer surface of the component layer. The plurality of passages may be configured to flow a cooling medium to the plurality of channels and provide impingement cooling to the cover layer. The plurality of channels may be configured to flow cooling medium therethrough, cooling the cover layer.

  17. Modelling of thermohydraulic emergency core cooling phenomena

    International Nuclear Information System (INIS)

    Yadigaroglu, G.; Andreani, M.; Lewis, M.J.

    1990-10-01

    The codes used in the early seventies for safety analysis and licensing were based either on the homogeneous model of two-phase flow or on the so-called separate-flow models, which are mixture models accounting, however, for the difference in average velocity between the two phases. In both cases the behavior of the mixture is prescribed a priori as a function of local parameters such as the mass flux and the quality. The modern best-estimate codes used for analyzing LWR LOCA's and transients are often based on a two-fluid or 6-equation formulation of the conservation equations. In this case the conservation equations are written separately for each phase; the mixture is allowed to evolve on its own, governed by the interfacial exchanges of mass, momentum and energy between the phases. It is generally agreed that such relatively sophisticated 6-equation formulations of two-phase flow are necessary for the correct modelling of a number of phenomena and situations arising in LWR accidental situations. They are in particular indispensible for the analysis of stratified or countercurrent flows and of situations in which large departures from thermal and velocity equilibrium exist. This report will be devoted to a discussion of the need for, the capacity and the limitations of the two-phase flow models (with emphasis on the 6-equation formulations) in modelling these two-phase flow and heat transfer phenomena and/or different core cooling situations. 18 figs., 1 tab., 72 refs

  18. Reference core design Mark-I and -II of the experimental, multi-purpose, high-temperature, gas-cooled reactor

    International Nuclear Information System (INIS)

    Shindo, Ryuiti; Hirano, Mitsumasa; Aruga, Takeo; Yasukawa, Sigeru

    1977-10-01

    Reactivity worth of the control rods and power distribution in the initial hot-clean core of reference core design Mark-I and -II have been studied. The need for burnable poison was confirmed, because of the limitations in number, diameter and reactivity worth of the control rods due to structures of pressure vessel and fuel element and to safety of the core. While the initial excess reactivity is reduced by use of the burnable poison, the recovery of core reactivity with burnup of the burnable poison requires a complicated withdrawal sequence of the control rods. The radial power gradient in the core is not large, due to orifice control of the coolant helium flow, effectiveness of the reflector in the small core and continuous distribution of burnup in the core by one-batch refuelling scheme. The local peaking factor in unit orifice regions, therefore, is the most important core design. Control of the axial power distribution is necessary to reduce the maximum fuel temperature and the exponential power distribution peaked toward the inlet of the core is most suitable. However, insertion of the control rods from top of the core disturbs the axial power distribution, so this effect must be considered in design of the withdrawal sequence of control rods. Nuclear properties of the core were revealed from results of the study for the initial hot-clean core. (auth.)

  19. Homogenization of some radiative heat transfer models: application to gas-cooled reactor cores; Homogeneisation de modeles de transferts thermiques et radiatifs: application au coeur des reacteurs a caloporteur gaz

    Energy Technology Data Exchange (ETDEWEB)

    El Ganaoui, K

    2006-09-15

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

  20. Thermodynamic properties of helium in the range from 20 to 15000C and 1 to 100 bar. Reactor core design of high-temperature gas-cooled reactors. Pt. 1

    International Nuclear Information System (INIS)

    Kipke, H.E.; Stoehr, A.; Banerjea, A.; Hammeke, K.; Huepping, N.

    1978-12-01

    The following report presents in tabular form the safety standard of the nuclear safety standard commission (KTA) on reactor core design of high-temperature gas-cooled reactors. Part 1: Calculation of thermodynamic properties of helium The basis of the present work is the data and formulae given by H. Petersen for the calculation of density, specific heat, thermal conductivity and dynamic viscosity of helium together with the formula for their standard deviations in the range of temperature and pressure stated above. The relations for specific enthalpy and specific entropy have been derived from density and specific heat, whereby specific heat is assumed constant over the given range of temperature and pressure. The latter section of this report contains tables of thermodynamic properties of helium calculated from the equations stated earlier in this paper. (orig.) [de

  1. STAR FORMATION EFFICIENCY IN THE COOL CORES OF GALAXY CLUSTERS

    International Nuclear Information System (INIS)

    McDonald, Michael; Veilleux, Sylvain; Mushotzky, Richard; Reynolds, Christopher; Rupke, David S. N.

    2011-01-01

    We have assembled a sample of high spatial resolution far-UV (Hubble Space Telescope Advanced Camera for Surveys/Solar Blind Channel) and Hα (Maryland-Magellan Tunable Filter) imaging for 15 cool core galaxy clusters. These data provide a detailed view of the thin, extended filaments in the cores of these clusters. Based on the ratio of the far-UV to Hα luminosity, the UV spectral energy distribution, and the far-UV and Hα morphology, we conclude that the warm, ionized gas in the cluster cores is photoionized by massive, young stars in all but a few (A1991, A2052, A2580) systems. We show that the extended filaments, when considered separately, appear to be star forming in the majority of cases, while the nuclei tend to have slightly lower far-UV luminosity for a given Hα luminosity, suggesting a harder ionization source or higher extinction. We observe a slight offset in the UV/Hα ratio from the expected value for continuous star formation which can be modeled by assuming intrinsic extinction by modest amounts of dust (E(B - V) ∼ 0.2) or a top-heavy initial mass function in the extended filaments. The measured star formation rates vary from ∼0.05 M sun yr -1 in the nuclei of non-cooling systems, consistent with passive, red ellipticals, to ∼5 M sun yr -1 in systems with complex, extended, optical filaments. Comparing the estimates of the star formation rate based on UV, Hα, and infrared luminosities to the spectroscopically determined X-ray cooling rate suggests a star formation efficiency of 14 +18 -8 %. This value represents the time-averaged fraction, by mass, of gas cooling out of the intracluster medium, which turns into stars and agrees well with the global fraction of baryons in stars required by simulations to reproduce the stellar mass function for galaxies. This result provides a new constraint on the efficiency of star formation in accreting systems.

  2. IAEA high temperature gas cooled reactor activities

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2001-01-01

    IAEA activities on high temperature gas cooled reactors are conducted with the review and support of Member States, primarily through the International Working Group on Gas Cooled Reactors (IWGGCR). This paper summarises the results of the IAEA gas cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products, and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (author)

  3. Ship exhaust gas plume cooling

    NARCIS (Netherlands)

    Schleijpen, H.M.A.; Neele, P.P.

    2004-01-01

    The exhaust gas plume is an important and sometimes dominating contributor to the infrared signature of ships. Suppression of the infrared ship signatures has been studied by TNO for the Royal Netherlands Navy over considerable time. This study deals with the suppression effects, which can be

  4. French activities on gas cooled reactors

    International Nuclear Information System (INIS)

    Bastien, D.

    1996-01-01

    The gas cooled reactor programme in France originally consisted of eight Natural Uranium Graphite Gas Cooled Reactors (UNGG). These eight units, which are now permanently shutdown, represented a combined net electrical power of 2,375 MW and a total operational history of 163 years. Studies related to these reactors concern monitoring and dismantling of decommissioned facilities, including the development of methods for dismantling. France has been monitoring the development of HTRs throughout the world since 1979, when it halted its own HTR R and D programme. France actively participates in three CRPs set up by the IAEA. (author). 1 tab

  5. Estimation of gas turbine blades cooling efficiency

    NARCIS (Netherlands)

    Moskalenko, A.B.; Kozhevnikov, A.

    2016-01-01

    This paper outlines the results of the evaluation of the most thermally stressed gas turbine elements, first stage power turbine blades, cooling efficiency. The calculations were implemented using a numerical simulation based on the Finite Element Method. The volume average temperature of the blade

  6. Advanced gas-cooled reactors (AGR)

    Energy Technology Data Exchange (ETDEWEB)

    Yeomans, R. M. [South of Scotland Electricity Board, Hunterston Power Station, West Kilbride, Ayshire, UK

    1981-01-15

    The paper describes the advanced gas-cooled reactor system, Hunterston ''B'' power station, which is a development of the earlier natural uranium Magnox type reactor. Data of construction, capital cost, operating performance, reactor safety and also the list of future developments are given.

  7. Gas-cooled reactor technology: a bibliography

    International Nuclear Information System (INIS)

    Raleigh, H.D.

    1981-09-01

    Included are 3358 citations on gas-cooled reactor technology contained in the DOE Energy Data Base for the period January 1978 through June 1981. The citations include reports, journal articles, books, conference papers, patents, and monographs. Corporate, Personal Author, Subject, Contract Number, and Report Number Indexes are provided

  8. Gas Cooled Fast Reactors: Recent advances and prospects

    International Nuclear Information System (INIS)

    Poette, C.; Guedeney, P.; Stainsby, R.; Mikityuk, K.; Knol, S.

    2013-01-01

    Gas Cooled Fast Reactors: Conclusion - GFR: an attractive longer term option allowing to combine Fast spectrum & Helium coolant benefits; • Innovative SiC fuel cladding solutions were found; • A first design confirming the encouraging potential of the reactor system Design improvements are nevertheless recommended and interesting tracks have been identified (core & system design, DHR system); • The GFR requires large R&D needs to confirm its potential (fuel & core materials, specific Helium technology); • ALLEGRO prototype studies are the first step and are drawing the R&D priorities

  9. Neutronic of heterogenous gas cooled reactors

    International Nuclear Information System (INIS)

    Maturana, Roberto Hernan

    2008-01-01

    At present, one of the main technical features of the advanced gas cooled reactor under development is its fuel element concept, which implies a neutronic homogeneous design, thus requiring higher enrichment compared with present commercial nuclear power plants.In this work a neutronic heterogeneous gas cooled reactor design is analyzed by studying the neutronic design of the Advanced Gas cooled Reactor (AGR), a low enrichment, gas cooled and graphite moderated nuclear power plant.A search of merit figures (some neutronic parameter, characteristic dimension, or a mixture of both) which are important and have been optimized during the reactor design stage is been done, to aim to comprise how a gas heterogeneous reactor is been design, given that semi-infinity arrangement criteria of rods in LWRs and clusters in HWRs can t be applied for a solid moderator and a gas refrigerator.The WIMS code for neutronic cell calculations is been utilized to model the AGR fuel cell and to calculate neutronic parameters such as the multiplication factor and the pick factor, as function of the fuel burnup.Also calculation is been done for various nucleus characteristic dimensions values (fuel pin radius, fuel channel pitch) and neutronic parameters (such as fuel enrichment), around the design established parameters values.A fuel cycle cost analysis is carried out according to the reactor in study, and the enrichment effect over it is been studied.Finally, a thermal stability analysis is been done, in subcritical condition and at power level, to study this reactor characteristic reactivity coefficients.Present results shows (considering the approximation used) a first set of neutronic design figures of merit consistent with the AGR design. [es

  10. 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)

  11. 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.

  12. Cooling process in separation devices of krypton gas

    Energy Technology Data Exchange (ETDEWEB)

    Kimura, S; Sugimoto, K

    1975-06-11

    To prevent entry of impurities into purified gases and to detect leaks of heat exchanger in a separation and recovering device of krypton gas by means of liquefaction and distillation, an intermediate refrigerant having the same or slightly higher boiling point than that of gas to be cooled is used between the gas to be cooled (process gas) and refrigerant (nitrogen), and the pressure of the gas to be cooled is controlled to have a pressure higher than the intermediate refrigerant to cool the gas to be cooled.

  13. The modular high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Lutz, D.E.; Lipps, A.J.

    1984-01-01

    Due to relatively high operating temperatures, the gas-cooled reactor has the potential to serve a wide variety of energy applications. This paper discusses the energy applications which can be served by the modular HTGR, the magnitude of the potential markets, and the HTGR product cost incentives relative to fossil fuel competition. Advantages of the HTGR modular systems are presented along with a description of the design features and performance characteristics of the current reference HTGR modular systems

  14. Gas-cooled breeder reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Chermanne, J.; Burgsmueller, P. [Societe Belge pour l' Industrie Nucleaire, Brussels

    1981-01-15

    The European Association for the Gas-cooled Breeder Reactor (G B R A), set-up in 1969 prepared between 1972 and 1974 a 1200 MWe Gas-cooled Breeder Reactor (G B R) commercial reference design G B R 4. It was then found necessary that a sound and neutral appraisal of the G B R licenseability be carried out. The Commission of the European Communities (C E C) accepted to sponsor this exercise. At the beginning of 1974, the C E C convened a group of experts to examine on a Community level, the safety documents prepared by the G B R A. A working party was set-up for that purpose. The experts examined a ''Preliminary Safety Working Document'' on which written questions and comments were presented. A ''Supplement'' containing the answers to all the questions plus a detailed fault tree and reliability analysis was then prepared. After a final study of this document and a last series of discussions with G B R A representatives, the experts concluded that on the basis of the evidence presented to the Working Party, no fundamental reasons were identified which would prevent a Gas-cooled Breeder Reactor of the kind proposed by the G B R A achieving a satisfactory safety status. Further work carried out on ultimate accident have confirmed this conclusion. One can therefore claim that the overall safety risk associated with G B R s compares favourably with that of any other reactor system.

  15. Development of local heat transfer and pressure drop models for pebble bed high temperature gas-cooled reactor cores - HTR2008-58296

    International Nuclear Information System (INIS)

    McLaughlin, B.; Worsley, M.; Stainsby, R.; Grief, A.; Dennier, A.; Macintosh, S.; Van Heerden, E.

    2008-01-01

    This paper describes pressure drop and heat transfer coefficient predictions for a typical coolant flow within the core of a pebble bed reactor (PBR) by examining a representative group of pebbles remote from the reflector region. The three- dimensional steady state flow and heat transfer predictions utilized in this work are obtained from a computational fluid dynamics (CFD) model created in the commercial software ANSYS FLUENT TM . This work utilizes three RANS turbulence models and the Chilton-Colburn analogy for heat transfer. A methodology is included in this paper for creating a quality unstructured mesh with prismatic surface layers on a random arrangement of touching pebbles. The results of the model are validated by comparing them with the correlations of the German KTA rules for a PBR. (authors)

  16. Development of the prediction technology of cable disconnection of in-core neutron detector for the future high-temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Shimazaki, Yosuke; Sawahata, Hiroaki; Kawamoto, Taiki; Suzuki, Hisashi; Shinohara, Masanori; Honda, Yuki; Katsuyama, Kozo; Takada, Shoji; Sawa, Kazuhiro

    2015-01-01

    Maintenance technologies for the reactor system have been developed by using the high-temperature engineering test reactor (HTTR). One of the important purposes of development is to accumulate the experiences and data to satisfy the availability of operation up to 90% by shortening the duration of the periodical maintenance for the future HTGRs by shifting from the time-based maintenance to condition-based maintenance. The technical issue of the maintenance of in-core neutron detector, wide range monitor (WRM), is to predict the malfunction caused by cable disconnection to plan the replacement schedule. This is because that it is difficult to observe directly inside of the WRM in detail. The electrical inspection method was proposed to detect and predict the cable disconnection of the WRM by remote monitoring from outside of the reactor by using the time domain reflectometry and so on. The disconnection position, which was specified by the electrical method, was identified by non-destructive and destructive inspection. The accumulated data is expected to be contributed for advanced maintenance of future HTGRs. (author)

  17. 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

  18. Gas-cooled fast reactor safety

    International Nuclear Information System (INIS)

    Rickard, C.L.; Simon, R.H.; Buttemer, D.R.

    1977-01-01

    Initial conceptual design work on the GCFR began in the USA in the early 1960s and since the later 1960s has proceeded with considerable international cooperation. A 300 MWe GCFR demonstration plant employing three main cooling loops is currently being developed at General Atomic. A major preapplication licensing review of this demonstration plant was initiated in 1971 leading in 1974 to publication of a Safety Evaluation Report by the USAEC Directorate of Licensing. The preapplication review is continuing by addressing areas of concern identified in this report such that a major part of the work necessary to support the actual licensing of a GCFR demonstration plant has been established. The safety performance of the GCFR demonstration plant is based upon its inherent safety characteristics among which are the single phase and chemically inert coolant which is not activated and has a low reactivity worth, the negative core power and temperature reactivity coefficients and the small and negative steam reactivity worth. Recent studies of larger core designs indicate that as the reactor size increases central fuel, clad and coolant reactivity worths decrease and the Doppler coefficient becomes more negative. These inherent safety characteristics are complemented by safety design features such as enclosing the entire primary coolant system within a prestressed concrete pressure vessel (PCRV), providing two independent and diverse shutdown systems and residual heat removal (RHR) systems, limiting the worth of control rods to less than $1, employing pressure-equalized fuel rods, a core supported rigidly at its upper end and otherwise unrestrained and coolant downflow within the core to enhance debris removal should local melting occur. The structurally redundant PCRV design allows the potential depressurization leak area to be controlled and, since the PCRV is located within a containment building, coolant is present even after a depressurization accident and each RHR

  19. Study on plant concept for gas cooled fast reactor

    International Nuclear Information System (INIS)

    Moribe, Takeshi; Kubo, Shigenobu; Saigusa, Toshiie; Konomura, Mamoru

    2003-05-01

    In 'Feasibility Study on Commercialized Fast Reactor Cycle System', technological options including various coolant (sodium, heavy metal, gas, water, etc.), fuel type (MOX, metal, nitride) and output power are considered and classified, and commercialized FBR that have economical cost equal to LWR are pursued. In conceptual study on gas cooled FBR in FY 2002, to identify the prospect of the technical materialization of the helium cooled FBR using coated particle fuel which is an attractive concept extracted in the year of FY2001, the preliminary conceptual design of the core and entire plant was performed. This report summarizes the results of the plant design study in FY2002. The results of study is as follows. 1) For the passive core shutdown equipment, the curie point magnet type self-actuated device was selected and the device concept was set up. 2) For the reactor block, the concept of the core supporting structure, insulators and liners was set up. For the material of the heat resistant structure, SiC was selected as a candidate. 3) For the seismic design of the plant, it was identified that a design concept with three-dimensional base isolation could be feasible taking the severe seismic condition into account. 4) For the core catcher, an estimation of possible event sequences under severe core damage condition was made. A core catcher concept which may suit the estimation was proposed. 5) The construction cost was roughly estimated based on the amount of materials and its dependency on the plant output power was evaluated. The value for a small sized plant exceeds the target construction cost about 20%. (author)

  20. Behaviour of gas cooled reactor fuel under accident conditions

    International Nuclear Information System (INIS)

    1991-11-01

    The Specialists Meeting on Behaviour of Gas Cooled Reactor Fuel under Accident Conditions was convened by the International Atomic Energy Agency on the recommendation of the International Working Group on Gas Cooled Reactors. The purpose of the meeting was to provide an international forum for the review of the development status and for the discussion on the behaviour of gas cooled reactor fuel under accident conditions and to identify areas in which additional research and development are still needed and where international co-operation would be beneficial for all involved parties. The meeting was attended by 45 participants from France, Germany, Japan, Switzerland, the Union of Soviet Socialists Republics, the United Kingdom, the United States of America, CEC and the IAEA. The meeting was subdivided into five technical sessions: Summary of Current Research and Development Programmes for Fuel; Fuel Manufacture and Quality Control; Safety Requirements; Modelling of Fission Product Release - Part I and Part II; Irradiation Testing/Operational Experience with Fuel Elements; Behaviour at Depressurization, Core Heat-up, Power Transients; Water/Steam Ingress - Part I and Part II. 22 papers were presented. A separate abstract was prepared for each of these papers. At the end of the meeting a round table discussion was held on Directions for Future R and D Work and International Co-operation. Refs, figs and tabs

  1. Gas-cooled reactor power systems for space

    International Nuclear Information System (INIS)

    Walter, C.E.

    1987-01-01

    Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system

  2. Design codes for gas cooled reactor components

    International Nuclear Information System (INIS)

    1990-12-01

    High-temperature gas-cooled reactor (HTGR) plants have been under development for about 30 years and experimental and prototype plants have been operated. The main line of development has been electricity generation based on the steam cycle. In addition the potential for high primary coolant temperature has resulted in research and development programmes for advanced applications including the direct cycle gas turbine and process heat applications. In order to compare results of the design techniques of various countries for high temperature reactor components, the IAEA established a Co-ordinated Research Programme (CRP) on Design Codes for Gas-Cooled Reactor Components. The Federal Republic of Germany, Japan, Switzerland and the USSR participated in this Co-ordinated Research Programme. Within the frame of this CRP a benchmark problem was established for the design of the hot steam header of the steam generator of an HTGR for electricity generation. This report presents the results of that effort. The publication also contains 5 reports presented by the participants. A separate abstract was prepared for each of these reports. Refs, figs and tabs

  3. British Columbia natural gas: Core market policy

    International Nuclear Information System (INIS)

    1988-06-01

    The core market for natural gas in British Columbia is defined as all natural gas consumers in the residential, institutional, commercial, and industrial sectors not currently purchasing natural gas directly and not exempted from the core market by the British Columbia Utilities Commission (BCUC). The intent of the definition is to include all customers who must be protected by contracts which ensure long-term security of supply and stable prices. Core market customers are excluded from direct natural gas purchase and will be served by distribution utilities. A customer may apply to BCUC to leave the core market; such an application may be approved if it is demonstrated that the customer has adequate long-term natural gas supplies or alternative fuel supplies to protect him from supply interruptions. The non-core market is defined as all large industrial customers who elect to make their own natural gas supply arrangements and who can demonstrate to the BCUC sufficient long-term natural gas supply protection or alternative fuel capability to ensure security of the industry. Non-core market customers have full and open access to the competitive natural gas market. The British Columbia government will not apply its core market policy to other jurisdictions through Energy Removal Certificates

  4. CEA programme on gas cooled reactors

    International Nuclear Information System (INIS)

    Carre, F.; Fiorini, G.L.; Chapelot, Ph.; Gauthier, J.C.

    2002-01-01

    Future nuclear energy systems studies conducted by the CEA aim at investigating and developing promising technologies for future reactors, fuels and fuel cycles, for nuclear power to play a major part in sustainable energy policies. Reactors and fuel cycles are considered as integral parts of a nuclear system to be optimised as a whole. Major goals assigned to future nuclear energy systems are the following: reinforced economic competitiveness with other electricity generation means, with a special emphasis on reducing the investment cost; enhanced reliability and safety, through an improved management of reactor operation in normal and abnormal plant conditions; minimum production of long lived radioactive waste; resource saving through an effective and flexible use of the available resources of fissile and fertile materials; enhanced resistance to proliferation risks. The three latter goals are essential for the sustainability of nuclear energy in the long term. Additional considerations such as the potentialities for other applications than electricity generation (co-generation, production of hydrogen, sea water desalination) take on an increasing importance. Sustainability goals call for fast neutron spectra (to transmute nuclear waste and to breed fertile fuel) and for recycling actinides from the spent fuel (plutonium and minor actinides). New applications and economic competitiveness call for high temperature technologies (850 deg C), that afford high conversion efficiencies and hence less radioactive waste production and discharged heat. These orientations call for breakthroughs beyond light water reactors. Therefore, as a result of a screening review of candidate technologies, the CEA has selected an innovative concept of high temperature gas cooled reactor with a fast neutron spectrum, robust refractory fuel, direct conversion with a gas turbine, and integrated on-site fuel cycle as a promising system for a sustainable energy development. This objective

  5. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki

    1982-07-01

    Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

  6. Gravity driven emergency core cooling experiments with the PACTEL facility

    International Nuclear Information System (INIS)

    Munther, R.; Kalli, H.; Kouhia, J.

    1996-01-01

    PACTEL (Parallel Channel Test Loop) is an experimental out-of-pile facility designed to simulated the major components and system behaviour of a commercial Pressurized Water Reactor (PWR) during different postulated LOCAs and transients. The reference reactor to the PACTEL facility is Loviisa type WWER-440. The recently made modifications enable experiments to be conducted also on the passive core cooling. In these experiments the passive core cooling system consisted of one core makeup tank (CMT) and pressure balancing lines from the pressurizer and from a cold leg connected to the top of the CMT in order to maintain the tank in pressure equilibrium with the primary system during ECC injection. The line from the pressurizer to the core makeup tank was normally open. The ECC flow was provided from the CMT located at a higher elevation than the main part of the primary system. A total number of nine experiments have been performed by now. 4 refs, 7 figs, 3 tabs

  7. Gravity driven emergency core cooling experiments with the PACTEL facility

    Energy Technology Data Exchange (ETDEWEB)

    Munther, R; Kalli, H [University of Technology, Lappeenranta (Finland); Kouhia, J [Technical Research Centre of Finland, Lappeenranta (Finland)

    1996-12-01

    PACTEL (Parallel Channel Test Loop) is an experimental out-of-pile facility designed to simulated the major components and system behaviour of a commercial Pressurized Water Reactor (PWR) during different postulated LOCAs and transients. The reference reactor to the PACTEL facility is Loviisa type WWER-440. The recently made modifications enable experiments to be conducted also on the passive core cooling. In these experiments the passive core cooling system consisted of one core makeup tank (CMT) and pressure balancing lines from the pressurizer and from a cold leg connected to the top of the CMT in order to maintain the tank in pressure equilibrium with the primary system during ECC injection. The line from the pressurizer to the core makeup tank was normally open. The ECC flow was provided from the CMT located at a higher elevation than the main part of the primary system. A total number of nine experiments have been performed by now. 4 refs, 7 figs, 3 tabs.

  8. Medium-size high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Peinado, C.O.; Koutz, S.L.

    1980-08-01

    This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760 0 C (1400 0 F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics [a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant] and engineered safety features

  9. Isolated core vs. superficial cooling effects on virtual maze navigation.

    Science.gov (United States)

    Payne, Jennifer; Cheung, Stephen S

    2007-07-01

    Cold impairs cognitive performance and is a common occurrence in many survival situations. Altered behavior patterns due to impaired navigation abilities in cold environments are potential problems in lost-person situations. We investigated the separate effects of low core temperature and superficial cooling on a spatially demanding virtual navigation task. There were 12 healthy men who were passively cooled via 15 degrees C water immersion to a core temperature of 36.0 degrees C, then transferred to a warm (40 degrees C) water bath to eliminate superficial shivering while completing a series of 20 virtual computer mazes. In a control condition, subjects rested in a thermoneutral (approximately 35 degrees C) bath for a time-matched period before being transferred to a warm bath for testing. Superficial cooling and distraction were achieved by whole-body immersion in 35 degree water for a time-matched period, followed by lower leg immersion in 10 degree C water for the duration of the navigational tests. Mean completion time and mean error scores for the mazes were not significantly different (p > 0.05) across the core cooling (16.59 +/- 11.54 s, 0.91 +/- 1.86 errors), control (15.40 +/- 8.85 s, 0.82 +/- 1.76 errors), and superficial cooling (15.19 +/- 7.80 s, 0.77 +/- 1.40 errors) conditions. Separately reducing core temperature or increasing cold sensation in the lower extremities did not influence performance on virtual computer mazes, suggesting that navigation is more resistive to cooling than other, simpler cognitive tasks. Further research is warranted to explore navigational ability at progressively lower core and skin temperatures, and in different populations.

  10. Emergency core cooling systems in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1981-12-01

    This report contains the responses by the Advisory Committee on Nuclear Safety to three questions posed by the Atomic Energy Control Board concerning the need for Emergency Core Cooling Systems (ECCS) in CANDU nuclear power plants, the effectiveness requirement for such systems, and the extent to which experimental evidence should be available to demonstrate compliance with effectiveness standards

  11. Design of a PWR emergency core cooling simulator loop

    International Nuclear Information System (INIS)

    Melo, C.A. de.

    1982-12-01

    The preliminary design of a PWR Emergency Core Cooling Simulator Loop for investigations of the phenomena involved in a postulated Loss-of-Coolant Accident, during the Reflooding Phase, is presented. The functions of each component of the loop, the design methods and calculations, the specification of the instrumentation, the system operation sequence, the materials list and a cost assessment are included. (Author) [pt

  12. Fuel arrangement for high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Tobin, J.M.

    1978-01-01

    Disclosed is a fuel arrangement for a high temperature gas cooled reactor including fuel assemblies with separate directly cooled fissile and fertile fuel elements removably inserted in an elongated moderator block also having a passageway for control elements

  13. Overview of gas cooled reactors' applications with CATHARE

    International Nuclear Information System (INIS)

    Genevieve Geffraye; Fabrice Bentivoglio; Anne Messie; Alain Ruby; Manuel Saez; Nicolas Tauveron; Ola Widlund

    2005-01-01

    Full text of publication follows: For about four years, CEA has launched feasibility studies of future nuclear advanced systems in a consistent series of Gas Cooled Reactors (GCR) ranging from thermal reactors, as the Very High Temperature Reactor (VHTR) for the mid term, to fast reactors (GFR) for the long term. Thermal hydraulic performances are a key issue for the core design, the evaluation of the thermal stresses on the structures and the decay heat removal systems. This analysis requires a 1D code able to simulate the whole reactor, including the core, the vessel, the piping and the components (turbine, compressors, heat exchangers). CATHARE is the reference code developed and extensively validated in collaboration between CEA, EDF, IRSN and FRAMATOME-ANP for the French Pressurized Water Reactors. CATHARE has the capabilities to model a Gas Cooled Reactor using standard 0D and 1D modules with some adaptations to treat the specificities of the GCR designs. In this paper, the different adaptations are presented and discussed. The direct coupling of a Gas Cooled Reactor with a closed gas-turbine cycle leads to a specific dynamic plant behaviour and a specific turbomachinery module has been developed. The thermal reactors' core consists of hexagonal graphite blocks with an annular-fueled region surrounded by reflectors and a special attention is paid on the thermal modeling of such a core leading to a quasi-2D thermal description. First designs of the VHTR are proposed and are based on an indirect cycle concept with a primary circuit, cooled by helium, and containing the core and a circulator. The core power is transmitted to the secondary circuit via an intermediate heat exchanger (IHX). The secondary circuit contains a turbine and a compressor coupled on a single shaft. It uses a mixture of helium and nitrogen, in order to benefit from both the favourable thermal properties of helium for the heat exchanger, and from existing experience of turbomachines using

  14. Evaluation of advanced cooling therapy's esophageal cooling device for core temperature control.

    Science.gov (United States)

    Naiman, Melissa; Shanley, Patrick; Garrett, Frank; Kulstad, Erik

    2016-05-01

    Managing core temperature is critical to patient outcomes in a wide range of clinical scenarios. Previous devices designed to perform temperature management required a trade-off between invasiveness and temperature modulation efficiency. The Esophageal Cooling Device, made by Advanced Cooling Therapy (Chicago, IL), was developed to optimize warming and cooling efficiency through an easy and low risk procedure that leverages heat transfer through convection and conduction. Clinical data from cardiac arrest, fever, and critical burn patients indicate that the Esophageal Cooling Device performs very well both in terms of temperature modulation (cooling rates of approximately 1.3°C/hour, warming of up to 0.5°C/hour) and maintaining temperature stability (variation around goal temperature ± 0.3°C). Physicians have reported that device performance is comparable to the performance of intravascular temperature management techniques and superior to the performance of surface devices, while avoiding the downsides associated with both.

  15. Method of injecting cooling water in emergency core cooling system (ECCS) of PWR type reactor

    International Nuclear Information System (INIS)

    Sobajima, Makoto; Adachi, Michihiro; Tasaka, Kanji; Suzuki, Mitsuhiro.

    1979-01-01

    Purpose: To provide a cooling water injection method in an ECCS, which can perform effective cooling of the reactor core. Method: In a method of injecting cooling water in an ECCS as a countermeasure against a rupture accident of a pwr type reactor, cooling water in the first pressure storage injection system is injected into the upper plenum of the reactor pressure vessel at a set pressure of from 50 to 90 atg. and a set temperature of from 80 to 200 0 C, cooling water in the second pressure storage injection system is injected into the lower plenum of the reactor pressure vessel at a pressure of from 25 to 60 atg. which is lower than the set pressure and a temperature less than 60 0 C, and further in combination with these procedures, cooling water of less than 60 0 C is injected into a high-temperature side piping, in the high-pressure injection system of upstroke of 100 atg. by means of a pump and the low-pressure injection system of upstroke of 20 atg. also by means of a pump, thereby cooling the reactor core. (Aizawa, K.)

  16. 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.)

  17. Status of national gas cooled reactor programmes

    International Nuclear Information System (INIS)

    1991-08-01

    This report has been compiled as a central source of summary-level information on the present status of High Temperature Gas-Cooled Reactor (HTGR) programmes in the world and on future plans for the continued development and deployment of HTGRs. Most of the information concerns the programmes in the United States, Germany, Japan and the Soviet Union, countries that have had large programmes related to HTGR technology for several years. Summary-level information is also provided in the report on HTGR-related activities in several other countries who either have an increasing interest in the technology and/or who are performing some development efforts related to HTGR technology. The report contains a summary-level update on the MAGNOX and AGR programmes. This is the twelfth issue of the document, the first of which was issued in March, 1979. The report has been prepared in the IAEA Nuclear Power Technology Development Section. Figs and tabs

  18. Heavy water moderated gas-cooled reactors

    International Nuclear Information System (INIS)

    Bailly du Bois, B.; Bernard, J.L.; Naudet, R.; Roche, R.

    1964-01-01

    France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [fr

  19. Seismic behaviour of gas cooled reactor components

    International Nuclear Information System (INIS)

    1990-08-01

    On invitation of the French Government the Specialists' Meeting on the Seismic Behaviour of Gas-Cooled Reactor Components was held at Gif-sur-Yvette, 14-16 November 1989. This was the second Specialists' Meeting on the general subject of gas-cooled reactor seismic design. There were 27 participants from France, the Federal Republic of Germany, Israel, Japan, Spain, Switzerland, the United Kingdom, the Soviet Union, the United States, the CEC and IAEA took the opportunity to present and discuss a total of 16 papers reflecting the state of the art of gained experiences in the field of their seismic qualification approach, seismic analysis methods and of the capabilities of various facilities used to qualify components and verify analytical methods. Since the first meeting, the sophistication and expanded capabilities of both the seismic analytical methods and the test facilities are apparent. The two main methods for seismic analysis, the impedance method and the finite element method, have been computer-programmed in several countries with the capability of each of the codes dependent on the computer capability. The correlations between calculation and tests are dependent on input assumptions such as boundary conditions, soil parameters and various interactions between the soil, the buildings and the contained equipment. The ability to adjust these parameters and match experimental results with calculations was displayed in several of the papers. The expanded capability of some of the new test facilities was graphically displayed by the description of the SAMSON vibration test facility at Juelich, FRG, capable of dynamically testing specimens weighing up to 25 tonnes, and the TAMARIS facility at the CEA laboratories in Gif-sur-Yvette where the largest table is capable of testing specimens weighing up to 100 tonnes. The proceedings of this meeting contain all 16 presented papers. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  20. 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)

  1. Thermal hydraulic analysis of gas-cooled reactors with annular fuel rods

    International Nuclear Information System (INIS)

    Han, Kyu Hyun; Chang, Soon Heung

    2005-01-01

    More than half of the world's energy is used in industrial processes and for heating applications which have hardly been touched by the nuclear industry. Nuclear power could be brought into a wide range of applications for industrial processes, provided that gas outlet temperatures of gascooled reactors are sufficiently high. The most limiting core design requirement which controls the core outlet temperature is the maximum acceptable fuel compact temperature. An innovative fuel design is required for a significant decrease in the fuel temperature. This study investigated the possibilities of implementing internally and externally cooled annular fuel rods in a gas-cooled reactor

  2. Status of and prospects for gas-cooled reactors

    International Nuclear Information System (INIS)

    1984-01-01

    The IAEA International Working Group on Gas-Cooled Reactors (IWGGCR) (see Annex I), which was established in 1978, recommended to the Agency that a report be prepared in order to provide an up-to-date summary of gas-cooled reactor technology. The present Technical Report is based mainly on submissions of Member Countries of the IWGGCR and consists of four main sections. Beside some general information about the gas-cooled reactor line, section 1 contains a description of the incentives for the development and deployment of gas-cooled reactors in various Agency Member States. These include both electricity generation and process steam and process heat production for various branches of industry. The historical development of gas-cooled reactors is reviewed in section 2. In this section information is provided on how, when and why gas-cooled reactors have been developed in various Agency Member States and, in addition, a detailed description of the different gas-cooled reactor lines is presented. Section 3 contains information about the technical status of gas-cooled reactors and their applications. Gas-cooled reactors that are under design or construction or in operation are listed and shortly described, together with an outlook for future reactor designs. In this section the various applications for gas-cooled reactors are described in detail. These include both electricity generation and process steam and process heat production. The last section (section 4) is entitled ''Special features of gas-cooled reactors'' and contains information about the technical performance, fuel utilization, safety characteristics and environmental impact, such as radiation exposure and heat rejection

  3. Gas-cooled reactor for space power systems

    International Nuclear Information System (INIS)

    Walter, C.E.; Pearson, J.S.

    1987-05-01

    Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors

  4. 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

  5. Evaluating the income and employment impacts of gas cooling technologies

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, P.J. [Oak Ridge National Lab., TN (United States); Laitner, S.

    1995-03-01

    The purpose of this study is to estimate the potential employment and income benefits of the emerging market for gas cooling products. The emphasis here is on exports because that is the major opportunity for the U.S. heating, ventilating, and air-conditioning (HVAC) industry. But domestic markets are also important and considered here because without a significant domestic market, it is unlikely that the plant investments, jobs, and income associated with gas cooling exports would be retained within the United States. The prospects for significant gas cooling exports appear promising for a variety of reasons. There is an expanding need for cooling in the developing world, natural gas is widely available, electric infrastructures are over-stressed in many areas, and the cost of building new gas infrastructure is modest compared to the cost of new electric infrastructure. Global gas cooling competition is currently limited, with Japanese and U.S. companies, and their foreign business partners, the only product sources. U.S. manufacturers of HVAC products are well positioned to compete globally, and are already one of the faster growing goods-exporting sectors of the U.S. economy. Net HVAC exports grew by over 800 percent from 1987 to 1992 and currently exceed $2.6 billion annually (ARI 1994). Net gas cooling job and income creation are estimated using an economic input-output model to compare a reference case to a gas cooling scenario. The reference case reflects current policies, practices, and trends with respect to conventional electric cooling technologies. The gas cooling scenario examines the impact of accelerated use of natural gas cooling technologies here and abroad.

  6. 78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

    Science.gov (United States)

    2013-10-25

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

  7. Gas Mixtures for Welding with Micro-Jet Cooling

    Directory of Open Access Journals (Sweden)

    Węgrzyn T.

    2015-04-01

    Full Text Available Welding with micro-jet cooling after was tested only for MIG and MAG processes. For micro-jet gases was tested only argon, helium and nitrogen. A paper presents a piece of information about gas mixtures for micro-jet cooling after in welding. There are put down information about gas mixtures that could be chosen both for MAG welding and for micro-jet process. There were given main information about influence of various micro-jet gas mixtures on metallographic structure of steel welds. Mechanical properties of weld was presented in terms of various gas mixtures selection for micro-jet cooling.

  8. THE RELATION BETWEEN COOL CLUSTER CORES AND HERSCHEL-DETECTED STAR FORMATION IN BRIGHTEST CLUSTER GALAXIES

    Energy Technology Data Exchange (ETDEWEB)

    Rawle, T. D.; Egami, E.; Rex, M.; Fiedler, A.; Haines, C. P.; Pereira, M. J.; Portouw, J.; Walth, G. [Steward Observatory, University of Arizona, 933 N. Cherry Ave., Tucson, AZ 85721 (United States); Edge, A. C. [Institute for Computational Cosmology, Durham University, South Road, Durham DH1 3LE (United Kingdom); Smith, G. P. [School of Physics and Astronomy, University of Birmingham, Edgbaston, Birmingham B15 2TT (United Kingdom); Altieri, B.; Valtchanov, I. [Herschel Science Centre, ESAC, ESA, P.O. Box 78, Villanueva de la Canada, 28691 Madrid (Spain); Perez-Gonzalez, P. G. [Departamento de Astrofisica, Facultad de CC. Fisicas, Universidad Complutense de Madrid, E-28040 Madrid (Spain); Van der Werf, P. P. [Sterrewacht Leiden, Leiden University, P.O. Box 9513, 2300 RA, Leiden (Netherlands); Zemcov, M., E-mail: trawle@as.arizona.edu [Department of Physics, Mathematics and Astronomy, California Institute of Technology, Pasadena, CA 91125 (United States)

    2012-03-01

    We present far-infrared (FIR) analysis of 68 brightest cluster galaxies (BCGs) at 0.08 < z < 1.0. Deriving total infrared luminosities directly from Spitzer and Herschel photometry spanning the peak of the dust component (24-500 {mu}m), we calculate the obscured star formation rate (SFR). 22{sup +6.2}{sub -5.3}% of the BCGs are detected in the far-infrared, with SFR = 1-150 M{sub Sun} yr{sup -1}. The infrared luminosity is highly correlated with cluster X-ray gas cooling times for cool-core clusters (gas cooling time <1 Gyr), strongly suggesting that the star formation in these BCGs is influenced by the cluster-scale cooling process. The occurrence of the molecular gas tracing H{alpha} emission is also correlated with obscured star formation. For all but the most luminous BCGs (L{sub TIR} > 2 Multiplication-Sign 10{sup 11} L{sub Sun }), only a small ({approx}<0.4 mag) reddening correction is required for SFR(H{alpha}) to agree with SFR{sub FIR}. The relatively low H{alpha} extinction (dust obscuration), compared to values reported for the general star-forming population, lends further weight to an alternate (external) origin for the cold gas. Finally, we use a stacking analysis of non-cool-core clusters to show that the majority of the fuel for star formation in the FIR-bright BCGs is unlikely to originate from normal stellar mass loss.

  9. Development of high temperature gas cooled reactor in China

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Wentao [Paul Scherrer Institute, Villigen (Switzerland). Dept. of Nuclear Energy and Safety; Schorer, Michael [Swiss Nuclear Forum, Olten (Switzerland)

    2018-02-15

    High temperature gas cooled reactor (HTGR) is one of the six Generation IV reactor types put forward by Generation IV International Forum (GIF) in 2002. This type of reactor has high outlet temperature. It uses Helium as coolant and graphite as moderator. Pebble fuel and ceramic reactor core are adopted. Inherit safety, good economy, high generating efficiency are the advantages of HTGR. According to the comprehensive evaluation from the international nuclear community, HTGR has already been given the priority to the research and development for commercial use. A demonstration project of the High Temperature Reactor-Pebble-�bed Modules (HTR-PM) in Shidao Bay nuclear power plant in China is under construction. In this paper, the development history of HTGR in China and the current situation of HTR-PM will be introduced. The experiences from China may be taken as a reference by the international nuclear community.

  10. Windscale advanced gas-cooled reactor (WAGR) decommissioning project overview

    International Nuclear Information System (INIS)

    Pattinson, A.

    2003-01-01

    The current BNFL reactor decommissioning projects are presented. The projects concern power reactor sites at Berkely, Trawsfynydd, Hunterstone, Bradwell, Hinkley Point; UKAEA Windscale Pile 1; Research reactors within UK Scottish Universities at East Kilbride and ICI (both complete); WAGR. The BNFL environmental role include contract management; effective dismantling strategy development; implementation and operation; sentencing, encapsulation and transportation of waste. In addition for the own sites it includes strategy development; baseline decommissioning planning; site management and regulator interface. The project objectives for the Windscale Advanced Gas-Cooled Reactor (WAGR) are 1) Safe and efficient decommissioning; 2) Building of good relationships with customer; 3) Completion of reactor decommissioning in 2005. The completed WAGR decommissioning campaigns are: Operational Waste; Hot Box; Loop Tubes; Neutron Shield; Graphite Core and Restrain System; Thermal Shield. The current campaign is Lower Structures and the remaining are: Pressure vessel and Insulation; Thermal Columns and Outer Vault Membrane. An overview of each campaign is presented

  11. 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

  12. Advanced gas cooled reactors - Designing for safety

    International Nuclear Information System (INIS)

    Keen, Barry A.

    1990-01-01

    The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme

  13. Advanced gas cooled reactors - Designing for safety

    Energy Technology Data Exchange (ETDEWEB)

    Keen, Barry A [Engineering Development Unit, NNC Limited, Booths Hall, Knutsford, Cheshire (United Kingdom)

    1990-07-01

    The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme.

  14. Core melt retention and cooling concept of the ERP

    Energy Technology Data Exchange (ETDEWEB)

    Weisshaeupl, H [SIEMENS/KWU, Erlangen (Germany); Yvon, M [Nuclear Power International, Paris (France)

    1996-12-01

    For the French/German European Pressurized Water Reactor (EPR) mitigative measures to cope with the event of a severe accident with core melt down are considered already at the design stage. Following the course of a postulated severe accident with reactor pressure vessel melt through one of the most important features of a future design must be to stabilize and cool the melt within the containment by dedicated measures. This measures should - as far as possible - be passive. One very promising solution for core melt retention seems to be a large enough spreading of the melt on a high temperature resistant protection layer with water cooling from above. This is the favorite concept for the EPR. In dealing with the retention of a molten core outside of the RPV several ``steps`` from leaving the RPV to finally stabilize the melt have to gone through. These steps are: collection of the melt; transfer of the melt; distribution of the melt; confining; cooling and stabilization. The technical features for the EPR solution of a large spreading of the melt are: Dedicated spreading chamber outside the reactor pit (area about 150 m{sup 2}); high temperature resistant protection layers (e.g. Zirconia bricks) at the bottom and part of the lateral structures (thus avoiding melt concrete interaction); reactor pit and spreading compartment are connected via a discharge channel which has a slope to the spreading area and is closed by a steel plate, which will resist the core melt for a certain time in order to allow a collection of the melt; the spreading compartments is connected with the In-Containment Refuelling Water Storage Tank (IRWST) with pipes for water flooding after spreading. These pipes are closed and will only be opened by the hot melt itself. It is shown how the course of the different steps mentioned above is processed and how each of these steps is automatically and passively achieved. (Abstract Truncated)

  15. A gas-cooled reactor surface power system

    International Nuclear Information System (INIS)

    Lipinski, R.J.; Wright, S.A.; Lenard, R.X.; Harms, G.A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars. copyright 1999 American Institute of Physics

  16. A gas-cooled reactor surface power system

    International Nuclear Information System (INIS)

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars

  17. A Gas-Cooled Reactor Surface Power System

    Energy Technology Data Exchange (ETDEWEB)

    Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

    1998-11-09

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  18. Response of fuel, fuel elements and gas cooled reactor cores under accidental air or water ingress conditions. Proceedings of a technical committee meeting held in Beijing, China, 25-27 October 1993

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-01-01

    The meeting was convened by the IAEA on the recommendation of the IAEA`s International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Netherlands, Switzerland, the Russian Federation, the United Kingdom and the United States of America. The meeting covered the following topics: experimental investigations of the effects of air and water ingress; predicted response of fuel, graphite and other reactor components; options for minimizing or mitigating the effects of air or water ingress. 19 papers were presented at the meeting. A separate abstract was prepared for each of these papers. Refs, figs and tabs.

  19. 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.)

  20. CAUSE AND EFFECT OF FEEDBACK: MULTIPHASE GAS IN CLUSTER CORES HEATED BY AGN JETS

    International Nuclear Information System (INIS)

    Gaspari, M.; Ruszkowski, M.; Sharma, P.

    2012-01-01

    Multiwavelength data indicate that the X-ray-emitting plasma in the cores of galaxy clusters is not cooling catastrophically. To a large extent, cooling is offset by heating due to active galactic nuclei (AGNs) via jets. The cool-core clusters, with cooler/denser plasmas, show multiphase gas and signs of some cooling in their cores. These observations suggest that the cool core is locally thermally unstable while maintaining global thermal equilibrium. Using high-resolution, three-dimensional simulations we study the formation of multiphase gas in cluster cores heated by collimated bipolar AGN jets. Our key conclusion is that spatially extended multiphase filaments form only when the instantaneous ratio of the thermal instability and free-fall timescales (t TI /t ff ) falls below a critical threshold of ≈10. When this happens, dense cold gas decouples from the hot intracluster medium (ICM) phase and generates inhomogeneous and spatially extended Hα filaments. These cold gas clumps and filaments 'rain' down onto the central regions of the core, forming a cold rotating torus and in part feeding the supermassive black hole. Consequently, the self-regulated feedback enhances AGN heating and the core returns to a higher entropy level with t TI /t ff > 10. Eventually, the core reaches quasi-stable global thermal equilibrium, and cold filaments condense out of the hot ICM whenever t TI /t ff ∼< 10. This occurs despite the fact that the energy from AGN jets is supplied to the core in a highly anisotropic fashion. The effective spatial redistribution of heat is enabled in part by the turbulent motions in the wake of freely falling cold filaments. Increased AGN activity can locally reverse the cold gas flow, launching cold filamentary gas away from the cluster center. Our criterion for the condensation of spatially extended cold gas is in agreement with observations and previous idealized simulations.

  1. Survey on Cooled-Vessel Designs in High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Kim, Min-Hwan; Lee, Won-Jae

    2006-01-01

    The core outlet temperature of the coolant in the high temperature gas-cooled reactors (HTGR) has been increased to improve the overall efficiency of their electricity generation by using the Brayton cycle or their nuclear hydrogen production by using thermo-chemical processes. The increase of the outlet temperature accompanies an increase of the coolant inlet temperature. A high coolant inlet temperature results in an increase of the reactor pressure vessel (RPV) operation temperature. The conventional steels, proven vessel material in light water reactors, cannot be used as materials for the RPV in the elevated temperatures which necessitate its design to account for the creep effects. Some ferritic or martensitic steels like 2 1/4Cr-1Mo and 9Cr-1Mo-V are very well established creep resistant materials for a temperature range of 400 to 550 C. Although these materials have been used in a chemical plant, there is limited experience with using these materials in nuclear reactors. Even though the 2 1/4Cr-1Mo steel was used to manufacture the RPV for HTR-10 of Japan Atomic Energy Agency(JAEA), a large RPV has not been manufactured by using this material or 9Cr-1Mo-V steel. Due to not only its difficulties in manufacturing but also its high cost, the JAEA determined that they would exclude these materials from the GTHTR design. For the above reasons, KAERI has been considering a cooled-vessel design as an option for the RPV design of a NHDD plant (Nuclear Hydrogen Development and Demonstration). In this study, we surveyed several HTGRs, which adopt the cooled-vessel concept for their RPV design, and discussed their design characteristics. The survey results in design considerations for the NHDD cooled-vessel design

  2. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies

  3. Testing Numerical Models of Cool Core Galaxy Cluster Formation with X-Ray Observations

    Science.gov (United States)

    Henning, Jason W.; Gantner, Brennan; Burns, Jack O.; Hallman, Eric J.

    2009-12-01

    Using archival Chandra and ROSAT data along with numerical simulations, we compare the properties of cool core and non-cool core galaxy clusters, paying particular attention to the region beyond the cluster cores. With the use of single and double β-models, we demonstrate a statistically significant difference in the slopes of observed cluster surface brightness profiles while the cluster cores remain indistinguishable between the two cluster types. Additionally, through the use of hardness ratio profiles, we find evidence suggesting cool core clusters are cooler beyond their cores than non-cool core clusters of comparable mass and temperature, both in observed and simulated clusters. The similarities between real and simulated clusters supports a model presented in earlier work by the authors describing differing merger histories between cool core and non-cool core clusters. Discrepancies between real and simulated clusters will inform upcoming numerical models and simulations as to new ways to incorporate feedback in these systems.

  4. Buffer gas cooling and mixture analysis

    Science.gov (United States)

    Patterson, David S.; Doyle, John M.

    2018-03-06

    An apparatus for spectroscopy of a gas mixture is described. Such an apparatus includes a gas mixing system configured to mix a hot analyte gas that includes at least one analyte species in a gas phase into a cold buffer gas, thereby forming a supersaturated mixture to be provided for spectroscopic analysis.

  5. The Role of Cerenkov Radiation in the Pressure Balance of Cool Core Clusters of Galaxies

    Energy Technology Data Exchange (ETDEWEB)

    Lieu, Richard [Department of Physics, University of Alabama, Huntsville, AL 35899 (United States)

    2017-03-20

    Despite the substantial progress made recently in understanding the role of AGN feedback and associated non-thermal effects, the precise mechanism that prevents the core of some clusters of galaxies from collapsing catastrophically by radiative cooling remains unidentified. In this Letter, we demonstrate that the evolution of a cluster's cooling core, in terms of its density, temperature, and magnetic field strength, inevitably enables the plasma electrons there to quickly become Cerenkov loss dominated, with emission at the radio frequency of ≲350 Hz, and with a rate considerably exceeding free–free continuum and line emission. However, the same does not apply to the plasmas at the cluster's outskirts, which lacks such radiation. Owing to its low frequency, the radiation cannot escape, but because over the relevant scale size of a Cerenkov wavelength the energy of an electron in the gas cannot follow the Boltzmann distribution to the requisite precision to ensure reabsorption always occurs faster than stimulated emission, the emitting gas cools before it reheats. This leaves behind the radiation itself, trapped by the overlying reflective plasma, yet providing enough pressure to maintain quasi-hydrostatic equilibrium. The mass condensation then happens by Rayleigh–Taylor instability, at a rate determined by the outermost radius where Cerenkov radiation can occur. In this way, it is possible to estimate the rate at ≈2 M {sub ⊙} year{sup −1}, consistent with observational inference. Thus, the process appears to provide a natural solution to the longstanding problem of “cooling flow” in clusters; at least it offers another line of defense against cooling and collapse should gas heating by AGN feedback be inadequate in some clusters.

  6. High temperature gas-cooled reactor: gas turbine application study

    International Nuclear Information System (INIS)

    1980-12-01

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project

  7. High temperature gas-cooled reactor: gas turbine application study

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

  8. Cooling performance of helium-gas/water coolers in HENDEL

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Takada, Shoji; Hayashi, Haruyoshi; Kobayashi, Toshiaki; Ohta, Yukimaru; Shimomura, Hiroaki; Miyamoto, Yoshiaki

    1994-01-01

    The helium engineering demonstration loop (HENDEL) has four helium-gas/water coolers where the cooling water flows in the tubes and helium gas on the shell side. Their cooling performance was studied using the operational data from 1982 to 1991. The heat transfer of helium gas on the shell was obtained for segmental and step-up baffle type coolers. Also, the change with operation time was investigated. The cooling performance was lowered by the graphite powder released from the graphite components for several thousand hours and thereafter recovered because the graphite powder from the components was reduced and the powder in the cooler shell was blown off during the operation. (orig.)

  9. Cause and Effect of Feedback: Multiphase Gas in Cluster Cores Heated by AGN Jets

    Science.gov (United States)

    Gaspari, M.; Ruszkowski, M.; Sharma, P.

    2012-02-01

    Multiwavelength data indicate that the X-ray-emitting plasma in the cores of galaxy clusters is not cooling catastrophically. To a large extent, cooling is offset by heating due to active galactic nuclei (AGNs) via jets. The cool-core clusters, with cooler/denser plasmas, show multiphase gas and signs of some cooling in their cores. These observations suggest that the cool core is locally thermally unstable while maintaining global thermal equilibrium. Using high-resolution, three-dimensional simulations we study the formation of multiphase gas in cluster cores heated by collimated bipolar AGN jets. Our key conclusion is that spatially extended multiphase filaments form only when the instantaneous ratio of the thermal instability and free-fall timescales (t TI/t ff) falls below a critical threshold of ≈10. When this happens, dense cold gas decouples from the hot intracluster medium (ICM) phase and generates inhomogeneous and spatially extended Hα filaments. These cold gas clumps and filaments "rain" down onto the central regions of the core, forming a cold rotating torus and in part feeding the supermassive black hole. Consequently, the self-regulated feedback enhances AGN heating and the core returns to a higher entropy level with t TI/t ff > 10. Eventually, the core reaches quasi-stable global thermal equilibrium, and cold filaments condense out of the hot ICM whenever t TI/t ff fashion. The effective spatial redistribution of heat is enabled in part by the turbulent motions in the wake of freely falling cold filaments. Increased AGN activity can locally reverse the cold gas flow, launching cold filamentary gas away from the cluster center. Our criterion for the condensation of spatially extended cold gas is in agreement with observations and previous idealized simulations.

  10. Update Knowledge Base for Long-term Core Cooling Reliability

    International Nuclear Information System (INIS)

    Agrell, Maria; Sandervag, Oddbjoern; Amri, Abdallah; ); Bang, Young S.; Blomart, Philippe; Broecker, Annette; Pointner, Winfried; Ganzmann, Ingo; Lenogue, Bruno; Guzonas, David; Herer, Christophe; Mattei, Jean-Marie; Tricottet, Matthieu; Masaoka, Hideaki; Soltesz, Vojtech; Tarkiainen, Seppo; Ui, Atsushi; Villalba, Cristina; Zigler, Gilbert

    2013-11-01

    This revision of the Knowledge Base for Emergency Core Cooling System Recirculation Reliability (NEA/CSNI/R (95)11) describes the current status (late 2012) of the knowledge base on emergency core cooling system (ECCS) and containment spray system (CSS) suction strainer performance and long-term cooling in operating power reactors. New reactors, such as the AP1000, EPR and APR1400 that are under construction in some Organization for Economic Co-operation and Development (OECD) member countries, are not addressed in detail in this revision. The containment sump (also known as the emergency or recirculation sump in pressurized water reactors (PWRs) and pressurized heavy water reactors (PHWRs) or the suppression pools or wet wells in boiling water reactors (BWRs)) and associated ECCS strainers are parts of the ECCS in both reactor types. All nuclear power plants (NPPs) are required to have an ECCS that is capable of mitigating a design basis accident (DBA). The containment sump collects reactor coolant, ECCS injection water, and containment spray solutions, if applicable, after a loss-of-coolant accident (LOCA). The sump serves as the water source to support long-term recirculation for residual heat removal, emergency core cooling, and containment atmosphere clean-up. This water source, the related pump suction inlets, and the piping between the source and inlets are important safety-related components. In addition, if fibrous material is deposited at the fuel element spacers, core cooling can be endangered. The performance of ECCS/CSS strainers was recognized many years ago as an important regulatory and safety issue. One of the primary concerns is the potential for debris generated by a jet of high-pressure coolant during a LOCA to clog the strainer and obstruct core cooling. The issue was considered resolved for all reactor types in the mid-1990s and the OECD/NEA/CSNI published report NEA/CSNI/R(95)11 in 1996 to document the state of knowledge of ECCS performance

  11. Exergy analysis of a gas-hydrate cool storage system

    International Nuclear Information System (INIS)

    Bi, Yuehong; Liu, Xiao; Jiang, Minghe

    2014-01-01

    Based on exergy analysis of charging and discharging processes in a gas-hydrate cool storage system, the formulas for exergy efficiency at the sensible heat transfer stage and the phase change stage corresponding to gas-hydrate charging and discharging processes are obtained. Furthermore, the overall exergy efficiency expressions of charging, discharging processes and the thermodynamic cycle of the gas-hydrate cool storage system are obtained. By using the above expressions, the effects of number of transfer units, the inlet temperatures of the cooling medium and the heating medium on exergy efficiencies of the gas-hydrate cool storage system are emphatically analyzed. The research results can be directly used to evaluate the performance of gas-hydrate cool storage systems and design more efficient energy systems by reducing the sources of inefficiency in gas-hydrate cool storage systems. - Highlights: • Formulas for exergy efficiency at four stages are obtained. • Exergy efficiency expressions of two processes and one cycle are obtained. • Three mainly influencing factors on exergy efficiencies are analyzed. • With increasing the inlet temperature of cooling medium, exergy efficiency increases. • With decreasing the inlet temperature of heating medium, exergy efficiency increases

  12. Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts

    International Nuclear Information System (INIS)

    K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05)

  13. Testing the Large-scale Environments of Cool-core and Non-cool-core Clusters with Clustering Bias

    Energy Technology Data Exchange (ETDEWEB)

    Medezinski, Elinor; Battaglia, Nicholas; Cen, Renyue; Gaspari, Massimo; Strauss, Michael A.; Spergel, David N. [Department of Astrophysical Sciences, 4 Ivy Lane, Princeton, NJ 08544 (United States); Coupon, Jean, E-mail: elinorm@astro.princeton.edu [Department of Astronomy, University of Geneva, ch. dEcogia 16, CH-1290 Versoix (Switzerland)

    2017-02-10

    There are well-observed differences between cool-core (CC) and non-cool-core (NCC) clusters, but the origin of this distinction is still largely unknown. Competing theories can be divided into internal (inside-out), in which internal physical processes transform or maintain the NCC phase, and external (outside-in), in which the cluster type is determined by its initial conditions, which in turn leads to different formation histories (i.e., assembly bias). We propose a new method that uses the relative assembly bias of CC to NCC clusters, as determined via the two-point cluster-galaxy cross-correlation function (CCF), to test whether formation history plays a role in determining their nature. We apply our method to 48 ACCEPT clusters, which have well resolved central entropies, and cross-correlate with the SDSS-III/BOSS LOWZ galaxy catalog. We find that the relative bias of NCC over CC clusters is b = 1.42 ± 0.35 (1.6 σ different from unity). Our measurement is limited by the small number of clusters with core entropy information within the BOSS footprint, 14 CC and 34 NCC clusters. Future compilations of X-ray cluster samples, combined with deep all-sky redshift surveys, will be able to better constrain the relative assembly bias of CC and NCC clusters and determine the origin of the bimodality.

  14. Testing the Large-scale Environments of Cool-core and Non-cool-core Clusters with Clustering Bias

    International Nuclear Information System (INIS)

    Medezinski, Elinor; Battaglia, Nicholas; Cen, Renyue; Gaspari, Massimo; Strauss, Michael A.; Spergel, David N.; Coupon, Jean

    2017-01-01

    There are well-observed differences between cool-core (CC) and non-cool-core (NCC) clusters, but the origin of this distinction is still largely unknown. Competing theories can be divided into internal (inside-out), in which internal physical processes transform or maintain the NCC phase, and external (outside-in), in which the cluster type is determined by its initial conditions, which in turn leads to different formation histories (i.e., assembly bias). We propose a new method that uses the relative assembly bias of CC to NCC clusters, as determined via the two-point cluster-galaxy cross-correlation function (CCF), to test whether formation history plays a role in determining their nature. We apply our method to 48 ACCEPT clusters, which have well resolved central entropies, and cross-correlate with the SDSS-III/BOSS LOWZ galaxy catalog. We find that the relative bias of NCC over CC clusters is b = 1.42 ± 0.35 (1.6 σ different from unity). Our measurement is limited by the small number of clusters with core entropy information within the BOSS footprint, 14 CC and 34 NCC clusters. Future compilations of X-ray cluster samples, combined with deep all-sky redshift surveys, will be able to better constrain the relative assembly bias of CC and NCC clusters and determine the origin of the bimodality.

  15. BRIGHTEST CLUSTER GALAXIES AND CORE GAS DENSITY IN REXCESS CLUSTERS

    International Nuclear Information System (INIS)

    Haarsma, Deborah B.; Leisman, Luke; Donahue, Megan; Bruch, Seth; Voit, G. Mark; Boehringer, Hans; Pratt, Gabriel W.; Pierini, Daniele; Croston, Judith H.; Arnaud, Monique

    2010-01-01

    We investigate the relationship between brightest cluster galaxies (BCGs) and their host clusters using a sample of nearby galaxy clusters from the Representative XMM-Newton Cluster Structure Survey. The sample was imaged with the Southern Observatory for Astrophysical Research in R band to investigate the mass of the old stellar population. Using a metric radius of 12 h -1 kpc, we found that the BCG luminosity depends weakly on overall cluster mass as L BCG ∝ M 0.18±0.07 cl , consistent with previous work. We found that 90% of the BCGs are located within 0.035 r 500 of the peak of the X-ray emission, including all of the cool core (CC) clusters. We also found an unexpected correlation between the BCG metric luminosity and the core gas density for non-cool-core (non-CC) clusters, following a power law of n e ∝ L 2.7±0.4 BCG (where n e is measured at 0.008 r 500 ). The correlation is not easily explained by star formation (which is weak in non-CC clusters) or overall cluster mass (which is not correlated with core gas density). The trend persists even when the BCG is not located near the peak of the X-ray emission, so proximity is not necessary. We suggest that, for non-CC clusters, this correlation implies that the same process that sets the central entropy of the cluster gas also determines the central stellar density of the BCG, and that this underlying physical process is likely to be mergers.

  16. Improvement of Cooling Technology through Atmosphere Gas Management

    Energy Technology Data Exchange (ETDEWEB)

    Renard, Michel; Dosogne, Edgaar; Crutzen, Jean Pierre; Raick, Jean Mare [DREVER INTERNATIONAL S.A., Liege (Belgium); Ji, Ma Jia; Jun, Lv; Zhi, Ma Bing [SHOUGANG Cold Rolling Mill Headquarter, Beijin (China)

    2009-12-15

    The production of advanced high strength steels requires the improvement of cooling technology. The use of high cooling rates allows relatively low levels of expensive alloying additions to ensure sufficient hardenability. In classical annealing and hot-dip galvanizing lines a mixing station is used to provide atmosphere gas containing 3-5% hydrogen and 97-95% nitrogen in the various sections of the furnace, including the rapid cooling section. Heat exchange enhancement in this cooling section can be insured by the increased hydrogen concentration. Driver international developed a patented improvement of cooling technology based on the following features: pure hydrogen gas is injected only in the rapid cooling section whereas the different sections of the furnace are supplied with pure nitrogen gas: the control of flows through atmosphere gas management allows to get high hydrogen concentration in cooling section and low hydrogen content in the other furnace zones. This cooling technology development insures higher cooling rates without additional expensive hydrogen gas consumption and without the use of complex sealing equipment between zones. In addition reduction in electrical energy consumption is obtained. This atmosphere control development can be combined with geometrical design improvements in order to get optimised cooling technology providing high cooling rates as well as reduced strip vibration amplitudes. Extensive validation of theoretical research has been conducted on industrial lines. New lines as well as existing lines, with limited modifications, can be equipped with this new development. Up to now this technology has successfully been implemented on 6 existing and 7 new lines in Europe and Asia.

  17. Basic criticality relations for gas core design

    International Nuclear Information System (INIS)

    Tanner, J.E.

    1992-01-01

    Minimum critical fissile concentrations are calculated for U-233, U-235, Pu-239, and Am-242m mixed homogeneously with hydrogen at temperatures to 15,000K. Minimum critical masses of the same mixtures in a 1000 liter sphere are also calculated. It is shown that propellent efficiencies of a gas core fizzler engine using Am-242m as fuel would exceed those in a solid core engine as small as 1000L operating at 100 atmospheres pressure. The same would be true for Pu-239 and possibly U-233 at pressures of 1000 atm. or at larger volumes

  18. Modelling the radiolysis of RSG-GAS primary cooling water

    Science.gov (United States)

    Butarbutar, S. L.; Kusumastuti, R.; Subekti, M.; Sunaryo, G. R.

    2018-02-01

    Water chemistry control for light water coolant reactor required a reliable understanding of radiolysis effect in mitigating corrosion and degradation of reactor structure material. It is known that oxidator products can promote the corrosion, cracking and hydrogen pickup both in the core and in the associated piping components of the reactor. The objective of this work is to provide the radiolysis model of RSG GAS cooling water and further more to predict the oxidator concentration which can lead to corrosion of reactor material. Direct observations or measurements of the chemistry in and around the high-flux core region of a nuclear reactor are difficult due to the extreme conditions of high temperature, pressure, and mixed radiation fields. For this reason, chemical models and computer simulations of the radiolysis of water under these conditions are an important route of investigation. FACSIMILE were used to calculate the concentration of O2 formed at relatively long-time by the pure water γ and neutron irradiation (pH=7) at temperature between 25 and 50 °C. This simulation method is based on a complex chemical reaction kinetic. In this present work, 300 MeV-proton were used to mimic γ-rays radiolysis and 2 MeV fast neutrons. Concentration of O2 were calculated at 10-6 - 106 s time scale.

  19. Gas turbine bucket with impingement cooled platform

    Science.gov (United States)

    Jones, Raphael Durand

    2002-01-01

    In a turbine bucket having an airfoil portion and a root portion, with a substantially planar platform at an interface between the airfoil portion and root portion, a platform cooling arrangement including at least one bore in the root portion and at least one impingement cooling tube seated in the bore, the tube extending beyond the bore with an outlet in close proximity to a targeted area on an underside of the platform.

  20. Continuous greenhouse gas measurements from ice cores

    DEFF Research Database (Denmark)

    Stowasser, Christopher

    Ice cores offer the unique possibility to study the history of past atmospheric greenhouse gases over the last 800,000 years, since past atmospheric air is trapped in bubbles in the ice. Since the 1950s, paleo-scientists have developed a variety of techniques to extract the trapped air from...... individual ice core samples, and to measure the mixing ratio of greenhouse gases such as carbon dioxide, methane and nitrous oxide in the extracted air. The discrete measurements have become highly accurate and reproducible, but require relatively large amounts of ice per measured species and are both time......-consuming and labor-intensive. This PhD thesis presents the development of a new method for measurements of greenhouse gas mixing ratios from ice cores based on a melting device of a continuous flow analysis (CFA) system. The coupling to a CFA melting device enables time-efficient measurements of high resolution...

  1. Safety aspects of forced flow cooldown transients in modular high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kroeger, P.G.

    1992-01-01

    During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs) the main Heat Transport System (HTS) and the Shutdown Cooling System (SCS), are assumed to have failed. Decay heat is then removed by the passive Reactor Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This paper uses the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits

  2. Gas core nuclear rocket feasibility project

    International Nuclear Information System (INIS)

    Howe, S.D.; DeVolder, B.; Thode, L.; Zerkle, D.

    1997-09-01

    The next giant leap for mankind will be the human exploration of Mars. Almost certainly within the next thirty years, a human crew will brave the isolation, the radiation, and the lack of gravity to walk on and explore the Red planet. However, because the mission distances and duration will be hundreds of times greater than the lunar missions, a human crew will face much greater obstacles and a higher risk than those experienced during the Apollo program. A single solution to many of these obstacles is to dramatically decrease the mission duration by developing a high performance propulsion system. The gas core nuclear rocket (GCNR) has the potential to be such a system. The gas core concept relies on the use of fluid dynamic forces to create and maintain a vortex. The vortex is composed of a fissile material which will achieve criticality and produce high power levels. By radiatively coupling to the surrounding fluids, extremely high temperatures in the propellant and, thus, high specific impulses can be generated. The ship velocities enabled by such performance may allow a 9 month round trip, manned Mars mission to be considered. Alternatively, one might consider slightly longer missions in ships that are heavily shielded against the intense Galactic Cosmic Ray flux to further reduce the radiation dose to the crew. The current status of the research program at the Los Alamos National Laboratory into the gas core nuclear rocket feasibility will be discussed

  3. Cooling system of the core of a nuclear reactor while it is being stopped or normally operating

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1986-01-01

    The present invention proposes a cooling system with intermediate gas flow which ensures the reactor core cooling when the primary pumps are stopped either directly by means of main heat-exchange circuits ensuring normally the reactor operation, or by means of separated loops, these ones being able so to operate in an autonomous way for they produce their own electricity needs and also an excedent which is added to the power plant production. The cooling circuit and the heat exchanger are described in detail [fr

  4. Gas cooled fast reactor benchmarks for JNC and Cea neutronic tools assessment

    International Nuclear Information System (INIS)

    Rimpault, G.; Sugino, K.; Hayashi, H.

    2005-01-01

    In order to verify the adequacy of JNC and Cea computational tools for the definition of GCFR (gas cooled fast reactor) core characteristics, GCFR neutronic benchmarks have been performed. The benchmarks have been carried out on two different cores: 1) a conventional Gas-Cooled fast Reactor (EGCR) core with pin-type fuel, and 2) an innovative He-cooled Coated-Particle Fuel (CPF) core. Core characteristics being studied include: -) Criticality (Effective multiplication factor or K-effective), -) Instantaneous breeding gain (BG), -) Core Doppler effect, and -) Coolant depressurization reactivity. K-effective and coolant depressurization reactivity at EOEC (End Of Equilibrium Cycle) state were calculated since these values are the most critical characteristics in the core design. In order to check the influence due to the difference of depletion calculation systems, a simple depletion calculation benchmark was performed. Values such as: -) burnup reactivity loss, -) mass balance of heavy metals and fission products (FP) were calculated. Results of the core design characteristics calculated by both JNC and Cea sides agree quite satisfactorily in terms of core conceptual design study. Potential features for improving the GCFR computational tools have been discovered during the course of this benchmark such as the way to calculate accurately the breeding gain. Different ways to improve the accuracy of the calculations have also been identified. In particular, investigation on nuclear data for steel is important for EGCR and for lumped fission products in both cores. The outcome of this benchmark is already satisfactory and will help to design more precisely GCFR cores. (authors)

  5. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

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

  6. Unravelling the core microbiome of biofilms in cooling tower systems.

    Science.gov (United States)

    Di Gregorio, L; Tandoi, V; Congestri, R; Rossetti, S; Di Pippo, F

    2017-11-01

    In this study, next generation sequencing and catalyzed reporter deposition fluorescence in situ hybridization, combined with confocal microscopy, were used to provide insights into the biodiversity and structure of biofilms collected from four full-scale European cooling systems. Water samples were also analyzed to evaluate the impact of suspended microbes on biofilm formation. A common core microbiome, containing members of the families Sphingomonadaceae, Comamonadaceae and Hyphomicrobiaceae, was found in all four biofilms, despite the water of each coming from different sources (river and groundwater). This suggests that selection of the pioneer community was influenced by abiotic factors (temperature, pH) and tolerances to biocides. Members of the Sphingomonadaceae were assumed to play a key role in initial biofilm formation. Subsequent biofilm development was driven primarily by light availability, since biofilms were dominated by phototrophs in the two studied 'open' systems. Their interactions with other microbial populations then shaped the structure of the mature biofilm communities analyzed.

  7. A comparative design study of PB-BI cooled reactor cores with forced and natural convection cooling

    International Nuclear Information System (INIS)

    Mizuno, Tomoyasu; Enuma, Yasuhiro; Tanji, Mikio

    2003-01-01

    A comparative core design study is performed on Pb-Bi cooled reactors with forced and natural convection (FC and NC) cooling. Major interests of the study are core performance and core safety features. The designed core concepts with nitride fuel achieve reasonable breeding capability. The results of unprotected event analyses such as UTOP and ULOF show that both of concepts have possible features to withstand unprotected events due to negative reactivity feedback by Doppler effect, control rod drive line expansion, etc. These results lead to a conclusion that both of concepts have possible capability as one of future promising core concepts. A FC cooling core concept has more advantage if fuel recycle viewpoint is emphasized. (author)

  8. Gas Mixtures for Welding with Micro-Jet Cooling

    OpenAIRE

    Węgrzyn T.

    2015-01-01

    Welding with micro-jet cooling after was tested only for MIG and MAG processes. For micro-jet gases was tested only argon, helium and nitrogen. A paper presents a piece of information about gas mixtures for micro-jet cooling after in welding. There are put down information about gas mixtures that could be chosen both for MAG welding and for micro-jet process. There were given main information about influence of various micro-jet gas mixtures on metallographic structure of steel welds. Mechani...

  9. Compatibility of gas turbine materials with steam cooling

    Energy Technology Data Exchange (ETDEWEB)

    Desai, V.; Tamboli, D.; Patel, Y. [Univ. of Central Florida, Orlando, FL (United States)

    1995-10-01

    Gas turbines had been traditionally used for peak load plants and remote locations as they offer advantage of low installation costs and quick start up time. Their use as a base load generator had not been feasible owing to their poor efficiency. However, with the advent of gas turbines based combined cycle plants (CCPs), continued advances in efficiency are being made. Coupled with ultra low NO{sub x} emissions, coal compatibility and higher unit output, gas turbines are now competing with conventional power plants for base load power generation. Currently, the turbines are designed with TIT of 2300{degrees}F and metal temperatures are maintained around 1700{degrees}F by using air cooling. New higher efficiency ATS turbines will have TIT as high as 2700{degrees}F. To withstand this high temperature improved materials, coatings, and advances in cooling system and design are warranted. Development of advanced materials with better capabilities specifically for land base applications are time consuming and may not be available by ATS time frame or may prove costly for the first generation ATS gas turbines. Therefore improvement in the cooling system of hot components, which can take place in a relatively shorter time frame, is important. One way to improve cooling efficiency is to use better cooling agent. Steam as an alternate cooling agent offers attractive advantages because of its higher specific heat (almost twice that of air) and lower viscosity.

  10. 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.)

  11. Design evaluation of emergency core cooling systems using Axiomatic Design

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Gyunyoung [Massachusetts Institute of Technology, Department of Mechanical Engineering, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)]. E-mail: gheo@mit.edu; Lee, Song Kyu [Korea Advanced Institute of Science and Technology, Department of Nuclear and Quantum Engineering, 373-1 Guseong-dong, Yuseong-gu, Daejeon (Korea, Republic of)

    2007-01-15

    In designing nuclear power plants (NPPs), the evaluation of safety is one of the important issues. As a measure for evaluating safety, this paper proposes a methodology to examine the design process of emergency core cooling systems (ECCSs) in NPPs using Axiomatic Design (AD). This is particularly important for identifying vulnerabilities and creating solutions. Korean Advanced Power Reactor 1400 MWe (APR1400) adopted the ECCS, which was improved to meet the stronger safety regulations than that of the current Optimized Power Reactor 1000 MWe (OPR1000). To improve the performance and safety of the ECCS, the various design strategies such as independency or redundancy were implemented, and their effectiveness was confirmed by calculating core damage frequency. We suggest an alternative viewpoint of evaluating the deployment of design strategies in terms of AD methodology. AD suggests two design principles and the visualization tools for organizing design process. The important benefit of AD is that it is capable of providing suitable priorities for deploying design strategies. The reverse engineering driven by AD has been able to show that the design process of the ECCS of APR1400 was improved in comparison to that of OPR1000 from the viewpoint of the coordination of design strategies.

  12. Post-implementation review of inadequate core cooling instrumentation

    International Nuclear Information System (INIS)

    Anderson, J.L.; Anderson, R.L.; Hagen, E.W.; Morelock, T.C.; Huang, T.L.; Phillips, L.E.

    1988-01-01

    Studies of Three Mile Island (TMI) accident identified the need for additional instrumentation to detect inadequate core cooling (ICC) in nuclear power plants. Industry studies by plant owners and reactor vendors supported the conclusion that improvements were needed to help operators diagnose the approach to or existence of ICC and to provide more complete information for operator control of safety injection, flow to minimize the consequences of such an accident. In 1980, the US Nuclear Regulatory Commission (NRC) required further studies by the industry and described ICC instrumentation design requirements that included human factors and environmental considerations. On December 10, 1982, NRC issued to Babcock and Wilcox (BandW) licensees' orders for Modification of License and transmitted to all pressurized water reactor (PWR) licensees Generic Letter 82-28 to inform them of the revised NRC requirements. The instrumentation requirements for detection of ICC include upgraded subcooling margin monitors (SMMs), upgraded core exit thermocouples (CETs), and installation of a reactor coolant inventory tracking system (RCITS)

  13. 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

  14. Validation of CATHARE for gas-cooled reactors

    International Nuclear Information System (INIS)

    Fabrice Bentivoglio; Ola Widlund; Manuel Saez

    2005-01-01

    Full text of publication follows: Extensively validated and qualified for light-water reactor safety studies, the thermo-hydraulics code CATHARE has been adapted to deal also with gas-cooled reactor applications. In order to validate the code for these novel applications, CEA (Commissariat a l'Energie Atomique) has initiated an ambitious long-term experimental program. The foreseen experimental facilities range from small-scale loops for physical correlations, to component technology and system demonstration loops. In the short-term perspective, CATHARE is being validated against existing experimental data, in particular from the German power plant Oberhausen II and the South African Pebble-Bed Micro Model (PBMM). Oberhausen II, operated by the German utility EVO, is a 50 MW(e) direct-cycle Helium turbine plant. The power source is a gas burner rather than a nuclear reactor core, but the power conversion system resembles those of the GFR (Gas-cooled Fast Reactor) and other high-temperature reactor concepts. Oberhausen II was operated for more than 100 000 hours between 1974 and 1988. Design specifications, drawings and experimental data have been obtained through the European HTR project, offering a unique opportunity to validate CATHARE on a large-scale Brayton cycle. Available measurements of temperatures, pressures and mass flows throughout the circuit have allowed a very comprehensive thermohydraulic description of the plant, in steady-state conditions as well as during transients. The Pebble-Bed Micro Model (PBMM) is a small-scale model conceived to demonstrate the operability and control strategies of the South African PBMR concept. The model uses Nitrogen instead of Helium, and an electrical heater with a maximum rating of 420 kW. As the full-scale PBMR, the PBMM loop features three turbines and two compressors on the primary circuit, located on three separate shafts. The generator, however, is modelled by a third compressor on a separate circuit, with a

  15. Removal of tritium from gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Nieder, R.

    1976-01-01

    Tritium contained in the coolant gas in the primary circuit of a gas cooled nuclear reactor together with further tritium adsorbed on the graphite used as a moderator for the reactor is removed by introducing hydrogen or a hydrogen-containing compound, for example methane or ammonia, into the coolant gas. The addition of the hydrogen or hydrogen-containing compound to the coolant gas causes the adsorbed tritium to be released into the coolant gas and the tritium is then removed from the coolant gas by passing the mixture of coolant gas and hydrogen or hydrogen-containing compound through a gas purification plant before recirculating the coolant gas through the reactor. 14 claims, 1 drawing figure

  16. Reassessment of debris ingestion effects on emergency core cooling-system pump performance

    International Nuclear Information System (INIS)

    Sciacca, F.W.; Rao, D.V.

    2004-01-01

    A study sponsored by the United States (US) Nuclear Regulatory Commission (NRC) was performed to reassess the effects of ingesting loss of coolant accident (LOCA) generated materials into emergency core cooling system (ECCS) pumps and the subsequent impact of this debris on the pumps' ability to provide long-term cooling to the reactor core. ECCS intake systems have been designed to screen out large post-LOCA debris materials. However, small-sized debris can penetrate these intake strainers or screens and reach critical pump components. Prior NRC-sponsored evaluations of possible debris and gas ingestion into ECCS pumps and attendant impacts on pump performance were performed in the early 1980's. The earlier study focused primarily on pressurised water reactor (PWR) ECCS pumps. This issue was revisited both to factor in our improved knowledge of LOCA generated debris and to address specifically both boiling water reactor (BWR) and PWR ECCS pumps. This study discusses the potential effects of ingested debris on pump seals, bearing assemblies, cyclone debris separators, and seal cooling water subsystems. This assessment included both near-term (less than one hour) and long-term (greater than one hour) effects introduced by the postulated LOCA. The work reported herein was performed during 1996-1997. (authors)

  17. A 3.55 keV line from DM →a→γ: predictions for cool-core and non-cool-core clusters

    Energy Technology Data Exchange (ETDEWEB)

    Conlon, Joseph P.; Powell, Andrew J. [Rudolf Peierls Centre for Theoretical Physics, University of Oxford, 1 Keble Road, Oxford, OX1 3NP (United Kingdom)

    2015-01-13

    We further study a scenario in which a 3.55 keV X-ray line arises from decay of dark matter to an axion-like particle (ALP), that subsequently converts to a photon in astrophysical magnetic fields. We perform numerical simulations of Gaussian random magnetic fields with radial scaling of the magnetic field magnitude with the electron density, for both cool-core 'Perseus' and non-cool-core 'Coma' electron density profiles. Using these, we quantitatively study the resulting signal strength and morphology for cool-core and non-cool-core clusters. Our study includes the effects of fields of view that cover only the central part of the cluster, the effects of offset pointings on the radial decline of signal strength and the effects of dividing clusters into annuli. We find good agreement with current data and make predictions for future analyses and observations.

  18. Use of thorium for high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Guimarães, Cláudio Q., E-mail: claudio_guimaraes@usp.br [Universidade de São Paulo (USP), SP (Brazil). Instituto de Física; Stefani, Giovanni L. de, E-mail: giovanni.stefani@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Santos, Thiago A. dos, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil)

    2017-07-01

    The HTGR ( High Temperature Gas-cooled Reactor) is a 4{sup th} generation nuclear reactor and is fuelled by a mixture of graphite and fuel-bearing microspheres. There are two competitive designs of this reactor type: The German “pebble bed” mode, which is a system that uses spherical fuel elements, containing a graphite-and-fuel mixture coated in a graphite shell; and the American version, whose fuel is loaded into precisely located graphite hexagonal prisms that interlock to create the core of the vessel. In both variants, the coolant consists of helium pressurised. The HTGR system operates most efficiently with the thorium fuel cycle, however, so relatively little development has been carried out in this country on that cycle for HTGRs. In the Nuclear Engineering Centre of IPEN (Instituto de Pesquisas Energéticas e Nucleares), a study group is being formed linked to thorium reactors, whose proposal is to investigate reactors using thorium for {sup 233}U production and rejects burning. The present work intends to show the use of thorium in HTGRs, their advantages and disadvantages and its feasibility. (author)

  19. Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

    2004-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  20. Evaluation of long-term post-accident core cooling of Three Mile Island Unit 2

    Energy Technology Data Exchange (ETDEWEB)

    None

    1979-04-15

    On the basis of current understanding of the accident scenario and available data, the staff reports here on its evaluation of the condition of the core and the core flow resistance as it might affect ability to cool the core by natural circulation. The natural circulation cooling capability of TMI-2 for the estimated core flow resistance and a variety of other conditions is evaluated and a comparison of the Base Case and off-nominal plant configurations is presented. The potential for and effects of natural convection core cooling are addressed, and the staff recommendations for reactor performance acceptance criteria upon initiation of natural convection are presented.

  1. IAEA high temperature gas-cooled reactor activities

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2000-01-01

    The IAEA activities on high temperature gas-cooled reactors are conducted with the review and support of the Member states, primarily through the International Working Group on Gas-Cooled Reactors (IWG-GCR). This paper summarises the results of the IAEA gas-cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (authors)

  2. Interotex-innovative gas equipment for heating and cooling

    Energy Technology Data Exchange (ETDEWEB)

    Winnington, T.L. [Interotex Ltd. (United Kingdom); Moore, N. [British Gas plc (United Kingdom); Valle, F.; Sanz, J. I. [Gas Natural SDG S.A. (Spain); Chavarri, J.M. [Fagor Electrodomesticos S. Coop. (Spain); Uselton, R. [Lennox Industries Inc. (United States)

    1997-10-01

    Conventionally, cooling technology for the residential market is provided by electrically driven vapour re-compression systems. But lately, due to the Montreal Protocol - restricting the utilisation of ozone depleting substances - and to the high peak demand in electricity, created by electrical air conditioning systems, there is a commercial opportunity for gas fired air conditioning appliances. This paper describes the development programme for a radical new absorption technology, from the theoretical studies, through the experimental programme, to the building, commissioning and installation of demonstration machines. It also includes an analysis of the world-wide residential cooling market and the opportunities available to manufacturers and gas utilities to introduce new gas heating and cooling technology, capable of competing effectively with electrical systems. (au)

  3. Design requirements, operation and maintenance of gas-cooled reactors

    International Nuclear Information System (INIS)

    1989-06-01

    At the invitation of the Government of the USA the Technical Committee Meeting on Design Requirements, Operation and Maintenance of Gas-Cooled Reactors, was held in San Diego on September 21-23, 1988, in tandem with the GCRA Conference. Both meetings attracted a large contingent of foreign participants. Approximately 100 delegates from 18 different countries participated in the Technical Committee meeting. The meeting was divided into three sessions: Gas-cooled reactor user requirement (8 papers); Gas-cooled reactor improvements to facilitate operation and maintenance (10 papers) and Safety, environmental impacts and waste disposal (5 papers). A separate abstract was prepared for each of these 23 papers. Refs, figs and tabs

  4. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Lap-Yan, C.; Wie, T. Y. C.

    2009-01-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

  5. Simulation for temperature changing investigation at RSG-GAS cooling system

    International Nuclear Information System (INIS)

    Utaja

    2002-01-01

    The RSG-GAS cooling system considers of primary and secondary system, is used for heat rejection from reactor core to the atmosphere. For temperature changing investigation cause by atmospherics condition changing or coolant flow rate changing, is more safe done by simulation. This paper describes the simulation for determine the RSG-GAS coolant temperature changing base on heat exchange and cooling tower characteristic. The simulation is done by computer programme running under WINDOWS 95 or higher. The temperature changing is based on heat transfer process on heat exchanger and cooling tower. The simulation will show the water tank temperature changing caused by the temperature and humidity of the atmosphere or by coolant flow rate changing. For example the humidity changing from 60% to 80% atmospherics temperature 30 oC and 32400 k Watt power will change the tank temperature from 37,97 oC to 40,03 oC

  6. Heat removal in gas-cooled fuel rod clusters

    International Nuclear Information System (INIS)

    Rehme, K.

    1975-01-01

    For a thermo- and fluid-dynamic analysis of fuel rod cluster subchannels for gas-cooled breeder reactors, the following values must be verified: a) friction coefficient as flow parameter; b) Stanton number as heat transfer parameter; c) influence of spacers on friction coefficient and Stanton number; d) heat and mass exchange between subchannels with different temperatures. These parameters are established by combining results of single experiments and of integral experiments. Mention is made of further studies to be performed in order to determine the heat removal from gas-cooled fast breeder fuel elements. (HR) [de

  7. Gas-cooled reactors for advanced terrestrial applications

    International Nuclear Information System (INIS)

    Kesavan, K.; Lance, J.R.; Jones, A.R.; Spurrier, F.R.; Peoples, J.A.; Porter, C.A.; Bresnahan, J.D.

    1986-01-01

    Conceptual design of a power plant on an inert gas cooled nuclear coupled to an open, air Brayton power conversion cycle is presented. The power system, called the Westinghouse GCR/ATA (Gas-Cooled Reactors for Advanced Terrestrial Applications), is designed to meet modern military needs, and offers the advantages of secure, reliable and safe electrical power. The GCR/ATA concept is adaptable over a range of 1 to 10 MWe power output. Design descriptions of a compact, air-transportable forward base unit for 1 to 3 MWe output and a fixed-base, permanent installation for 3 to 10 MWe output are presented

  8. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in

  9. Simulation Of The Secondary Cooling System Failed For One Line Mode Of RSG-GAS

    International Nuclear Information System (INIS)

    Dibyo, Sukmanto; Susyadi; Sembiring, Tagor M; Isnaeni, Darwis

    2003-01-01

    Recently, an assessment of 15 MW power reactor RSG-GAS operated using one line cooling mode is under carried out, in which is in the same manner as BA TAN policy. At the power above mentioned, requirement for the research as well as isotop production has been fulfilled. To obtain the transient condition of 1 line-cooling mode, the simulation using RELAP5.MOD3.2 code was carried out. The simulation parameters interesting known are the inlet of primary coolant temperature after failed the secondary cooling system. At the first, reactor is operated at 15 MW steady state condition using 1 line-cooling mode. Primary coolant flow rate of 430 kg/s and secondary of 550 kg/s respectively. After that the decreasing is occurred due to stop of secondary cooling pump. Therefore the primary cooling inlet temperature to the core increase cause scram reactor by inserted control rod. During the transient occur, the characteristic of primary cooling temperature pattern change were obtained. The simulation result shows that the temperature increase (ΔT) temperature to the reactor is 5,1 o C at the second of 85.5. Here is lower than ΔT for the two cooling mode of 10 o C. That temperature characteristic still tolerable against acceptable safety margin to the flow instability

  10. Optimization of advanced gas-cooled reactor fuel performance by a stochastic method

    International Nuclear Information System (INIS)

    Parks, G.T.

    1987-01-01

    A brief description is presented of a model representing the in-core behaviour of a single advanced gas-cooled reactor fuel channel, developed specifically for optimization studies. The performances of the only suitable Numerical Algorithms Group (NAG) library package and a Metropolis algorithm routine on this problem are discussed and contrasted. It is concluded that, for the problem in question, the stochastic Metropolis algorithm has distinct advantages over the deterministic NAG routine. (author)

  11. A solution to level 3 dismantling of gas-cooled reactors: Graphite incineration

    International Nuclear Information System (INIS)

    Dubourg, M.

    1993-01-01

    This paper presents an approach developed to solve the specific decommissioning problems of the G2 and G3 gas cooled reactors at Marcoule and the strategy applied with emphasis in incinerating the graphite core components, using a fluidized-bed incinerator developed jointly between the CEA and FRAMATOME. The incineration option was selected over subsurface storage for technical and economic reasons. Studies have shown that gaseous incineration releases are environmentally acceptable

  12. Cooling of gas turbines IX : cooling effects from use of ceramic coatings on water-cooled turbine blades

    Science.gov (United States)

    Brown, W Byron; Livingood, John N B

    1948-01-01

    The hottest part of a turbine blade is likely to be the trailing portion. When the blades are cooled and when water is used as the coolant, the cooling passages are placed as close as possible to the trailing edge in order to cool this portion. In some cases, however, the trailing portion of the blade is so narrow, for aerodynamic reasons, that water passages cannot be located very near the trailing edge. Because ceramic coatings offer the possibility of protection for the trailing part of such narrow blades, a theoretical study has been made of the cooling effect of a ceramic coating on: (1) the blade-metal temperature when the gas temperature is unchanged, and (2) the gas temperature when the metal temperature is unchanged. Comparison is also made between the changes in the blade or gas temperatures produced by ceramic coatings and the changes produced by moving the cooling passages nearer the trailing edge. This comparison was made to provide a standard for evaluating the gains obtainable with ceramic coatings as compared to those obtainable by constructing the turbine blade in such a manner that water passages could be located very near the trailing edge.

  13. Superconducting cable cooling system by helium gas at two pressures

    International Nuclear Information System (INIS)

    Dean, J.W.

    1977-01-01

    Thermally contacting, oppositely streaming, cryogenic fluid streams in the same enclosure in a closed cycle changes the fluid from a cool high pressure helium gas to a cooler reduced pressure helium gas in an expander so as to be at different temperature ranges and pressures respectively in go and return legs that are in thermal contact with each other and in thermal contact with a longitudinally extending superconducting transmission line enclosed in the same cable enclosure that insulates the line from the ambient at a temperature T 1 . By first circulating the fluid from a refrigerator at one end of the line as a cool gas at a temperature range T 2 to T 3 in the go leg, then circulating the gas through an expander at the other end of the line where the gas becomes a cooler gas at a reduced pressure and at a reduced temperature T 4 and finally by circulating the cooler gas back again to the refrigerator in a return leg at a temperature range T 4 to T 5 , while in thermal contact with the gas in the go leg, and in the same enclosure therewith for compression into a higher pressure gas at T 2 in a closed cycle, where T 2 greater than T 3 and T 5 greater than T 4 , the fluid leaves the enclosure in the go leg as a gas at its coldest point in the go leg, and the temperature distribution is such that the line temperature decreases along its length from the refrigerator due to the cooling from the gas in the return leg

  14. 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.)

  15. The effects of aging on BWR core isolation cooling systems

    International Nuclear Information System (INIS)

    Lee, B.S.

    1994-10-01

    A study was performed to assess the effects of aging on the Reactor Core Isolation Cooling (RCIC) system in commercial Boiling Water Reactors (BWRs). This study is part of the Nuclear Plant Aging Research (NPAR) program sponsored by the US Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. The failure data from national databases, as well as plant specific data were reviewed and analyzed to understand the effects of aging on the RCIC system. This analysis identified important components that should receive the highest priority in terms of aging management. The aging characterization provided information on the effects of aging on component failure frequency, failure modes, and failures causes. Current inspection, surveillance, and monitoring practices were also reviewed

  16. Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

    Science.gov (United States)

    Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

    2017-12-01

    The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

  17. Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Grodzki Marcin

    2017-12-01

    Full Text Available The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an ‘early design’ variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit. A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

  18. Fuel Development For Gas-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  19. International working group on gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-15

    The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

  20. Periodic reviews of structural integrity of gas-cooled reactors

    International Nuclear Information System (INIS)

    Banks, P.J.; Stokoe, T.Y.; Thomas, D.L.

    1995-01-01

    Nuclear Electric operates 12 gas-cooled reactor power stations which have been in service for between 5 and 30 years. Periodically, comprehensive reviews of the safety cases are carried out for each station. The approach followed in these reviews in respect of structural integrity is outlined with the use of illustrative examples. (author)

  1. Gas-cooled reactor safety and accident analysis

    International Nuclear Information System (INIS)

    1985-12-01

    The Specialists' Meeting on Gas-Cooled Reactor Safety and Accident Analysis was convened by the International Atomic Energy Agency in Oak Ridge on the invitation of the Department of Energy in Washington, USA. The meeting was hosted by the Oak Ridge National Laboratory. The purpose of the meeting was to provide an opportunity to compare and discuss results of safety and accident analysis of gas-cooled reactors under development, construction or in operation, to review their lay-out, design, and their operational performance, and to identify areas in which additional research and development are needed. The meeting emphasized the high safety margins of gas-cooled reactors and gave particular attention to the inherent safety features of small reactor units. The meeting was subdivided into four technical sessions: Safety and Related Experience with Operating Gas-Cooled Reactors (4 papers); Risk and Safety Analysis (11 papers); Accident Analysis (9 papers); Miscellaneous Related Topics (5 papers). A separate abstract was prepared for each of these papers

  2. French gas cooled reactor experience with moisture ingress

    International Nuclear Information System (INIS)

    Bastien, D.; Brie, M.

    1995-01-01

    During the history of operation of six gas cooled reactors in France, some experience has been gained with accidental water ingress into the primary system. This occurred as a result of leaks in steam generators. This paper describes the cause of the leaks, and the resulting consequences. (author). 2 refs, 8 figs

  3. Detection of fuel element failures at the gas cooled reactors

    International Nuclear Information System (INIS)

    Bremsak, P.; Ranzel, S.; Ogrin, R.; Pisanski, C.

    1964-01-01

    This paper describes the system for measuring the concentration of released Kr and Xe in the cooling gas. The system developed in the Jozef Stefan Institute is efficient and suitable for application The method is based on electrostatic collection of daughter elements from radioactive decay of Xe and Kr

  4. IAEA activities in gas-cooled reactor technology development

    International Nuclear Information System (INIS)

    Cleveland, J.; Kupitz, J.

    1992-01-01

    The International Atomic Energy Agency (IAEA) has the charter to ''foster the exchange of scientific and technical information'', and ''encourage and assist research on, and development and practical application of, atomic energy for peaceful uses throughout the world''. This paper describes the Agency's activities in Gas-cooled Reactor (GCR) technology development

  5. Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

    International Nuclear Information System (INIS)

    1988-01-01

    The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics

  6. Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-07-01

    The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics.

  7. Hypothetical air ingress scenarios in advanced modular high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Kroeger, P.G.

    1988-01-01

    Considering an extremely hypothetical scenario of complete cross duct failure and unlimited air supply into the reactor vessel of a modular high temperature gas cooled ractor, it is found that the potential air inflow remains limited due to the high friction pressure drop through the active core. All incoming air will be oxidized to CO and some local external burning would be temporarily possible in such a scenario. The accident would have to continue with unlimited air supply for hundreds of hours before the core structural integrity would be jeopardized

  8. Heat transport and afterheat removal for gas cooled reactors under accident conditions

    International Nuclear Information System (INIS)

    2001-01-01

    The Co-ordinated Research Project (CRP) on Heat Transport and Afterheat Removal for Gas Cooled Reactors Under Accident Conditions was organized within the framework of the International Working Group on Gas Cooled Reactors (IWGGCR). This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs) and supports the conduct of these activities. Advanced GCR designs currently being developed are predicted to achieve a high degree of safety through reliance on inherent safety features. Such design features should permit the technical demonstration of exceptional public protection with significantly reduced emergency planning requirements. For advanced GCRs, this predicted high degree of safety largely derives from the ability of the ceramic coated fuel particles to retain the fission products under normal and accident conditions, the safe neutron physics behaviour of the core, the chemical stability of the core and the ability of the design to dissipate decay heat by natural heat transport mechanisms without reaching excessive temperatures. Prior to licensing and commercial deployment of advanced GCRs, these features must first be demonstrated under experimental conditions representing realistic reactor conditions, and the methods used to predict the performance of the fuel and reactor must be validated against these experimental data. Within this CRP, the participants addressed the inherent mechanisms for removal of decay heat from GCRs under accident conditions. The objective of this CRP was to establish sufficient experimental data at realistic conditions and validated analytical tools to confirm the predicted safe thermal response of advance gas cooled reactors during accidents. The scope includes experimental and analytical investigations of heat transport by natural convection conduction and thermal

  9. Availability analysis of the AP600 passive core cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Syarip, M [National Atomic Energy Research Agency, Yogyakarta (Indonesia); Subki, I R [BATAN Head Office, Jakarta (Indonesia); Canton, M H [Westinghouse Electric Corp. (United States)

    1996-12-01

    The reliability analysis of the AP600 Passive Core Cooling System (PXS) has been done. The fault tree analysis method was used for the quantitative analysis. The PXS can be grouped to several sub-systems i.e.: Reactor Coolant System (RCS) Injection Subsystem, Emergency Core Decay Heat Removal Subsystem, and Containment Sump pH Control Subsystem. The quantitative analysis results indicates that the system unavailability is highly dependent on the valves configuration of the Automatic Depressurization System (ADS). If the ADS valves is arranged in Option-1, the system unavailability is 2.347E-03, this means that the yearly contribution to plant down time can be estimated to be about 20.56 hours per year. Whereas, by using Option-2 of fourth stage ADS valves, the system unavailability is reduced to be 9.877E-04 or 8.65 hours per year and this value is consistent with the allocated goal value (8.0 hours per year). The ADS contributes 66.89% to the system unavailability if it is arranged in Option-1, and will reduced to be about 21.21% if its fourth stages are arranged in Option-2. If the ADS is not included as a subsystem of the PXS (relocate to RCS as a subsystem of RCS), then the PXS unavailability will be reduced to about 7.784E-04 or 6.82 hours per year; this is less then the allocated goal value. The major contributors to the system unavailability are mostly dominated by Stage-4 ADS valves (air piston operated valves and squib valves), inservice testing valves of ADS (solenoid operated valves), solenoid valves of Nitrogen Supply to Accumulator, and Passive Residual Heat Removal actuation valves (air operated valves). It is recommended that those valves be analyzed more detail to gain the improvement in its reliability. It is also recommended that the fourth stage of ADS valves should be arranged according to Option-2, i.e. one 10-inch normally open motor operated gate valve in series with one 10-inch normally closed squib valve. (author). 13 refs, 3 figs, 3 tabs.

  10. Fuel assembly for gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Yellowlees, J.M.

    1976-01-01

    A fuel assembly is described for gas-cooled nuclear reactor which consists of a wrapper tube within which are positioned a number of spaced apart beds in a stack, with each bed containing spherical coated particles of fuel; each of the beds has a perforated top and bottom plate; gaseous coolant passes successively through each of the beds; through each of the beds also passes a bypass tube; part of the gas travels through the bed and part passes through the bypass tube; the gas coolant which passes through both the bed and the bypass tube mixes in the space on the outlet side of the bed before entering the next bed

  11. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Barsell, A.W.

    1980-05-01

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core

  12. Long-term prospects for the gas-cooled reactor

    International Nuclear Information System (INIS)

    Tan, W.P.S.

    1982-01-01

    Towards the second half of a fifty-year time span the market for gas-cooled reactors as sources of high temperature process heat and as highly fuel efficient electricity producers should be reasonably bright, given a fair degree of technological maturity and consequent realisation of inherent economic advantages. Declining fossil resources and increasing prices, initially in oil and gas later in open-cast coal, provide the economic impetus towards substitution of nuclear for coal heat, not only in the generally accepted processes of coal conversion and steel-making but also for oil shale pyrolysis and electrothermal aluminium smelting. Around 2010, if not sooner, the need for uranium conservation should allow the market penetration of breeders and thorium-cycle reactors for which gas cooling has a potential techno-economic edge. (author)

  13. Long-term prospects for the gas-cooled reactor

    International Nuclear Information System (INIS)

    Tan, W.P.S.

    1983-01-01

    Towards the second half of a 50-year time span the market for gas-cooled reactors as sources of high-temperature process heat and as highly fuel-efficient electricity producers should be reasonably bright, given a fair degree of technological maturity and consequent realization of inherent economic advantages. Declining fossil resources and increasing prices, initially in oil and gas, later in open-cast coal, provide the economic impetus towards substitution of nuclear for coal heat, not only in the generally accepted processes of coal conversion and steel making but also for oil shale pyrolysis and electrothermal aluminium smelting. Around 2010, if not sooner, the need for uranium conservation should allow the market penetration of breeders and thorium-cycle reactors for which gas cooling has a potential techno-economic edge. (author)

  14. The dynamic characteristics of HTGR (High Temperature Gas Cooled Reactor) system, (2)

    International Nuclear Information System (INIS)

    Kudo, Kazuhiko; Ohta, Masao; Kawasaki, Hidenori

    1979-01-01

    The dynamic characteristics of a HTGR plant, which has two cooling loops, was investigated. The analytical model consists of the core with fuel sleeves, coolant channels and blocks, the upper and lower reflectors, the high and low temperature plenums, two double wall pipings, two intermediate heat exchangers and the secondary system. The key plant parameters for calculation were as follows: the core outlet gas temperature 1000 deg C, the reactor thermal output 50 MW, the flow rate of primary coolant gas 7.96 kg/sec-loop and the pressure of primary coolant gas 40 kg/cm 2 at the rated operating condition. The calculating parameters were fixed as follows: the time interval for core characteristic analysis 0.1 sec, the time interval for thermal characteristic analysis 5.0 sec, the number of division of fuel channels 130, and the number of division of an intermediate heat exchanger 200. The assumptions for making the model were evaluated especially for the power distribution in the core and the heat transmission coefficients in the core, the double wall piping and the intermediate heat exchangers. Concerning the analytical results, the self-control to the outer disturbance of reactivity and the plant dynamic behavior due to the change of flow rate of primary and secondary coolants, and the change of gas temperature of secondary coolant at the inlet of intermediate heat exchangers, are presented. (Nakai, Y.)

  15. Simulation of Two-Phase Natural Circulation Loop for Core Cather Cooling Using Air Water

    International Nuclear Information System (INIS)

    Revankar, S. T.; Huang, S. F.; Song, K. W.; Rhee, B. W.; Park, R. J.; Song, J. H.

    2012-01-01

    A closed loop natural circulation system employs thermally induced density gradients in single phase or two-phase liquid form to induce circulation of the working fluid thereby obviating the need for any mechanical moving parts such as pumps and pump controls. This increases the reliability and safety of the cooling system and reduces installation, operation and maintenance costs. That is the reason natural circulation cooling has been considered in advanced reactor core cooling and in engineered safety systems. Natural circulation cooling has been proposed to remove reactor decay heat by external vessel cooling for in-vessel core retention during sever accident scenario. Recently in APR1400 reactor core catcher design natural circulation cooling is proposed to stabilize and cool the corium ejected from the reactor vessel following core melt and breach of reactor vessel. The natural circulation flow is similar to external vessel cooling where water flows through an inclined narrow gap below hot surface and is heated to produce boiling. The two-phase natural circulation enables cooling of the corium pool collected on core catcher. Due to importance of this problem this paper focuses simulation of the two-phase natural circulation through inclined gap using air-water system. Scaling criteria for air-water loop are derived that enable simulation of the flow regimes and natural circulation flow rates in such systems using air-water system

  16. Natural convection in closed vertical cylinders with particular reference to gas cooled reactor standpipes

    International Nuclear Information System (INIS)

    Spence, I.D.

    1975-09-01

    The access to the core for fuel assemblies and control rods of the Advanced Gas Cooled Reactor is through the top cap by means of standpipes. The standpipe is essentially a cylindrical, vertical tube with cooled side wall, closed upper end and an orifice at the lower end which is exposed to the hot core fluid. This creates confined natural convection flow in the empty standpipe and this is the subject of this thesis. The investigation is carried out using analytical and experimental methods. For the analytical work, solution of laminar and turbulent flow is attempted using finite-difference computer techniques. The laminar flow performance is evaluated using two different finite-difference procedures, and the results are compared to each other and to existing analytical and experimental results for the open thermosyphon with cool inflow and hot sidewall, i.e. the complementary problem to the present one. For turbulent flow a two equation turbulence model is employed which provides transport equations for the kinetic energy of turbulence and its dissipation rate. The experimental rig is a full scale replica of the Advanced Gas Cooled Reactor control rod mechanism standpipe. Carbon dioxide and helium are used as the working fluids for the series of tests. (author)

  17. Operation method and operation control device for emergency core cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Kinoshita, Shoichiro; Takahashi, Toshiyuki; Fujii, Tadashi [Hitachi Ltd., Tokyo (Japan); Mizutani, Akira

    1996-05-07

    The present invention provides a method of reducing continuous load capacity of an emergency cooling system of a BWR type reactor and a device reducing a rated capacity of an emergency power source facility. Namely, the emergency core cooling system comprises a first cooling system having a plurality of power source systems based on a plurality of emergency power sources and a second cooling system having a remaining heat removing function. In this case, when the first cooling system is operated the manual starting under a predetermined condition that an external power source loss event should occur, a power source division different from the first cooling system shares the operation to operate the secondary cooling system simultaneously. Further, the first cooling system is constituted as a high pressure reactor core water injection system and the second cooling system is constituted as a remaining heat removing system. With such a constitution, a high pressure reactor core water injection system for manual starting and a remaining heat removing system of different power source division can be operated simultaneously before automatic operation of the emergency core cooling system upon loss of external power source of a nuclear power plant. (I.S.)

  18. Fundamental conceptual design of the experimental multi-purpose high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shimokawa, Junichi; Yasuno, Takehiko; Yasukawa, Shigeru; Mitake, Susumu; Miyamoto, Yoshiaki

    1975-06-01

    The fundamental conceptual design of the experimental multi-purpose very high-temperature gas-cooled reactor (experimental VHTR of thermal output 50 MW with reactor outlet-gas temperature 1,000 0 C) has been carried out to provide the operation modes of the system consisting of the reactor and the heat-utilization system, including characteristics and performance of the components and safety of the plant system. For the heat-utilization system of the plant, heat distribution, temperature condition, cooling system constitution, and the containment facility are specified. For the operation of plant, testing capability of the reactor and controlability of the system are taken into consideration. Detail design is made of the fuel element, reactor core, reactivity control and pressure vessel, and also the heat exchanger, steam reformer, steam generator, helium circulator, helium-gas turbine, and helium-gas purification, fuel handling, and engineered safety systems. Emphasis is placed on providing the increase of the reactor outlet-gas temperature. Fuel element design is directed to the prismatic graphite blocks of hexagonal cross-section accommodating the hollow or tubular fuel pins sheathed in graphite sleeve. The reactor core is composed of 73 fuel columns in 7 stages, concerning the reference design MK-II. Orificing is made in the upper portion of core; one orifice for every 7 fuel columns. Average core power density is 2.5 watts/cm 3 . Fuel temperature is kept below 1,300 0 C in rated power. The main components, i.e. pressure vessel, reformer, gas turbine and intermediate heat exchanger are designed in detail; the IHX is of a double-shell and helically-wound tube coils, the reformer is of a byonet tube type, and the turbine-compressor unit is of an axial flow type (turbine in 6 stages and compressor in 16 stages). (auth.)

  19. THE GROWTH OF COOL CORES AND EVOLUTION OF COOLING PROPERTIES IN A SAMPLE OF 83 GALAXY CLUSTERS AT 0.3 < z < 1.2 SELECTED FROM THE SPT-SZ SURVEY

    Energy Technology Data Exchange (ETDEWEB)

    McDonald, M.; Bautz, M. W. [Kavli Institute for Astrophysics and Space Research, Massachusetts Institute of Technology, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States); Benson, B. A.; Bleem, L. E.; Carlstrom, J. E.; Chang, C. L.; Crawford, T. M.; Crites, A. T. [Kavli Institute for Cosmological Physics, University of Chicago, 5640 South Ellis Avenue, Chicago, IL 60637 (United States); Vikhlinin, A.; Stalder, B.; Ashby, M. L. N.; Bayliss, M. [Harvard-Smithsonian Center for Astrophysics, 60 Garden Street, Cambridge, MA 02138 (United States); De Haan, T. [Department of Physics, McGill University, 3600 Rue University, Montreal, Quebec H3A 2T8 (Canada); Lin, H. W. [Caddo Parish Magnet High School, Shrevport, LA 71101 (United States); Aird, K. A. [University of Chicago, 5640 South Ellis Avenue, Chicago, IL 60637 (United States); Bocquet, S.; Desai, S. [Department of Physics, Ludwig-Maximilians-Universitaet, Scheinerstr. 1, D-81679 Muenchen (Germany); Brodwin, M. [Department of Physics and Astronomy, University of Missouri, 5110 Rockhill Road, Kansas City, MO 64110 (United States); Cho, H. M. [NIST Quantum Devices Group, 325 Broadway Mailcode 817.03, Boulder, CO 80305 (United States); Clocchiatti, A., E-mail: mcdonald@space.mit.edu [Departamento de Astronomia y Astrosifica, Pontificia Universidad Catolica (Chile); and others

    2013-09-01

    We present first results on the cooling properties derived from Chandra X-ray observations of 83 high-redshift (0.3 < z < 1.2) massive galaxy clusters selected by their Sunyaev-Zel'dovich signature in the South Pole Telescope data. We measure each cluster's central cooling time, central entropy, and mass deposition rate, and compare these properties to those for local cluster samples. We find no significant evolution from z {approx} 0 to z {approx} 1 in the distribution of these properties, suggesting that cooling in cluster cores is stable over long periods of time. We also find that the average cool core entropy profile in the inner {approx}100 kpc has not changed dramatically since z {approx} 1, implying that feedback must be providing nearly constant energy injection to maintain the observed ''entropy floor'' at {approx}10 keV cm{sup 2}. While the cooling properties appear roughly constant over long periods of time, we observe strong evolution in the gas density profile, with the normalized central density ({rho}{sub g,0}/{rho}{sub crit}) increasing by an order of magnitude from z {approx} 1 to z {approx} 0. When using metrics defined by the inner surface brightness profile of clusters, we find an apparent lack of classical, cuspy, cool-core clusters at z > 0.75, consistent with earlier reports for clusters at z > 0.5 using similar definitions. Our measurements indicate that cool cores have been steadily growing over the 8 Gyr spanned by our sample, consistent with a constant, {approx}150 M{sub Sun} yr{sup -1} cooling flow that is unable to cool below entropies of 10 keV cm{sup 2} and, instead, accumulates in the cluster center. We estimate that cool cores began to assemble in these massive systems at z{sub cool}=1.0{sup +1.0}{sub -0.2}, which represents the first constraints on the onset of cooling in galaxy cluster cores. At high redshift (z {approx}> 0.75), galaxy clusters may be classified as ''cooling flows

  20. Core debris cooling with flooded vessel or core-catcher. Heat exchange coefficients under natural convection

    International Nuclear Information System (INIS)

    Rouge, S.; Seiler, J.M.

    1994-09-01

    External cooling by natural water circulation is necessary for molten core retention in LWR lower head or in a core-catcher. Considering the expected heat flux levels (between 0.2 to 1.5 MW/m 2 ) film boiling should be avoided. This rises the question of the knowledge of the level of the critical heat flux for the considered geometries and flow paths. The document proposes a state of the art of the research in this field. Mainly small scale experiments have been performed in a very recent past. These experiments are not sufficient to extrapolate to large scale reactor structures. Limited large scale experimental results exist. These results together with some theoretical investigations show that external cooling by natural water circulation may be considered as a reasonable objective of severe accident R and D. Recently (in fact since the beginning of 1994) new results are available from large scale experiments (CYBL, ULPU 2000, SULTAN). These results indicate that CHF larger than 1 MW/m 2 can be obtained under natural water circulation conditions. In this report, emphasis is given to the pursuit of finding predictive models for the critical heat flux in large, naturally convective channels with thick walls. This theoretical understanding is important for the capability to extrapolate to different situations (various geometries, flow paths....). The outcome of this research should be the ability to calculate Boundary Layer Boiling situations (2D), channelling boiling situations (1D) and related CHF conditions. However, a more straightforward approach can be used for the analysis of specific designs. Today there are already some CHF data available for hemispherical geometry and these data can be used before a mechanistic understanding is achieved

  1. Gas-Cooled Reactors: the importance of their development

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1978-01-01

    Gas-Cooled Reactors are considered to have a significant future impact on the application of fission energy. The specific types are the steam-cycle High-Temperature Gas-Cooled Reactor, the Gas-Cooled Fast Breeder Reactor, the gas-turbine HTGR, and the Very High-Temperature Process Heat Reactor. The importance of developing the above systems is discussed relative to alternative fission power systems involving Light Water Reactors, Heavy Water Reactors, Spectral Shift Controlled Reactors, and Liquid-Metal-Cooled Fast Breeder Reactors. A primary advantage of developing GCRs as a class lies in the technology and cost interrelations, permitting cost-effective development of systems having diverse applications. Further, HTGR-type systems have highly proliferation-resistant characteristics and very attractive safety features. Finally, such systems and GCFRs are mutally complementary. Overall, GCRs provide interrelated systems that serve different purposes and needs; their development can proceed in stages that provide early benefits while contributing to future needs. It is concluded that the long-term importance of the various GCRs is as follows: HTGR, providing a technology for economic GCFRs and HTGR-GTs, while providing a proliferation-resistant reactor system having early economic and fuel utilization benefits; GCFR, providing relatively low cost fissile fuel and reducing overall separative work needs at capital costs lower than those for LMFBRs; HTGR-GT (in combination with a bottoming cycle), providing a very high thermal efficiency system having low capital costs and improved fuel utilization and technology pertinent to VHTRs; HTGR-GT, providing a power system well suited for dry cooling conditions for low-temperature process heat needs; and VHTR, providing a high-temperature heat source for hydrogen production processes

  2. High-temperature gas-cooled reactors and process heat

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1980-01-01

    High-Temperature Gas-Cooled Reactors (HTGRs) are fueled with ceramic-coated microspheres of uranium and thorium oxides/carbides embedded in graphite blocks which are cooled with helium. Promising areas of HTGR application are in cogeneration, energy transport using Heat Transfer Salt, recovery of oils from oil shale, steam reforming of methane for chemical production, coal gasification, and in energy transfer using chemical heat jpipes in the long term. Further, HTGRs could be used as the energy source for hydrogen production through thermochemical water splitting in the long term. The potential market for Process Heat HTGRs is 100-200 large units by about the year 2020

  3. Absorber rod driving into a gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Elter, C.; Schmitt, H.; Schoening, J.

    1987-01-01

    The absorber rod consists of a hollow cylinder which has a layer of absorber material applied on its inside circumferential surface. The absorber rod is held via a guide sleeve, which is supported centrally in a hole in the side reflector. The guidance within the sleeve is provided by flanges on the hollow cylinder. The movement of the hollow cylinder is carried out hydraulically or pneumatically. A flow of cooling gas is used for cooling, which is passed through the inner central areas of the hollow cylinder and the guide sleeve. (DG) [de

  4. Analysis of a sustainable gas cooled fast breeder reactor concept

    International Nuclear Information System (INIS)

    Kumar, Akansha; Chirayath, Sunil S.; Tsvetkov, Pavel V.

    2014-01-01

    Highlights: • A Thorium-GFBR breeder for actinide recycling ability, and thorium fuel feasibility. • A mixture of 232 Th and 233 U is used as fuel and LWR used fuel is used. • Detailed neutronics, fuel cycle, and thermal-hydraulics analysis has been presented. • Run this TGFBR for 20 years with breeding of 239 Pu and 233 U. • Neutronics analysis using MCNP and Brayton cycle for energy conversion are used. - Abstract: Analysis of a thorium fuelled gas cooled fast breeder reactor (TGFBR) concept has been done to demonstrate the self-sustainability, breeding capability, actinide recycling ability, and thorium fuel feasibility. Simultaneous use of 232 Th and used fuel from light water reactor in the core has been considered. Results obtained confirm the core neutron spectrum dominates in an intermediate energy range (peak at 100 keV) similar to that seen in a fast breeder reactor. The conceptual design achieves a breeding ratio of 1.034 and an average fuel burnup of 74.5 (GWd)/(MTHM) . TGFBR concept is to address the eventual shortage of 235 U and nuclear waste management issues. A mixture of thorium and uranium ( 232 Th + 233 U) is used as fuel and light water reactor used fuel is utilized as blanket, for the breeding of 239 Pu. Initial feed of 233 U has to be obtained from thorium based reactors; even though there are no thorium breeders to breed 233 U a theoretical evaluation has been used to derive the data for the source of 233 U. Reactor calculations have been performed with Monte Carlo radiation transport code, MCNP/MCNPX. It is determined that this reactor has to be fuelled once every 5 years assuming the design thermal power output as 445 MW. Detailed analysis of control rod worth has been performed and different reactivity coefficients have been evaluated as part of the safety analysis. The TGFBR concept demonstrates the sustainability of thorium, viability of 233 U as an alternate to 235 U and an alternate use for light water reactor used fuel as a

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

    International Nuclear Information System (INIS)

    Kiefhaber, E.

    1978-10-01

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

  6. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

  7. Empirical evaluation of the radiative cooling coefficient for krypton gas in the FTU plasma

    International Nuclear Information System (INIS)

    Fournier, K.B.; Pacella, D.; Mazzitelli, G.; Stutman, D.; Soukanovskii, V.; Goldstein, W.H.

    1997-01-01

    For future fusion reactors, a careful balance must be achieved between the cooling of the outer plasma via impurity radiation and the deleterious effects of inevitable core penetration by impurity ions. We have injected krypton gas into the Frascati Tokamak Upgrade (FTU) plasma. The measured visible bremsstrahlung and bolometric signals from krypton have been inverted and the resulting radial impurity density profile and power loss profile for krypton gas are extracted. Using the measured electron density and temperature profiles, the radiative cooling coefficient for krypton is derived. The level of intrinsic impurities (Mo, Cr, Mn and Fe) in the plasma during the krypton puffing is monitored with a VUV SPRED spectrometer. Models for krypton emissivity from the literature are compared to our measured results. 7 figs

  8. Numerical investigation of heat transfer in high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chen, g.; Anghaie, S. [Univ. of Florida, Gainesville, FL (United States)

    1995-09-01

    This paper proposes a computational model for analysis of flow and heat transfer in high-temperature gas-cooled reactors. The formulation of the problem is based on using the axisymmetric, thin layer Navier-Stokes equations. A hybrid implicit-explicit method based on finite volume approach is used to numerically solve the governing equations. A fast converging scheme is developed to accelerate the Gauss-Siedel iterative method for problems involving the wall heat flux boundary condition. Several cases are simulated and results of temperature and pressure distribution in the core are presented. Results of a parametric analysis for the assessment of the impact of power density on the convective heat transfer rate and wall temperature are discussed. A comparative analysis is conducted to identify the Nusselt number correlation that best fits the physical conditions of the high-temperature gas-cooled reactors.

  9. Experimental investigation of temperature rise in bone drilling with cooling: A comparison between modes of without cooling, internal gas cooling, and external liquid cooling.

    Science.gov (United States)

    Shakouri, Ehsan; Haghighi Hassanalideh, Hossein; Gholampour, Seifollah

    2018-01-01

    Bone fracture occurs due to accident, aging, and disease. For the treatment of bone fractures, it is essential that the bones are kept fixed in the right place. In complex fractures, internal fixation or external methods are used to fix the fracture position. In order to immobilize the fracture position and connect the holder equipment to it, bone drilling is required. During the drilling of the bone, the required forces to chip formation could cause an increase in the temperature. If the resulting temperature increases to 47 °C, it causes thermal necrosis of the bone. Thermal necrosis decreases bone strength in the hole and, subsequently, due to incomplete immobilization of bone, fracture repair is not performed correctly. In this study, attempts have been made to compare local temperature increases in different processes of bone drilling. This comparison has been done between drilling without cooling, drilling with gas cooling, and liquid cooling on bovine femur. Drilling tests with gas coolant using direct injection of CO 2 and N 2 gases were carried out by internal coolant drill bit. The results showed that with the use of gas coolant, the elevation of temperature has limited to 6 °C and the thermal necrosis is prevented. Maximum temperature rise reached in drilling without cooling was 56 °C, using gas and liquid coolant, a maximum temperature elevation of 43 °C and 42 °C have been obtained, respectively. This resulted in decreased possibility of thermal necrosis of bone in drilling with gas and liquid cooling. However, the results showed that the values obtained with the drilling method with direct gas cooling are independent of the rotational speed of drill.

  10. Ambient air cooling arrangement having a pre-swirler for gas turbine engine blade cooling

    Science.gov (United States)

    Lee, Ching-Pang; Tham, Kok-Mun; Schroeder, Eric; Meeroff, Jamie; Miller, Jr., Samuel R; Marra, John J

    2015-01-06

    A gas turbine engine including: an ambient-air cooling circuit (10) having a cooling channel (26) disposed in a turbine blade (22) and in fluid communication with a source (12) of ambient air: and an pre-swirler (18), the pre-swirler having: an inner shroud (38); an outer shroud (56); and a plurality of guide vanes (42), each spanning from the inner shroud to the outer shroud. Circumferentially adjacent guide vanes (46, 48) define respective nozzles (44) there between. Forces created by a rotation of the turbine blade motivate ambient air through the cooling circuit. The pre-swirler is configured to impart swirl to ambient air drawn through the nozzles and to direct the swirled ambient air toward a base of the turbine blade. The end walls (50, 54) of the pre-swirler may be contoured.

  11. Contributions to the neutronic analysis of a gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Martin-del-Campo, Cecilia; Reyes-Ramirez, Ricardo; Francois, Juan-Luis; Reinking-Cejudo, Arturo G.

    2011-01-01

    Highlights: → Differences on reactivity with MCNPX and TRIPOLI-4 are negligible. → Fuel lattice and core criticality calculations were done. → A higher Doppler coefficient than coolant density coefficient. → Zirconium carbide is a better reflector than silicon carbide. → Adequate active height, radial size and reflector thickness were obtained. - Abstract: In this work the Monte Carlo codes MCNPX and TRIPOLI-4 were used to perform the criticality calculations of the fuel assembly and the core configuration of a gas-cooled fast reactor (GFR) concept, currently in development. The objective is to make contributions to the neutronic analysis of a gas-cooled fast reactor. In this study the fuel assembly is based on a hexagonal lattice of fuel-pins. The materials used are uranium and plutonium carbide as fuel, silicon carbide as cladding, and helium gas as coolant. Criticality calculations were done for a fuel assembly where the axial reflector thickness was varied in order to find the optimal thickness. In order to determine the best material to be used as a reflector, in the reactor core with neutrons of high energy spectrum, criticality calculations were done for three reflector materials: zirconium carbide, silicon carbide and natural uranium. It was found that the zirconium carbide provides the best neutron reflection. Criticality calculations using different active heights were done to determine the optimal height, and the reflector thickness was adjusted. Core criticality calculations were performed with different radius sizes to determine the active radial dimension of the core. A negative temperature coefficient of reactivity was verified for the fuel. The effect on reactivity produced by changes in the coolant density was also evaluated. We present the main neutronic characteristics of a preliminary fuel and core designs for the GFR concept. ENDF-VI cross-sections libraries were used in both the MCNPX and TRIPOLI-4 codes, and we verified that the

  12. Power Conversion Study for High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Chang Oh; Richard Moore; Robert Barner

    2005-01-01

    The Idaho National Laboratory (INL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. There are some technical issues to be resolved before the selection of the final design of the high temperature gas cooled reactor, called as a Next Generation Nuclear Plant (NGNP), which is supposed to be built at the INEEL by year 2017. The technical issues are the selection of the working fluid, direct vs. indirect cycle, power cycle type, the optimized design in terms of a number of intercoolers, and others. In this paper, we investigated a number of working fluids for the power conversion loop, direct versus indirect cycle, the effect of intercoolers, and other thermal hydraulics issues. However, in this paper, we present part of the results we have obtained. HYSYS computer code was used along with a computer model developed using Visual Basic computer language

  13. Optimal design of gas adsorption refrigerators for cryogenic cooling

    Science.gov (United States)

    Chan, C. K.

    1983-01-01

    The design of gas adsorption refrigerators used for cryogenic cooling in the temperature range of 4K to 120K was examined. The functional relationships among the power requirement for the refrigerator, the system mass, the cycle time and the operating conditions were derived. It was found that the precool temperature, the temperature dependent heat capacities and thermal conductivities, and pressure and temperature variations in the compressors have important impacts on the cooling performance. Optimal designs based on a minimum power criterion were performed for four different gas adsorption refrigerators and a multistage system. It is concluded that the estimates of the power required and the system mass are within manageable limits in various spacecraft environments.

  14. Gas turbine cooling modeling - Thermodynamic analysis and cycle simulations

    Energy Technology Data Exchange (ETDEWEB)

    Jordal, Kristin

    1999-02-01

    Considering that blade and vane cooling are a vital point in the studies of modern gas turbines, there are many ways to include cooling in gas turbine models. Thermodynamic methods for doing this are reviewed in this report, and, based on some of these methods, a number of model requirements are set up and a Cooled Gas Turbine Model (CGTM) for design-point calculations of cooled gas turbines is established. Thereafter, it is shown that it is possible to simulate existing gas turbines with the CGTM. Knowledge of at least one temperature in the hot part of the turbine (TET, TRIT or possibly TIT) is found to be vital for a complete heat balance over the turbine. The losses, which are caused by the mixing of coolant and main flow, are in the CGTM considered through a polytropic efficiency reduction factor S. Through the study of S, it can be demonstrated that there is more to gain from coolant reduction in a small and/or old turbine with poor aerodynamics, than there is to gain in a large, modern turbine, where the losses due to interaction between coolant and main flow are, relatively speaking, small. It is demonstrated, at the design point (TET=1360 deg C, {pi}=20) for the simple-cycle gas turbine, that heat exchanging between coolant and fuel proves to have a large positive impact on cycle efficiency, with an increase of 0.9 percentage points if all of the coolant passes through the heat exchanger. The corresponding improvement for humidified coolant is 0.8 percentage points. A design-point study for the HAT cycle shows that if all of the coolant is extracted after the humidification tower, there is a decrease in coolant requirements of 7.16 percentage points, from 19.58% to 12.52% of the compressed air, and an increase in thermal efficiency of 0.46 percentage points, from 53.46% to 53.92%. Furthermore, it is demonstrated with a TET-parameter variation, that the cooling of a simple-cycle gas turbine with humid air can have a positive effect on thermal efficiency

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

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

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

    International Nuclear Information System (INIS)

    1997-11-01

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

  17. Hypothetical accident scenario analyses for a 250-MW(t) modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

    1985-11-01

    This paper describes calculations performed to characterize the inherent safety of a 250-MW(t), 100-MW(e), pebble bed modular high temperature gas-cooled reactor (HTGR) design with vertical in-line arrangement (i.e., upflow core with steam generators directly above the core). A variety of postulated accident sequences involving combinations of loss of forced primary coolant (helium) circulation, loss of primary coolant pressurization, and loss of heat sink were studied and were discussed

  18. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  19. Multidisciplinary design optimization of film-cooled gas turbine blades

    OpenAIRE

    Shashishekara S. Talya; J. N. Rajadas; A. Chattopadhyay

    1999-01-01

    Design optimization of a gas turbine blade geometry for effective film cooling toreduce the blade temperature has been done using a multiobjective optimization formulation. Three optimization formulations have been used. In the first, the average blade temperature is chosen as the objective function to be minimized. An upper bound constraint has been imposed on the maximum blade temperature. In the second, the maximum blade temperature is chosen as the objective function to be minimized with ...

  20. The status of graphite development for gas cooled reactors

    International Nuclear Information System (INIS)

    1993-02-01

    The meeting was convened by the IAEA on the recommendation of the International Working Group on Gas Cooled Reactors. It was attended by 61 participants from 6 countries. The meeting covered the following subjects: overview of national programs; design criteria, fracture mechanisms and component test; materials development and properties; non-destructive examination, inspection and surveillance. The participants presented 33 papers on behalf of their countries. A separate abstract was prepared for each of these papers. Refs, figs, tabs, photos and diagrams

  1. The Formation and Physical Origin of Highly Ionized Cooling Gas

    Energy Technology Data Exchange (ETDEWEB)

    Bordoloi, Rongmon [MIT-Kavli Center for Astrophysics and Space Research, 77 Massachusetts Avenue, Cambridge, MA, 02139 (United States); Wagner, Alexander Y. [University of Tsukuba, Center for Computational Sciences, Tennodai 1-1-1, Tsukuba, Ibaraki (Japan); Heckman, Timothy M.; Norman, Colin A., E-mail: bordoloi@mit.edu, E-mail: bordoloi@mit.edu [Department of Physics and Astronomy, John Hopkins University, 21218, Baltimore, MD (United States)

    2017-10-20

    We present a simple model that explains the origin of warm, diffuse gas seen primarily as highly ionized absorption-line systems in the spectra of background sources. We predict the observed column densities of several highly ionized transitions such as O vi, O vii, Ne viii, N v, and Mg x, and we present a unified comparison of the model predictions with absorption lines seen in the Milky Way disk, Milky Way halo, starburst galaxies, the circumgalactic medium, and the intergalactic medium at low and high redshifts. We show that diffuse gas seen in such diverse environments can be simultaneously explained by a simple model of radiatively cooling gas. We show that most such absorption-line systems are consistent with being collisionally ionized, and we estimate the maximum-likelihood temperature of the gas in each observation. This model satisfactorily explains why O vi is regularly observed around star-forming low- z L* galaxies, and why N v is rarely seen around the same galaxies. We further present some consequences of this model in quantifying the dynamics of the cooling gas around galaxies and predict the shock velocities associated with such flows. A unique strength of this model is that while it has only one free (but physically well-constrained) parameter, it nevertheless successfully reproduces the available data on O vi absorbers in the interstellar, circumgalactic, intragroup, and intergalactic media, as well as the available data on other absorption lines from highly ionized species.

  2. Risk Based Inspection of Gas-Cooling Heat Exchanger

    Directory of Open Access Journals (Sweden)

    Dwi Priyanta

    2017-09-01

    Full Text Available On October 2013, Pertamina Hulu Energi Offshore North West Java (PHE – ONWJ platform personnel found 93 leaking tubes locations in the finfan coolers/ gas-cooling heat exchanger. After analysis had been performed, the crack in the tube strongly indicate that stress corrosion cracking was occurred by chloride. Chloride stress corrosion cracking (CLSCC is the cracking occurred by the combined influence of tensile stress and a corrosive environment. CLSCC is the one of the most common reasons why austenitic stainless steel pipework or tube and vessels deteriorate in the chemical processing, petrochemical industries and maritime industries. In this thesis purpose to determine the appropriate inspection planning for two main items (tubes and header box in the gas-cooling heat exchanger using risk based inspection (RBI method. The result, inspection of the tubes must be performed on July 6, 2024 and for the header box inspection must be performed on July 6, 2025. In the end, RBI method can be applicated to gas-cooling heat exchanger. Because, risk on the tubes can be reduced from 4.537 m2/year to 0.453 m2/year. And inspection planning for header box can be reduced from 4.528 m2/year to 0.563 m2/year.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  4. Core-in-shell sorbent for hot coal gas desulfurization

    Science.gov (United States)

    Wheelock, Thomas D.; Akiti, Jr., Tetteh T.

    2004-02-10

    A core-in-shell sorbent is described herein. The core is reactive to the compounds of interest, and is preferably calcium-based, such as limestone for hot gas desulfurization. The shell is a porous protective layer, preferably inert, which allows the reactive core to remove the desired compounds while maintaining the desired physical characteristics to withstand the conditions of use.

  5. Natural convection cooling of LEU cores for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Khan, L.A.; Bokhari, I.H.; Akhtar, K.M.

    1991-08-01

    The first high power and equilibrium LEU cores of PARR-1 have been analysed to assess the maximum operating power based on natural convection cooling, need for forced cooling to remove the decay heat and to estimate safety margins that commensurate with the predetermined power limit. Computer code NATCON and standard correlations have been used for the analysis. The parameters studied includes coolant velocity, temperature distribution in the core, heat fluxes at onset of nucleate boiling, pulsed boiling and burnup. (author)

  6. Evaluation of tritium production rate in a gas-cooled reactor with continuous tritium recovery system for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matsuura, Hideaki, E-mail: mat@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Nakaya, Hiroyuki; Nakao, Yasuyuki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki 311-1393 (Japan); Nishikawa, Masabumi [Graduate School of Engineering Science, Kyushu University, 6-10-1 Hakozaki, Fukuoka 812-8581 (Japan)

    2013-10-15

    Highlights: • The performance of a gas-cooled reactor as a tritium production system was studied. • A continuous tritium recovery using helium gas was considered. • Gas-cooled reactors with 3 GW output in all can produce ∼6 kg of tritium in a year • Performance of the system was examined for Li{sub 4}SiO{sub 4}, Li{sub 2}TiO{sub 3} and LiAlO{sub 2} compounds. -- Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600 MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3 GW in all can produce ∼6 kg of tritium maximum in a year.

  7. Spitzer mid-infrared spectra of cool-core galaxy clusters

    NARCIS (Netherlands)

    de Messières, G.E.; O'Connell, R.W.; McNamara, B.R.; Donahue, M.; Nulsen, P.E.J.; Voit, G.M.; Wise, M.W.; Smith, B.; Higdon, J.; Higdon, S.; Bastian, N.

    2010-01-01

    We have obtained mid-infrared spectra of nine cool-core galaxy clusters with the Infrared Spectrograph aboard the Spitzer Space Telescope. X-ray, ultraviolet and optical observations have demonstrated that each of these clusters hosts a cooling flow which seems to be fueling vigorous star formation

  8. Microscopical examination of carbon deposits formed in the Windscale advanced gas cooled reactor

    International Nuclear Information System (INIS)

    Livesey, D.J.; Chatwin, W.H.; Pearce, J.H.

    1980-12-01

    Methods are described of sampling and examining carbon deposits on fuel cladding in the Windscale advanced gas-cooled reactor. Deposition is observed on fuel cladding in both the reactor core and experimental loops in carbon dioxide coolants containing various amounts of carbon monoxide and methane. Deposit distribution over the cladding surface indicated that nucleation is dependent on local surface conditions. Microscopical examination showed that deposit thickness increases by carbon filament growth into the coolant gas stream and that the process can be markedly influenced by metallic impurities. There is evidence that nickel can play a particularly significant role in deposition in loop experiments but similar effects have not been observed in the reactor core. (author)

  9. Gas Hydrate-Sediment Morphologies Revealed by Pressure Core Analysis

    Science.gov (United States)

    Holland, M.; Schultheiss, P.; Roberts, J.; Druce, M.

    2006-12-01

    Analysis of HYACINTH pressure cores collected on IODP Expedition 311 and NGHP Expedition 1 showed gas hydrate layers, lenses, and veins contained in fine-grained sediments as well as gas hydrate contained in coarse-grained layers. Pressure cores were recovered from sediments on the Cascadia Margin off the North American West Coast and in the Krishna-Godavari Basin in the Western Bay of Bengal in water depths of 800- 1400 meters. Recovered cores were transferred to laboratory chambers without loss of pressure and nondestructive measurements were made at in situ pressures and controlled temperatures. Gamma density, P-wave velocity, and X-ray images showed evidence of grain-displacing and pore-filling gas hydrate in the cores. Data highlights include X-ray images of fine-grained sediment cores showing wispy subvertical veins of gas hydrate and P-wave velocity excursions corresponding to grain-displacing layers and pore-filling layers of gas hydrate. Most cores were subjected to controlled depressurization experiments, where expelled gas was collected, analyzed for composition, and used to calculate gas hydrate saturation within the core. Selected cores were stored under pressure for postcruise analysis and subsampling.

  10. Nuclear design for high temperature gas cooled reactor (GTHTR300C) using MOX fuel

    International Nuclear Information System (INIS)

    Mouri, Tomoaki; Kunitomi, Kazuhiko

    2008-01-01

    A design study of the hydrogen cogeneration high temperature gas cooled reactor (GTHTR300C) that can produce both electricity and hydrogen has been carried out in Japan Atomic Energy Agency. The GTHTR300C is the system with thermal power of 600MW and reactor outlet temperature of 950degC, which is expected to supply the hydrogen to fuel cell vehicles after 2020s. In future, the full deployment of fast reactor cycle without natural uranium will demand the use of Mixed-Oxide (MOX) fuels in the GTHTR300C. Therefore, a nuclear design was performed to confirm the feasibility of the reactor core using MOX fuels. The designed reactor core has high performance and meets safety requirements. In this paper, the outline of the GTHTR300C and the nuclear design of the reactor core using MOX fuels are described. (author)

  11. Analysis Of The Heat Exchanger Capability At One Line Cooling System Operation Mode Of The RSG-GAS

    International Nuclear Information System (INIS)

    Dibyo, Sukmanto; Kuntoro, Iman

    2000-01-01

    In the frame of minimizing the operation lost of the RSG-GAS reactor, operation using one line cooling system at certain power range is being evaluated. Analysis the performance of cooling system for determining maximum power should be carried out. Analysis was carried out based on heat exchanger calculation using actual operation data. Constraints imposed to the analysis are that inlet cooling system to the reactor core shall be less than 42 o C. The result shows that by using one line of primary and secondary coolant flow of 1780 m exp. 3/hr and 2000 m 3 /hr and secondary coolant temperature from the cooling tower of 38 o C, the primary coolant to the core will be reach 42 o C if reactor operated at power of 16 MW

  12. Design of a spherical fuel element for a gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Van Rooijen, W.F.G.; Kloosterman, J.L.; Van Dam, H.; Van der Hagen, T.H.J.J.

    2004-01-01

    A study is undertaken to develop a fuel cycle for a gas-cooled fast reactor (GCFR). The design goals are: highly efficient use of (depleted) uranium, application of Pu recycled from LWR discharge as fissile material, high temperature output and simplicity of design. The design focuses on spherical TRISO-like fuel elements, a homogeneous core at start-up, providing for easy fuel fabrication, and self-breeding capability with a flat k eff with burn-up. Nitride fuel ( 15 N > 99%) has been selected because of its favourable thermal conductivity, high heavy metal density and compatibility with PUREX reprocessing. Two core concepts have been studied: one with coated particles embedded inside fuel pebbles, and one with coated particles cooled directly by helium. The result is that a flat k eff can be achieved for a long period of time, using coated particles cooled directly, with a homogeneous core at, start-up, with a closed fuel cycle and a simple refuelling and reprocessing scheme. (author)

  13. Gas Hydrate Investigations Using Pressure Core Analysis: Current Practice

    Science.gov (United States)

    Schultheiss, P.; Holland, M.; Roberts, J.; Druce, M.

    2006-12-01

    Recently there have been a number of major gas hydrate expeditions, both academic and commercially oriented, that have benefited from advances in the practice of pressure coring and pressure core analysis, especially using the HYACINTH pressure coring systems. We report on the now mature process of pressure core acquisition, pressure core handling and pressure core analysis and the results from the analysis of pressure cores, which have revealed important in situ properties along with some remarkable views of gas hydrate morphologies. Pressure coring success rates have improved as the tools have been modified and adapted for use on different drilling platforms. To ensure that pressure cores remain within the hydrate stability zone, tool deployment, recovery and on-deck handling procedures now mitigate against unwanted temperature rises. Core analysis has been integrated into the core transfer protocol and automated nondestructive measurements, including P-wave velocity, gamma density, and X-ray imaging, are routinely made on cores. Pressure cores can be subjected to controlled depressurization experiments while nondestructive measurements are being made, or cores can be stored at in situ conditions for further analysis and subsampling.

  14. Heat transfer problems in gas-cooled solid blankets

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.

    1976-01-01

    In all fusion reactors using the deuterium-tritium fuel cycle, a large fraction approximately 80 percent of the fusion energy will be released as approximately 14 MeV neutrons which must be slowed down in a relatively thick blanket surrounding the plasma, thereby, converting their kinetic energy to high temperature heat which can be continuously removed by a coolant stream and converted in part to electricity in a conventional power turbine. Because of the primary goal of achieving minimum radioactivity, to date Brookhaven blanket concepts have been restricted to the use of some form of solid lithium, with inert gas-cooling and in some design cases, water-cooling of the shell structure. Aluminum and graphite have been identified as very promising structural materials for fusion blankets, and conceptual designs based on these materials have been made. Depending on the thermal loading on the ''first'' wall which surrounds the plasma as well as blanket design, heat transfer problems may be noticeably different in gas-cooled solid blankets. Approaches to solution of heat removal problems as well as explanation of: (a) the after-heat problems in blankets; (b) tritium breeding in solids; and (c) materials selection for radiation shields relative to the minimum activity blanket efforts at Brookhaven are discussed

  15. Verification of Thermal Models of Internally Cooled Gas Turbine Blades

    Directory of Open Access Journals (Sweden)

    Igor Shevchenko

    2018-01-01

    Full Text Available Numerical simulation of temperature field of cooled turbine blades is a required element of gas turbine engine design process. The verification is usually performed on the basis of results of test of full-size blade prototype on a gas-dynamic test bench. A method of calorimetric measurement in a molten metal thermostat for verification of a thermal model of cooled blade is proposed in this paper. The method allows obtaining local values of heat flux in each point of blade surface within a single experiment. The error of determination of local heat transfer coefficients using this method does not exceed 8% for blades with radial channels. An important feature of the method is that the heat load remains unchanged during the experiment and the blade outer surface temperature equals zinc melting point. The verification of thermal-hydraulic model of high-pressure turbine blade with cooling allowing asymmetrical heat removal from pressure and suction sides was carried out using the developed method. An analysis of heat transfer coefficients confirmed the high level of heat transfer in the leading edge, whose value is comparable with jet impingement heat transfer. The maximum of the heat transfer coefficients is shifted from the critical point of the leading edge to the pressure side.

  16. Review of RSG-GAS secondary cooling pump performance

    International Nuclear Information System (INIS)

    Marsahala, Y.B.

    1999-01-01

    The control system of RSG-GAS secondary pump is the study for the operation existence of RSG-GAS secondary pump. The research is about characteristic of the secondary pump and its control system. The measuring of characteristic parameter of secondary cooling pump was being done while the pump running. The pump was loading with capacity 1950 m3/hr. with ambient temperature 28.5 oC. The fault effect of public grid (PLN) such as the fluctuation of both voltage and frequency likes voltage drops (dip). Supply block out that effect of the electric motor performances directly will be analyzed. How far those faults will effect the overall performance of secondary cooling system. Analyzing. Will be done according to the control system was installed. Has be done to find the direct effects of the motor performances against the motor rotation fluctuation which run from 1450 rpm to 1475 rpm. The using of start-delta starting method with delay time about 6 seconds, is enough or not to reduce the inrush starting current also analyzed in this paper. From the research can be obtained that in the steady state condition , the electric motor runs with both power and current are still under tolerances permitted. According to the analyzed data above, it will be consider that the control system of secondary pump would be modified or not. Therefore the analyzed data can show the characteristic curve of the secondary cooling system performance

  17. Gas-Cooled Thorium Reactor with Fuel Block of the Unified Design

    Directory of Open Access Journals (Sweden)

    Igor Shamanin

    2015-01-01

    Full Text Available Scientific researches of new technological platform realization carried out in Russia are based on ideas of nuclear fuel breeding in closed fuel cycle and physical principles of fast neutron reactors. Innovative projects of low-power reactor systems correspond to the new technological platform. High-temperature gas-cooled thorium reactors with good transportability properties, small installation time, and operation without overloading for a long time are considered perspective. Such small modular reactor systems at good commercial, competitive level are capable of creating the basis of the regional power industry of the Russian Federation. The analysis of information about application of thorium as fuel in reactor systems and its perspective use is presented in the work. The results of the first stage of neutron-physical researches of a 3D model of the high-temperature gas-cooled thorium reactor based on the fuel block of the unified design are given. The calculation 3D model for the program code of MCU-5 series was developed. According to the comparison results of neutron-physical characteristics, several optimum reactor core compositions were chosen. The results of calculations of the reactivity margins, neutron flux distribution, and power density in the reactor core for the chosen core compositions are presented in the work.

  18. Conceptual design study for the enhanced gas cooled reactor (EGCR)

    International Nuclear Information System (INIS)

    Nakano, M.; Sadahiro, D.; Ozaki, H.; Bryant, S.D.; Cheyne, A.; Gilroy, J.E.; Hulme, G.; Lennox, T.A.; Sunderland, R.E.; Beaumont, H.M.; Kida, M.; Nomura, M.

    2001-01-01

    The preliminary concept of the carbon dioxide cooled fast reactor EGCR has been studied as a Generation IV system. EGCR with MOX fuel has a very good core performance, a breeding ratio over 1.2, a long operating cycle of 24 months, and a high burnup of 150 GWd/t. The plant system is based on the successful AGR experience but provides 3600 MWth. Enhanced passive safety features are provided and a debris tray included. Preliminary costing studies show that EGCR can be competitive to LWRs and can be constructed on a similar schedule. This EGCR concept also shows development potential. (author)

  19. The modular high-temperature gas-cooled reactor (MHTGR) in the US

    International Nuclear Information System (INIS)

    Neylan, A.J.; Graf, D.F.; Millunzi, A.C.

    1987-01-01

    GA Technologies Inc. and other U.S. corporations, in a cooperative program with the U.S. Department of Energy, is developing a Modular High-Temperature Gas-Cooled Reactor (MHTGR) that will provide highly reliable, economic, nuclear power. The MHTGR system assures maximum safety to the public, the owner/operator, and the environment. The MHTGR is being designed to meet and exceed rigorous requirements established by the user industry for availability, operation and maintenance, plant investment protection, safety and licensing, siting flexibility and economics. The plant will be equally attractive for deployment and operation in the U.S., other major industrialized nations including Korea, Japan, and the Republic of China, as well as the developing nations. The High-Temperature Gas-Cooled Reactor (HTGR) is an advanced, third generation nuclear power system which incorporates distinctive technical features, including the use of pressurized helium as a coolant, graphite as the moderator and core structural material, and fuel in the form of ceramic coated uranium particles. The modular HTGR builds upon generic gas-cooled reactor experience and specific HTGR programs and projects. The MHTGR offers unique technological features and the opportunity for the cooperative international development of an advanced energy system that will help assure adaquate world energy resources for the future. Such international joint venturing of energy development can offer significant benefits to participating industries and governments and also provides a long term solution to the complex problems of the international balance of payments

  20. 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.

  1. Neutron spectrum effects on TRU recycling in Pb-Bi cooled fast reactor core

    International Nuclear Information System (INIS)

    Kim, Yong Nam; Kim, Jong Kyung; Park, Won Seok

    2003-01-01

    This study is intended to evaluate the dependency of TRU recycling characteristics on the neutron spectrum shift in a Pb-Bi cooled core. Considering two Pb-Bi cooled cores with the soft and the hard spectrum, respectively, various characteristics of the recycled core are carefully examined and compared with each other. Assuming very simplified fuel cycle management with the homogeneous and single region fuel loading, the burnup calculations are performed until the recycled core reached to the (quasi-) equilibrium state. The mechanism of TRU recycling toward the equilibrium is analyzed in terms of burnup reactivity and the isotopic compositions of TRU fuel. In the comparative analyses, the difference in the recycling behavior between the two cores is clarified. In addition, the basic safety characteristics of the recycled core are also discussed in terms of the Doppler coefficient, the coolant loss reactivity coefficient, and the effective delayed neutron fraction

  2. Heat removal capability of core-catcher with inclined cooling channels

    International Nuclear Information System (INIS)

    Suzuki, Y.; Tahara, M.; Kurita, T.; Hamazaki, R.; Morooka, S.

    2009-01-01

    A core-catcher is one of the mitigation systems that provide functions of molten corium cooling and stabilization during a severe accident. Toshiba has been developing a compact core-catcher to be placed at the lower drywell floor in the containment vessel for the next generation BWR as well as near term ABWR. This paper presents the evaluation of heat removal capability of the core-catcher with inclined cooling channels, our verification status and plan. The heat removal capability of the core-catcher is analyzed by using the newly developed two-phase flow analysis code which incorporates drift flux parameters for inclined channels and the CHF correlation obtained from SULTAN tests. Effects of geometrical parameters such as the inclination and the gap size of the cooling channel on the heat removal capability are also evaluated. These results show that the core-catcher has sufficient capability to cool the molten corium during a severe accident. Based on the analysis, it has been shown that the core-catcher has an efficient capability of heat removal to cool the molten corium. (author)

  3. Passive decay heat removal by sump cooling after core meltdown

    International Nuclear Information System (INIS)

    Knebel, J.U.; Mueller, U.

    1996-01-01

    This article presents the basic physical phenomena and scaling criteria of decay heat removal from a large coolant pool by single-phase and two-phase natural circulation flow. The physical significance of the dimensionless similarity groups derived is evaluated. The above results are applied to the SUCO program that is performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of a sump cooling concept for future light water reactors. The sump cooling concept is based on passive safety features within the containment. The work is supported by the German utilities and the Siemens AG. The article gives first measurement results of the 1:20 linearly scaled plane two-dimensional SUCOS-2D test facility. The experimental results of the model geometry are transformed to prototype conditions

  4. Gas-cooled reactor application for a university campus

    International Nuclear Information System (INIS)

    Colak, Ue.; Kadiroghlu, O.K.; Soekmen, C.N.; Schmitt, H.

    1991-01-01

    Large urban areas with unfavourable topographic and meteorological conditions suffer severe air pollution during the winter months. Use of low grade lignites, imported higher quality coal or imported fuel oil are the sources of air pollution in the form of sulphur dioxide, fly ash and soot. Large housing complexes or old and historical locations within the city are in need of pollution free centralized district heating systems. Natural gas imported from the Soviet Union is a solution for this problem. Lack of gas distribution network for high pressure gas within the city is the main bottle-neck for the heating systems utilizing natural gas. Concern of the safety of flammable high pressure gas circulating within the city is another drawback for the natural gas heating systems. Nuclear district heating is an environmentally viable option worth looking into it. Localized urban nuclear heating is an interesting solution for large urban areas with old and historical character. The results of a feasibility study on the HGR application for the Hacettepe University presented here, summarizes the concept of gas-cooled heating reactors specially designed for urban centers. The inherently safe characteristics of the pebble bed heating reactor makes localized urban nuclear heating a viable alternative to other heat sources. An economical analysis of various heat sources with equal power levels is done for the Beytepe campus of Hacettepe University in Ankara. Under special boundary conditions, the price for heat generation can be much lower for nuclear heating with GHR 20 than for hard coal or fuel oil. It is also possible that if the price escalation rate for natural gas exceeds 3%, then nuclear heating with GHR can be more competitive. It is concluded that the nuclear heating of Beytepe campus with a GHR 20 is feasible and economical. (author) 3 figs., 5 refs

  5. Thermodynamic analysis of turbine blade cooling on the performance of gas turbine cycle

    International Nuclear Information System (INIS)

    Sarabchi, K.; Shokri, M.

    2002-01-01

    Turbine inlet temperature strongly affects gas turbine performance. Today blade cooling technologies facilitate the use of higher inlet temperatures. Of course blade cooling causes some thermodynamic penalties that destroys to some extent the positive effect of higher inlet temperatures. This research aims to model and evaluate the performance of gas turbine cycle with air cooled turbine. In this study internal and transpiration cooling methods has been investigated and the penalties as the result of gas flow friction, cooling air throttling, mixing of cooling air flow with hot gas flow, and irreversible heat transfer have been considered. In addition, it is attempted to consider any factor influencing actual conditions of system in the analysis. It is concluded that penalties due to blade cooling decrease as permissible temperature of the blade surface increases. Also it is observed that transpiration method leads to better performance of gas turbine comparing to internal cooling method

  6. Gas-cooled reactor coolant circulator and blower technology

    International Nuclear Information System (INIS)

    1988-08-01

    In the previous 17 meetings held within the framework of the International Working Group on Gas-Cooled Reactors, a wide variety of topics and components have been addressed, but the San Diego meeting represented the first time that a group of specialists had been convened to discuss circulator and blower related technology. A total of 20 specialists from 6 countries attended the meeting in which 15 technical papers were presented in 5 sessions: circulator operating experience I and II (6 papers); circulator design considerations I and II (6 papers); bearing technology (3 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  7. Refueling system for the gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Hawke, B.C.

    1980-05-01

    Criteria specifically related to the handling of Gas-Cooled Fast Breeder Reactor (GCFR) fuel are briefly reviewed, and the most significant requirements with which the refueling system must comply are discussed. Each component of the refueling system is identified, and a functional description of the fuel handling machine is presented. An illustrated operating sequence describing the various functions involved in a typical refueling cycle is presented. The design status of components and subsystems selected for conceptual development is reviewed, and anticipated refueling time frames are given

  8. The UK gas-cooled reactor programme - Progress report 1988

    International Nuclear Information System (INIS)

    Askew, J.R.

    1989-01-01

    This paper summarises key developments during 1988 on the 26 Magnox reactors and 14 AGRs now operating in the UK. Details are given of long-term safety reviews of the Berkeley and Bradwell Magnox stations which resulted in a decision by CEGB to cease generation at Berkeley but to continue operation at Bradwell. The summary of operating experience with the AGRs concentrates on the completion of construction and successful commissioning of the second generation AGRs at Heysham 2 and Torness. An appended article by John Wilson, Deputy Director of the UKAEA's gas-cooled reactor R and D programme, gives details of the aims and achievements of the programme during 1988. (author)

  9. Graphite development for gas-cooled reactors in the USA

    International Nuclear Information System (INIS)

    Burchell, T.D.

    1991-01-01

    This document discusses Modular High-Temperature Gas-Cooled Reactor (MHTGR) graphite activities in the USA which currently include the following research and development tasks: coke examination; effects of irradiation; variability of physical properties (mechanical, thermal-physical, and fracture); fatigue behavior, oxidation behavior; NDE techniques; structural design criteria; and carbon-carbon composite control rod clad materials. These tasks support nuclear grade graphite manufacturing technology including nondestructive examination of billets and components. Moreover, data shall be furnished to support design and licensing of graphite components for the MHTGR

  10. Beam cooling using a gas-filled RFQ ion guide

    CERN Document Server

    Henry, S; De Saint-Simon, M; Jacotin, M; Képinski, J F; Lunney, M D

    1999-01-01

    A radiofrequency quadrupole mass filter is being developed for use as a high-transmission beam cooler by operating it in buffer gas at high pressure. Such a device will increase the sensitivity of on-line experiments that make use of weakly produced radioactive ion beams. We present simulations and some preliminary measurements for a device designed to cool the beam for the MISTRAL RF mass spectrometer on- line at ISOLDE. The work is carried out partly within the frame of the European Community research network: EXOTRAPS. (9 refs).

  11. Perspectives on understanding and verifying the safety terrain of modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, Donald E., E-mail: donald@carlsonperin.net [11221 Empire Lane, Rockville, MD 20852 (United States); Ball, Sydney J., E-mail: beckysyd@comcast.net [100 Greywood Place, Oak Ridge, TN 37830 (United States)

    2016-09-15

    The passive safety characteristics of modular high temperature gas-cooled reactors (HTGRs) are conceptually well known and are largely supported by insights from past and ongoing research. This paper offers perspectives on selected issues in areas where further analysis and testing achievable within existing research and demonstration programs could help address residual uncertainties and better support the analysis of safety performance and the regulatory assessment of defense in depth. Areas considered include the evaluation of normal and anomalous core operating conditions and the analysis of accidents involving loss of forced cooling, coolant depressurization, air ingress, moisture ingress, and reactivity events. In addition to discussing associated uncertainties and potential measures to address them, this paper also proposes supplemental “safety terrain” studies that would use realistic assessments of postulated extreme event sequences to establish a more comprehensive understanding of the inherent behaviors and ultimate safety capabilities of modular HTGRs.

  12. Modular high-temperature gas-cooled reactor simulation using parallel processors

    International Nuclear Information System (INIS)

    Ball, S.J.; Conklin, J.C.

    1989-01-01

    The MHPP (Modular HTGR Parallel Processor) code has been developed to simulate modular high-temperature gas-cooled reactor (MHTGR) transients and accidents. MHPP incorporates a very detailed model for predicting the dynamics of the reactor core, vessel, and cooling systems over a wide variety of scenarios ranging from expected transients to very-low-probability severe accidents. The simulations routines, which had originally been developed entirely as serial code, were readily adapted to parallel processing Fortran. The resulting parallelized simulation speed was enhanced significantly. Workstation interfaces are being developed to provide for user (operator) interaction. In this paper the benefits realized by adapting previous MHTGR codes to run on a parallel processor are discussed, along with results of typical accident analyses

  13. 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.

  14. Numerical simulation and geometry optimization of hot-gas mixing in lower plenum of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Hang; Wang Jie; Laurien, E.

    2010-01-01

    The lower plenum in high temperature gas-cooled reactor was designed to mix the gas of different temperatures from the reactor core. Previous researches suggest the current geometry of the lower plenum to be improved for better mixing capability and lower pressure drop. In the presented work, a series of varied geometries were investigated with numerical simulation way. The choice of appropriate mesh type and size used in the geometry variation was discussed with the reference of experimental data. The original thin ribs in the current design were merged into thicker ones, and a junction located at the starting end of the outlet pipe was introduced. After comparing several potential optimization methods, an improved geometry was selected with the merged ribs increasing the pre-defined mixing coefficient and the junction reducing the pressure drop. Future work was discussed based on the simulation of real reactor case. The work shows a direction for design improvements of the lower plenum geometry. (authors)

  15. Commercial sector gas cooling technology frontier and market share analysis

    International Nuclear Information System (INIS)

    Pine, G.D.; Mac Donald, J.M.; McLain, H.A.

    1990-01-01

    This paper describes a method, developed for the Gas Research Institute of the United States, that can assist planning for commercial sector natural gas cooling systems R and D. These systems are higher in first cost than conventional electric chillers. Yet, engine-driven chiller designs exist which are currently competitive in U.S. markets typified by high electricity or demand charges. Section II describes a scenario analysis approach used to develop and test the method. Section III defines the technology frontier, a conceptual tool for identifying new designs with sales potential. Section IV describes a discrete choice method for predicting market shares of technologies with sales potential. Section V shows how the method predicts operating parameter, cost, and/or performance goals for technologies without current sales potential (or for enhancing a frontier technology's sales potential). Section VI concludes with an illustrative example for the Chicago office building retrofit market

  16. High temperature friction and seizure in gas cooled nuclear reactors

    International Nuclear Information System (INIS)

    Cousseran, P.; Febvre, A.; Martin, R.; Roche, R.

    1978-01-01

    One of the most delicate problems encountered in the gas cooled nuclear reactors is the friction without lubrication in a dry and hot (800 0 C /1472 0 F) helium atmosphere even at very small velocity. The research and development programs are described together with special tribometers that operate at mode than 1000 0 C (1832 0 F) in dry helium. The most interesting test conditions and results are given for gas nitrited steels and for strongly alloyed Ni-Cr steels coated with chromium carbide by plasma sprayed. The effects of parameters live velocity, travelled distance, contact pressure, roughness, temperature and prolonged stops under charge are described together with the effects of negative phenomena like attachment and chattering [fr

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

    International Nuclear Information System (INIS)

    Dalle Donne, M.

    1980-05-01

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

  18. Gas-cooled reactors: the importance of their development

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1979-06-01

    The nearest term GCR is the steam-cycle HTGR, which can be used for both power and process steam production. Use of SC-HTGRs permits timely introduction of thorium fuel cycles and of high-thermal-efficiency reactors, decreasing the need for mined U 3 O 8 before arrival of symbiotic fueling of fast-thermal reactor systems. The gas-turbine HTGR offers prospects of lower capital costs than other nuclear reactors, but it appears to require longer and more costly development than the SC-HTGR. Accelerated development of the GT-HTGR is needed to gain the advantages of timely introduction. The Gas-Cooled Fast Breeder Reactor (GCFR) offers the possibility of fast breeder reactors with lower capital costs and with higher breeding ratios from oxide fuels. The VHTR provides high-temperature heat for hydrogen production

  19. 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

  20. Multidisciplinary design optimization of film-cooled gas turbine blades

    Directory of Open Access Journals (Sweden)

    Talya Shashishekara S.

    1999-01-01

    Full Text Available Design optimization of a gas turbine blade geometry for effective film cooling toreduce the blade temperature has been done using a multiobjective optimization formulation. Three optimization formulations have been used. In the first, the average blade temperature is chosen as the objective function to be minimized. An upper bound constraint has been imposed on the maximum blade temperature. In the second, the maximum blade temperature is chosen as the objective function to be minimized with an upper bound constraint on the average blade temperature. In the third formulation, the blade average and maximum temperatures are chosen as objective functions. Shape optimization is performed using geometric parameters associated with film cooling and blade external shape. A quasi-three-dimensional Navier–Stokes solver for turbomachinery flows is used to solve for the flow field external to the blade with appropriate modifications to incorporate the effect of film cooling. The heat transfer analysis for temperature distribution within the blade is performed by solving the heat diffusion equation using the finite element method. The multiobjective Kreisselmeier–Steinhauser function approach has been used in conjunction with an approximate analysis technique for optimization. The results obtained using both formulations are compared with reference geometry. All three formulations yield significant reductions in blade temperature with the multiobjective formulation yielding largest reduction in blade temperature.

  1. Latest developments in prestressed concrete vessels for gas-cooled reactors

    International Nuclear Information System (INIS)

    Ople, F.S. Jr.

    1979-01-01

    This paper is an update of the design development of prestressed concrete vessels, commonly referred to as 'PCRVs' starting with the first single-cavity PCRV for the Fort St. Vrain Nuclear Generating Station to the latest multi-cavity PCRV configurations being utilized as the primary reactor vessels for both the High Temperature Gas-Cooled Reactor (HTGR) and the Gas-Cooled Fast Breeder Reactor (GCFR) in the U.S.A. The complexity of PCRV design varies not only due to the type of vessel configuration (single versus multi-cavity) but also on the application to the specific type of reactor concept. PCRV technology as applied to the Steam Cycle HTGR is fairly well established; however, some significant technical complexities are associated with PCRV design for the Gas Turbine HTGR and the GCFR. For the Gas Turbine HTGR, for instance, the fluid dynamics of the turbo-machinery cause multi-pressure conditions to exist in various portions of the power conversion loops during operation. This condition complicates the design approach and the proof test specification for the PCRV. The geometric configuration of the multi-cavity PCRV is also more complex due to the introduction of large horizontal cylindrical cavities (housing the turbo/machines for the Gas Turbine HTGR and circulators for the GCFR) in addition to the vertical cylindrical cavities for the core and heat exchangers. Because of this complex geometry, it becomes difficult to achieve an optimum prestressing arrangement for the PCRV. Other novel features of the multi-cavity PCRV resulting from the continuing design optimization effort are the incorporation of an asymmetric (offset core) configuration and the use of large vessel cavity/penetration concrete closures directly held down by prestressing tendons for both economic and safety reasons. (orig.)

  2. Application of Hastelloy X in gas-cooled reactor systems

    International Nuclear Information System (INIS)

    Brinkman, C.R.; Rittenhouse, P.L.; Corwin, W.R.; Strizak, J.P.; Lystrup, A.; DiStefano, J.R.

    1976-10-01

    Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data are reported. Properties of concern include tensile, creep, creep-rupture, fatigue, creep-fatigue interaction, subcritical crack growth, thermal stability, and the influence of helium environments with controlled amounts of impurities on these properties. In order to develop these properties in helium environments that are expected to be prototypic of HTGR operating conditions, it was necessary to construct special environmental test systems. Details of construction and operating parameters are described. Interim results from tests designed to determine the above properties are presented. To date a fairly extensive amount of information has been generated on this material at Oak Ridge National Laboratory and elsewhere concerning behavior in air, which is reviewed. However, only limited data are available from tests conducted in helium. Comparisons of the fatigue and subcritical growth behavior in air between Hastelloy X and a number of other structural alloys are given

  3. Modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shepherd, L.R.

    1988-01-01

    The high financial risk involved in building large nuclear power reactors has been a major factor in halting investment in new plant and in bringing further technical development to a standstill. Increased public concern about the safety of nuclear plant, particularly after Chernobyl, has contributed to this stagnation. Financial and technical risk could be reduced considerably by going to small modular units, which would make it possible to build up power station capacity in small steps. Such modular plant, based on the helium-cooled high temperature reactor (HTR), offers remarkable advantages in terms of inherent safety characteristics, partly because of the relatively small size of the individual modules but more on account of the enormous thermal capacity and high temperature margins of the graphitic reactor assemblies. Assessments indicate that, in the USA, the cost of power from the modular systems would be less than that from conventional single reactor plant, up to about 600 MW(e), and only marginally greater above that level, a margin that should be offset by the shorter time required in bringing the modular units on line to earn revenue. The modular HTR would be particularly appropriate in the UK, because of the considerable British industrial background in gas-cooled reactors, and could be a suitable replacement for Magnox. The modular reactor would be particularly suited to combined heat and power schemes and would offer great potential for the eventual development of gas turbine power conversion and the production of high-temperature process heat. (author)

  4. Evaluation of the RSG-GAS cooling tower performance

    International Nuclear Information System (INIS)

    Suroso

    2003-01-01

    Utilization of RSG-GAS reactor should be operated as efficiently as possible, so that reactor operation planning using one line primary coolant can be anticipated. To analyze the performance of the RSG-GAS cooling tower with one line primary coolant doing by using same data from 10 MW thermal reactor operation. The result were then compare to those achieved using CATHENA code. The results indicated that, for design condition the ratio of water flowrate to air is (L/G) 1.52 and number transfer unit (NTU) is 0.348. For operation condition, the average of L/G and NTU are respectively 1.37 and 0,348. Moreover the results achieved by the code showed that L/G and NTU are respectively 1.35 and 0,302. The performance of cooling tower achieved operation condition and the code results are respectively 91% and 72%. This means that the calculated results are lower than measurement results

  5. Study of risk reduction by improving operation of reactor core isolation cooling system

    International Nuclear Information System (INIS)

    Watanabe, Yamato; Tazai, Ayuko; Yamagishi, Shohei; Muramatsu, Ken; Muta, Hitoshi

    2014-01-01

    The Fukushima Daiichi nuclear power plant fell into a station blackout (SBO) due to the earthquake and tsunami in which most of the core cooling systems were disabled. In the units 2 and 3, water injection to the core was performed only by water injection system with turbine driven pumps. In particular, it is inferred from observed plant parameters that the reactor core isolation cooling system (RCIC) continued its operation much longer than it was originally expected (8 hours). Since the preparation of safety measures did not work, the reactor core damaged. With a view to reduce risk of station blackout events in a BWR by accident management, this study investigated the efficacy of operation procedures that takes advantage of RCIC which can be operated with only equipment inside reactor building and does not require an AC power source. The efficacy was assessed in this study by two steps. The first step is a thermal hydraulic analysis with the RETRAN3D code to estimate the potential extension of duration of core cooling by RCIC and the second step is the estimation of time required for recovery of off-site power from experiences at nuclear power stations under the 3.11 earthquake. This study showed that it is possible to implement more reliable measures for accident termination and to greatly reduce the risk of SBO by the installation of accident management measures with use of RCIC for extension of core cooling under SBO conditions. (author)

  6. A review of helium gas turbine technology for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    No, Hee Cheon; Kim, Ji Hwan; Kim, Hyeun Min

    2007-01-01

    Current High-Temperature Gas-cooled Reactors (HTGRs) are based on a closed brayton cycle with helium gas as the working fluid. Thermodynamic performance of the axial-flow helium gas turbines is of critical concern as it considerably affects the overall cycle efficiency. Helium gas turbines pose some design challenges compared to steam or air turbomachinery because of the physical properties of helium and the uniqueness of the operating conditions at high pressure with low pressure ratio. This report present a review of the helium Brayton cycle experiences in Germany and in Japan. The design and availability of helium gas turbines for HTGR are also presented in this study. We have developed a new throughflow calculation code to calculate the design-point performance of helium gas turbines. Use of the method has been illustrated by applying it to the GTHTR300 reference

  7. Investigations on sump cooling after core melt down

    Energy Technology Data Exchange (ETDEWEB)

    Knebel, J.U. [Forschungeszentrum Karlsruhe - Technik und Umwelt Institut fuer Angewandte Thermo- und Fluiddynamik, Karlsruhe (Germany)

    1995-09-01

    This article presents the basic physical phenomena and scaling criteria of decay heat removal from a large coolant pool by single-phase and two-phase natural circulation flow. The physical significance of the dimensionless similarity groups derived is evaluated. The above results are applied to the SUCO program that is performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of an optional sump cooling concept for the European Pressurized Water Reactor EPR. This concept is entirely based on passive safety features within the containment. The work is supported by the German utilities and the Siemens dimensional SUCOS-2D test facility. The experimental results of the model geometry are transformed to prototypic conditions.

  8. Investigations on sump cooling after core melt down

    International Nuclear Information System (INIS)

    Knebel, J.U.

    1995-01-01

    This article presents the basic physical phenomena and scaling criteria of decay heat removal from a large coolant pool by single-phase and two-phase natural circulation flow. The physical significance of the dimensionless similarity groups derived is evaluated. The above results are applied to the SUCO program that is performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of an optional sump cooling concept for the European Pressurized Water Reactor EPR. This concept is entirely based on passive safety features within the containment. The work is supported by the German utilities and the Siemens dimensional SUCOS-2D test facility. The experimental results of the model geometry are transformed to prototypic conditions

  9. Dynamics and inherent safety features of small modular high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

    1986-01-01

    Investigations were made at Oak Ridge National Laboratory to characterize the dynamics and inherent safety features of various modular high temperature gas-cooled reactor (HTGR) designs. This work was sponsored by the US Nuclear Regulatory Commission's HTGR Safety Research program. The US Department of Energy (DOE) and the Gas Cooled Reactor Associates (GCRA) have sponsored studies of several modular HTGR concepts, each having it own unique advantageous economic and inherent safety features. The DOE design team has recently choses a 350-MW(t) annular core with prismatic, graphite matrix fuel for its reference plant. The various safety features of this plant and of the pebble-bed core designs similar to those currently being developed and operated in the Federal Republic of Germany (FRG) are described. A varity of postulated accident sequences involving combinations of loss of forced circulation of the helium primary coolant, loss of primary coolant pressurization, and loss of normal and backup heat sinks were studied and are discussed. Results demonstrate that each concept can withstand an uncontrolled heatup accident without reaching excessive peak fuel temperatures. Comparisons of calculated and measured response for a loss of forced circulation test on the FRG reactor, AVR, are also presented. 10 refs

  10. Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

    Science.gov (United States)

    Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

    2017-01-01

    The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

  11. The role of the IAEA in gas-cooled reactor development and application

    Energy Technology Data Exchange (ETDEWEB)

    Cleveland, J.; Kupitz, J. [International Atomic Energy Agency, Vienna (Austria). Div. of Nuclear Power

    1995-03-01

    The IAEA`s activities in gas-cooled reactor development focus on the four technical areas which are predicted to provide advanced HTGRs with a high degree of safety, but which must be proven. These are: (a) the safe neutronic behaviour of the core, (b) reliance on ceramic coated fuel particles to retain fission products even under extreme conditions, (c) the ability to dissipate decay heat by natural heat transport mechanisms, and (d) the safe behaviour of the fuel and reactor core under chemical attack (air or water ingress). The first three are subjects of Coordinated Research Programmes (CRPs) and the last was recently addressed in an information exchange meeting. CRPs are 3 to 6 years in duration, and often involve experimental activities. CRPs allow a sharing of efforts on an international basis and benefit from the experience and expertise of researchers from the participating institutes. (J.P.N.).

  12. Evaluation of safety test needs for the gas cooled breeder reactors

    International Nuclear Information System (INIS)

    Emon, D.E.; Buttemer, D.R.; Sevy, R.H.

    1976-01-01

    This paper deals with the process used in determining the safety test needs for the Gas Cooled Fast Breeder Reactor (GCFR), reports existing tentative conclusions, and indicates the direction that the process is taking at this time. The process is based upon two ideas: (1) that the safety information needs will be identified through risk analysis directly dependent on the various design features of the GCFR and (2) that the safety program will be determined by a safety review committee. The paper limits itself to presenting thoughts on the safety test needs directly associated with the GCFR core during severe beyond design basis accident situations involving the loss of coolable core geometry. Representative event sequence diagrams are reported for the three generic classes of accidents considered. The following categories of information are identified: safety information needs, safety tests required to fulfill these information needs, and the facilities required to perform the tests

  13. Research on enhancement of natural circulation capability in lead–bismuth alloy cooled reactor by using gas-lift pump

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, Juanli, E-mail: Jenyzuo@163.com; Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn; Chen, Ronghua, E-mail: ronghua.chen@stu.xjtu.edu.cn; Qiu, Suizheng; Su, Guanghui, E-mail: ghsu@mail.xjtu.edu.cn

    2013-10-15

    Highlights: • The gas-lift pump has been adopted to enhance the natural circulation capability. • LENAC code is developed in my study. • The calculation results by LENAC code show good agreement with experiment results. • Gas mass flow rate, bubble diameter, rising pipe length are important parameters. -- Abstract: The gas-lift pump has been adopted to enhance the natural circulation capability in the type of lead–bismuth alloy cooled reactors such as Accelerator Driven System (ADS) and Liquid–metal Fast Reactor (LMFR). The natural circulation ability and the system safety are obviously influenced by the two phase flow characteristics of liquid metal–inert gas. In this study, LENAC (LEad bismuth alloy NAtural Circulation capability) code has been developed to evaluate the natural circulation capability of lead–bismuth cooled ADS with gas-lift pump. The drift flow theory, void fraction prediction model and friction pressure drop prediction model have been incorporated into LENAC code. The calculation results by LENAC code show good agreement with experiment results of CIRCulation Experiment (CIRCE) facility. The effects of the gas mass flow rate, void fraction, gas quality, bubble diameter and the rising pipe height or the potential difference between heat exchanger and reactor core on natural circulation capability of gas-lift pump have been analyzed. The results showed that in bubbly flow pattern, for a fixed value of gas mass flow rate, the natural circulation capability increased with the decrease of the bubble diameter. In the bubbly flow, slug flow, churn flow and annular flow pattern, with the gas mass flow rate increasing, the natural circulation capability initially increased and then declined. And the flow parameters influenced the thermal hydraulic characteristics of the reactor core significantly. The present work is helpful for revealing the law of enhancing the natural circulation capability by gas-lift pump, and providing theoretical

  14. Provision of reliable core cooling in vessel-type boiling reactors

    International Nuclear Information System (INIS)

    Alferov, N.S.; Balunov, B.F.; Davydov, S.A.

    1987-01-01

    Methods for providing reliable core cooling in vessel-type boiling reactors with natural circulation for heat supply are analysed. The solution of this problem is reduced to satisfaction of two conditions such as: water confinement over the reactor core necessary in case of an accident and confinement of sufficient coolant flow rate through the bottom cross section of fuel assemblies for some time. The reliable fuel element cooling under conditions of a maximum credible accident (brittle failure of a reactor vessel) is shown to be provided practically in any accident, using the safety vessel in combination with the application of means of standard operation and minimal composition and capacity of ECCS

  15. Fundamental design bases for independent core cooling in Swedish nuclear power reactors

    International Nuclear Information System (INIS)

    Jelinek, Tomas

    2015-01-01

    New regulations on design and construction of nuclear power plants came into force in 2005. The need of an independent core cooling system and if the regulations should include such a requirement was discussed. The Swedish Radiation Safety authority (SSM) decided to not include such a requirement because of open questions about the water balance and started to investigate the consequences of an independent core cooling system. The investigation is now finished and SSM is also looking at the lessons learned from the accident in Fukushima 2011. One of the most important measures in the Swedish national action plan is the implementation of an independent core cooling function for all Swedish power plants. SSM has investigated the basic design criteria for such a function where some important questions are the level of defence in depth and the acceptance criteria. There is also a question about independence between the levels of defence in depth that SSM have included in the criteria. Another issue that has to be taken into account is the complexity of the system and the need of automation where independence and simplicity are very strong criteria. In the beginning of 2014 a memorandum was finalized regarding fundamental design bases for independent core cooling in Swedish nuclear power reactors. A decision based on this memorandum with an implementation plan will be made in the first half of 2014. Sweden is also investigating the possibility to have armed personnel on site, which is not allowed currently. The result from the investigation will have impact on the possibility to use mobile equipment and the level of protection of permanent equipment. In this paper, SSM will present the memorandum for design bases for independent core cooling in Swedish nuclear power reactors that was finalized in March 20147 that also describe SSM's position regarding independence and automation of the independent core cooling function. This memorandum describes the Swedish

  16. 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

  17. The development of advanced gas cooled reactor iodine adsorber systems

    International Nuclear Information System (INIS)

    Meddings, P.

    1986-01-01

    Advanced Gas Cooled Reactors (AGRs) are provided with plants to process the carbon dioxide coolant prior to its discharge to atmosphere. Included in these are beds of granular activated charcoal, contained within a suitable pressure vessel, through which the high pressure carbon dioxide is passed for the purpose of retaining iodine and iodine-containing compounds. Carry-over carbon dust from the adsorption beds was identified during active in-situ commissioning testing, radio-iodine being transported with the particulate material due to gross disturbance of the adsorber carbon bed and displacement of the vessel internals. The methods used to identify the causes of the problems and find solutions are described. A development programme for the Heysham-2 and Torness reactors iodine adsorber units was set up to identify a method of de-dusting granular charcoal and develop it for full-scale use, of assess the effect under conditions of high gas density of approach velocity on charcoal fines production and to establish the pressure drop characteristics of a packed granular bed and to develop an effective design of inlet gas diffuser manifold to ensure an acceptable velocity distribution. This has involved the construction of a small scale high pressure carbon dioxide rig and development of an air flow model. This work is described. (UK)

  18. RBMK-1500 accident management for loss of long-term core cooling

    International Nuclear Information System (INIS)

    Uspuras, E.; Kaliatka, A.

    2001-01-01

    Results of the Level 1 probabilistic safety assessment of the Ignalina NPP has shown that in topography of the risk, transients dominate above the accidents with LOCAs and failure of the core long-term cooling are the main factors to frequency of the core damage. Previous analyses have shown, that after initial event, as a rule, the reactivity control, as well as short-term and intermediate cooling are provided. However, the acceptance criteria of the long-term cooling are not always carried out. It means that from this point of view the most dangerous accident scenarios are the scenarios related to loss of the core long-term cooling. On the other hand, the transition to the core condition due to loss of the long-term cooling specifies potential opportunities for the management of the accident consequences. Hence, accident management for the mitigation of the accident consequences should be considered and developed. The most likely initiating event, which probably leads to the loss of long term cooling accident, is station blackout. The station blackout is the loss of normal electrical power supply for local needs with an additional failure on start-up of all diesel generators. In the case of loss of electrical power supply MCPs, the circulating pumps of the service water system and MFWPs are switched-off. At the same time, TCV of both turbines are closed. Failure of diesel generators leads to the non-operability of the ECCS long-term cooling subsystem. It means the impossibility to feed MCC by water. The analysis of the station blackout for Ignalina NPP was performed using RELAP5 code. (author)

  19. Measurement of sulphur-35 in the coolant gas of the Windscale Advanced Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Sandalls, F.J.

    1978-03-01

    Sulphur is an important element in some food chains and the release of radioactive sulphur to the environment must be closely controlled if the chemical form is such that it is available or potentially available for entering food chains. The presence of sulphur-35 in the coolant gas of the Windscale Advanced Gas-Cooled Reactor warranted a study to assess the quantity and chemical form of the radioactive sulphur in order to estimate the magnitude of the potential environmental hazard which might arise from the release of coolant gas from Civil Advanced Gas-Cooled Reactors. A combination of gas chromatographic and radiochemical analyses revealed carbonyl sulphide to be the only sulphur-35 compound present in the coolant gas of the Windscale Reactor. The concentration of carbonyl sulphide was found to lie in the range 40 to 100 x 10 -9 parts by volume and the sulphur-35 specific activity was about 20 mCi per gramme. The analytical techniques are described in detail. The sulphur-35 appears to be derived from the sulphur and chlorine impurities in the graphite. A method for the preparation of carbonyl sulphide labelled with sulphur-35 is described. (author)

  20. Auxiliary bearing design considerations for gas cooled reactors

    International Nuclear Information System (INIS)

    Penfield, S.R. Jr.; Rodwell, E.

    2001-01-01

    The need to avoid contamination of the primary system, along with other perceived advantages, has led to the selection of electromagnetic bearings (EMBs) in most ongoing commercial-scale gas cooled reactor (GCR) designs. However, one implication of magnetic bearings is the requirement to provide backup support to mitigate the effects of failures or overload conditions. The demands on these auxiliary or 'catcher' bearings have been substantially escalated by the recent development of direct Brayton cycle GCR concepts. Conversely, there has been only limited directed research in the area of auxiliary bearings, particularly for vertically oriented turbomachines. This paper explores the current state-of-the-art for auxiliary bearings and the implications for current GCR designs. (author)

  1. Advanced gas cooled nuclear reactor materials evaluation and development program

    International Nuclear Information System (INIS)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed

  2. Acoustical environment of gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Blevins, R.D.

    1986-01-01

    Methods for acoustical analysis of gas-cooled nuclear reactors in terms of the sources of sound, the propagation of sound about the coolant circuit and the response of reactor structures to sound, are described. Sources of sound that are considered are circulators, jets, vortex shedding and separated flow. Circulators are generally the dominant source of sound. At low frequency the sound propagates one dimensionally through the ducts and cavities of the reactor. At high frequency the sound excites closely spaced two- and three-dimensional acoustic modes, and the resultant sound field can be described only statistically. The sound excites plate and shell structures within the coolant circuit. Secondary steam piping can also be excited by pumps and valves. Formulations are presented for the resultant vibration. Vibration-induced damage is also reviewed. (author)

  3. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

    2012-12-01

    This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

  4. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

    2012-06-01

    This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

  5. Safety analysis of a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shimazu, Akira; Morimoto, Toshio

    1975-01-01

    In recent years, in order to satisfy the social requirements of environment and safety and also to cope with the current energy stringency, the installation of safe nuclear power plants is indispensable. Herein, safety analysis and evaluation to confirm quantitatively the safety design of a nuclear power plant become more and more important. The safety analysis and its methods for a high temperature gas-cooled reactor are described, with emphasis placed on the practices by Fuji Electric Manufacturing Co. Fundamental rule of securing plant safety ; safety analysis in normal operation regarding plant dynamic characteristics and radioactivity evaluation ; and safety analysis at the time of accidents regarding plant response to the accidents and radioactivity evaluation are explained. (Mori, K.)

  6. Description of the magnox type of gas cooled reactor (MAGNOX)

    International Nuclear Information System (INIS)

    Jensen, S.E.; Nonboel, E.

    1999-05-01

    The present report comprises a technical description of the MAGNOX type of reactor as it has been build in Great Britain. The Magnox reactor is gas cooled (CO 2 ) with graphite moderators. The fuels is natural uranium in metallic form, canned with a magnesium alloy called 'Magnox'. The Calder Hall Magnox plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other stations are given in tables with a summary of design data. Special design features are also shortly described. Where specific data for Calder Hall Magnox has not been available, corresponding data from other Magnox plants has been used. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 sub-project 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au)

  7. Gas cooled fast breeder reactors using mixed carbide fuel

    International Nuclear Information System (INIS)

    Kypreos, S.

    1976-09-01

    The fast reactors being developed at the present time use mixed oxide fuel, stainless-steel cladding and liquid sodium as coolant (LMFBR). Theoretical and experimental designing work has also been done in the field of gas-cooled fast breeder reactors. The more advanced carbide fuel offers greater potential for developing fuel systems with doubling times in the range of ten years. The thermohydraulic and physics performance of a GCFR utilising this fuel is assessed. One question to be answered is whether helium is an efficient coolant to be coupled with the carbide fuel while preserving its superior neutronic performance. Also, an assessment of the fuel cycle cost in comparison to oxide fuel is presented. (Auth.)

  8. A He-gas Cooled, Stationary Granular Target

    CERN Document Server

    Pugnat, P

    2003-01-01

    In the CERN approach to the design of a neutrino factory, the repetition frequency of the proton beam is high enough to consider stationary solid targets as a viable solution for multi-MW beams. The target consists of high density tantalum spheres of 2 mm diameter which can efficiently be cooled by passing a high mass flow He-gas stream through the voids between the Ta-granules. Very small thermal shocks and stresses will arise in this fine grained structure due to the relatively long burst of 3.3 ms from the SPL-proton linac. In a quadruple target system where each target receives only one quarter of the total beam power of 4 MW, conservative temperature levels and adequate lifetimes of the target are estimated in its very high radiation environment. A conceptual design of the integration of the target into the magnetic horn-pion-collector is presented.

  9. Development of failure detection system for gas-cooled reactor

    International Nuclear Information System (INIS)

    Feirreira, M.P.

    1990-01-01

    This work presents several kinds of Failure Detection Systems for Fuel Elements, stressing their functional principles and major applications. A comparative study indicates that the method of electrostatic precipitation of the fission gases Kr and Xe is the most efficient for fuel failure detection in gas-cooled reactors. A detailed study of the physical phenomena involved in electrostatic precipitation led to the derivation of an equation for the measured counting rate. The emission of fission products from the fuel and the ion recombination inside the chamber are evaluated. A computer program, developed to simulate the complete operation of the system, relates the counting rate to the concentration of Kr and Xe isotopes. The project of a mock-up is then presented. Finally, the program calculations are compared to experimental data, available from the literature, yielding a close agreement. (author)

  10. Gas-cooled reactor commercialization study. Interim report

    International Nuclear Information System (INIS)

    1977-01-01

    This report of the gas-cooled reactor commercialization study completes the technical and cost evaluation portions of this study contract. A final report in December will update the status of the incentive analyses and the issues of commercialization. This study was designed to bring together potential industry participants (utilities and suppliers) to evaluate the commercial potential of the HTGR-SC and to build channels of communication among the participating organizations at the same time that technical, economic and institutional issues were being evaluated. RAMCO, Inc., in suggesting and using this study approach, believes its application extends to any commercialization problem involving multi-party involvement in high capital, intensive, high risk energy technologies

  11. Thermal-hydraulic evaluation study of the effectiveness of emergency core cooling system for light water reactors

    International Nuclear Information System (INIS)

    Sobajima, Makoto

    1985-08-01

    In order to evaluate the core cooling capability of the emergeny core cooling system, which is a safety guard system of light water reactors for a loss-of-coolant accident, a variety of large scale test were performed. Through the results, many phenomena were investigated and the predictabity of analytical codes were examined. The tests conducted were a single-vessel blowdown test, emergency core cooling test in a PWR simulation facility, spray cooling test for a BWR, large scale reflood test and a separate effect test on countercurrent flow. These test results were examined to clarify thermal-hydraulic phenomena and the effect of various test parameters and were utilized to improve predictability of the analytical codes. Some models for flow behavior in the upper core were also developed. By evaluating the effectiveness of various emergency core cooling system configurations, more effective cooling system than the current one was proposed and demonstrated. (author)

  12. Shivering heat production and body fat protect the core from cooling during body immersion, but not during head submersion: a structural equation model.

    Science.gov (United States)

    Pretorius, Thea; Lix, Lisa; Giesbrecht, Gordon

    2011-03-01

    Previous studies showed that core cooling rates are similar when only the head or only the body is cooled. Structural equation modeling was used on data from two cold water studies involving body-only, or whole body (including head) cooling. Exposure of both the body and head increased core cooling, while only body cooling elicited shivering. Body fat attenuates shivering and core cooling. It is postulated that this protection occurs mainly during body cooling where fat acts as insulation against cold. This explains why head cooling increases surface heat loss with only 11% while increasing core cooling by 39%. Copyright © 2011 Elsevier Ltd. All rights reserved.

  13. Safety aspects of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)

    International Nuclear Information System (INIS)

    Silady, F.A.; Millunzi, A.C.

    1989-08-01

    The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes the basic high-temperature gas-cooled reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The qualitative top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. The MHTGR safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles has been evaluated. A broad range of challenges to core heat removal have been examined which include a loss of helium pressure and a simultaneous loss of forced cooling of the core. The challenges to control of heat generation have considered not only the failure to insert the reactivity control systems, but the withdrawal of control rods. Finally, challenges to control chemical attack of the ceramic coated fuel have been considered, including catastrophic failure of the steam generator allowing water ingress or of the pressure vessels allowing air ingress. The plant's response to these extreme challenges is not dependent on operator action and the events considered encompass conceivable operator errors. In the same vein, reliance on radionuclide retention within the full particle and on passive features to perform a few key functions to maintain the fuel within acceptable conditions also reduced susceptibility to external events, site-specific events, and to acts of sabotage and terrorism. 4 refs., 14 figs., 1 tab

  14. Metaphysics methods development for high temperature gas cooled reactor analysis

    International Nuclear Information System (INIS)

    Seker, V.; Downar, T. J.

    2007-01-01

    Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts in the Generation-IV technology road map. Considerable research has been performed on the design and safety analysis of these reactors. However, the calculational tools being used to perform these analyses are not state-of-the-art and are not capable of performing detailed three-dimensional analyses. This paper presents the results of an effort to develop an improved thermal-hydraulic solver for the pebble bed type high temperature gas cooled reactors. The solution method is based on the porous medium approach and the momentum equation including the modified Ergun's resistance model for pebble bed is solved in three-dimensional geometry. The heat transfer in the pebble bed is modeled considering the local thermal non-equilibrium between the solid and gas, which results in two separate energy equations for each medium. The effective thermal conductivity of the pebble-bed can be calculated both from Zehner-Schluender and Robold correlations. Both the fluid flow and the heat transfer are modeled in three dimensional cylindrical coordinates and can be solved in steady-state and time dependent. The spatial discretization is performed using the finite volume method and the theta-method is used in the temporal discretization. A preliminary verification was performed by comparing the results with the experiments conducted at the SANA test facility. This facility is located at the Institute for Safety Research and Reactor Technology (ISR), Julich, Germany. Various experimental cases are modeled and good agreement in the gas and solid temperatures is observed. An on-going effort is to model the control rod ejection scenarios as described in the OECD/NEA/NSC PBMR-400 benchmark problem. In order to perform these analyses PARCS reactor simulator code will be coupled with the new thermal-hydraulic solver. Furthermore, some of the other anticipated accident scenarios in the benchmark

  15. Portable tester for determining gas content within a core sample

    Science.gov (United States)

    Garcia, F. Jr.; Schatzel, S.J.

    1998-04-21

    A portable tester is provided for reading and displaying the pressure of a gas released from a rock core sample stored within a sealed container and for taking a sample of the released pressurized gas for chemical analysis thereof for subsequent use in a modified direct method test which determines the volume of gas and specific type of gas contained within the core sample. The portable tester includes a pair of low and high range electrical pressure transducers for detecting a gas pressure; a pair of low and high range display units for displaying the pressure of the detected gas; a selector valve connected to the low and high range pressure transducers and a selector knob for selecting gas flow to one of the flow paths; control valve having an inlet connection to the sealed container; and outlets connected to: a sample gas canister, a second outlet port connected to the selector valve means for reading the pressure of the gas from the sealed container to either the low range or high range pressure transducers, and a connection for venting gas contained within the sealed container to the atmosphere. A battery is electrically connected to and supplies the power for operating the unit. The pressure transducers, display units, selector and control valve means and the battery is mounted to and housed within a protective casing for portable transport and use. 5 figs.

  16. Multicycle Optimization of Advanced Gas-Cooled Reactor Loading Patterns Using Genetic Algorithms

    International Nuclear Information System (INIS)

    Ziver, A. Kemal; Carter, Jonathan N.; Pain, Christopher C.; Oliveira, Cassiano R.E. de; Goddard, Antony J. H.; Overton, Richard S.

    2003-01-01

    A genetic algorithm (GA)-based optimizer (GAOPT) has been developed for in-core fuel management of advanced gas-cooled reactors (AGRs) at HINKLEY B and HARTLEPOOL, which employ on-load and off-load refueling, respectively. The optimizer has been linked to the reactor analysis code PANTHER for the automated evaluation of loading patterns in a two-dimensional geometry, which is collapsed from the three-dimensional reactor model. GAOPT uses a directed stochastic (Monte Carlo) algorithm to generate initial population members, within predetermined constraints, for use in GAs, which apply the standard genetic operators: selection by tournament, crossover, and mutation. The GAOPT is able to generate and optimize loading patterns for successive reactor cycles (multicycle) within acceptable CPU times even on single-processor systems. The algorithm allows radial shuffling of fuel assemblies in a multicycle refueling optimization, which is constructed to aid long-term core management planning decisions. This paper presents the application of the GA-based optimization to two AGR stations, which apply different in-core management operational rules. Results obtained from the testing of GAOPT are discussed

  17. observer-based diagnostics and monitoring of vibrations in nuclear reactor core cooling system

    International Nuclear Information System (INIS)

    Siry, S.A K.

    2007-01-01

    analysis and diagnostics of vibration in industrial systems play a significant rule to prevent severe severe damages . drive shaft vibration is a complicated phenomenon composed of two independent forms of vibrations, translational and torsional. translational vibration measurements in case of the reactor core cooling system are introduced. the system under study consists of the three phase induction motor, flywheel, centrifugal pump, and two coupling between motor-flywheel, and flywheel-pump. this system structure is considered to be one where the blades are pegged into the discs fitting into the shafts. a non-linear model to simulate vibration in the reactor core cooling system will be introduced. simulation results of an operating reactor core cooling system using the actual parameters will be presented to validate the accuracy and reliability of the proposed analytical method the accuracy in analyzing the results depends on the system model. the shortcomings of the conventional model will be avoided through the use of that accurate nonlinear model which improve the simulation of the reactor core cooling system

  18. Core design of a high breeding fast reactor cooled by supercritical pressure light water

    Energy Technology Data Exchange (ETDEWEB)

    Someya, Takayuki, E-mail: russell@ruri.waseda.jp; Yamaji, Akifumi

    2016-01-15

    Highlights: • Core design concept of supercritical light water cooled fast breeding reactor is developed. • Compound system doubling time (CSDT) is applied for considering an appropriate target of breeding performance. • Breeding performance is improved by reducing fuel rod diameter of the seed assembly. • Core pressure loss is reduced by enlarging the coolant channel area of the seed assembly. - Abstract: A high breeding fast reactor core concept, cooled by supercritical pressure light water has been developed with fully-coupled neutronics and thermal-hydraulics core calculations, which takes into account the influence of core pressure loss to the core neutronics characteristics. Design target of the breeding performance has been determined to be compound system doubling time (CSDT) of less than 50 years, by referring to the relationship of energy consumption and economic growth rate of advanced countries such as the G7 member countries. Based on the past design study of supercritical water cooled fast breeder reactor (Super FBR) with the concept of tightly packed fuel assembly (TPFA), further improvement of breeding performance and reduction of core pressure loss are investigated by considering different fuel rod diameters and coolant channel geometries. The sensitivities of CSDT and the core pressure loss with respect to major core design parameters have been clarified. The developed Super FBR design concept achieves fissile plutonium surviving ratio (FPSR) of 1.028, compound system doubling time (CSDT) of 38 years and pressure loss of 1.02 MPa with positive density reactivity (negative void reactivity). The short CSDT indicates high breeding performance, which may enable installation of the reactors at a rate comparable to energy growth rate of developed countries such as G7 member countries.

  19. Assessment and status report High-Temperature Gas-Cooled Reactor gas-turbine technology

    International Nuclear Information System (INIS)

    1981-01-01

    Purpose of this report is to present a brief summary assessment of the High Temperature Gas-Cooled Reactor - Gas Turbine (HTGR-GT) technology. The focal point for the study was a potential 2000 MW(t)/800 MW(e) HTGR-GT commercial plant. Principal findings of the study were that: the HTGR-GT is feasible, but with significantly greater development risk than the HTGR-SC (Steam Cycle). At the level of performance corresponding to the reference design, no incremental economic incentive can be identified for the HTGR-GT to offset the increased development costs and risk relative to the HTGR-SC. The relative economics of the HTGR-GT and HTGR-SC are not significantly impacted by dry cooling considerations. While reduced cycel complexity may ultimately result in a reliability advantage for the HTGR-GT, the value of that potential advantage was not quantified

  20. Low pressure cooling seal system for a gas turbine engine

    Science.gov (United States)

    Marra, John J

    2014-04-01

    A low pressure cooling system for a turbine engine for directing cooling fluids at low pressure, such as at ambient pressure, through at least one cooling fluid supply channel and into a cooling fluid mixing chamber positioned immediately downstream from a row of turbine blades extending radially outward from a rotor assembly to prevent ingestion of hot gases into internal aspects of the rotor assembly. The low pressure cooling system may also include at least one bleed channel that may extend through the rotor assembly and exhaust cooling fluids into the cooling fluid mixing chamber to seal a gap between rotational turbine blades and a downstream, stationary turbine component. Use of ambient pressure cooling fluids by the low pressure cooling system results in tremendous efficiencies by eliminating the need for pressurized cooling fluids for sealing this gap.

  1. A study on different thermodynamic cycle schemes coupled with a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Qu, Xinhe; Yang, Xiaoyong; Wang, Jie

    2017-01-01

    Highlights: • The features of three different power generation schemes, including closed Brayton cycle, non-reheating combined cycle and reheating combined cycle, coupled with high temperature gas-cooled reactor (HTGR) were investigated and compared. • The effects and mechanism of reactor core outlet temperature, compression ratio and other key parameters over cycle characteristics were analyzed by the thermodynamic models.. • It is found that reheated combined cycle has the highest efficiency. Reactor outlet temperature and main steam parameters are key factors to improve the cycle’s performance. - Abstract: With gradual increase in reactor outlet temperature, the efficient power conversion technology has become one of developing trends of (very) high temperature gas-cooled reactors (HTGRs). In this paper, different cycle power generation schemes for HTGRs were systematically studied. Physical and mathematical models were established for these three cycle schemes: closed Brayton cycle, simple combined cycle, and reheated combined cycle. The effects and mechanism of key parameters such as reactor core outlet temperature, reactor core inlet temperature and compression ratio on the features of these cycles were analyzed. Then, optimization results were given with engineering restrictive conditions, including pinch point temperature differences. Results revealed that within the temperature range of HTGRs (700–900 °C), the reheated combined cycle had the highest efficiency, while the simple combined cycle had the lowest efficiency (900 °C). The efficiencies of the closed Brayton cycle, simple combined cycle and reheated combined cycle are 49.5%, 46.6% and 50.1%, respectively. These results provide insights on the different schemes of these cycles, and reveal the effects of key parameters on performance of these cycles. It could be helpful to understand and develop a combined cycle coupled with a high temperature reactor in the future.

  2. Advanced Small-Safe Long-Life Lead Cooled Reactor Cores for Future Nuclear Energy

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jin Hyeong; Hong, Ser Gi [Kyung Hee University, Seoul (Korea, Republic of)

    2014-10-15

    One of the reasons for use of the lead or lead-bismuth alloy coolants is the high boiling temperature that avoids the possibility of coolant voiding. Also, these coolants are compatible with air, steam, and water. Therefore, intermediate coolant loop is not required as in the sodium cooled reactors 3. Lead is considered to be more attractive coolant than lead-bismuth alloy because of its higher availability, lower price, and much lower amount of polonium activity by factor of 104 relatively to lead. On the other hand, lead has higher melting temperature of 601K than that of lead-bismuth (398K), which narrows the operating temperature range and also leads to the possibility of freezing and blockage in fresh cores. Neutronically, the lead and lead-bismuth have very similar characteristics to each other. The lead-alloy coolants have lower moderating power and higher scattering without increasing moderation for neutrons below 0.5MeV, which reduces the leakage of the neutrons through the core and provides an excellent reflecting capability for neutrons. Due to the above features of lead or lead-alloy coolants, there have been lots of studies on the small lead cooled core designs. In this paper, small-safe long-life lead cooled reactor cores having high discharge burnup are designed and neutronically analyzed.. The cores considered in this work rates 110MWt (36.7MWe). In this work, the long-life with high discharge burnup was achieved by using thorium or depleted uranium blanket loaded in the central region of the core. Also, we considered a reference core having no blanket for the comparison. This paper provides the detailed neutronic analyses for these small long-life cores and the detailed analyses of the reactivity coefficients and the composition changes in blankets. The results of the core design and analyses show that our small long-life cores can be operated without refueling over their long-lives longer than 45EFPYs (Effective Full Power Year). In this work

  3. Improving Fuel Cycle Design and Safety Characteristics of a Gas Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Rooijen, W.F.G. van

    2006-01-01

    The Gas Cooled Fast Reactor (GCFR)is one of the Generation IV reactor concepts. This concept specifically targets sustainability of nuclear power generation. In nuclear reactors fertile material is converted to fissile fuel. If the neutrons inducing fission are highly energetic, the opportunity exists to convert more than one fertile nucleus per fission, thereby effectively breeding new nuclear fuel. Reactors operating on this principle are called ‘Fast Breeder Reactor’. Since natural uranium contains 99.3%of the fertile isotope 238 U, breeding increases the energy harvested from the nuclear fuel. If nuclear energy is to play an important role as a source of energy in the future, fast breeder reactors are essential for breeding nuclear fuel. Fast neutrons are also more efficient to destruct heavy (Minor Actinide, MA) isotopes, such as Np, Am and Cm isotopes, which dominate the long-term radioactivity of nuclear waste. So the waste life-time can be shortened if the MA nuclei are destroyed. An important prerequisite of sustainable nuclear energy is the closed fuel cycle, where only fission products are discharged to a final repository, and all Heavy Metal (HM) are recycled. The reactor should breed just enough fissile material to allow refueling of the same reactor, adding only fertile material to the recycled material. Other key design choices are highly efficient power conversion using a direct cycle gas turbine, and better safety through the use of helium, a chemically inert coolant which cannot have phase changes in the reactor core. Because the envisaged core temperatures and operating conditions are similar to thermal-spectrum High Temperature Reactor (HTR) concepts, the research for this thesis initially focused on a design based on existing HTR fuel technology: coated particle fuel, assembled into fuel assemblies. It was found that such a fuel concept could not meet the Generation IV criteria set for GCFR: self-breeding is difficult, the temperature

  4. Thermodynamic assessment of impact of inlet air cooling techniques on gas turbine and combined cycle performance

    International Nuclear Information System (INIS)

    Mohapatra, Alok Ku; Sanjay

    2014-01-01

    The article is focused on the comparison of impact of two different methods of inlet air cooling (vapor compression and vapor absorption cooling) integrated to a cooled gas turbine based combined cycle plant. Air-film cooling has been adopted as the cooling technique for gas turbine blades. A parametric study of the effect of compressor pressure ratio, compressor inlet temperature (T i , C ), turbine inlet temperature (T i , T ), ambient relative humidity and ambient temperature on performance parameters of plant has been carried out. Optimum T i , T corresponding to maximum plant efficiency of combined cycle increases by 100 °C due to the integration of inlet air cooling. It has been observed that vapor compression cooling improves the efficiency of gas turbine cycle by 4.88% and work output by 14.77%. In case of vapor absorption cooling an improvement of 17.2% in gas cycle work output and 9.47% in gas cycle efficiency has been observed. For combined cycle configuration, however, vapor compression cooling should be preferred over absorption cooling in terms of higher plant performance. The optimum value of compressor inlet temperature has been observed to be 20 °C for the chosen set of conditions for both the inlet air cooling schemes. - Highlights: • Inlet air cooling improves performance of cooled gas turbine based combined cycle. • Vapor compression inlet air cooling is superior to vapor absorption inlet cooling. • For every turbine inlet temperature, there exists an optimum pressure ratio. • The optimum compressor inlet temperature is found to be 293 K

  5. Reactor core design optimization of the 200 MWt Pb-Bi cooled fast reactor for hydrogen production

    International Nuclear Information System (INIS)

    Bahrum, Epung Saepul; Su'ud, Zaki; Waris, Abdul; Fitriyani, Dian; Wahjoedi, Bambang Ari

    2008-01-01

    In this study reactor core geometrical optimization of 200 MWt Pb-Bi cooled long life fast reactor for hydrogen production has been conducted. The reactor life time is 20 years and the fuel type is UN-PuN. Geometrical core configurations considered in this study are balance, pancake and tall cylindrical cores. For the hydrogen production unit we adopt steam membrane reforming hydrogen gas production. The optimum operating temperature for the catalytic reaction is 540degC. Fast reactor design optimization calculation was run by using FI-ITB-CHI software package. The design criteria were restricted by the multiplication factor that should be less than 1.002, the average outlet coolant temperature 550degC and the maximum coolant outlet temperature less than 700degC. By taking into account of the hydrogen production as well as corrosion resulting from Pb-Bi, the balance cylindrical geometrical core design with diameter and height of the active core of 157 cm each, the inlet coolant temperature of 350degC and the coolant flow rate of 7000 kg/s were preferred as the best design parameters. (author)

  6. Nuclear reactor core support incorporating also a cooling fluid flow system

    International Nuclear Information System (INIS)

    Pennell, W.E.

    1975-01-01

    A description is given of a core bearing plate with several modular intake units having cooling fluid intake openings on their lower extensions, and on their upper ends located above the bearing plate, at least one fuel assembly which is thus in communication with the area under the bearing plate through the modular intake unit. The means for introducing the cooling fluid into the reactor vessel area are located under the bearing plate. The lower ends of the modular intake have ribs arranged essentially on a plane and join together with openings provided between the seals, in such a manner that the ribs form a barrier. The cooling fluid intake openings are located above this barrier, so that the cooling fluid is compelled to cross it before penetrating into the modular intake units [fr

  7. Gas expulsion vs gas retention in young stellar clusters II: effects of cooling and mass segregation

    Science.gov (United States)

    Silich, Sergiy; Tenorio-Tagle, Guillermo

    2018-05-01

    Gas expulsion or gas retention is a central issue in most of the models for multiple stellar populations and light element anti-correlations in globular clusters. The success of the residual matter expulsion or its retention within young stellar clusters has also a fundamental importance in order to understand how star formation proceeds in present-day and ancient star-forming galaxies and if proto-globular clusters with multiple stellar populations are formed in the present epoch. It is usually suggested that either the residual gas is rapidly ejected from star-forming clouds by stellar winds and supernova explosions, or that the enrichment of the residual gas and the formation of the second stellar generation occur so rapidly, that the negative stellar feedback is not significant. Here we continue our study of the early development of star clusters in the extreme environments and discuss the restrictions that strong radiative cooling and stellar mass segregation provide on the gas expulsion from dense star-forming clouds. A large range of physical initial conditions in star-forming clouds which include the star-forming cloud mass, compactness, gas metallicity, star formation efficiency and effects of massive stars segregation are discussed. It is shown that in sufficiently massive and compact clusters hot shocked winds around individual massive stars may cool before merging with their neighbors. This dramatically reduces the negative stellar feedback, prevents the development of the global star cluster wind and expulsion of the residual and the processed matter into the ambient interstellar medium. The critical lines which separate the gas expulsion and the gas retention regimes are obtained.

  8. Antiproton Powered Gas Core Fission Rocket

    International Nuclear Information System (INIS)

    Kammash, Terry

    2005-01-01

    Extensive research in recent years has demonstrated that 'at rest' annihilation of antiprotons in the uranium isotope U238 leads to fission at nearly 100% efficiency. The resulting highly-ionizing, energetic fission fragments can heat a suitable medium to very high temperatures, making such a process particularly suitable for space propulsion applications. Such an ionized medium, which would serve as a propellant, can be confined by a magnetic field during the heating process, and subsequently ejected through a magnetic nozzle to generate thrust. The gasdynamic mirror (GDM) magnetic configuration is especially suited for this application since the underlying confinement principle is that the plasma be of such density and temperature as to make the ion-ion collision mean free path shorter than the plasma length. Under these conditions the plasma behaves like a fluid, and its escape from the system is analogous to the flow of a gas into vacuum from a vessel with a hole. For the system we propose we envisage radially injecting atomic or U238 plasma beam at a pre-determined position and axially pulsing an antiproton beam which upon interaction with the uranium target gives rise to near isotropic ejection of fission fragments with a total mass of 212 amu and total energy of about 160 MeV. These particles, along with the annihilation products (i.e. pions and muons) will heat the background U238 gas - inserted into the chamber just prior to the release of the antiproton - to one keV temperature. Preliminary analysis reveals that such a propulsion system can produce a specific impulse of about 3000 seconds at a thrust of about 50 kN. When applied to a round trip Mars mission, we find that such a journey can be accomplished in about 142 days with 2 days of thrusting and requiring only one gram of antiprotons to achieve it

  9. 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

  10. Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

    International Nuclear Information System (INIS)

    Hong, S. G.; Kim, Y. H.; Venneri, F.

    2010-01-01

    In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

  11. The modular high-temperature gas-cooled reactor - a new production reactor

    International Nuclear Information System (INIS)

    Nulton, J.D.

    1990-01-01

    One of the reactor concepts being considered for application as a new production reactor (NPR) is a 350-MW(thermal) modular high-temperature gas-cooled reactor (MHTGR). The proposed MHTGR-NPR is based on the design of the commercial MHTGR and is being developed by a team that includes General Atomics and Combustion Engineering. The proposed design includes four modules combined into a production block that includes a shared containment, a spent-fuel storage facility, and other support facilities. The MHTGR has a helium-cooled, graphite-moderated, graphite-reflected annular core formed from prismatic graphite fuel blocks. The MHTGR fuel consists of highly enriched uranium oxycarbide (UCO) microsphere fuel particles that are coated with successive layers of pyrolytic carbon (PyC) and silicon carbide (SiC). Tritium-producing targets consist of enriched 6 Li aluminate microsphere target particles that are coated with successive layers of PyC and SiC similar to the fuel microspheres. Normal reactivity control is implemented by articulated control rods that can be inserted into channels in the inner and outer reflector blocks. Shutdown heat removal is accomplished by a single shutdown heat exchanger and electric motor-driven circulator located in the bottom of the reactor vessel. Current plans are to stack spent fuel elements in dry, helium-filled, water-cooled wells and store them for ∼1 yr before reprocessing. All phases of MHTGR fuel reprocessing have been demonstrated

  12. Assessment of gas cooled fast reactor with indirect supercritical CO2 cycle

    International Nuclear Information System (INIS)

    Hejzlar, P.; Driscoll, M. J.; Dostal, V.; Dumaz, P.; Poullennec, G.; Alpy, N.

    2006-01-01

    Various indirect power cycle options for a helium cooled Gas cooled Fast Reactor (GFR) with particular focus on a supercritical CO 2 (SCO 2 ) indirect cycle are investigated as an alternative to a helium cooled direct cycle GFR. The Balance Of Plant (BOP) options include helium-nitrogen Brayton cycle, supercritical water Rankine cycle, and SCO 2 recompression Brayton power cycle in three versions: (1) basic design with turbine inlet temperature of 550 .deg. C, (2) advanced design with turbine inlet temperature of 650 .deg. C and (3) advanced design with the same turbine inlet temperature and reduced compressor inlet temperature. The indirect SCO 2 recompression cycle is found attractive since in addition to easier BOP maintenance it allows significant reduction of core outlet temperature, making design of the primary system easier while achieving very attractive efficiencies comparable to or slightly lower than, the efficiency of the reference GFR direct cycle design. In addition, the indirect cycle arrangement allows significant reduction of the GFR 'proximate-containment' and the BOP for the SCO 2 cycle is very compact. Both these factors will lead to reduced capital cost

  13. Performance comparison of liquid metal and gas cooled ATW system point designs

    International Nuclear Information System (INIS)

    Yang, W.S.; Taiwo, T.A.; Hill, R.N.; Khalil, H.S.; Wade, D.C.

    2001-01-01

    As part of the Advanced Accelerator Application (AAA) program in the U.S., preliminary design studies have been performed at Argonne National Laboratory (ANL) and Los Alamos National Laboratory (LANL) to define and compare candidate Accelerator Transmutation of Waste (ATW) systems. The studies at ANL have focused primarily on the transmutation blanket component of the overall system. Lead-bismuth eutectic (LBE), sodium, and gas cooled systems are among the blanket technology options currently under consideration. This paper summarizes the results from neutronics trade studies performed at ANL. Core designs have been developed for LBE and sodium cooled 840 MWt fast spectrum accelerator driven systems employing re-cycle. Additionally, neutronics analyses have been performed for a helium-cooled 600 MWt hybrid thermal and fast spectrum system proposed by General Atomics (GA), which is operated in the critical mode for three cycles and in a subcritical accelerator driven mode for a subsequent single cycle. For these three point designs, isotopic inventories, consumption rates, and annual burnup rates are compared. The mass flows and the ultimate loss of transuranic (TRU) isotopes to the waste stream per unit of heat generated during transmutation are also compared on a consistent basis. (author)

  14. The influence of liquid-gas velocity ratio on the noise of the cooling tower

    Science.gov (United States)

    Yang, Bin; Liu, Xuanzuo; Chen, Chi; Zhao, Zhouli; Song, Jinchun

    2018-05-01

    The noise from the cooling tower has a great influence on psychological performance of human beings. The cooling tower noise mainly consists of fan noise, falling water noise and mechanical noise. This thesis used DES turbulence model with FH-W model to simulate the flow and sound pressure field in cooling tower based on CFD software FLUENT and analyzed the influence of different kinds noise, which affected by diverse factors, on the cooling tower noise. It can be concluded that the addition of cooling water can reduce the turbulence and vortex noise of the rotor fluid field in the cooling tower at some extent, but increase the impact noise of the liquid-gas two phase. In general, the cooling tower noise decreases with the velocity ratio of liquid to gas increasing, and reaches the lowest when the velocity ratio of liquid to gas is close to l.

  15. Comparative Experiments to Assess the Effects of Accumulator Nitrogen Injection on Passive Core Cooling During Small Break LOCA

    Directory of Open Access Journals (Sweden)

    Li Yuquan

    2017-02-01

    Full Text Available The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA, the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility—the advanced core-cooling mechanism experiment (ACME—was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA transient. Two comparison test groups—a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI line break—were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the

  16. Comparative experiments to assess the effects of accumulator nitrogen injection on passive core cooling during small break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Li, YuQuan; Hao, Botao; Zhong, Jia; Wan Nam [State Nuclear Power Technology R and D Center, South Park, Beijing Future Science and Technology City, Beijing (China)

    2017-02-15

    The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA), the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility—the advanced core-cooling mechanism experiment (ACME)—was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA) transient. Two comparison test groups—a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI) line break—were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the potential negative

  17. Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Kuniyoshi, E-mail: takamatsu.kuniyoshi@jaea.go.jp; Yan, Xing L.; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

    2014-05-01

    It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9 MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9 MW to 0 MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3 h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7 h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the

  18. The early history of high-temperature helium gas-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    Simnad, M.T.; California Univ., San Diego, La Jolla, CA

    1991-01-01

    The original concepts in the proposals for high-temperature helium gas-cooled power reactors by Farrington Daniels, during the decade 1944-1955, are summarized. The early research on the development of the helium gas-cooled power reactors is reviewed, and the operational experiences with the first generation of HTGRs are discussed. (author)

  19. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

  20. Role of gas cooling in tomorrow`s energy services industry

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, P.J.

    1997-04-01

    This article discusses the marketing approach and opportunities for suppliers and manufacturers of gas cooling equipment to partner with energy service companies (ESCOs). The author`s viewpoint is that in educating and partnering with ESCOs the gas cooling industry enables their technology to reach its potential in the projects that the ESCOs develop.

  1. Gas cooled fast reactor 2400 MWTh, status on the conceptual design studies and preliminary safety analysis

    International Nuclear Information System (INIS)

    Malo, J.Y.; Alpy, N.; Bentivoglio, F.

    2009-01-01

    The Gas cooled Fast Reactor (GFR) is considered by the French Commissariat a l'Energie Atomique as a promising concept, combining the benefits of fast spectrum and high temperature, using Helium as coolant. A status on the GFR preliminary viability was made at the end of 2007, ending the pre-conceptual design phase. A consistent overall systems arrangement was proposed and a preliminary safety analysis based on operating transient calculations and a simplified PSA had established a global confidence in the feasibility and safety of this baseline concept. Its potential for attractive performances had been pointed out. Compare to the more mature Sodium Fast Reactor technology, no demonstrator has ever been built and the feasibility demonstration will required a longer lead time. The next main project milestone is related to the GFR viability, scheduled in 2012. The current studies consist in revisiting the reactor reference design options as selected at the end of 2007. Most of them are being consolidated by going more in depth in the analysis. Some possible alternatives are assessed. The paper will give a status on the last studies performed on the core design and corresponding neutronics and cycle performance, the Decay Heat Removal strategy and preliminary safety analysis, systems design and balance of plant... This paper is complementary to the Icapp'09 papers 9062 dealing with the Gas cooled Fast Reactor Demonstrator ALLEGRO and 9378 related to GFR transients analysis. (author)

  2. Development of components for the gas-cooled fast breeder reactor program

    International Nuclear Information System (INIS)

    Dee, J.B.; Macken, T.

    1977-01-01

    The gas-cooled fast breeder reactor (GCFR) component development program is based on an extension of high temperature gas-cooled reactor (HTGR) component technology; therefore, the GCFR development program is addressed primarily to components which differ in design and requirements from HTGR components. The principal differences in primary system components are due to the increase in helium coolant pressure level, which benefits system size and efficiency in the GCFR, and differences in the reactor internals and fuel handling systems due to the use of the compact metal-clad core. The purpose of this paper is to present an overview of the principal component design differences between the GCFR and HTGR and the consequent influences of these differences on GCFR component development programs. Development program plans are discussed and include those for the prestressed concrete reactor vessel (PCRV), the main helium circulator and its supporting systems, the steam generators, the reactor thermal shielding, and the fuel handling system. Facility requirements to support these development programs are also discussed. Studies to date show that GCFR component development continues to appear to be incremental in nature, and the required tests are adaptations of related HTGR test programs. (Auth.)

  3. Analysis of loss of coolant accident and emergency core cooling system

    International Nuclear Information System (INIS)

    Abe, Kiyoharu; Kobayashi, Kenji; Hayata, Kunihisa; Tasaka, Kanji; Shiba, Masayoshi

    1977-01-01

    In this paper, the analysis for the performance evaluation of emergency core cooling system is described, which is the safety protection device to the loss of coolant accidents due to the break of primary cooling pipings of light water reactors. In the LOCA analysis for the performance evaluation of ECCS, it must be shown that a reactor core keeps the form which can be cooled with the ECCS in case of LOCA, and the overheat of the core can be prevented. Namely, the shattering of fuel cladding tubes is never to occur, and for the purpose, the maximum temperature of Zircaloy 2 or 4 cladding tubes must be limited to 1200 deg C, and the relative thickness of oxide film must be below 15%. The calculation for determining the temperature of cladding tubes in case of the LOCA in BWRs and PWRs is explained. First, the primary cooling system, the ECCS and the related installations of BWRs and PWRs are outlined. The code systems for LOCA/ECCS analysis are divid ed into several steps, such as blowdown process, reflooding process and heatup calculation. The examples of the sensitivity analysis of the codes are shown. The LOCA experiments carried out so far in Japan and foreign countries and the LOCA analysis of a BWR with RELAP-4J code are described. The guidance for the performance evaluation of ECCS was established in 1975 by the Reactor Safety Deliberation Committee in Japan, and the contents are quoted. (Kako, I.)

  4. Heat transfer tests of ribbed surfaces for gas-cooled reactors

    International Nuclear Information System (INIS)

    Klepper, O.H.

    1975-07-01

    The performance of gas-cooled reactors is often limited by the heat transfer in the reactor core. Means for modifying core heat transfer surfaces to improve their performance were investigated. The 0.3-in.-OD stainless steel clad heater rods were photo-etched to produce external ribs 0.006 in. high and 0.12 in. wide with a pitch of 0.072 in. Helical ribs with a helix angle of 37 0 (to promote interchannel flow mixing in a multirod array) were provided on one surface. For comparison purposes, a transversely ribbed surface and a smooth rod were also studied. The test surfaces were 49 in. long with a 24-in. heated region, concentrically arranged inside a smooth 0.602-in.-ID stainless steel tube. Nitrogen gas at pressures up to 400 psig was used as the coolant; the linear heat rating ranged to 6.8 kW/ft at surface temperatures up to 1400 0 F; T/sub w/T/sub b/ varied from 1.2 to 2.4 at Re values up to 450,000. Annulus results were recalculated for rod geometry using two different transformations. Good agreement was observed with applicable literature values. The effectiveness of the surfaces was assessed as the ratio E of the heat transfer coefficients of the roughened rods to that of a smooth rod at the same pumping power. The effectiveness of the spiral ribs ranged from 1.3 to 1.4, and from 1.2 to 1.4 for the transverse ribs, spanning Re values from 60,000 to 400,000. These data include variations introduced by alternate transformation methods that were used to make annulus test results applicable to rod geometry. The surfaces investigated in these tests were considered for fast gas-cooled reactors; however, the range of parameters studied also applies to heat transfer from ribbed rod-type fuel elements in thermal gas-cooled reactors. (U.S.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-01-01

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

  6. The effect of lower body cooling on the changes in three core temperature indices

    International Nuclear Information System (INIS)

    Basset, F A; Cahill, F; Handrigan, G; DuCharme, M B; Cheung, S S

    2011-01-01

    Rectal (T re ), ear canal (T ear ) and esophageal (T es ) temperatures have been used in the literature as core temperature indices in humans. The aim of the study was to investigate if localized lower body cooling would have a different effect on each of these measurements. We hypothesized that prolonged lower body surface cooling will result in a localized cooling effect for the rectal temperature not reflected in the other core measurement sites. Twelve participants (mean ± SD; 26.8 ± 6.0 years; 82.6 ± 13.9 kg; 179 ± 10 cm, BSA = 2.00 ± 0.21 m 2 ) attended one experimental session consisting of sitting on a rubberized raft floor surface suspended in 5 °C water in a thermoneutral air environment (∼21.5 ± 0.5 °C). Experimental conditions were (a) a baseline phase during which participants were seated for 15 min in an upright position on an insulated pad (1.408 K . m 2 . W −1 ); (b) a cooling phase during which participants were exposed to the cooling surface for 2 h, and (c) an insulation phase during which the baseline condition was repeated for 1 h. Temperature data were collected at 1 Hz, reduced to 1 min averages, and transformed from absolute values to a change in temperature from baseline (15 min average). Metabolic data were collected breath-by-breath and integrated over the same temperature epoch. Within the baseline phase no significant change was found between the three indices of core temperature. By the end of the cooling phase, T re was significantly lower (Δ = −1.0 ± 0.4 °C) from baseline values than from T ear (Δ = −0.3 ± 0.3 °C) and T es (Δ = −0.1 ± 0.3 °C). T re continued to decrease during the insulation phase from Δ −1.0 ± 0.4 °C to as low as Δ −1.4 ± 0.5 °C. By the end of the insulation phase T re had slightly risen back to Δ −1.3 ± 0.4 °C but remained significantly different from baseline values and from the other two core measures. Metabolic data showed no variation throughout the experiment. In

  7. Gas cooled reactor assessment. Volume II. Final report, February 9, 1976--June 30, 1976

    International Nuclear Information System (INIS)

    1976-08-01

    This report was prepared to document the estimated power plant capital and operating costs, and the safety and environmental assessments used in support of the Gas Cooled Reactor Assessment performed by Arthur D. Little, Inc. (ADL), for the U.S. Energy Research and Development Administration. The gas-cooled reactor technologies investigated include: the High Temperature Gas Reactor Steam Cycle (HTGR-SC), the HTGR Direct Cycle (HTGR-DC), the Very High Temperature Reactor (VHTR) and the Gas Cooled Fast Reactor (GCFR). Reference technologies used for comparison include: Light Water Reactors (LWR), the Liquid Metal Fast Breeder Reactor (LMFBR), conventional coal-fired steam plants, and coal combustion for process heat

  8. A system for the discharge of gas bubbles from the coolant flow of a nuclear reactor cooled by forced circulation

    International Nuclear Information System (INIS)

    Markfort, D.; Kaiser, A.; Dohmen, A.

    1975-01-01

    In a reactor cooled by forced circulation the gas bubbles carried along with the coolant flow are separated before entering the reactor core or forced away into the external zones. For this purpose the coolant is radially guided into a plenum below the core and deflected to a tangential direction by means of flow guide elements. The flow runs spirally downwards. On the bubbles, during their dwell time in this channel, the buoyant force and a force towards the axis of symmetry of the tank are exerted. The major part of the coolant is directed into a radial direction by means of a guiding apparatus in the lower section of the channel and guided through a chimney in the plenum to the center of the reactor core. This inner chimney is enclosed by an outer chimney for the core edge zones through which coolant with a small share of bubbles is taken away. (RW) [de

  9. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    Science.gov (United States)

    Ilham, Muhammad; Su'ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  10. 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.

  11. A novel nuclear combined power and cooling system integrating high temperature gas-cooled reactor with ammonia–water cycle

    International Nuclear Information System (INIS)

    Luo, Chending; Zhao, Fuqiang; Zhang, Na

    2014-01-01

    Highlights: • We propose a novel nuclear ammonia–water power and cooling cogeneration system. • The high temperature reactor is inherently safe, with exhaust heat fully recovered. • The thermal performances are improved compared with nuclear combined cycle. • The base case attains an energy efficiency of 69.9% and exergy efficiency of 72.5%. • Energy conservation and emission reduction are achieved in this cogeneration way. - Abstract: A nuclear ammonia–water power and refrigeration cogeneration system (NAPR) has been proposed and analyzed in this paper. It consists of a closed high temperature gas-cooled reactor (HTGR) topping Brayton cycle and a modified ammonia water power/refrigeration combined bottoming cycle (APR). The HTGR is an inherently safe reactor, and thus could be stable, flexible and suitable for various energy supply situation, and its exhaust heat is fully recovered by the mixture of ammonia and water in the bottoming cycle. To reduce exergy losses and enhance outputs, the ammonia concentrations of the bottoming cycle working fluid are optimized in both power and refrigeration processes. With the HTGR of 200 MW thermal capacity and 900 °C/70 bar reactor-core-outlet helium, the system achieves 88.8 MW net electrical output and 9.27 MW refrigeration capacity, and also attains an energy efficiency of 69.9% and exergy efficiency of 72.5%, which are higher by 5.3%-points and 2.6%-points as compared with the nuclear combined cycle (NCC, like a conventional gas/steam power-only combined cycle while the topping cycle is a closed HTGR Brayton cycle) with the same nuclear energy input. Compared with conventional separate power and refrigeration generation systems, the fossil fuel saving (based on CH 4 ) and CO 2 emission reduction of base-case NAPR could reach ∼9.66 × 10 4 t/y and ∼26.6 × 10 4 t/y, respectively. The system integration accomplishes the safe and high-efficiency utilization of nuclear energy by power and refrigeration

  12. Crossflow characteristics of flange type fuel element for very high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Takizuka, Takakazu; Kaburaki, Hideo; Suzuki, Kunihiko; Nakamura, Masahide.

    1987-01-01

    Fuel element design incorporating mating flanges at block end faces has the potential to improve thermal hydraulic performance of a VHTR (very high temperature gas-cooled reactor) core. As part of research and development efforts to establish flange type fuel element design, experiments and analyses were carried out on crossflow through interface gap between elements. Air at atmospheric pressure and ambient temperature was used as a fluid. Crossflow loss coefficient factors were obtained with three test models, having different flange mating clearances, for various interface gap configurations, gap widths and block misalignments. It was found that crossflow loss coefficient factors for flange type fuel element were much larger than those for conventional flat-faced element. Numerical analyses were also made using a simple model devised to represent the crossflow path at the fuel element interface. The close agreement between numerical results and experimental data indicated that this model could predict well the crossflow characteristics of the flange type fuel element. (author)

  13. Safety and licensing of MHTGR [Modular High Temperature Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Silady, F.A.; Millunzi, A.C.; Kelley, A.P. Jr.; Cunliffe, J.

    1987-07-01

    The Modular High Temperature Gas Cooled Reactor (MHTGR) design meets stringent top-level regulatory and user safety requirements that require that the normal and off-normal operation of the plant not disturb the public's day-to-day activities. Quantitative, top-level regulatory criteria have been specified from US NRC and EPA sources to guide the design. The user/utility group has further specified that these criteria be met at the plant boundary. The focus of the safety approach has then been centered on retaining the radionuclide inventory within the fuel by removing core heat, controlling chemical attack, and by controlling heat generation. The MHTGR is shown to passively meet the stringent requirements with margin. No operator action is required and the plant is insensitive to operator error

  14. Benchmark Analysis Of The High Temperature Gas Cooled Reactors Using Monte Carlo Technique

    International Nuclear Information System (INIS)

    Nguyen Kien Cuong; Huda, M.Q.

    2008-01-01

    Information about several past and present experimental and prototypical facilities based on High Temperature Gas-Cooled Reactor (HTGR) concepts have been examined to assess the potential of these facilities for use in this benchmarking effort. Both reactors and critical facilities applicable to pebble-bed type cores have been considered. Two facilities - HTR-PROTEUS of Switzerland and HTR-10 of China and one conceptual design from Germany - HTR-PAP20 - appear to have the greatest potential for use in benchmarking the codes. This study presents the benchmark analysis of these reactors technologies by using MCNP4C2 and MVP/GMVP Codes to support the evaluation and future development of HTGRs. The ultimate objective of this work is to identify and develop new capabilities needed to support Generation IV initiative. (author)

  15. The rate of diffusion into advanced gas cooled reactor moderator bricks: an equivalent cylinder model

    International Nuclear Information System (INIS)

    Kyte, W.S.

    1980-01-01

    The graphite moderator bricks which make up the moderator of an advanced gas-cooled nuclear reactor (AGR) are of many different and complex shapes. Many physico-chemical processes that occur within these porous bricks include a diffusional step and thus to model these processes it is necessary to solve the diffusion equation (with chemical reaction) in a porous medium of complex shape. A finite element technique is applied to calculating the rate at which nitrogen diffuses into and out of the porous moderator graphite during operation of a shutdown procedure for an AGR. However, the finite element method suffers from several disadvantages that undermine its general usefulness for calculating rates of diffusion in AGR moderator cores. A model which overcomes some of these disadvantages is presented (the equivalent cylinder model) and it is shown that this gives good results for a variety of different boundary and initial conditions

  16. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    International Nuclear Information System (INIS)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor

  17. Perspectives on Understanding and Verifying the Safety Terrain of Modular High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Carlson, Donald E.

    2014-01-01

    The inherent safety characteristics of modular high temperature gas-cooled reactors (HTGRs) are conceptually well known and are largely supported by insights from past and ongoing research. This paper offers perspectives on selected issues in areas where further analysis and testing achievable within existing research and demonstration programs could help address residual uncertainties and better support the analysis of safety performance and the regulatory assessment of defense in depth. Areas considered include the evaluation of normal and anomalous core operating conditions and the analysis of accidents involving coolant depressurization, air ingress, moisture ingress, and reactivity insertion. In addition to discussing associated uncertainties and potential measures to address them, the paper also proposes supplemental “safety terrain” studies that would use realistic assessments of postulated extreme event sequences to establish a more comprehensive understanding of the inherent behaviors and ultimate safety capabilities of modular HTGRs. (author)

  18. Reactors Dynamic analysis Due to Reactivity of The RSG-Gas at One Line Cooling Mode

    International Nuclear Information System (INIS)

    Hastuti, Endiah Puji

    2003-01-01

    In the frame of minimizing the operation-cost, operation mode using one line cooling system is being evaluated. Maximum reactor power has been determined and steady state and LOFA transient analysis have also been done. To complete those analyses, the reactivity analysis was done by means of a core dynamic and thermal hydraulic code, PARET-ANL. Accident simulation was done. by a ramp reactivity accident due to control rod withdrawal. Reactivity analysis was carried out at two power range i.e. low and high power level, by imposing one line mode reactor protection limits. The results show that technically, the RSG-Gas can be operated safely using one line mode

  19. Material development for gas-cooled high temperature reactors for the production of nuclear process heat

    International Nuclear Information System (INIS)

    Nickel, H.

    1977-04-01

    In the framework of the material development for gas-cooled high temperature reactors, considerable investigations of the materials for the reactor core and the primary cicuit are being conducted. Concerning the core components, the current state-of-the-art and the objectives of the development work on the spherical fuel elements, coated particles and structural graphite are discussed. As an example of the structural graphite, the non-replaceable reflector of the process heat reactor is discussed. The primary circuit will be constructed mainly from metallic materials, although some ceramics are also being considered. Components of interest are hot gas ducts, liners, methane reformer tubes and helium-helium intermediate heat exchangers. The gaseous impurities present in the helium coolant may cause oxidation and carburization of the nickel-base and iron-base alloys envisaged for use in these components, with a possible associated adverse effect on the mechanical properties such as creep and fatigue. Test capacity has therefore been installed to investigate materials behaviour in simulated reactor helium under both constant and alternating stress conditions. The first results on the creep behaviour of several alloys in impure helium are presented and discussed. (orig./GSC) [de

  20. New deployment of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Tsuchie, Yasuo; Kunitomi, Kazuhiko; Shiozawa, Shusaku; Konuki, Kaoru; Inagaki, Yoshiyuki; Hayakawa, Hitoshi

    2002-01-01

    The high temperature gas-cooled reactor (HTGR) is now under a condition difficult to know it well, because of considering not only power generation, but also diverse applications of its nuclear heat, of having extremely different safe principle from that of conventional reactors, of having two types of pebble-bed and block which are extremely different types, of promoting its construction plan in South Africa, of including its application to disposition of Russian surplus weapons plutonium of less reporting HTTR in Japan in spite of its full operation, and so on. However, HTGR is expected for an extremely important nuclear reactor aiming at the next coming one of LWR. HTGR which is late started and developed under complete private leading, is strongly conscious at environmental problem since its beginning. Before 30 years when large scale HTGR was expected to operate, it advertised a merit to reduce wasted heat because of its high temperature. As ratio occupied by electricity expands among application of energies, ratio occupied by the other energies are larger. When considering applications except electric power, high temperature thermal energy from HTGR can be thought wider applications than that from LWR and so on. (G.K.)

  1. Description of the advanced gas cooled type of reactor (AGR)

    Energy Technology Data Exchange (ETDEWEB)

    Nonboel, E. [Risoe National Lab., Roskilde (Denmark)

    1996-11-01

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: `Reactors in Nordic Surroundings`, which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs.

  2. Description of the advanced gas cooled type of reactor (AGR)

    International Nuclear Information System (INIS)

    Nonboel, E.

    1996-11-01

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs

  3. Inherently safe high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Yamada, Masao; Hayakawa, Hitoshi

    1987-01-01

    It is recognized in general that High Temperature Gas-cooled Reactors have remarkable characteristics in inherent safety and it is well known that credits of the time margin have been admitted for accident evaluation in the licensing of the currently operating prototype HTGRs (300 MWe class). Recently, more inherently safe HTGRs are being developed in various countries and drawing attention on their possibility for urban siting. The inherent safety characteristics of these HTRs differ each other depending on their design philosophy and on the features of the components/structures which constitute the plant. At first, the specific features/characteristics of the elemental components/structures of the HTRs are explained one by one and then the overall safety features/characteristics of these HTR plants are explained in connection with their design philosophy and combination of the elemental features. Taking the KWU/Interatom Modular Reactor System as an example, the particular design philosophy and safety characteristics of the inherently safe HTR are explained with a result of preliminary evaluation on the possibility of siting close to densely populated area. (author)

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

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

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

  5. Small high temperature gas-cooled reactors with innovative nuclear burning

    International Nuclear Information System (INIS)

    Liem, Peng Hong; Ismail; Sekimoto, Hiroshi

    2008-01-01

    Since the innovative concept of CANDLE (Constant Axial shape of Neutron Flux, nuclide densities and power shape During Life of Energy producing reactor) burning strategy was proposed, intensive research works have been continuously conducted to evaluate the feasibility and the performance of the burning strategy on both fast and thermal reactors. We learned that one potential application of the burning strategy for thermal reactors is for the High Temperature Gas-Cooled Reactors (HTGR) with prismatic/block-type fuel elements. Several characteristics of CANDLE burning strategy such as constant reactor characteristics during burn-up, no need for burn-up reactivity control mechanism, proportionality of core height with core lifetime, sub-criticality of fresh fuel elements, etc. enable us to design small sized HTGR with a high degree of safety easiness of operation and maintenance, and long core lifetime which are required for introducing the reactors into remote areas or developing countries with limited infrastructures and resources. In the present work, we report our evaluation results on small sized block-type HTGR designs with CANDLE burning strategy and compared with other existing small HTGR designs including the ones with pebble fuel elements, under both uranium and thorium fuel cycles. (author)

  6. Severe water ingress accident analysis for a Modular High Temperature Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Scherer, Winfried

    1997-01-01

    This paper analyzes the severe water ingress accidents in the SIEMENS 200MW Modular High Temperature Gas Cooled Reactor (HTR-Module) under the assumption of no active safety protection systems in order to find the safety margin of the current HTR-Module design. A water, steam and helium multi-phase cavity model is originally developed and implemented in the DSNP simulation system. The developed DSNP system is used to simulate the primary circuit of HTR-Module power plant. The comparisons of the models with the TINTE calculations validate the current simulation. After analyzing the effects of blower separation on water droplets, the wall heat storage, etc., it is found that the maximum H 2 O density increase rate in the reactor core is smaller than 0.3 kg/(m 3 s). The liquid water vaporization in the steam generator and H 2 O transport from the steam generator to the reactor core reduces the impulse of the H 2 O in the reactor core. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600degC was not reached in any case. (author)

  7. Utilization of multi-purpose high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kawada, Osamu; Onuki, Yoshiaki; Wasaoka, Takeshi.

    1974-01-01

    Concerning the utilization of multi-purpose high temperature gas-cooled reactors, the electric power generation with gas turbines is described: features of HTR-He gas turbine power plants; the state of development of He gas turbines; and combined cycle with gas turbines and steam turbines. The features of gas turbines concern heat dissipation into the environment and the mode of load operation. Outstanding work in the development of He gas turbines is that in Hochtemperatur Helium-Turbine Project in West Germany. The power generation with combined gas turbines and steam turbines appears to be superior to that with gas turbines alone. (Mori, K.)

  8. ETDR, The European Union's Experimental Gas-Cooled Fast Reactor Project

    International Nuclear Information System (INIS)

    Poette, Christian; Brun-Magaud, Valerie; Morin, Franck; Dor, Isabelle; Pignatel, Jean-Francois; Bertrand, Frederic; Stainsby, Richard; Pelloni, Sandro; Every, Denis; Da Cruz, Dirceu

    2008-01-01

    In the Gas-Cooled Fast Reactor (GFR) development plan, the Experimental Technology Demonstration Reactor (ETDR) is the first necessary step towards the electricity generating prototype GFR. It is a low power (∼50 MWth) Helium cooled fast reactor. The pre-conceptual design of the ETDR is shared between European partners through the GCFR Specifically Targeted Research Project (STREP) within the European Commission's 6. R and D Framework Program. After recalling the place of ETDR in the GFR development plan, the main reactor objectives, the role of the European partners in the different design and safety tasks, the paper will give an overview of the current design with recent progresses in various areas like: - Sub-assembly technology for the starting core (pin bundle with MOX fuel and stainless steel cladding). - The design of experimental advanced ceramic GFR fuel sub-assemblies included in several locations of the starting core. - Starting Core reactivity management studies model including experimental GFR sub-assemblies. - Neutron and radiation shielding calculations using a specific MCNP model. The model allows evaluation of the neutron doses for the vessel and internals and radiation doses for maintenance operations. - System design and safety considerations, with a reactor architecture largely influenced by the Decay Heat Removal strategy (DHR) for de-pressurized accidents. The design of the reactor raises a number of issues in terms of fuel, neutronics, thermal-hydraulics codes qualification as well as critical components (blowers, IHX, thermal barriers) qualification. An overview of the R and D development on codes and technology qualification program is presented. Finally, the status of international collaborations and their perspectives for the ETDR are mentioned. (authors)

  9. Analysis and prevention of water hammer for the emergency core cooling system

    International Nuclear Information System (INIS)

    Zhao Jun

    2008-01-01

    Emergency core cooling system (ECCS) is an engineered safety feature of nuclear power plant. If the water hammer happens during ECCS injection, the piping system may be broken. It will cause loss of ECC system and affect the safety of reactor core. Based on the functions and characteristics of ECCS and the theory of water hammer, the paper analyzed the potential risk of water hammer in ECCS in Qinshan III, and proposed modifications to prevent the water-hammer damage during ECCS injection. (authors)

  10. Application of a steam injector for passive emergency core cooling during a station blackout

    International Nuclear Information System (INIS)

    Heinze, D.; Behnke, L.; Schulenberg, T.

    2012-01-01

    One of the basic protection targets of reactor safety is the safe heat removal during normal operation but also following shut-down. Since the reactor accident in Fukushima an optimization of the plant robustness in case of beyond-design accident is performed. Special attention is given to the increase of time available for starting appropriate measures for emergency core cooling in case of a station blackout. The state-of the art in engineering and research is presented. Investigations on the applicability of a steam injector for passive emergency core cooling during a station blackout in BWR-type reactors have progressed, experiments on dynamic behavior of the injector are described. A precise design with respect to the thermal hydraulic boundary conditions has been performed.

  11. Improvement of the decay heat removal characteristics of the generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Epiney, A. S.

    2010-09-01

    The majority of NPPs worldwide are currently light water reactors, using ordinary water as both coolant and moderator. (...) For the longer-term future, viz. beyond the year 2030, Research and Development is currently ongoing on Generation IV NPPs, aimed at achieving closure of the nuclear fuel cycle, and hence both drastically improved utilization of fuel resources and minimization of long-lived radioactive wastes. Since the very beginning of the international cooperation on Generation IV, viz. the year 2000, the main research interest in Europe as regards the advanced fast-spectrum systems needed for achieving complete fuel cycle closure, has been for the Sodium-cooled Fast Reactor (SFR). However, the Gas-cooled Fast Reactor (GFR) is currently considered as the main back-up solution. Like the SFR, the GFR is an efficient breeder, also able to work as iso-breeder using simply natural uranium as feed and producing waste which is predominantly in the form of fission products. The main drawback of the GFR is the difficulty to evacuate decay heat following a loss-of-coolant accident (LOCA) due to the low thermal inertia of the core, as well as to the low coolant density. The present doctoral research focuses on the improvement of decay heat removal (DHR) for the Generation-IV GFR. The reference GFR system design considered in the thesis is the 2006 CEA concept, with a power of 2400 MWth. The CEA 2006 DHR strategy foresees, in all accidental cases (independent of the system pressure), that the reactor is shut down. For high pressure events, dedicated DHR loops with blowers and heat exchangers are designed to operate when the power conversion system cannot be used to provide acceptable core temperatures under natural convection conditions. For depressurized events, the strategy relies on a dedicated small containment (called the guard containment) providing an intermediate back-up pressure. The DHR blowers, designed to work under these pressure conditions, need to be

  12. New Sodium Cooled Long-Life Cores with Axially Multi-Driver Regions

    International Nuclear Information System (INIS)

    Hyun, Hae Ri; Hong, Ser Gi

    2014-01-01

    In this concept of long-life core (they are sometimes called B-B (Breed and Burn)), tall blanket is placed above the relatively short driver fuel. In the initial stage of burning, the power by fission is mostly generated in the driver region and it moves into the blanket region. The power and flux distributions that are highly peaked in the axial direction propagates slowly from the driver into the blanket region. This concept of long-life core fully utilizes the breeding of blanket in the fast spectra and it can achieve very high burnup of fuel. In this work, we introduce new sodium cooled longlife cores rating 600MWe (1800MWt). In these cores, the driver regions are heterogeneously placed into blanket region so as to achieve stabilized and less peaked axial power distribution as depletion proceeds. At present, our study is focused on only two axial driver regions but this concept can be easily extended onto the multi-driver region concept. The cores designed in this paper have two axial driver regions so as to have stabilized and less peaked axial power distributions as depletion proceeds. The results of the core design and analyses show that the cores have very long-lives longer than -49EFPYs and high discharge burnup higher than 200GWD/kg. Additionally, we considered a long-life core having no blanket. As expected, it was shown that these cores have stabilized and less peaked axial power distribution as the fuel depletes. However, the study shows that the cores having two driver regions still show high initial peaking of the axial power distributions and the core can be optimized by changing the driver fuel height

  13. Control rod shadowing and anti-shadowing effects in a large gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Girardin, G.; Chawla, R.; Rimpault, G.; Coddington, P.

    2007-01-01

    An investigation of control rod shadowing and anti-shadowing (interaction) effects has been carried out in the context of a design study of the control rod pattern for the large 2400 MWth Generation IV Gas-cooled Fast Reactor (GFR). For the calculations, the deterministic code system ERANOS-2.0 has been used, in association with a full core model including a European Fast Reactor (EFR)-type pattern for the control rods. More specifically, the core contains a total of 33 control (CSD) and safety (DSD) rods implemented in three banks: -1) a first bank of 6 CSD rods, placed at 64 cm from core centre in the inner fuel zone (Pu content 16.3 % vol.), -2) a safety bank consisting of 9 DSD rods, at an average distance of 118 cm, and -3) a third bank with 18 CSD rods, placed at 171 cm, i.e. at the interface between the inner and outer (Pu content 19.2 % vol.) core regions. Each control rod has been modelled as a homogeneous material containing 90%-enriched B 4 C, steel and helium. Considerable shadowing effects have been observed between the first bank and the safety bank, as also between individual rods within the first bank. Large anti-shadowing effects take place in an even greater number of the studied rod configurations. The largest interaction is between the two CSD banks, the anti-shadowing value being 46% in this case, implying that the total rod worth is increased by a factor of almost 2 when compared to the sum of the individual bank values. Additional investigations have been performed, in particular the computation of the first order eigenvalue and the eigenvalue separation. The main finding is that the interactions are lower when one of the control rod banks is located at a radial position corresponding to half the core radius. (authors)

  14. Transient Analysis of a Gas-cooled Fast Reactor for Single Control Assembly Withdrawal

    International Nuclear Information System (INIS)

    Choi, Hangbok

    2014-01-01

    The Energy Multiplier Module (EMZ) system response has been evaluated for control assembly withdrawal transients. Currently the EM2 core is equipped with six cylindrical drum-type control assemblies in the reflector zone for excess reactivity control and power maneuvering during the operating core life. This study investigates the system response to the control assembly withdrawal accident with various rotational speeds and reactivity worth to determine feasible control assembly design requirements from the physics viewpoint. The simulations have been conducted for single control assembly withdrawal transients without scram by a gas-cooled reactor plant simulator, which is based on a simplified plant nodal model, including the point reactor kinetics, single channel core thermal-fluid model, and a turbo-machinery performance model. Simulations were conducted for the middle-of- cycle core, when the excess reactivity of the core is the highest. Control assembly withdrawal times were varied from 1 (runaway) to 180 sec and reactivity worth was varied from 100 to 400 pcm. For a single control assembly withdrawal, the simulation has shown that the peak fuel temperature is expected to be ~1820°C when the assembly worth is 200 pcm and the runaway time is 1 sec per 180 degree rotation. The peak temperature could be reduced to ~1780°C if the assembly is rotated out in a moderate speed such as 1 degree/sec. These peak temperatures give a thermal margin of 22 to 24% to the melting point of uranium carbide fuel. The results also indicate that the current design with a single control assembly worth of 314 pcm may need adjustments in the future design. (author)

  15. The core design of ALFRED, a demonstrator for the European lead-cooled reactors

    International Nuclear Information System (INIS)

    Grasso, G.; Petrovich, C.; Mattioli, D.; Artioli, C.; Sciora, P.; Gugiu, D.; Bandini, G.; Bubelis, E.; Mikityuk, K.

    2014-01-01

    Highlights: • The design for the lead fast reactor is conceived in a comprehensive approach. • Neutronic, thermal-hydraulic, and transient analyses show promising results. • The system is designed to withstand even design extension conditions accidents. • Activation products in lead, including polonium, are evaluated. - Abstract: The European Union has recently co-funded the LEADER (Lead-cooled European Advanced DEmonstration Reactor) project, in the frame of which the preliminary designs of an industrial size lead-cooled reactor (1500 MW th ) and of its demonstrator reactor (300 MW th ) were developed. The latter is called ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) and its core, as designed and characterized in the project, is presented here. The core parameters have been fixed in a comprehensive approach taking into account the main technological constraints and goals of the system from the very beginning: the limiting temperature of the clad and of the fuel, the Pu enrichment, the achievement of a burn-up of 100 GWd/t, the respect of the integrity of the system even in design extension conditions (DEC). After the general core design has been fixed, it has been characterized from the neutronic point of view by two independent codes (MCNPX and ERANOS), whose results are compared. The power deposition and the reactivity coefficient calculations have been used respectively as input for the thermal-hydraulic analysis (TRACE, CFD and ANTEO codes) and for some preliminary transient calculations (RELAP, CATHARE and SIM-LFR codes). The results of the lead activation analysis are also presented (FISPACT code). Some issues of the core design are to be reviewed and improved, uncertainties are still to be evaluated, but the verifications performed so far confirm the promising safety features of the lead-cooled fast reactors

  16. The core design of ALFRED, a demonstrator for the European lead-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grasso, G., E-mail: giacomo.grasso@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Petrovich, C., E-mail: carlo.petrovich@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Mattioli, D., E-mail: davide.mattioli@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Artioli, C., E-mail: carlo.artioli@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Sciora, P., E-mail: pierre.sciora@cea.fr [CEA (Alternative Energies and Atomic Energy Commission), DEN, DER, 13108 St Paul lez Durance (France); Gugiu, D., E-mail: daniela.gugiu@nuclear.ro [RATEN-ICN (Institute for Nuclear Research), Cod 115400 Mioveni, Str. Campului, 1, Jud. Arges (Romania); Bandini, G., E-mail: giacomino.bandini@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Bubelis, E., E-mail: evaldas.bubelis@kit.edu [KIT (Karlsruhe Institute of Technology), Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [PSI (Paul Scherrer Institute), OHSA/D11, 5232 Villigen PSI (Switzerland)

    2014-10-15

    Highlights: • The design for the lead fast reactor is conceived in a comprehensive approach. • Neutronic, thermal-hydraulic, and transient analyses show promising results. • The system is designed to withstand even design extension conditions accidents. • Activation products in lead, including polonium, are evaluated. - Abstract: The European Union has recently co-funded the LEADER (Lead-cooled European Advanced DEmonstration Reactor) project, in the frame of which the preliminary designs of an industrial size lead-cooled reactor (1500 MW{sub th}) and of its demonstrator reactor (300 MW{sub th}) were developed. The latter is called ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) and its core, as designed and characterized in the project, is presented here. The core parameters have been fixed in a comprehensive approach taking into account the main technological constraints and goals of the system from the very beginning: the limiting temperature of the clad and of the fuel, the Pu enrichment, the achievement of a burn-up of 100 GWd/t, the respect of the integrity of the system even in design extension conditions (DEC). After the general core design has been fixed, it has been characterized from the neutronic point of view by two independent codes (MCNPX and ERANOS), whose results are compared. The power deposition and the reactivity coefficient calculations have been used respectively as input for the thermal-hydraulic analysis (TRACE, CFD and ANTEO codes) and for some preliminary transient calculations (RELAP, CATHARE and SIM-LFR codes). The results of the lead activation analysis are also presented (FISPACT code). Some issues of the core design are to be reviewed and improved, uncertainties are still to be evaluated, but the verifications performed so far confirm the promising safety features of the lead-cooled fast reactors.

  17. Mount Elbert Gas Hydrate Stratigraphic Test Well, Alaska North Slope: Coring operations, core sedimentology, and lithostratigraphy

    Science.gov (United States)

    Rose, K.; Boswell, R.; Collett, T.

    2011-01-01

    In February 2007, BP Exploration (Alaska), the U.S. Department of Energy, and the U.S. Geological Survey completed the BPXA-DOE-USGS Mount Elbert Gas Hydrate Stratigraphic Test Well (Mount Elbert well) in the Milne Point Unit on the Alaska North Slope. The program achieved its primary goals of validating the pre-drill estimates of gas hydrate occurrence and thickness based on 3-D seismic interpretations and wireline log correlations and collecting a comprehensive suite of logging, coring, and pressure testing data. The upper section of the Mount Elbert well was drilled through the base of ice-bearing permafrost to a casing point of 594??m (1950??ft), approximately 15??m (50??ft) above the top of the targeted reservoir interval. The lower portion of the well was continuously cored from 606??m (1987??ft) to 760??m (2494??ft) and drilled to a total depth of 914??m. Ice-bearing permafrost extends to a depth of roughly 536??m and the base of gas hydrate stability is interpreted to extend to a depth of 870??m. Coring through the targeted gas hydrate bearing reservoirs was completed using a wireline-retrievable system. The coring program achieved 85% recovery of 7.6??cm (3??in) diameter core through 154??m (504??ft) of the hole. An onsite team processed the cores, collecting and preserving approximately 250 sub-samples for analyses of pore water geochemistry, microbiology, gas chemistry, petrophysical analysis, and thermal and physical properties. Eleven samples were immediately transferred to either methane-charged pressure vessels or liquid nitrogen for future study of the preserved gas hydrate. Additional offsite sampling, analyses, and detailed description of the cores were also conducted. Based on this work, one lithostratigraphic unit with eight subunits was identified across the cored interval. Subunits II and Va comprise the majority of the reservoir facies and are dominantly very fine to fine, moderately sorted, quartz, feldspar, and lithic fragment-bearing to

  18. Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

    International Nuclear Information System (INIS)

    Mella, R.; Wenman, M.R.

    2013-01-01

    Thermo-mechanical contributions to pellet–clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS’s well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used

  19. Benchmark for Neutronic Analysis of Sodium-cooled Fast Reactor Cores with Various Fuel Types and Core Sizes

    International Nuclear Information System (INIS)

    Stauff, N.E.; Kim, T.K.; Taiwo, T.A.; Buiron, L.; Rimpault, G.; Brun, E.; Lee, Y.K.; Pataki, I.; Kereszturi, A.; Tota, A.; Parisi, C.; Fridman, E.; Guilliard, N.; Kugo, T.; Sugino, K.; Uematsu, M.M.; Ponomarev, A.; Messaoudi, N.; Lin Tan, R.; Kozlowski, T.; Bernnat, W.; Blanchet, D.; Brun, E.; Buiron, L.; Fridman, E.; Guilliard, N.; Kereszturi, A.; Kim, T.K.; Kozlowski, T.; Kugo, T.; Lee, Y.K.; Lin Tan, R.; Messaoudi, N.; Parisi, C.; Pataki, I.; Ponomarev, A.; Rimpault, G.; Stauff, N.E.; Sugino, K.; Taiwo, T.A.; Tota, A.; Uematsu, M.M.; Monti, S.; Yamaji, A.; Nakahara, Y.; Gulliford, J.

    2016-01-01

    One of the foremost Generation IV International Forum (GIF) objectives is to design nuclear reactor cores that can passively avoid damage of the reactor when control rods fail to scram in response to postulated accident initiators (e.g. inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core. Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS), the Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force was proposed to evaluate core performance characteristics of several Generation IV Sodium-cooled Fast Reactor (SFR) concepts. A set of four numerical benchmark cases was initially developed with different core sizes and fuel types in order to perform neutronic characterisation, evaluation of the feedback coefficients and transient calculations. Two 'large' SFR core designs were proposed by CEA: those generate 3 600 MW(th) and employ oxide and carbide fuel technologies. Two 'medium' SFR core designs proposed by ANL complete the set. These medium SFR cores generate 1 000 MW(th) and employ oxide and metallic fuel technologies. The present report summarises the results obtained by the WPRS for the neutronic characterisation benchmark exercise proposed. The benchmark definition is detailed in Chapter 2. Eleven institutions contributed to this benchmark: Argonne National Laboratory (ANL), Commissariat a l'energie atomique et aux energies alternatives (CEA of Cadarache), Commissariat a l'energie atomique et aux energies alternatives (CEA of Saclay), Centre for Energy Research (CER-EK), Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Helmholtz Zentrum Dresden Rossendorf (HZDR), Institute of Nuclear Technology and Energy Systems (IKE), Japan Atomic Energy Agency (JAEA), Karlsruhe Institute of Technology (KIT

  20. RCCS Experiments and Validation for High Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Chang Oh; Cliff Davis; Goon C. Park

    2007-01-01

    A reactor cavity cooling system (RCCS), an air-cooled helical coil RCCS unit immersed in the water pool, was proposed to overcome the disadvantages of the weak cooling ability of air-cooled RCCS and the complex structure of water-cooled RCCS for the high temperature gas-cooled reactor (HTGR). An experimental apparatus was constructed to investigate the various heat transfer phenomena in the water pool type RCCS, such as the natural convection of air inside the cavity, radiation in the cavity, the natural convection of water in the water pool and the forced convection of air in the cooling pipe. The RCCS experimental results were compared with published correlations. The CFX code was validated using data from the air-cooled portion of the RCCS. The RELAP5 code was validated using measured temperatures from the reactor vessel and cavity walls

  1. Gas-fired cogeneration and cooling: new study identifies major benefits

    International Nuclear Information System (INIS)

    Watt, G.

    2001-01-01

    A research paper- 'Gas Fired Cogeneration and Cooling: Markets, Technologies and Greenhouse Gas Savings'- launched at last month's Australian Gas Association 2001 Convention, reveals that gas cooling could replace 25 PJ of electricity summer demand, and reduce greenhouse gas emissions by 58 percent compared with electrical technologies. Commissioned by the AGA's Gas Cooling Task Force and supported by the Sustainable Energy Authority of Victoria and the Sustainable Energy Development Authority of NSW, the study examined market opportunities and environmental outcomes for the combined gas cogeneration and cooling technologies. It shows that the penetration of gas into the distributed cooling and power generation market is being driven by the following developments: the uncertainty and volatility of electricity costs, particularly during summer, electricity market structural changes which encourage distributed generation, high and uncertain world oil prices, the relative stability of Australian gas prices, the encouragement of demand and energy management strategies by regulators, greenhouse gas emission reduction policies, indoor air quality issues, product and productivity improvements in industry and CFC phase-out opportunities

  2. Proposal of a fast gas-cooled reactor using transuranics

    International Nuclear Information System (INIS)

    Macedo, Anderson Altair Pinheiro de

    2016-01-01

    In the last two decades, nations that have invested in research and energy generation through nuclear source have devoted part of their efforts in developing new technologies for nuclear reactors. Part of this investment focuses on new material testing, particularly regarding new fuels. In a world view that breaths sustainability, the reprocess and reuse of spent fuel from conventional reactors comes alive in nuclear technology, presenting itself as a real alternative of energy source for the latest generation of reactors. Different concepts of fourth generation reactors have been proposed and must meet some basic requirements, such as: extended burnup, improvement of passive safety, better radioactive waste management, possibility to use reprocessed fuel and proliferation resistance. In this context, the GFR (Gas-cooled Fast Reactor) is one of the future promises, presenting satisfactory neutronic results on the use of type of fuel (U, Pu) C. In the present work, the fuel of a traditional GFR reactor that uses (U, Pu)C was sub was replaced by a transuranic reprocessed fuel (TRU), obtained by non-proliferation reprocessing technology. The UO 2 fuel initially enriched by 3.1% was burned in a standard PWR, with full burn of 33,000 MWd/T. Afterward it was left in a pool for 5 years and finally reprocessed by UREX + method. Two fuels were studied and evaluated, one diluted with depleted uranium (U, TRU)C, and the other diluted in thorium (Th, TRU)C. Assessments were done in steady state and as well as during burning and were compared with results obtained using the standard fuel, (U, Pu) C. The outcome shows that the use of TRU as a fuel, in GFR type reactors, is a real possibility. The research was done using the SCALE 6.0 code modules. (author)

  3. Improved spacers for high temperature gas-cooled heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Nordstroem, L A [Swiss Federal Institute for Reactor Research, Wuerenlingen (Switzerland)

    1984-07-01

    Experimental and analytical investigations in the field of heat exchanger thermohydraulics have been performed at EIR for many years, Basic studies have been carried out on heat transfer and pressure loss for tube bundles of different geometries and tube surfaces. As a part of this overall R+D programme for heat exchangers, investigations have been carried out on spacer pressure loss in bundles with longitudinal flow. An analytical spacer pressure loss model was developed which could handle different types of subchannel within the bundle. The model has been evaluated against experiments, using about 25 spacers of widely differing geometries. In a gas-cooled reactor it is important to keep the pressure loss over the primary circuit heat exchangers to a minimum. In exchangers with grid spacers these contribute a significant proportion of the overall bundle losses. For example, in the HHT Recuperator, with a shell-side pressure loss of 3.5 % of the inlet pressure, the spacers cause about one half of this loss. Reducing the loss to, say, 2.5 % results in an overall increase in plant efficiency by more than 1 % - a significant improvement Preliminary analysis identified 5 geometries in particular which were chosen for experimental evaluation as part of a joint project with the SULZER Company, to develop a low pressure-loss spacer for HHT heat exchangers (longitudinal counter-flow He/He and He/H{sub 2}O designs). The aim of the tests was to verify the low pressure-loss characteristics of these spacer grid types, as well as the quality of the results calculated by the computer code analytical model. The experimental and analytical results are compared in this report.

  4. 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).

  5. 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)

  6. Automated-process gas-chromatograph system for use in accelerated corrosion testing of HTGR core-support posts

    International Nuclear Information System (INIS)

    Harper, R.E.; Herndon, P.G.

    1982-01-01

    An automated-process gas chromatograph is the heart of a gaseous-impurities-analysis system developed for the Oak Ridge National Laboratory Core Support Performance Test, at which graphite core-support posts for high-temperature gas-cooled fission reactors are being subjected to accelerated corrosion tests under tightly controlled conditions of atmosphere and temperature. Realistic estimation of in-core corrosion rates is critically dependent upon the accurate measurement of low concentrations of CO, CO 2 , CH 4 , H 2 , and O 2 in the predominantly helium atmosphere. In addition, the capital and labor investment associated with each test puts a premium upon the reliability of the analytical system, as excessive downtime or failure to obtain accurate data would result in unacceptable costs and schedule delays. After an extensive survey of available measurement techniques, gas chromatography was chosen for reasons of accuracy, flexibility, good-performance record, and cost

  7. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS's heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis

  8. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.

  9. Improvement of the decay heat removal characteristics of the generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Epiney, A.S.

    2010-01-01

    Gas cooling in nuclear power plants (NPPs) has a long history, the corresponding reactor types developed in France, the UK and the US having been thermal neutron spectrum systems using graphite as the moderator. The majority of NPPs worldwide, however, are currently light water reactors, using ordinary water as both coolant and moderator. These NPPs - of the so-called second generation - will soon need replacement, and a third generation is now being made available, offering increased safety while still based on light water technology. For the longer-term future, viz. beyond the year 2030, R and D is currently ongoing on Generation IV NPPs, aimed at achieving closure of the nuclear fuel cycle, and hence both drastically improved utilization of fuel resources and minimization of long-lived radioactive wastes. Like the SFR, the GFR is an efficient breeder, also able to work as iso-breeder using simply natural uranium as feed and producing waste which is predominantly in the form of fission products. The main drawback of the GFR is the difficulty to evacuate decay heat following a loss-of-coolant accident (LOCA) due to the low thermal inertia of the core, as well as to the low coolant density. The present doctoral research focuses on the improvement of decay heat removal (DHR) for the Generation-IV GFR. The reference GFR system design considered in the thesis is the 2006 CEA concept, with a power of 2400 MWth. The CEA 2006 DHR strategy foresees, in all accidental cases (independent of the system pressure), that the reactor is shut down. For high pressure events, dedicated DHR loops with blowers and heat exchangers are designed to operate when the power conversion system cannot be used to provide acceptable core temperatures under natural convection conditions. For de-pressurized events, the strategy relies on a dedicated small containment (called the guard containment) providing an intermediate back-up pressure. The DHR blowers, designed to work under these pressure

  10. Testing and analyses of a high temperature thermal barrier for gas-cooled reactors

    International Nuclear Information System (INIS)

    Black, W.E.; Betts, W.S.; Felten, P.

    1979-01-01

    A full size, multi-panel section of a thermal barrier system was fabricated from a nickel-base superalloy and a combination of fibrous blanket insulation materials for specific application in a steam cycle gas-cooled nuclear reactor. The 2.4 m square array was representative of the sidewall of the lower core outlet plenum and included coverplates, attachments, seals, and a simulated water-cooled liner. Testing was conducted in a reactor grade, helium-filled chamber at 816 0 C for 100 hours, which established a normal (baseline) condition; 982 0 C for 10 hours, which satisfied an emergency condition; 1093 0 C for 1 hour, which simulated a faulted condition; and 1260 0 C, which was a non-design condition test to demonstrate the temperature overshoot capability of the system. Post-test examination indicated: (1) an acceptable performance by the anti-friction chromium carbide (Cr 3 C 2 ) coating; (2) no significant galling between non-coated surfaces; (3) no distortion of attachment fixtures; (4) predictable coverplate deflection during the design conditions testing (normal, emergency, and faulted); and (5) considerable plastic deformation resulting from the near-incipient melting temperature. (orig.)

  11. Interim Status Report on the Design of the Gas-Cooled Fast Reactor (GFR)

    International Nuclear Information System (INIS)

    Weaver, K. D.

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report outlines the current design status of the GFR, and includes work done in the areas mentioned above

  12. Gas cooled fast reactor materials: compatibility and reaction kinetics of fuel/matrices couples

    International Nuclear Information System (INIS)

    Lechelle, J.; Aufore, L.; Basini, V.; Belin, R.; Vaudez, S.

    2004-01-01

    Fourth Generation Gas cooled Fast Reactor concept implies a fast neutron spectrum and aims to lead to an iso-generation of minor actinides. Criteria have been defined for these fuels such as: high core filling factor, efficient fuel cooling, low operation temperature, i.e. 400-850 deg C, good fission product retention, burn-ups in the range of 5-8 atom%, Pu content in the range of 15-25%. Materials matching this demand are considered: mixed uranium - plutonium nitrides and carbides as fuels, whereas TiN, TiC, ZrN, ZrC, SiC are investigated as inert matrices. Thermo-chemical compatibility studies have been carried out, mostly for (U,Pu)N/SiC and (U,Pu)N/TiN couples. They have been associated to matching diffusional studies. For the first studies, accidental reactor conditions have been chosen (1600 deg C) so as to select a couple. Results are presented in terms of nature and quantity of resulting phases identified by XRD and SEM for thermodynamical equilibrium experiments. (authors)

  13. Current design efforts for the gas-cooled fast reactor (GFR)

    International Nuclear Information System (INIS)

    Weaver, K.D.

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GCFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFC I) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GCFR: a helium-cooled, direct Brayton cycle power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GCFR. These are EURATOM (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, EURATOM (including the United Kingdom), France, Japan, and Switzerland have active research activities with respect to the GCFR. The research includes GCFR design and safety, and fuels/in-core materials/fuel cycle projects. This paper outlines the current design status of the GCFR, and includes work done in the areas mentioned above. (Author)

  14. A study of silver behavior in Gas-turbine High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Sawa, Kazuhiro; Tanaka, Toshiyuki

    1995-11-01

    A Gas-turbine High Temperature Gas-cooled Reactor (GT-HTGR) is one of the promising reactor systems of future HTGRs. In the design of GT-HTGR, behavior of fission products, especially of silver, is considered to be important from the view point of maintenance of gas-turbine. A study of silver behavior in the GT-HTGR was carried out based on current knowledge. The purposes of this study were to determine an importance of the silver problem quantitatively, countermeasures to the problem and items of future research and development which will be needed. In this study, inventory, fractional release from fuel, plateout in the primary circuit and radiation dose were evaluated, respectively. Based on this study, it is predicted that gamma-ray from plateout silver in gas-turbine system contributes about a half of total radiation dose after reactor shutdown. In future, more detail data for silver release from fuel, plateout behavior, etc. using the High Temperature Engineering Test Reactor (HTTR), for example, will be needed to carry out reasonable design. (author)

  15. Detection of gas-permeable fuel particles for highl 7490 temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Thiele, B.A.; Stinton, D.P.; Costanzo, D.A.

    1980-01-01

    Fuel for High-Temperature Gas-Cooled Reactors (HTGR) consists of uranium oxide-carbide and thoria microspheres coated with layers of pyrolytic carbon and silicon carbide. The pyrolytic carbon coatings must be gas-tight to perform properly during irradiation. Therefore, particles must be carefully characterized to determine the number of defective particles (ie bare kernels, and cracked or permeable coatings). Although techniques are available to determine the number of bare kernels or cracked coatings, no reliable technique has been available to measure coating permeability. This work describes a technique recently developed to determine whether coatings for a batch of particles are gas-tight or permeable. Although most of this study was performed on Biso-coated particles, the technique applies equally well to Triso-coated particles. About 150 randomly selected Biso-particle batches were studied in this work. These batches were first subjected to an 18-hr chlorination at 15000C, and the volatile thorium tetrachloride released through cracked or very permeable coatings was measured versus chlorination time. Chlorinated batches were also radiographed to detect any thorium that had migrated from the kernel into the coatings. From this work a technique was developed to determine coating permeability. This consists of an 18-hr chlorination of multiple samples without measurement of the heavy metal released. Each batch is then radiographed and the heavy metal diffusion within each particle is examined so it can be determined if a particle batch is permeable, slightly permeable, or gas-tight. (author)

  16. 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

  17. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  18. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Choi, Jin Ha; Kim, Myung Hyun

    2016-01-01

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  19. Coated particle fuel for high temperature gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl; Nabielek, Heinz [Research Center Julich (FZJ), Julich (Germany); Kendall, James M. [Global Virtual L1c, Prescott (United States)

    2007-10-15

    applications at 850-900 .deg. C and for process heat/hydrogen generation applications with 950 .deg. C outlet temperatures. There is a clear set of standards for modern high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a 500 {mu}m diameter UO{sub 2} kernel of 10% enrichment is surrounded by a 100 {mu}m thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of 35 {mu}m thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum 1600 .deg. C afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modern coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond 1600 .deg. C for a short period of time. This work should proceed at both national and international level.

  20. Coated particle fuel for high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Verfondern, Karl; Nabielek, Heinz; Kendall, James M.

    2007-01-01

    and for process heat/hydrogen generation applications with 950 .deg. C outlet temperatures. There is a clear set of standards for modern high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a 500 μm diameter UO 2 kernel of 10% enrichment is surrounded by a 100 μm thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of 35 μm thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum 1600 .deg. C afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modern coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond 1600 .deg. C for a short period of time. This work should proceed at both national and international level

  1. DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-TEMPERATURE GAS-COOLED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

    2016-11-01

    Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. The quantification of uncertainties in nuclear engineering has grown more important and the IAEA Coordinated Research Program (CRP) on High-Temperature Gas Cooled Reactor (HTGR) initiated in 2012 aims to investigate the various uncertainty quantification methodologies for this type of reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections co-variances. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations. Nominal results for the prismatic single block (Ex.I-2a) and super cell models (Ex.I-2c) have been obtained using the SCALE 6.1.3 two-dimensional lattice code NEWT coupled to the TRITON sequence for cross section generation. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex.I-2a and various models of Ex.I-2c is utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the INL coupled code PHISICS/RELAP5-3D, utilizing a fixed-temperature feedback for Ex. II-1a.. It is observed that the core power density does not vary significantly in shape, but the magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models.

  2. Numerical simulation of severe water ingress accidents in a modular high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Scherer, W.

    1996-01-01

    This report analyzes reverse water ingress accidents in the SIEMENS 200 MW Modular Pebble-Bed High Temperature Gas Cooled Reactor (HTR-MODULE) under the assumption of no active safety protection systems in order to find the safety margins of the current HTR-MODULE design and to realize a catastrophe-free nuclear technology. A water, steam and helium multi-phase cavity model is developed and implemented in the DSNP simulation system. The DSNP system is then used to simulate the primary and secondary circuit of a HTR-MODULE power plant. Comparisons of the model with experiments and with TINTE calculations serve as validation of the simulation. The analysis of the primary circuit tries to answer the question how fast the water enters the reactor core. It was found that the maximum H 2 O concentration increase in the reactor core is smaller than 0.3 kg/(m 3 s). The liquid water vaporization in the steam generator and H 2 O transport from the steam generator to the reactor core reduce the ingress velocity of the H 2 O into the reactor core. In order to answer the question how much water enters the primary circuit, the full cavitation of the feed water pumps is analyzed. It is found that if the secondary circuit is depressurized enough, the feed water pumps will be inherently stopped by the full cavitation. This limits the water to be pumped from the deaerator to the steam generator. A comprehensive simulation of the MODUL-HTR power plant then shows that the H 2 O inventory in the primary circuit can be limited to about 3000 kg. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600 C was not reached in any case. (orig.) [de

  3. 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

  4. Characteristic behaviour of Pebble Bed High Temperature Gas-cooled Reactors during water ingress events

    International Nuclear Information System (INIS)

    Khoza, Samukelisiwe N.; Serfontein, Dawid E.; Reitsma, Frederik

    2014-01-01

    The presence of water on the tube-side of the steam generators in high temperature gas-cooled reactors (HTGRs) with indirect cycle layouts presents a possibility for a penetration of neutron moderating steam into the core, which may cause a power excursion. This article presents results on the effect of water ingress into the core of the two South African Pebble Bed Modular Reactor design concepts, i.e. the PBMR-200 MW th and the PBMR-400 MW th developed by PBMR SOC Ltd. The VSOP 99/05 suite of codes was used for the simulation of this event. Partial steam vapour pressures were added in stages into the primary circuit in order to investigate the effect of water ingress on reactivity, power profiles and thermal neutron flux profiles. The effects of water ingress into the core are explained by increased neutron moderation, due to the addition of 1 H, which leads to a decrease in resonance capture by 238 U and therefore an increase in the multiplication factor. The more effective moderation of neutrons by definition reduces the fast neutron flux and increases the thermal flux in the core, i.e. leads to a softer spectrum. The more effective moderation also increases the average increase in lethargy between collisions of a neutron with successive fuel kernels, which reduces the probability for neutron capture in the radiative capture resonances of 238 U. The resulting higher resonance escape probability also increases the thermal flux in the core. The softening of the neutron spectrum leads to an increased effective microscopic fission cross section in the fissile isotopes and thus to increased neutron absorption for fission, which reduces the remaining number of neutrons that can diffuse into the reflectors. Therefore water ingress into the core leads to a reduced thermal neutron flux in the reflectors. The power density spatial distribution behaved similarly to the thermal neutron flux in the core. Analysis of possible mechanisms was conducted. The results show that

  5. Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors

    International Nuclear Information System (INIS)

    1994-08-01

    The meeting was convened by the International Atomic Energy Agency on the recommendation of the IAEA's International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Poland, the Russian Federation, Switzerland, the United Kingdom and the United States of America. The meeting was chaired by Prof. Dr. K. Kugeler and Prof. Dr. E. Hicken, Directors of the Institute for Safety Research Technology of the KFA Research Center, and covered the following: Design and licensing requirements for gas cooled reactors; concepts for decay heat removal in modern gas cooled reactors; analytical methods for predictions of thermal response, accuracy of predictions; experimental data for validation of predictive methods - operational experience from gas cooled reactors and experimental data from test facilities. Refs, figs and tabs

  6. Influence of precooling cooling air on the performance of a gas turbine combined cycle

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Ik Hwan; Kang, Do Won; Kang, Soo Young; Kim, Tong Seop [Inha Univ., Incheon (Korea, Republic of)

    2012-02-15

    Cooling of hot sections, especially the turbine nozzle and rotor blades, has a significant impact on gas turbine performance. In this study, the influence of precooling of the cooling air on the performance of gas turbines and their combined cycle plants was investigated. A state of the art F class gas turbine was selected, and its design performance was deliberately simulated using detailed component models including turbine blade cooling. Off design analysis was used to simulate changes in the operating conditions and performance of the gas turbines due to precooling of the cooling air. Thermodynamic and aerodynamic models were used to simulate the performance of the cooled nozzle and rotor blade. In the combined cycle plant, the heat rejected from the cooling air was recovered at the bottoming steam cycle to optimize the overall plant performance. With a 200K decrease of all cooling air stream, an almost 1.78% power upgrade due to increase in main gas flow and a 0.70 percent point efficiency decrease due to the fuel flow increase to maintain design turbine inlet temperature were predicted.

  7. CFD Analysis of the Fuel Temperature in High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    In, W. K.; Chun, T. H.; Lee, W. J.; Chang, J. H.

    2005-01-01

    High temperature gas-cooled reactors (HTGR) have received a renewed interest as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor (PBR) and a prismatic modular reactor (PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both PBR and PMR. The objective of this study is to predict the fuel temperature distributions in PBR and PMR using a computational fluid dynamics(CFD) code, CFX-5. The reference reactor designs used in this analysis are PBMR400 and GT-MHR600

  8. Gas Cooled Fast Reactor Research and Development in the European Union

    Directory of Open Access Journals (Sweden)

    Richard Stainsby

    2009-01-01

    Full Text Available Gas-cooled fast reactor (GFR research is directed towards fulfilling the ambitious goals of Generation IV (Gen IV, that is, to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. The research is directed towards developing the GFR as an economic electricity generator, with good safety and sustainability characteristics. Fast reactors maximise the usefulness of uranium resources by breeding plutonium and can contribute to minimising both the quantity and radiotoxicity nuclear waste by actinide transmutation in a closed fuel cycle. Transmutation is particularly effective in the GFR core owing to its inherently hard neutron spectrum. Further, GFR is suitable for hydrogen production and process heat applications through its high core outlet temperature. As such GFR can inherit the non-electricity applications that will be developed for thermal high temperature reactors in a sustainable manner. The Euratom organisation provides a route by which researchers in all European states, and other non-European affiliates, can contribute to the Gen IV GFR system. This paper summarises the achievements of Euratom's research into the GFR system, starting with the 5th Framework programme (FP5 GCFR project in 2000, through FP6 (2005 to 2009 and looking ahead to the proposed activities within the 7th Framework Programme (FP7.

  9. Plutonium-burn high temperature gas-cooled reactor for 3E+3S

    International Nuclear Information System (INIS)

    Okamoto, Koji

    2015-01-01

    The Nuclear Energy Development in Japan is facing a very difficult conditions after Fukushima-Daiichi NPP Accident. Nuclear Energy has strong advantages on 3E, i.e., Energy security, Economical efficiency and Environment. However, people does not believe the Safety 'S' of Nuclear Energy, now. The disadvantage of 'S' overrides the advantages of '3E'. In Nuclear Energy, 'S' is expanded into 3S, i.e., Safety, Security and Safeguards. Especially, the management of Plutonium inventory in Spent Fuel generated by the NPP operation is very important in the viewpoints of non-proliferation. The high-temperature gas cooled reactor (HTGR) is the solution of these disadvantages of '3S' in Nuclear Energy. The fuel of HTGR is composed by 1 mm spherical fuel particle, i.e., TRISO made by fuel, graphite and silicon-carbide. The silicon-carbide can confine the fission products in any conditions of fuel life cycle, i.e., during operation, accidents and disposal for 1 million years. The confinement of the radioactive materials can be confirmed by the TRISO. The HTGR core has strong negative feedback for temperature. So, the fission automatically stopped at the accidental conditions, such as loss of flow and LOCA. Also, the residual heat can be cooled by the radiation heat transfer to reactor vessel wall. The HTGR system usually has passive vessel wall cooling system. When the passive cooling system had been failed, the heat can be transferred to the land by heat conductions, and fuel does not reach the SiC broken temperature. The fission chain reaction has been stopped automatically by negative feedback, i.e., physics. The residual heat had been cooled automatically by radiation. The radioactive materials had been confined automatically by silicon-carbide. The HTGR is superior for 'S' safety. Plutonium can be burned by the HTGR. In the viewpoints of non-proliferation, the fuel should be made by YSZ-PuO 2 , stabilized buffer

  10. High Temperature Gas Cooled Reactor Fuels and Materials

    International Nuclear Information System (INIS)

    2010-03-01

    At the third annual meeting of the technical working group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), held in Vienna, in 2004, it was suggested 'to develop manuals/handbooks and best practice documents for use in training and education in coated particle fuel technology' in the IAEA's Programme for the year 2006-2007. In the context of supporting interested Member States, the activity to develop a handbook for use in the 'education and training' of a new generation of scientists and engineers on coated particle fuel technology was undertaken. To make aware of the role of nuclear science education and training in all Member States to enhance their capacity to develop innovative technologies for sustainable nuclear energy is of paramount importance to the IAEA Significant efforts are underway in several Member States to develop high temperature gas cooled reactors (HTGR) based on either pebble bed or prismatic designs. All these reactors are primarily fuelled by TRISO (tri iso-structural) coated particles. The aim however is to build future nuclear fuel cycles in concert with the aim of the Generation IV International Forum and includes nuclear reactor applications for process heat, hydrogen production and electricity generation. Moreover, developmental work is ongoing and focuses on the burning of weapon-grade plutonium including civil plutonium and other transuranic elements using the 'deep-burn concept' or 'inert matrix fuels', especially in HTGR systems in the form of coated particle fuels. The document will serve as the primary resource materials for 'education and training' in the area of advanced fuels forming the building blocks for future development in the interested Member States. This document broadly covers several aspects of coated particle fuel technology, namely: manufacture of coated particles, compacts and elements; design-basis; quality assurance/quality control and characterization techniques; fuel irradiations; fuel

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

    International Nuclear Information System (INIS)

    1985-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

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

  13. Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Verma, V., E-mail: vasudha.verma@physics.uu.se [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Barbot, L.; Filliatre, P. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Hellesen, C. [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Jammes, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Svärd, S. Jacobsson [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden)

    2017-07-11

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment. - Highlights: • Studied possibility of using SPNDs as in-core detectors in SFRs. • Study done to detect local power profile changes when reactor is at nominal power. • SPND with a Pt-emitter gives measurable prompt current of the order of 600 nA/m. • Dominant proportion of prompt response is maintained throughout the operation. • Detector signal gives dynamic information on the power fluctuations.

  14. Safeguarding of emergency core cooling in case of loss-of-coolant accidents with insulation material release

    International Nuclear Information System (INIS)

    Pointner, W.; Broecker, A.

    2012-01-01

    The report on safeguarding of emergency core cooling in case of loss-of-coolant accidents with insulation material release covers the following issues: assessment of the relevant status for PWR, evaluation of the national and international (USA, Canada, France) status, actualization of recommendations, transferability from PWR to BWR. Generic studies on the core cooling capability in case of insulation material release in BWR-type reactors were evaluated.

  15. Analysis of expediency to set regulators of high-pressure emergency core cooling system of WWER 1000 (B-320)

    International Nuclear Information System (INIS)

    Skalozubov, V.I.; Komarov, Yu.A.; Tikhonova, G.G.; Nikiforov, S.N.; Bogodist, V.V.; Fol'tov, I.M.; Khadzh Faradzhallakh Dabbakh, A.

    2011-01-01

    The work shows that setting regulative valves in high-pressure emergency core cooling system of WWER 1000/B-320 can be effective only involving the additional tuning to account traverse speed of operating elements of regulator and configuration of the systems providing cooling of primary loop.

  16. Analytical Modelling of the Effects of Different Gas Turbine Cooling Techniques on Engine Performance =

    Science.gov (United States)

    Uysal, Selcuk Can

    In this research, MATLAB SimulinkRTM was used to develop a cooled engine model for industrial gas turbines and aero-engines. The model consists of uncooled on-design, mean-line turbomachinery design and a cooled off-design analysis in order to evaluate the engine performance parameters by using operating conditions, polytropic efficiencies, material information and cooling system details. The cooling analysis algorithm involves a 2nd law analysis to calculate losses from the cooling technique applied. The model is used in a sensitivity analysis that evaluates the impacts of variations in metal Biot number, thermal barrier coating Biot number, film cooling effectiveness, internal cooling effectiveness and maximum allowable blade temperature on main engine performance parameters of aero and industrial gas turbine engines. The model is subsequently used to analyze the relative performance impact of employing Anti-Vortex Film Cooling holes (AVH) by means of data obtained for these holes by Detached Eddy Simulation-CFD Techniques that are valid for engine-like turbulence intensity conditions. Cooled blade configurations with AVH and other different external cooling techniques were used in a performance comparison study. (Abstract shortened by ProQuest.).

  17. Progress in the development of tooling and dismantling methodologies for the Windscale advanced gas cooled reactor (WAGR)

    International Nuclear Information System (INIS)

    Cross, M.T.; Wareing, M.I.; Dixon, C.

    1998-01-01

    Decommissioning of the Windscale Advanced Gas-Cooled Reactor (WAGR) is a major UK reactor decommissioning project co-funded by the UK Government, the European Commission and Magnox Electric. WAGR was a CO 2 cooled, graphite moderated reactor which served as a test bed for the development of Advanced Gas-Cooled Reactor technology in the UK. It operated from 1963 until shutdown in 1981. AEA Technology plc are currently the Managing Agents on behalf of UKAEA for the WAGR decommissioning project and are responsible for the co-ordination of the project up to the point when the contents of the reactor core and associated radioactive materials are removed and either disposed of or packaged for disposal at some time in the future. Decommissioning has progressed to the point where the reactor has been dismantled down to the level of the hot gas collection manifold with the removal of the top biological shield, the refuelling standpipes and the top section of the reactor pressure vessel. The 4 heat exchangers have also been removed and committed to shallow land burial. This paper describes the work carried out by AEA Technology under separate contracts of UKAEA in developing some of the equipment and deployment methods for the next phase of active operations required in preparation for the dismantling of the core structure. Most recent work has concentrated on the development of specialist tooling for removal of items of operational waste stored within the reactor core, equipment for cutting and removal of the highly radioactive stainless steel 'loop' pressure tubes, diamond wire cutting equipment for sectioning large diameter pipework, and equipment for dismantling the reactor neutron shield. The paper emphasises the process of adaptation and extension of existing technologies for cost-effective application in the decommissioning environment, the need for adequate forward planning of decommissioning methodologies together with large-scale 'mock-up' testing of equipment to

  18. The variation of particle gas-borne concentration with time in a gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reed, J; Hall, D; Reeks, M W [Central Electricity Generating Board, Berkeley Nuclear Laboratories (United Kingdom)

    1985-07-01

    If volatile fission products are released from fuel during a reactor fault, a significant fraction could become attached to small particles also present in the coolant. In such circumstances the retention of those particles by the reactor circuit will limit the level of gas-borne particle concentration and hence be important in reducing the potential release of fission product activity to the atmosphere. Clearly the retention of particles will be influenced by both the deposition and resuspension of particles from surfaces exposed to the coolant flow. In this paper we consider deposition and resuspension but pay particular attention to the role of resuspension, which in the past has been given little consideration. A recently developed model for the resuspension of small particles by a turbulent flow is outlined. Traditionally, resuspension has been interpreted as a force balance between the aerodynamic removal forces and the surface adhesive forces. In contrast, this new approach embodies an energy balance criterion for particle resuspension. Furthermore, the stochastic nature of this new model has shown that resuspension can be sub-divided into two regimes: (i) initial resuspension (resuspension occurring in times less than a second) which reduces the net deposition of particles to a surface; and (ii) longer term resuspension (resuspension after 1 second) which determines the asymptotic decay of particle gas-borne concentration. It is seen that the asymptotic decay varies almost inversely as the decay time. Force balance models are unsuccessful in accounting for the experimentally observed longer term resuspension. We show that a Volterra integro-differential equation best describes the variation of particle gas-borne concentration with time in a recirculating gas flow such as a gas cooled reactor. It is seen that the longer term resuspension has a major influence in the final decay of particle concentration. (author)

  19. The variation of particle gas-borne concentration with time in a gas cooled reactor

    International Nuclear Information System (INIS)

    Reed, J.; Hall, D.; Reeks, M.W.

    1985-01-01

    If volatile fission products are released from fuel during a reactor fault, a significant fraction could become attached to small particles also present in the coolant. In such circumstances the retention of those particles by the reactor circuit will limit the level of gas-borne particle concentration and hence be important in reducing the potential release of fission product activity to the atmosphere. Clearly the retention of particles will be influenced by both the deposition and resuspension of particles from surfaces exposed to the coolant flow. In this paper we consider deposition and resuspension but pay particular attention to the role of resuspension, which in the past has been given little consideration. A recently developed model for the resuspension of small particles by a turbulent flow is outlined. Traditionally, resuspension has been interpreted as a force balance between the aerodynamic removal forces and the surface adhesive forces. In contrast, this new approach embodies an energy balance criterion for particle resuspension. Furthermore, the stochastic nature of this new model has shown that resuspension can be sub-divided into two regimes: (i) initial resuspension (resuspension occurring in times less than a second) which reduces the net deposition of particles to a surface; and (ii) longer term resuspension (resuspension after 1 second) which determines the asymptotic decay of particle gas-borne concentration. It is seen that the asymptotic decay varies almost inversely as the decay time. Force balance models are unsuccessful in accounting for the experimentally observed longer term resuspension. We show that a Volterra integro-differential equation best describes the variation of particle gas-borne concentration with time in a recirculating gas flow such as a gas cooled reactor. It is seen that the longer term resuspension has a major influence in the final decay of particle concentration. (author)

  20. Feasibility of maintaining natural convection mode core cooling in research reactor power upgrades

    International Nuclear Information System (INIS)

    Ha, J.J.; Belhadj, M.; Aldemir, T.; Christensen, R.N.

    1987-01-01

    Two operational concerns for natural convection coooled research reactors using plate type fuels are: 1) pool top 16 N activity (PTNA), and 2) nucleate boiling in core channels. The feasibility assessment of a power upgrade while maintaining natural convection mode core cooling requires addressing these operational concerns. Previous studies have shown that: a) The conventional technique for reducing PTNA by plume dispersion may not be effective in a large power upgrade of research reactors with small pools. b) Currently used correlations to predict onset of nucleate boiling (ONB) in thin, rectangular core channels are not valid for low-velocity, upward flows such as encountered in natural convection cooling. The PTNA depends on the velocity distribution in the reactor pool. COMMIX-1A code is used to determine the three-dimensional velocity fields in The Ohio State University Research Reactor (OSURR) pool as a function of varying design conditions, following a power upgrade to 500 kW with LEU fuel. It is shown that a sufficiently deep stagnant water layer can be created below the pool top by properly choosing the disperser flow rate. The ONB heat flux is experimentally determined for channel gaps and upward flow velocities in the range 2mm-4mm and 3-16 cm/sec., respectively. Two alternatives to plume dispersion for reducing PTNA and a new correlation to determine the ONB heat flux in thin, rectangular channels under low-velocity, upward flow conditions are proposed. (Author)

  1. Utilization of control rod drive (CRD) system for long term core cooling

    International Nuclear Information System (INIS)

    Huerta B, A.

    1991-01-01

    In this paper we consider an application of Probabilistic Risk Assessment (PRA) to risk management. Foreseeable risk management strategies to prevent core damage are constrained by the availability of first line systems as well as support systems. The actual trend in the evaluation of risk management options can be performed in a number of ways. An example is the identification of back-up systems which could be used to perform the same safety functions. In this work we deal with the evaluation of the feasibility, for BWR's, to use the Control Rod Drive system to maintain an adequate reactor core long term cooling in some accident sequences. This preliminary evaluation is carried out as a part of the Internal Events Analysis for Laguna Verde Nuclear Power Plant (LVNPP) that is currently under way by the Mexican Nuclear Regulatory Body. This analysis addresses the evaluation and incorporation of all the systems, including the safety related and the back-up non safety related systems, that are available for the operator in order to prevent core damage. As a part of this analysis the containment venting capability is also evaluated as a back-up of the containment heat removal function. This will prevent the primary containment overpressurization and loss of certain core cooling systems. A selection of accident sequences in which the Control Rod Drive system could be used to mitigate the accident and prevent core damage are discussed. A personal computer transient analysis code is used to carry out thermohydraulic simulations in order to evaluate the Control Rod Drive system performance, the corresponding results are presented. Finally, some preliminary conclusions are drawn. (author). 9 refs, 5 figs

  2. Thermodynamic performance of a gas-core fission reactor

    International Nuclear Information System (INIS)

    Klein, W.

    1987-01-01

    The purpose of this thesis was to investigate the thermodynamic behaviour of a critical quantity of gaseous uranium-fluorides in chemical equilibrium with a graphite wall. From the very beginning a container was considered with cooled walls. As it was evident that a nuclear reactor working with gaseous fuel should run at much higher temperatures than classical LWR or HTGR reactors, most of the investigations were performed for walls with a surface temperature of 1800 to 2000 K. It was supposed that such a surface temperature would be technologically possible for a heat load between 1 and 5 MWatt m -2 . Cooling with high pressure helium-gas has to keep balance with this heat flux. The technical construction of such a wall will be a problem in itself. It is thought that the experiences with re-entry-vessels in space-technology can be used. A basic assumption in all the calculations is that the U-C-F reactor gas 'sees' a graphite wall, possibly graphite tiles supported by heat resistant materials like SiN 2 , SiC 2 and at a lower temperature level by niobium-steel. Such a gastight compound-system is not necessarily of high-tensile strength materials. It has to be surrounded by a cooled neutron moderator-reflector which in its turn must be supported by a steel-wall at room temperature holding pressure of the order of 100 bar (10 MPa). The design of such a compound-wall is a task for the future. 116 refs.; 28 figs.; 29 tabs

  3. Power distribution monitoring system in the boiling water cooled reactor core

    International Nuclear Information System (INIS)

    Leshchenko, Yu.I.; Sadulin, V.P.; Semidotskij, I.I.

    1987-01-01

    Consideration is being given to the system of physical power distribution monitoring, used during several years in the VK-50 tank type boiling water cooled reactor. Experiments were conducted to measure the ratios of detector prompt and activation currents, coefficients of detector relative sensitivity with respect to neutrons and effective cross sections of 103 Rh interaction with thermal and epithermal neutrons. Mobile self-powered detectors (SPD) with rhodium emitters are used as the power distribution detectors in the considered system. All detectors move simultaneously with constant rate in channels, located in fuel assembly central tubes, when conducting the measurements. It is concluded on the basis of analyzing the obtained data, that investigated system with calibrated SPD enables to monitor the absolute power distribution in fuel assemblies under conditions of boiling water cooled reactor and is independent of thermal engineering measurements conducted by in core instruments

  4. Parametric studies on different gas turbine cycles for a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Jie; Gu Yihua

    2005-01-01

    The high temperature gas-cooled reactor (HTGR) coupled with turbine cycle is considered as one of the leading candidates for future nuclear power plants. In this paper, the various types of HTGR gas turbine cycles are concluded as three typical cycles of direct cycle, closed indirect cycle and open indirect cycle. Furthermore they are theoretically converted to three Brayton cycles of helium, nitrogen and air. Those three types of Brayton cycles are thermodynamically analyzed and optimized. The results show that the variety of gas affects the cycle pressure ratio more significantly than other cycle parameters, however, the optimized cycle efficiencies of the three Brayton cycles are almost the same. In addition, the turbomachines which are required for the three optimized Brayton cycles are aerodynamically analyzed and compared and their fundamental characteristics are obtained. Helium turbocompressor has lower stage pressure ratio and more stage number than those for nitrogen and air machines, while helium and nitrogen turbocompressors have shorter blade length than that for air machine

  5. Analysis of emergency core cooling capability of direct vessel vertical injection using CFX

    International Nuclear Information System (INIS)

    Yoon, Sang H.; Yu, Yong H.; Suh, Kune Y.

    2003-01-01

    More reliable and efficient safety injection system is of utmost importance in the design of advanced reactors such as the APR1400 (Advanced Power Reactor 1400 MWe). In this work, a new idea is proposed to inject the Emergency Core Cooling (ECC) water utilizing a dedicated nozzle with a vertically downward elbow. The Direct Vessel Injection (DVI) system is located horizontally above the cold leg in the APR1400. However, the horizontal injection method may not always satisfy the ECC penetration requirement into the core on account of rather involved multidimensional thermal and hydraulic phenomena occurring in the annular reactor downcomer such as bypass, impingement, entrainment and sweepout, condensation oscillation, etc. Thus, a novel concept is called for from the reactor safety point of view. The Direct Vessel Vertical Injection (DVVI) system is one of these efforts to penetrate as much the ECC water through the downcomer into the core as is practically achievable. The DVVI system can increase the momentum of the downward flow, thus minimizing the effect of water impingement on the core barrel and the direct bypass though the break. To support the claim of increased downward momentum of flow in the DVVI system, computational fluid dynamics analyses were performed using CFX. The new concept of the DVVI system, which can certainly help increase the core thermal margin, is found to be more efficient than DVI. If the structural problem in the manufacturing process is properly solved, this concept can safely be applied in the advanced nuclear reactor design

  6. Development of variable width ribbon heating elements for liquid metal and gas-cooled fast breeder reactor fuel rod simulators

    International Nuclear Information System (INIS)

    McCulloch, R.W.; Lovell, R.T.; Post, D.W.; Snyder, S.D.

    1980-01-01

    Variable width ribbon heating elements have been fabricated which provide a chopped cosine, variable heat flux profile for fuel rod simulators used in test loops by the Breeder Reactor Program Thermal Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor Core Flow Test Loop. Thermal, mechanical, and electrical design considerations result in the derivation of an analytical expression for the ribbon contours. From this, the ribbons are machined and wound on numerically controlled equipment. Postprocessing and inspection results in a wound, variable width ribbon with the precise dimensional, electrical, and mechanical properties needed for use in fuel pin simulators

  7. Dynamic simulation for scram of high temperature gas-cooled reactor with indirect helium turbine cycle system

    International Nuclear Information System (INIS)

    Li Wenlong; Xie Heng

    2011-01-01

    A dynamic analysis code for this system was developed after the mathematical modeling and programming of important equipment of 10 MW High Temperature Gas Cooled Reactor Helium Turbine Power Generation (HTR-10GT), such as reactor core, heat exchanger and turbine-compressor system. A scram accident caused by a 0.1 $ reactivity injection at 5 second was simulated. The results show that the design emergency shutdown plan for this system is safe and reasonable and that the design of bypass valve has a large safety margin. (authors)

  8. Pressure transients analysis of a high-temperature gas-cooled reactor with direct helium turbine cycle

    Energy Technology Data Exchange (ETDEWEB)

    Dang, M.; Dupont, J. F.; Jacquemoud, P.; Mylonas, R. [Eidgenoessisches Inst. fuer Reaktorforschung, Wuerenlingen (Switzerland)

    1981-01-15

    The direct coupling of a gas cooled reactor with a closed gas turbine cycle leads to a specific dynamic plant behaviour, which may be summarized as follows: a) any operational transient involving a variation of the core mass flow rate causes a variation of the pressure ratio of the turbomachines and leads unavoidably to pressure and temperature transients in the gas turbine cycle; and b) very severe pressure equalization transients initiated by unlikely events such as the deblading of one or more turbomachines must be taken into account. This behaviour is described and illustrated through results gained from computer analyses performed at the Swiss Federal Institute for Reactor Research (EIR) in Wurenlingen within the scope of the Swiss-German HHT project.

  9. A Feasibility Study on Core Cooling of Reduced-Moderation PWR for the Large Break LOCA

    International Nuclear Information System (INIS)

    Hiroyuki Yoshida; Akira Ohnuki; Hajime Akimoto

    2002-01-01

    A design study of a reduced-moderation water reactor (RMWR) with tight lattice core is being carried out at the Japan Atomic Energy Research Institute (JAERI) as one candidate for future reactors. The concept is developed to achieve a conversion ratio greater than unity using the tight lattice core (volume ratio of moderator to fuel is around 0.5 and the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated). Under such tight configuration, the core thermal margin becomes smaller and should be evaluated in a normal operation and also during the reflood phase in a large break loss-of-coolant accident (LBLOCA) for PWR type reactors. In this study, we have performed a feasibility evaluation on core cooling of reduced moderation PWR for the LBLOCA (200% break). The evaluation was performed for the primary system after the break by the REFLA/TRAC code. The core thermal output of the reduced moderation PWR is 2900 MWt, the gap between adjacent fuel rods is 1 mm, and heavy water is used as the moderator and coolant. The present design adopts seed fuel assemblies (MOX fuel) and several blanket fuel assemblies. In the blanket fuel assemblies, power density is lower than that of the seed fuel assemblies. Then, we set a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because the possibility that the coolant boils in the seed fuel assemblies is very high. The pressure vessel diameter is bigger in comparison with a current PWR and core height is smaller than the current one. The current 4-loop PWR system is used, and, however, to fit into the bigger pressure vessel volume (about 1.5 times), we set up the capacity of the accumulator (1.5 times of the current PWR). Although the maximum clad temperature reached at about 1200 K in the position of 0.6 m from the lower core support plate, it is sufficiently lower than the design criteria of the current PWR (1500 K). The core cooling of the reduced moderation

  10. Hot gas path component cooling system having a particle collection chamber

    Science.gov (United States)

    Miranda, Carlos Miguel; Lacy, Benjamin Paul

    2018-02-20

    A cooling system for a hot gas path component includes a substrate having an outer surface and an inner surface. The inner surface defines at least one interior space. A passage is formed in the substrate between the outer surface and the inner surface. An access passage is formed in the substrate and extends from the outer surface to the inner space. The access passage is formed at a first acute angle to the passage and includes a particle collection chamber. The access passage is configured to channel a cooling fluid to the passage. Furthermore, the passage is configured to channel the cooling fluid therethrough to cool the substrate.

  11. Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

    Science.gov (United States)

    Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

    2017-07-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

  12. Quench cooling of superheated debris beds in containment during LWR core meltdown accidents

    International Nuclear Information System (INIS)

    Ginsberg, T.; Chen, J.C.

    1984-01-01

    Light water reactor core meltdown accident sequence studies suggest that superheated debris beds may settle on the concrete floor beneath the reactor vessel. A model for the heat transfer processes during quench of superheated debris beds cooled by an overlying pool of water has been presented in a prior paper. This paper discusses the coolability of decay-heated debris beds from the standpoint of their transient quench characteristics. It is shown that even though a debris bed configuration may be coolable from the point of view of steady-state decay heat removal, the quench behavior from an initially elevated temperature may lead to bed melting prior to quench of the debris

  13. Laser-cooled atoms inside a hollow-core photonic-crystal fiber

    DEFF Research Database (Denmark)

    Bajcsy, Michal; Hofferberth, S.; Peyronel, Thibault

    2011-01-01

    We describe the loading of laser-cooled rubidium atoms into a single-mode hollow-core photonic-crystal fiber. Inside the fiber, the atoms are confined by a far-detuned optical trap and probed by a weak resonant beam. We describe different loading methods and compare their trade-offs in terms...... of implementation complexity and atom-loading efficiency. The most efficient procedure results in loading of ∼30,000 rubidium atoms, which creates a medium with an optical depth of ∼180 inside the fiber. Compared to our earlier study this represents a sixfold increase in the maximum achieved optical depth...

  14. The effects of aging on Boiling Water Reactor core isolation cooling system

    International Nuclear Information System (INIS)

    Lee, Bom Soon.

    1994-01-01

    A study was performed to assess the effects of aging on the Reactor Core Isolation Cooling system in commercial Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research program sponsored by the US Nuclear Regulatory Commission. The failure data, from national databases, as well as plant specific data were reviewed and analyzed to understand the effects of aging on the RCIC system. This analysis identified important components that should receive the highest priority in terms of aging management. The aging characterization provided information on the effects of aging on component failure frequency, failure modes, and failure causes

  15. Strength analysis of fast gas cooled reactor fuel element in conditions of fuel-cladding interraction and non-uniform azimuthal heating

    International Nuclear Information System (INIS)

    Kulikov, I.S.; Tverkovkin, B.E.

    1984-01-01

    The technique and the PRORT mathematical program in FORTRAN language for determining mechanical properties of a fuel element with motionless fuel-cladding interaction taking into account circular temperature non-uniformity in gas-cooled fast reactor conditions are proposed. The calculation results of the fuel element of dissociating gas cooled fast reactor are presented for seven cross-sections over the height of the core. The obtained data testify to appreciable swelling of Cr16Ni15Mo3Nb steel fuel cladding in the conditions of dissociating gas cooled fast reactor through the allowance for the effect of stresses on this essential parameter shows, that its value is lower in comparison with swelling, wherein stresses are not taken into account

  16. The cooling history and the depth of detachment faulting at the Atlantis Massif oceanic core complex

    Science.gov (United States)

    Schoolmeesters, Nicole; Cheadle, Michael J.; John, Barbara E.; Reiners, Peter W.; Gee, Jeffrey; Grimes, Craig B.

    2012-10-01

    Oceanic core complexes (OCCs) are domal exposures of oceanic crust and mantle interpreted to be denuded to the seafloor by large slip oceanic detachment faults. We combine previously reported U-Pb zircon crystallization ages with (U-Th)/He zircon thermochronometry and multicomponent magnetic remanence data to determine the cooling history of the footwall to the Atlantis Massif OCC (30°N, MAR) and help establish cooling rates, as well as depths of detachment faulting and gabbro emplacement. We present nine new (U-Th)/He zircon ages for samples from IODP Hole U1309D ranging from 40 to 1415 m below seafloor. These data paired with U-Pb zircon ages and magnetic remanence data constrain cooling rates of gabbroic rocks from the upper 800 m of the central dome at Atlantis Massif as 2895 (+1276/-1162) °C Myr-1 (from ˜780°C to ˜250°C); the lower 600 m of the borehole cooled more slowly at mean rates of ˜500 (+125/-102) °C Myr-1(from ˜780°C to present-day temperatures). Rocks from the uppermost part of the hole also reveal a brief period of slow cooling at rates of ˜300°C Myr-1, possibly due to hydrothermal circulation to ˜4 km depth through the detachment fault zone. Assuming a fault slip rate of 20 mm/yr (from U-Pb zircon ages of surface samples) and a rolling hinge model for the sub-surface fault geometry, we predict that the 780°C isotherm lies at ˜7 km below the axial valley floor, likely corresponding both to the depth at which the semi-brittle detachment fault roots and the probable upper limit of significant gabbro emplacement.

  17. Neutronic design of the RSG-GAS silicide core

    Energy Technology Data Exchange (ETDEWEB)

    Sembiring, T.M.; Kuntoro, I.; Hastowo, H. [Center for Development of Research Reactor Technology National Nuclear Energy Agency BATAN, PUSPIPTEK Serpong Tangerang, 15310 (Indonesia)

    2002-07-01

    The objective of core conversion program of the RSG-GAS multipurpose reactor is to convert the fuel from oxide, U{sub 3}O{sub 8}-Al to silicide, U{sub 3}Si{sub 2}-Al. The aim of the program is to gain longer operation cycle by having, which is technically possible for silicide fuel, a higher density. Upon constraints of the existing reactor system and utilization, an optimal fuel density in amount of 3.55 g U/cc was found. This paper describes the neutronic parameter design of the silicide equilibrium core and the design of its transition cores as well. From reactivity control point of view, a modification of control rod system is also discussed. All calculations are carried out by means of diffusion codes, Batan-EQUIL-2D, Batan-2DIFF and -3DIFF. The silicide core shows that longer operation cycle of 32 full power days can be achieved without decreasing the safety criteria and utilization capabilities. (author)

  18. Assessments of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Dong Yujie; Scherer, Winfried

    2005-01-01

    Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemented in the dynamic simulator for nuclear power plants (DSNP) simulation system. Comparisons of the DSNP simulations incorporating these models with experiments and with calculations using the time-dependent neutronics and temperature dynamics code were made to validate the simulation. The analysis of the primary circuit showed that the maximum water concentration increase in the reactor core was 3 s). The water vaporization in the steam generator and characteristics of water transport from the steam generator to the reactor core would reduce the rate of water ingress into the reactor core. The analysis of a full cavitation of the feedwater pump showed that if the secondary circuit could be depressurized, the feedwater pump would be stopped by the full cavitation. This limits the water transported from the deaerator to the steam generator. A comprehensive simulation of the HTR-module power plant showed that the water inventory in the primary circuit was limited to ∼3000 kg. The nuclear reactivity increase caused by the water ingress would lead to a fast power excursion, which would be inherently counterbalanced by negative feedback effects. The integrity of the fuel elements, because the safety-relevant temperature limit of 1600 deg. C is not reached in any case, is not challenged

  19. Minor actinide transmutation in a board type sodium cooled breed and burn reactor core

    International Nuclear Information System (INIS)

    Zheng, Meiyin; Tian, Wenxi; Zhang, Dalin; Qiu, Suizheng; Su, Guanghui

    2015-01-01

    Highlights: • A 1250 MWt board type sodium cooled breed and burn reactor core is further designed. • MCNP–ORIGEN coupled code MCORE is applied to perform neutronics and depletion calculation. • Transmutation efficiency and neutronic safety parameters are compared under different MA weight fraction. - Abstract: In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0 wt.%, 4.0 wt.%, 6.0 wt.%, 8.0 wt.%, 10.0 wt.% and 12.0 wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library. In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22 years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void

  20. Thermal-hydraulic analysis of the OSURR pool for power upgrade with natural convection core cooling

    International Nuclear Information System (INIS)

    Ha, J.J.; Aldemir, T.

    1988-01-01

    Natural convection mode core cooling will be maintained in the LEU conversion/power upgrade of The Ohio State University Research Reactor (OSURR) to 250-500 kW. The pool water will be cooled by a water-glycol-air and a water-water heat exchanger. A plume disperser will be installed in the pool to minimize evaporation from the pool top and to maintain the dose rate due to N-16 activity within allowable levels. The minimization of the pool heat removal system operation costs necessitates maximizing the inlet temperature to the water-glycol-air heat exchanger. For the maximization process, the change in the pool temperature and velocity fields have to be investigated as a function of: location and orientation of the heat removal system components and the plume disperser in the pool; mass flow rate through the plume disperser. The velocity and temperature fields in the pool are determined using COMMIX-1A. The computational system model accounts for the presence of all the pool components (i.e. core, thermal column, beam ports, ion chamber, guide tubes, rabbit, neutron source etc.). The results show that: (1) Both the heat removal system inlet point and the plume disperser have to be located close to the top of the core. (2) Using a disperser system consisting of several pipes may be more feasible than a single unit. (3) For high disperser flow, the disperser jet has to be almost parallel to the top of the core to prevent flow reversal in coolant channels. (4) More than one disperser system may be necessary to create an inversion layer in the pool

  1. A global model for gas cooled reactors for the Generation-4: application to the Very High Temperature Reactor (VHTR)

    International Nuclear Information System (INIS)

    Limaiem, I.

    2006-12-01

    Gas cooled high temperature reactor (HTR) belongs to the new generation of nuclear power plants called Generation IV. The Generation IV gathers the entire future nuclear reactors concept with an effective deployment by 2050. The technological choices relating to the nature of the fuel, the moderator and the coolant as well as the annular geometry of the core lead to some physical characteristics. The most important of these characteristics is the very strong thermal feedback in both active zone and the reflectors. Consequently, HTR physics study requires taking into account the strong coupling between neutronic and thermal hydraulics. The work achieved in this Phd consists in modeling, programming and studying of the neutronic and thermal hydraulics coupling system for block type gas cooled HTR. The coupling system uses a separate resolution of the neutronic and thermal hydraulics problems. The neutronic scheme is a double level Transport (APOLLO2) /Diffusion (CRONOS2) scheme respectively on the scale of the fuel assembly and a reactor core scale. The thermal hydraulics model uses simplified Navier Stokes equations solved in homogeneous porous media in code CAST3M CFD code. A generic homogenization model is used to calculate the thermal hydraulics parameters of the porous media. A de-homogenization model ensures the link between the porous media temperatures of the temperature defined in the neutronic model. The coupling system is made by external procedures communicating between the thermal hydraulics and neutronic computer codes. This Phd thesis contributed to the Very High Temperature Reactor (VHTR) physics studies. In this field, we studied the VHTR core in normal operating mode. The studies concern the VHTR core equilibrium cycle with the control rods and using the neutronic and thermal hydraulics coupling system. These studies allowed the study of the equilibrium between the power, the temperature and Xenon. These studies open new perspective for core

  2. Review of coaxial flow gas core nuclear rocket fluid mechanics

    International Nuclear Information System (INIS)

    Weinstein, H.

    1976-01-01

    In a prematurely aborted attempt to demonstrate the feasibility of using a gas core nuclear reactor as a rocket engine, NASA initiated a number of studies on the relevant fluid mechanics problems. These studies were carried out at NASA laboratories, universities and industrial research laboratories. Because of the relatively sudden termination of most of this work, a unified overview was never presented which demonstrated the accomplishments of the program and pointed out the areas where additional work was required for a full understanding of the cavity flow. This review attempts to fulfill a part of this need in two important areas

  3. The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

    International Nuclear Information System (INIS)

    Mursid Djokolelono.

    1976-01-01

    Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

  4. Heat Transfer in Pebble-Bed Nuclear Reactor Cores Cooled by Fluoride Salts

    Science.gov (United States)

    Huddar, Lakshana Ravindranath

    With electricity demand predicted to rise by more than 50% within the next 20 years and a burgeoning world population requiring reliable emissions-free base-load electricity, can we design advanced nuclear reactors to help meet this challenge? At the University of California, Berkeley (UCB) Fluoride-salt-cooled High Temperature Reactors (FHR) are currently being investigated. FHRs are designed with better safety and economic characteristics than conventional light water reactors (LWR) currently in operation. These reactors operate at high temperature and low pressure making them more efficient and safer than LWRs. The pebble-bed FHR (PB-FHR) variant includes an annular nuclear reactor core that is filled with randomly packed pebble fuel. It is crucial to characterize the heat transfer within this unique geometry as this informs the safety limits of the reactor. The work presented in this dissertation focused on furthering the understanding of heat transfer in pebble-bed nuclear reactor cores using fluoride salts as a coolant. This was done through experimental, analytical and computational techniques. A complex nuclear system with a coolant that has never previously been in commercial use requires experimental data that can directly inform aspects of its design. It is important to isolate heat transfer phenomena in order to understand the underlying physics in the context of the PB-FHR, as well as to make decisions about further experimental work that needs to be done in support of developing the PB-FHR. Certain organic oils can simulate the heat transfer behaviour of the fluoride salt if relevant non-dimensional parameters are matched. The advantage of this method is that experiments can be done at a much lower temperature and at a smaller geometric scale compared to FHRs, thereby lowering costs. In this dissertation, experiments were designed and performed to collect data demonstrating similitude. The limitations of these experiments were also elucidated by

  5. Compact sodium cooled nuclear power plant with fast core (KNK II- Karlsruhe), Safety Report

    International Nuclear Information System (INIS)

    1977-09-01

    After the operation of the KNK plant with a thermal core (KNK I), the installation of a fast core (KNK II) had been realized. The planning of the core and the necessary reconstruction work was done by INTERATOM. Owner and customer was the Nuclear Research Center Karlsruhe (KfK), while the operating company was the Kernkraftwerk-Betriebsgesellschaft mbH (KBG) Karlsruhe. The main goals of the KNK II project and its special experimental test program were to gather experience for the construction, the licensing and operation of future larger plants, to develop and to test fuel and absorber assemblies and to further develop the sodium technology and the associated components. The present safety report consists of three parts. Part 1 contains the description of the nuclear plant. Hereby, the reactor and its components, the handling facilities, the instrumentation with the plant protection, the design of the plant including the reactor core and the nominal operation processes are described. Part 2 contains the safety related investigation and measures. This concerns the reactivity accidents, local cooling perturbations, radiological consequences with the surveillance measures and the justification of the choice of structural materials. Part three finally is the appendix with the figures, showing the different buildings, the reactor and its components, the heat transfer systems and the different auxiliary facilities [de

  6. Long term contracts in portfolios of core LDC gas supply

    International Nuclear Information System (INIS)

    John, F.E.

    1992-01-01

    This paper recommends that local distribution companies (LDCs) should use a portfolio approach for their gas supply strategy. The author recommends that LDCs not rely on spot supplies to meet the peak needs of the core residential and commercial markets. He recommends that a secure supply through long-term contracts are better sources than spot or even intermediate term suppliers. The paper provides a brief outline format of the advantages to the use of a portfolio approach which include the rapid restructuring of the market, general changes in the market, and general market performance. By maintaining a portfolio, a list of available natural gas suppliers is always available. This portfolio also acts to compare pricing between short, medium, and long-term pricing for the LDCs

  7. Development of cooling and cleaning systems for enhanced gas ...

    African Journals Online (AJOL)

    In order to address these tar related problems a cleaning and cooling system has been developed in house that facilitates tar removal to acceptable levels tolerated by the internal combustion (IC) engine and meets emission standards as well. The main objective of the present work is to reduce tar level and develop control ...

  8. Experiment of IEA-R1 reactor core cooling by air convection after pool water loss accident

    International Nuclear Information System (INIS)

    Torres, Walmir Maximo; Baptista Filho, Benedito Dias

    2000-01-01

    This paper presents a study of a Emergency Core Cooling to be applied to the IEA-R1 reactor. This system must have the characteristics of passive action, with water spraying over the core, and feeding by gravity from elevated reservoirs. In the evaluation, this system must demonstrate that when the reservoirs are emptied, the core cooling must assure to be fulfilled by air natural convection. This work presents the results of temperature distribution in a test section with plates electrically heated simulation the heat generation conditions on the most heated reactor element

  9. Noble gas, binary mixtures for commercial gas-cooled reactor systems

    International Nuclear Information System (INIS)

    El-Genk, M. S.; Tournier, J. M.

    2007-01-01

    Commercial gas cooled reactors employ helium as a coolant and working fluid for the Closed Brayton Cycle (CBC) turbo-machines. Helium has the highest thermal conductivity and lowest dynamic viscosity of all noble gases. This paper compares the relative performance of pure helium to binary mixtures of helium and other noble gases of higher molecular weights. The comparison is for the same molecular flow rate, and same operating temperatures and geometry. Results show that although helium is a good working fluid because of its high heat transfer coefficient and significantly lower pumping requirement, a binary gas mixture of He-Xe with M = 15 gm/mole has a heat transfer coefficient that is ∼7% higher than that of helium and requires only 25% of the number stages of the turbo-machines. The binary mixture, however, requires 3.5 times the pumping requirement with helium. The second best working fluid is He-Kr binary mixture with M = 10 gm/mole. It has 4% higher heat transfer coefficient than He and requires 30% of the number of stages in the turbo-machines, but requires twice the pumping power

  10. Recombining processes in a cooling plasma by mixing of initially heated gas

    International Nuclear Information System (INIS)

    Furukane, Utaro; Sato, Kuninori; Takiyama, Ken; Oda, Toshiatsu.

    1992-03-01

    A numerical investigation of recombining process in a high temperature plasma in a quasi-steady state is made in a gas contact cooling, in which the initial temperature effect of contact gas heated up by the hot plasma is considered as well as the gas cooling due to the surrounding neutral particles freely coming into the plasma. The calculation has shown that the electron temperature relaxes in accord with experimental results and that the occurrence of recombining region and the inverted populations almost agree with the experimental ones. (author)

  11. Numerical evaluation of flow through a prismatic very high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Barros Filho, Jose A.; Santos, Andre A.C.; Navarro, Moyses A.; Ribeiro, Felipe Lopes

    2011-01-01

    The High-temperature Gas-cooled reactor (HTGR) is a Next Generation Nuclear System that has a good chance to be used as energy generation source in the near future owing to its potential capacity to supply hydrogen without greenhouse gas emission for the future humanity. Recently, improvements in the HTGR design led to the Very High Temperature Reactor (VHTR) concept in which the outlet temperature of the coolant gas reaches to 1000 deg C increasing the efficiency of the hydrogen and electricity generation. Among the core concepts emerging in the VHTR development stands out the prismatic block which uses coated fuel microspheres named TRISO pressed into cylinders and assembled in hexagonal graphite blocks staked to form columns. The graphite blocks contain flow channels around the fuel cylinders for the helium coolant. In this study an analysis is performed using the CFD code CFX 13.0 on a prismatic fuel assembly in order to investigate its thermo-fluid dynamic performance. The simulations were made in a 1/12 fuel element model of the GT-MHR design which was developed by General Atomics. A numerical mesh verification process based on the Grid Convergence Index (GCI) was performed using five progressively refined meshes to assess the numerical uncertainty of the simulation and determine adequate mesh parameters. An analysis was also performed to evaluate different methods to define the inlet and outlet boundary conditions. In this study simulations of models with and without inlet and outlet plena were compared, showing that the presence of the plena offers a more realistic flow distribution. (author)

  12. CALIOP: a multichannel design code for gas-cooled fast reactors. Code description and user's guide

    International Nuclear Information System (INIS)

    Thompson, W.I.

    1980-10-01

    CALIOP is a design code for fluid-cooled reactors composed of parallel fuel tubes in hexagonal or cylindrical ducts. It may be used with gaseous or liquid coolants. It has been used chiefly for design of a helium-cooled fast breeder reactor and has built-in cross section information to permit calculations of fuel loading, breeding ratio, and doubling time. Optional cross-section input allows the code to be used with moderated cores and with other fuels

  13. Conceptual core design study for Japan sodium-cooled fast reactor: Review of sodium void reactivity worth evaluation

    International Nuclear Information System (INIS)

    Ohki, Shigeo

    2012-01-01

    The conceptual core design study for a large-scale Japan sodium-cooled fast reactor (JSFR) have been carried out in the framework of the FaCT project. The reference “High-internal conversion” core can satisfy the requirements for enhanced safety, as well as achieving economic competitiveness. In order to increase the design reliability, more rigorous uncertainty evaluation is important. Development of the verification and validation methodology of the core neutronic design method is currently underway. (author)

  14. DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-[TEMPERATURE GAS-COOLED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

    2016-11-01

    A point design for a graphite-moderated, high-temperature, gas-cooled test reactor (HTG TR) has been developed by Idaho National Laboratory (INL) as part of a United States (U.S.) Department of Energy (DOE) initiative to explore and potentially expand the existing U.S. test reactor capabilities. This paper provides a summary of the design and its main attributes. The 200 MW HTG TR is a thermal-neutron spectrum reactor composed of hexagonal prismatic fuel and graphite reflector blocks. Twelve fuel columns (96 fuel blocks total and 6.34 m active core height) are arranged in two hexagonal rings to form a relatively compact, high-power density, annular core sandwiched between inner, outer, top, and bottom graphite reflectors. The HTG-TR is designed to operate at 7 MPa with a coolant inlet/outlet temperature of 325°C/650°C, and utilizes TRISO particle fuel from the DOE AGR Program with 425 ?m uranium oxycarbide (UCO) kernels and an enrichment of 15.5 wt% 235U. The primary mission of the HTG TR is material irradiation and therefore the core has been specifically designed and optimized to provide the highest possible thermal and fast neutron fluxes. The highest thermal neutron flux (3.90E+14 n/cm2s) occurs in the outer reflector, and the maximum fast flux levels (1.17E+14 n/cm2s) are produced in the central reflector column where most of the graphite has been removed. Due to high core temperatures under accident conditions, all the irradiation test facilities have been located in the inner and outer reflectors where fast flux levels decline. The core features a large number of irradiation positions with large test volumes and long test lengths, ideal for thermal neutron irradiation of large test articles. The total available test volume is more than 1100 liters. Up to four test loop facilities can be accommodated with pressure tube boundaries to isolate test articles and test fluids (e.g., liquid metal, liquid salt, light water) from the helium primary coolant system.

  15. Nuclear powerplant with closed gas-cooling circuit

    International Nuclear Information System (INIS)

    Haferkamp, D.; Hodzic, A.; Winter, U.

    1976-01-01

    Disclosed is a nuclear power plant comprising a pressure-tight safety vessel surrounding the entire plant, an inner vessel of reinforced concrete, a high-temperature reactor contained in the inner vessel, a gas turbine assembly having a turbine and a high- and low-pressure compressor located in a horizontally oriented chamber below the reactor, a plurality of heat exchange units positioned in a plurality of vertically oriented pods spaced radially symmetrically about the reactor and suitable conduits for carrying the reactive gas between the system components. The conduits are arranged in generally horizontally and vertically oriented straight lines, and the conduits for carrying low-pressure gas comprise a horizontal system positioned beneath the turbine assembly having a plurality of coaxial connecting tubes, collectors and distributors as well as normal conduits, so that high pressure gas flows in the internal passage and low-pressure gas flows in the outer passage. 22 claims, 7 figures

  16. Hot gas path component having near wall cooling features

    Science.gov (United States)

    Miranda, Carlos Miguel; Kottilingam, Srikanth Chandrudu; Lacy, Benjamin Paul

    2017-11-28

    A method for providing micro-channels in a hot gas path component includes forming a first micro-channel in an exterior surface of a substrate of the hot gas path component. A second micro-channel is formed in the exterior surface of the hot gas path component such that it is separated from the first micro-channel by a surface gap having a first width. The method also includes disposing a braze sheet onto the exterior surface of the hot gas path component such that the braze sheet covers at least of portion of the first and second micro-channels, and heating the braze sheet to bond it to at least a portion of the exterior surface of the hot gas path component.

  17. Study on the nuclear heat application system with a high temperature gas-cooled reactor and its safety evaluation (Thesis)

    International Nuclear Information System (INIS)

    Inaba, Yoshitomo

    2008-03-01

    Aiming at the realization of the nuclear heat application system with a High Temperature Gas-cooled Reactor (HTGR), research and development on the whole evaluation of the system, the connection technology between the HTGR and a chemical plant such as the safety evaluation against the fire and explosion and the control technology, and the vessel cooling system of the HTGR were carried out. In the whole evaluation of the nuclear heat application system, an ammonia production system using nuclear heat was examined, and the technical subjects caused by the connection of the chemical plant to the HTGR were distilled. After distilling the subjects, the safety evaluation method against the fire and explosion to the reactor, the mitigation technology of thermal disturbance to the reactor, and the reactor core cooling by the vessel cooling system were discussed. These subjects are very important in terms of safety. About the fire and explosion, the safety evaluation method was established by developing the process and the numerical analysis code system. About the mitigation technology of the thermal disturbance, it was demonstrated that the steam generator, which was installed at the downstream of the chemical reactor in the chemical plant, could mitigate the thermal disturbance to the reactor. In order to enhance the safety of the reactor in accidents, the heat transfer characteristic of the passive indirect core cooling system was investigated, and the heat transfer equation considering both thermal radiation and natural convection was developed for the system design. As a result, some technical subjects related to safety in the nuclear heat application system were solved. (author)

  18. The Merging Galaxy Cluster A520 - A Broken-Up Cool Core, A Dark Subcluster, and an X-Ray Channel

    Science.gov (United States)

    Wang, Qian H.S.; Markevitch, Maxim; Giacintucci, Simona

    2016-01-01

    We present results from a deep Chandra X-ray observation of a merging galaxy cluster A520. A high-resolution gas temperature map reveals a long trail of dense, cool clumpsapparently the fragments of a cool core that has been stripped from the infalling subcluster by ram pressure. The clumps should still be connected by the stretched magnetic field lines. The observed temperature variations imply that thermal conductivity is suppressed by a factor greater than 100 across the presumed direction of the magnetic field (as found in other clusters), and is also suppressed along the field lines by a factor of several. Two massive clumps in the periphery of A520, visible in the weak-lensing mass map and the X-ray image, have apparently been completely stripped of gas during the merger, but then re-accreted the surrounding high-entropy gas upon exit from the cluster. The mass clump that hosted the stripped cool core is also re-accreting hotter gas. An X-ray hydrostatic mass estimate for the clump that has the simplest geometry agrees with the lensing mass. Its current gas mass to total mass ratio is very low, 1.5 percent to 3 percent, which makes it a "dark subcluster." We also found a curious low X-ray brightness channel (likely a low-density sheet in projection) going across the cluster along the direction of an apparent secondary merger. The channel may be caused by plasma depletion in a region of an amplified magnetic field (with plasma Beta approximately equal to 10-20). The shock in A520 will be studied in a separate paper.

  19. Performance and economic enhancement of cogeneration gas turbines through compressor inlet air cooling

    Science.gov (United States)

    Delucia, M.; Bronconi, R.; Carnevale, E.

    1994-04-01

    Gas turbine air cooling systems serve to raise performance to peak power levels during the hot months when high atmospheric temperatures cause reductions in net power output. This work describes the technical and economic advantages of providing a compressor inlet air cooling system to increase the gas turbine's power rating and reduce its heat rate. The pros and cons of state-of-the-art cooling technologies, i.e., absorption and compression refrigeration, with and without thermal energy storage, were examined in order to select the most suitable cooling solution. Heavy-duty gas turbine cogeneration systems with and without absorption units were modeled, as well as various industrial sectors, i.e., paper and pulp, pharmaceuticals, food processing, textiles, tanning, and building materials. The ambient temperature variations were modeled so the effects of climate could be accounted for in the simulation. The results validated the advantages of gas turbine cogeneration with absorption air cooling as compared to other systems without air cooling.

  20. MHD/gas turbine systems designed for low cooling water requirements

    International Nuclear Information System (INIS)

    Annen, K.D.; Eustis, R.H.

    1983-01-01

    The MHD/gas turbine combined-cycle system has been designed specifically for applications where the availability of cooling water is very limited. The base case systems which were studied consist of a coal-fired MHD plant with an air turbine bottoming plant and require no cooling water. In addition to the base case systems, systems were considered which included the addition of a vapor cycle bottoming plant to improve the thermal efficiency. These systems require a small amount of cooling water. The results show that the MHD/gas turbine systems have very good thermal and economic performances. The base case I MHD/gas turbine system (782 MW /SUB e/ ) requires no cooling water, has a heat rate which is 13% higher, and a cost of electricity which is only 7% higher than a comparable MHD/steam system (878 MW /SUB e/ ) having a cooling tower heat load of 720 MW. The case I vapor cycle bottomed systems have thermal and economic performances which approach and even exceed those of the MHD/steam system, while having substantially lower cooling water requirements. Performances of a second-generation MHD/gas turbine system and an oxygen-enriched, early commercial system are also evaluated. An analysis of nitric oxide emissions shows compliance with emission standards

  1. Effect of horizontal flow on the cooling of the moderator brick in the advanced gas-cooled reactor

    International Nuclear Information System (INIS)

    Ganesan, P.; He, S.; Hamad, F.; Gotts, J.

    2011-01-01

    The paper reports an investigation of the effect of the horizontal cross flow on the temperature of the moderator brick in UK Advanced Gas-cooled Reactor (AGR) using computational fluid dynamics (CFD) with a conjugate heat transfer model for the solid and fluid. The commercial software package of ANSYS Fluent is used for this purpose. The CFD model comprises the full axial length of one-half of a typical fuel channel (assuming symmetry) and part of neighbouring channels on either side. Two sets of simulations have been carried out, namely, one with cross flow and one without cross flow. The effect of cross flow has subsequently been derived by comparing the results from the two groups of simulations. The study shows that a small cross flow can have a significant effect on the cooling of the graphite brick, causing the peak temperature of the brick to reduce significantly. Two mechanisms are identified to be responsible for this. Firstly, the small cross flow causes a significant redistribution of the main axial downward flow and this leads to an enhancement of heat transfer in some of the small clearances, and an impairment in others although overall, the enhancement is dominant leading to a better cooling. Secondly, the cross flow makes effective use of the small clearances between the key/keyway connections which increases the effective heat transfer area, hence increasing the cooling. Under the conditions of no cross flow, these areas remain largely inactive in heat transfer. The study shows that the cooling of the moderator is significantly enhanced by the cross flow perpendicular to the main cooling flow. (author)

  2. 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

  3. Core Design and Deployment Strategy of Heavy Water Cooled Sustainable Thorium Reactor

    Directory of Open Access Journals (Sweden)

    Naoyuki Takaki

    2012-08-01

    Full Text Available Our previous studies on water cooled thorium breeder reactor based on matured pressurized water reactor (PWR plant technology concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array and using heavy water as coolant is appropriate for achieving better breeding performance and higher burn-up simultaneously [1–6]. One optimum core that produces 3.5 GW thermal energy using Th-233U oxide fuel shows a breeding ratio of 1.07 and averaged burn-up of about 80 GWd/t with long cycle length of 1300 days. The moderator to fuel volume ratio is 0.6 and required enrichment of 233U for the fresh fuel is about 7%. The coolant reactivity coefficient is negative during all cycles despite it being a large scale breeder reactor. In order to introduce this sustainable thorium reactor, three-step deployment scenario, with intermediate transition phase between current light water reactor (LWR phase and future sustainer phase, is proposed. Both in transition phase and sustainer phase, almost the same core design can be applicable only by changing fissile materials mixed with thorium from plutonium to 233U with slight modification in the fuel assembly design. Assuming total capacity of 60 GWe in current LWR phase and reprocessing capacity of 800 ton/y with further extensions to 1600 ton/y, all LWRs will be replaced by heavy water cooled thorium reactors within about one century then thorium reactors will be kept operational owing to its potential to sustain fissile fuels while reprocessing all spent fuels until exhaustion of massive thorium resource.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  6. Material for shutting down gas cooled nuclear reactors

    International Nuclear Information System (INIS)

    Jackson, F.

    1977-01-01

    Some disadvantage of conventional emergency shutdown means for nuclear reactors employing a supply of B steel shot or B powder are mentioned. With regard to B powder it is stated that there is some uncertainty as to whether the powder once dispersed into the core will settle in the active part of the core in sufficient quantities to ensure shutdown. The system described aims to avoid these disadvantages. Pellets are provided comprising a solid neutron poison material and a solid organic substance that remains solid at the relatively low temperature normally expected to prevail in the reactor coolant channel away from the reactor core. The organic substance melts at a higher temperature expected to prevail in the coolant channel within the core., and is adherent on melting to the coolant channel wall and to the solid neutron poison, being thus capable of causing adherence of the latter to the coolant channel wall in the reactor core. The pellets are preferably given a moisture resistant coating to prevent them sticking together and to impart free flowing characteristics. The neutron poison may consist of B, Cd, Gd, or their compounds, and for the coating a suitable polymer may be used. Steel filings may be incorporated in the pellets to aid easy flowing under gravity. Examples of manufacture of the pellets are given. (U.K.)

  7. A new small modular high-temperature gas-cooled reactor plant concept based on proven technology

    International Nuclear Information System (INIS)

    McDonald, C.F.; Goodjohn, A.J.

    1982-01-01

    Based on the established and proven high-temperature gas-cooled reactor (HTGR) technologies from the Peach Bottom 1 and Fort St. Vrain utility-operated units, a new small modular HTGR reactor is currently being evaluated. The basic nuclear reactor heat source, with a prismatic core, is being designed so that the decay heat can be removed by passive means (i.e., natural circulation). Although this concept is still in the preconceptual design stage, emphasis is being placed on establishing an inherently safe or benign concept which, when engineered, will have acceptable capital cost and power generation economics. The proposed new HTGR concept has a variety of applications, including electrical power generation, cogeneration, and high-temperature process heat. This paper discusses the simplest application, i.e., a steam Rankine cycle electrical power generating version. The gas-cooled modular reactor concepts presented are based on a graphite moderated prismatic core of low-power density (i.e., 4.1 W/cm 3 ) with a thermal rating of 250 MW(t). With the potential for inherently safe characteristics, a new small reactor could be sited close to industrial and urban areas to provide electrical power and thermal heating needs (i.e., district and space heating). Incorporating a multiplicity of small modular units to provide a larger power output is also discussed. The potential for a small, inherently safe HTGR reactor concept is highlighted

  8. Study of the oxidation mechanisms between impurities and surfaces applied to the future gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Duval, A.

    2010-01-01

    Inconel 617, main candidate for the heat exchangers of the gas-cooled next generation of nuclear reactors has been investigated. Two different problems occurring in the cooling system splits the study into two parts. Oxidizing impurities contained in the coolant can cause severe corrosion at 850 C. Radioactive impurities, coming from the fission reaction of the core can, in another hand contaminate the cooling loop and cause radioprotection problem for the maintenance and dismantling operations. Firstly, oxidizing gas partial pressure influence on oxidation of IN 617 at 850 C was investigated varying oxygen and water vapour partial pressure between 1.10 -5 mbar and 200 mbar. Oxide layers were characterized using XPS, SEM, EDX, GD-OES, XRD. Influence of partial pressure on layers structure and composition was determined. Effect of water vapour and partial pressure on growth mechanisms were also investigated. The second part of this study is focused on diffusion of Ag, stable isotope of Ag-110m in IN617 alloy and in the oxide layer forming at its surface at 850 C. Concentration profiles were obtained by GD-OES calibrated analysis. Diffusion coefficient could be obtained from these diffusion profiles: volume diffusion and grain boundary diffusion coefficients for the diffusion in the alloy, and an apparent diffusion coefficient for the diffusion in the oxide, due to the porosity of the structure. (author) [fr

  9. Gas-cooled fast reactors. Motivation and presentation of the ENIGMA program in the MASURCA experimental critical facility

    International Nuclear Information System (INIS)

    Tommasi, Jean; Jacqmin, Robert; Mellier, Frederic

    2005-01-01

    This paper describes a new experimental physics program in support of gas cooled fast reactor (GCFR) design studies, called ENIGMA, to be performed in the MASURCA critical facility at CEA-Cadarache, France. The prospective GCFR design studies at CEA are presented, as well as the specific neutronics features needing an extension of the validation of calculation tools and nuclear data. The relevant existing experiments are briefly reviewed and the need for new experimental data is pointed out. The first phase of the proposed new experiments includes a reference core with a representative spectrum, and a series of central core substitutions involving spectrum shifts, streaming studies, low-grade Pu substitutions, innovative material (Si, Zr) substitutions. Reflector substitution zones will include elements foreseen for the reflectors (Si, Zr, C). Subsequent phases will involve larger amounts of low-grade Pu or innovative materials, and configurations representative of experimental and demonstration GCFRs. (author)

  10. DELIGHT-B/REDEL, point reactivity burnup code for high-temperature gas-cooled reactor cells

    International Nuclear Information System (INIS)

    Shindo, Ryuiti; Watanabe, Takashi.

    1977-03-01

    Code DELIGHT-2 was previously developed to analyze cell burnup characteristics and to produce few-group constants for core burnup calculation in high-temperature gas-cooled reactors. In the code, burnup dependency of the burnable poison, boron-10, is considered with the homogeneous model of space. In actuality, however, the burnable poison is used as homogeneous rods or uniform rods of small granular poison and graphite, to control the reactivity and power distribution. Precise analysis of the burnup characteristics is thus difficult because of the heterogeneity due to the configuration of poison rods. In cell burnup calculation, the DELIGHT-B, which is a modification of DELIGHT-2, takes into consideration this heterogeneous effect. The auxiliary code REDEL, a reduction of DELIGHT-B, used in combination with 3 dimensional diffusion code CITATION, is for core burnup calculation with the macro-scopic cross section model. (auth.)

  11. EXCURS: a computing programme for analysis of core transient behaviour in a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Saito, Shinzo

    1977-09-01

    In the code EXCURS developed for core transient behaviour calculation of a sodium-cooled fast reactor, a one-channel model is used to represent thermal behaviour of the reactor core. Calculations are made for three different channels; i.e. average, hot and hottest. In the average channel the power density and coolant velocity are equal to the mean values of the whole core. In the hot channel, a maximum power density of the core and a specific coolant velocity are introduced. In the hottest channel, engineering hot channel factors are considered to the hot channel. A one-point neutron kinetics equation with six delayed neutron groups is used to calculate the time-dependent power behaviour. Externally introduced reactivity effect and control rod movement in the case of a scram are taken into account. In the feedback effects evaluated on the basis of the average channel temperatures are considered Doppler effect, fuel axial expansion, cladding expansion, coolant expansion and structure expansion. The decay heat after reactor scram is also considered. Heat balance is taken in each cross section, neglecting the axial heat transfer except for the coolant region. Temperature dependence of the physical properties of materials is considered by second-order polynomials approximation, and also the fuel melting process. Each channel can be divided into a maximum of 20 regions in both radially and axially. The reactor core transient behaviour due to reactivity insertion or loss-of-coolant flow can be studied by EXCURS. The calculated results are plotted optionally by connected code EXPLOT. (auth.)

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

    International Nuclear Information System (INIS)

    Monteith, H.C.

    1978-10-01

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

  13. The development of the gas cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Kernforschungszentrum Karlsruhe

    1975-01-01

    A survey of the present technological state is given on the basis of the developments made so far. Some milestones of development - e.g. the German gas breeder memorandum, the Gas Breeder Reactor Association the results of the BR-2 radiation experiments and of GfK-KWU design and safety studies - are described. The problems connected with a large store of plutonium are also discussed. (UA/AK) [de

  14. The design, safety and project development status of the modular high temperature gas-cooled reactor in the United States

    International Nuclear Information System (INIS)

    Mears, L.D.; Dean, R.A.

    1987-01-01

    The cooperative government and industry Modular High Temperature Gas-Cooled Reactor (MHTGR) Program in the United States has advanced a 350 MW(t) plant design through the conceptual development stage. The system incorporates an annular core of prismatic fuel elements within a steel pressure vessel connected, in a side-by-side arrangement, by a concentric duct to a second steel vessel containing a steam generator and helium coolant circulator. The reference plant design consists of four reactor modules installed in separate below-grade silos, providing steam to two conventional turbine generators. The nominal net plant output is 540 MW(e). The small reactor system takes unique advantage of the high temperature capability of the refractory coated fuel and the large thermal inertia of the graphite moderator to provide a design capable of withstanding a complete loss of active core cooling without causing excessive core heatup and significant release of fission products from the fuel. Present program activities are concentrated on interactions with the Nuclear Regulatory Commission aimed at obtaining a Licensability Statement. A project initiative to build a prototype plant which would demonstrate the MHTGR-unique licensing process, plant performance, costs and schedule plus establish an industrial infrastructure to proceed with follow-on commercial MHTGR plants by the turn of the century, is being undertaken by the utility/vendor participants (author)

  15. HEXEREI: a multi-channel heat conduction convection code for use in transient thermal hydraulic analysis of high-temperature, gas-cooled reactors. Interim report

    International Nuclear Information System (INIS)

    Giles, G.E.; DeVault, R.M.; Turner, W.D.; Becker, B.R.

    1976-05-01

    A description is given of the development and verification of a generalized coupled conduction-convection, multichannel heat transfer computer program to analyze specific safety questions involving high temperature gas-cooled reactors (HTGR). The HEXEREI code was designed to provide steady-state and transient heat transfer analysis of the HTGR active core using a basic hexagonal mesh and multichannel coolant flow. In addition, the core auxiliary cooling systems were included in the code to provide more complete analysis of the reactor system during accidents involving reactor trip and cooling down on the auxiliary systems. Included are brief descriptions of the components of the HEXEREI code and sample HEXEREI analyses compared with analytical solutions and other heat transfer codes

  16. A UKAEA review of gas-cooled reactors in the United Kingdom

    International Nuclear Information System (INIS)

    Heath, E.C.; Knowles, A.N.

    1983-01-01

    The commercial use of nuclear power for electrical generation commenced in the UK in the 1950s with the Calder Hall reactors. Based on this concept, eighteen commercial reactor units, with two further units outside the UK, were constructed and have been in operation for periods ranging from 10 to 19 years. The paper reviews this experience mainly from the aspects of safety and the achieved costs, which compare favourably with current figures for fossil fired generation. The further development of the gas-cooled system in the UK commenced with the construction of the Windscale AGR, which came into operation in 1962. This led to the ordering of 14 large commercial AGR units, 4 of which have been in service since 1976, 6 are at an advanced stage of construction and 4 are at an early stage of construction. The paper reviews the main safety features of the AGR and considers the costs, taking achieved costs for the units which are in service and a combination of historical costs and projected costs for the units under construction. Again a clear advantage over fossil fuelled stations is shown. The paper also includes a preliminary account of the use of the prototype AGR at Windscale for the series of experiments concerning plateout, over-temperature in the fuel and simulated fault transients in the core which were carried out earlier in 1981. (author)

  17. Design study of the experimental multi-purpose high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Tsunoda, Ryokichi

    1981-01-01

    In this paper, the design study carried out since 1973 is outlined. The basic conceptual design was performed in fiscal 1973. In this design, concept was established on the total system of the experimental high temperature gas-cooled reactor including heat-utilizing system. The first conceptual design was carried out in fiscal 1974. The range of design was limited to the experimental reactor and its direct heat-removing system. The part 2 of the first conceptual design was performed in fiscal 1975, and the system design concerning the plant characteristics was made. The part 1 of the adjustment design was carried out in fiscal 1976, and the subject was the adjustment design of plant systems. The part 2 was performed in fiscal 1977, and the characteristics of plant control system were analyzed. In fiscal 1978, the analysis of flow characteristics in the core was made. The integrated system design was carried out in fiscal 1979, and the design of the total plant system except heat-utilizing system was started again. The part 1 of the detailed design was performed in fiscal 1980, and in addition, the possibility of increasing power output was examined. The construction cost of the experimental reactor plant estimated in 1979 was far higher than that in 1973. (Kako, I.)

  18. Effect of Target Configuration on the Neutronic Performance of the Gas-Cooled ADS

    CERN Document Server

    Biss, K; Shetty, N; Nabbi, R

    2013-01-01

    With the utilization of nuclear energy transuranic elements like Pu, Am and Cm are produced causing high, long term radioactivity and radio toxicity, respectively. To reduce the radiological impact on the environment and to the repository Partitioning and Transmutation is considered as an efficient way. In this respect comprehensive research works are performed at different research institutes worldwide. The results show that the transmutation of TRU is achieved with fast neutrons due to the higher fission probability. Based on Accelerator Driven Systems (ADS) those neutrons are used in a particular system, in which mainly liquid metal eutectic (lead bismuth) is used as coolant. The neutronic performance of an ADS system based on gas cooling was studied in this work by using the simulation tool MCNPX. The usage of the Monte-Carlo method in MCNPX allows the simulation of the physical processes in a 3D-model of the core. In dependence of the spallation target material and design several parameters like the mult...

  19. Gas-Cooled Thermal Reactor Program. Semiannual technical progress report, October 1, 1982-March 3, 1983

    International Nuclear Information System (INIS)

    1983-06-01

    This report provides descriptions and results of the technical effort during the first half of FY 83 on the Gas-Cooled Thermal Reactor Program. The work on Integration and Management (WBS 01) includes the preparation of the Advanced Systems Concept Evaluation Plan and the Advanced Systems Technology Development Plan in addition to the program management activities. The Market Definition (WBS 03) efforts considered the application of the Modular Reactor System with reforming (MRS-R) to the production of methanol and ammonia and the refining of petroleum. Within the Plant Technology (WBS 13) task there were activities to develop anlytical methods for investigation of Coolant Transport Behavior and to define methods and criteria for High Temperature Structural Engineering design. In addition to the work on the advanced HTGR for process heat users, new activities were initiated in support of the HTGR-SC/C Lead plant Protect (WBS 30 and 31). The Plant Simulation task (WBS 31) was initiated to develop a computer code for simulation of plant operation and for plant transient systems analysis. The efforts on the advanced HTGR systems was performed under the Modular Systems task (WBS 41) to study the potential for multiple small reactors to provide lower costs, improved safety, and higher availability than the large monolithic core reactors

  20. Gas-cooled fast reactor fuel-cost assessment. Final report, October 1978-September 1979

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, M.L.

    1979-01-01

    This program, contracted to provide a Gas Cooled Fast Reactor (GCFR) fuel assembly fabrication cost assessment, comprised the following basic activities: establish agreement on the ground rules for cost assessment, prepare a fuel factory flow sheet, and prepare a cost assessment for fuel assembly fabrication. Two factory sizes, 250 and 25 MTHM/year, were considered for fuel assembly fabrication cost assessment. The work on this program involved utilizing GE LMFBR cost assessment and fuel factory studies experience to provide a cost assessment of GCFR fuel assembly fabrication. The recent impact of highly sensitive safety and safeguards environment policies on fuel factory containment, safety, quality assurance and safeguards costs are significantly higher than might have been expected just a few years ago. Fuel assembly fabrication costs are significant because they represent an estimated 30 to 60% of the total fuel cycle costs. In light of the relative high cost of fabrication, changes in the core and assembly design may be necessary in order to enhance the overall fuel cycle economics. Fabrication costs are based on similar operations and experience used in other fuel cycle studies. Because of extrapolation of present technology (e.g., remote fuel fabrication versus present contact fabrication) and regulatory requirements, conservative cost estimates were made.