Test plan for core sampling drill bit temperature monitor

International Nuclear Information System (INIS)

Francis, P.M.

1994-01-01

At WHC, one of the functions of the Tank Waste Remediation System division is sampling waste tanks to characterize their contents. The push-mode core sampling truck is currently used to take samples of liquid and sludge. Sampling of tanks containing hard salt cake is to be performed with the rotary-mode core sampling system, consisting of the core sample truck, mobile exhauster unit, and ancillary subsystems. When drilling through the salt cake material, friction and heat can be generated in the drill bit. Based upon tank safety reviews, it has been determined that the drill bit temperature must not exceed 180 C, due to the potential reactivity of tank contents at this temperature. Consequently, a drill bit temperature limit of 150 C was established for operation of the core sample truck to have an adequate margin of safety. Unpredictable factors, such as localized heating, cause this buffer to be so great. The most desirable safeguard against exceeding this threshold is bit temperature monitoring . This document describes the recommended plan for testing the prototype of a drill bit temperature monitor developed for core sampling by Sandia National Labs. The device will be tested at their facilities. This test plan documents the tests that Westinghouse Hanford Company considers necessary for effective testing of the system

Numerical simulation on coolant flow and heat transfer in core

International Nuclear Information System (INIS)

Yao Zhaohui; Wang Xuefang; Shen Mengyu

1997-01-01

To simulate the coolant flow and the heat transfer characteristics of a core, a computer code, THAPMA (Thermal Hydraulic Analysis Porous Medium Analysis) has been developed. In THAPMA code, conservation equations are based on a porous-medium formulation, which uses four parameters, i.e, volume porosity, directional surface porosity, distributed resistance, and distributed heat source (sink), to model the effects of fuel rods and other internal solid structures on flow and heat transfer. Because the scheme and the solution are very important in accuracy and speed of calculation, a new difference scheme (WSUC) has been used in the energy equation, and a modified PISO solution method have been employed to simulate the steady/transient states. The code has been proved reliable and can effectively solve the transient state problem by several numerical tests. According to the design of Qinshan NPP-II, the flow and heat transfer phenomena in reactor core have been numerically simulated. The distributions of the velocity and the temperature can provide a theoretical basis for core design and safety analysis

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Performance test of ex-core high temperature and high pressure water loop test equipment (Contract research)

International Nuclear Information System (INIS)

Nakano, Hiroko; Uehara, Toshiaki; Takeuchi, Tomoaki; Shibata, Hiroshi; Nakamura, Jinichi; Matsui, Yoshinori; Tsuchiya, Kunihiko

2016-03-01

In Japan Atomic Energy Agency, we started research and development so as to monitor the situations in the Nuclear Plant Facilities during a severe accident, such as a radiation-resistant monitoring camera, a radiation-resistant transmission system for conveying the in-core information, and a heat-resistant signal cable. As a part of developments of the heat-resistant signal cable, we prepared ex-core high-temperature and high-pressure water loop test equipment, which can simulate the conditions of BWRs and PWRs, for evaluating reliability and properties of sheath materials of the cable. This equipment consists of autoclave, water conditioning tank, high-pressure metering pump, preheater, heat exchanger and water purification equipment, etc. This report describes the basic design and the performance test results of ex-core high-temperature and high-pressure water loop test equipment. (author)

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

Science.gov (United States)

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

2006-01-01

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

Demonstration tests for HTGR fuel elements and core components with test sections in HENDEL

Energy Technology Data Exchange (ETDEWEB)

Miyamoto, Yoshiaki; Hino, Ryutaro; Inagaki, Yoshiyuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

1995-03-01

In the fuel stack test section (T{sub 1}) of the Helium Engineering Demonstration Loop (HENDEL), thermal and hydraulic performances of helium gas flows through a fuel rod channel and a fuel stack have been investigated for the High-Temperature Engineering Test Reactor (HTTR) core thermal design. The test data showed that the turbulent characteristics appearing in the Reynolds number above 2000: no typical behavior in the transition zone, and friction factors and heat transfer coefficients in the fuel channel were found to be higher than those in a smooth annular channel. Heat transfer behavior of gas flow in a fuel element channel with blockage and cross-flow through a gap between upper and lower fuel elements stacked was revealed using the mock-up models. On the other hand, demonstration tests have been performed to verify thermal and hydraulic characteristics and structural integrity related to the core bottom structure using a full-scale test facility named as the in-core structure test section (T{sub 2}). The sealing performance test revealed that the leakage of low-temperature helium gas through gaps between the permanent reflector blocks to the core was very low level compared with the HTTR design value and no change of the leakage flow rate were observed after a long term operation. The heat transfer tests including thermal transient at shutdown of gas circulators verified good insulating performance of core insulation structures in the core bottom structure and the hot gas duct; the temperature of the metal portion of these structure was below the design value. Examination of the thermal mixing characteristics indicated that the mixing of the hot helium gas started at a hot plenum and finished completely at downstream of the outlet hot gas duct. The present results obtained from these demonstration tests have been practically applied to the detailed design works and licensing procedures of the HTTR. (J.P.N.) 92 refs.

Evaluation report on SCTF Core-II test S2-19

International Nuclear Information System (INIS)

Ohnuki, Akira; Iwamura, Takamichi; Iguchi, Tadashi; Abe, Yutaka; Murao, Yoshio; Adachi, Hiromichi.

1991-03-01

Experimental studies using Slab Core Test Facility (SCTF) have revealed that the heat transfer enhancement in higher power bundles is mainly governed by the radial power ratio in core during the reflood in PWR-LOCA. As a physical mechanism for the heat transfer enhancement, it can be considered from the experimental evidence that the increase of upward steam flow rate in a higher power bundle which is caused by the higher steam production rate in the bundle gives the higher upward liquid flow rate in the bundle and the increase of the liquid flow rate gives the heat transfer enhancement. In order to develop a mechanistic model for the heat transfer enhancement based on this idea, the following relations should be identified quantitatively: (1) Relation between the steam production rate and the upward liquid flow rate, (2) Cross flow rate above the quench front and (3) Relation between the degree of heat transfer enhancement due to radial power ratio and the amount of increase of upward liquid flow rate. In this report, the above relation (3) was investigated experimentally as a step to develop the mechanistic model using the SCTF where the relation between the radial power ratio and the heat transfer enhancement has been made clear quantitatively. The degree of increase of heat transfer between two forced feed tests with the different flow rate in LPCI period was compared with the degree of heat transfer enhancement under a radial power ratio in the previous SCTF tests. The two forced feed tests were performed under the condition without any significant two-dimensional hydraulic behavior in core. The ratio of the mass flow rate between the two tests was about double. (author)

Evaluation report on CCTF core-II reflood test C2-6 (Run 64)

International Nuclear Information System (INIS)

Akimoto, Hajime; Iguchi, Tadashi; Sugimoto, Jun; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu.

1985-03-01

In order to evaluate the effect of the radial power profile on the system behavior and the core thermal hydraulic behavior during the reflood phase of a PWR LOCA, a test was performed using the Cylindrical Core Test Facility(CCTF) with the flat radial power profile. The test was conducted with the same total core power as that of the steep radial power test C2-5(Run 63). Through the comparisons of the results from these two tests, the following conclusions were obtained: (1) The radial power profile in the core has weak effect on the thermal hydraulic behavior in the primary system except the core. (2) Almost the same differential pressure was observed at various elevations in the periphery of the core regardless of different radial power profile. The result suggests that the core differential pressure is determined mainly by the total power and the total stored energy rather than by the local power and the local stored energy. (3) The test results support the single channel core model with the average power rod used in the reactor safety analysis codes such as REFLA-1DS, WREM for the evaluation of the overall system behavior. (4) In the steep radial power test, the heat transfer coefficient in the central(high power) region was higher than that in the peripheral(low power) region. The tendency was not explained by the estimation with the heat transfer correlation developed by Murao and Sugimoto assuming that the void fraction was uniform in a horizontal cross section. It is necessary to study more the dependency of core heat transfer on the radial power profile in the wide core. (author)

Development and test of a space-reactor-core heat pipe

International Nuclear Information System (INIS)

Merrigan, M.A.; Runyan, J.E.; Martinez, H.E.; Keddy, E.S.

1983-01-01

A heat pipe designed to meet the heat transfer requirements of a 100-kW/sub e/ space nuclear power system has been developed and tested. General design requirements for the device included an operating temperature of 1500 0 K with an evaporator radial flux density of 100 w/cm 2 . The total heat-pipe length of 2 m comprised an evaporator length of 0.3 m, a 1.2-m adiabatic section, and a condenser length of 0.5 m. A four-artery design employing screen arteries and distribution wicks was used with lithium serving as the working fluid. Molybdenum alloys were used for the screen materials and tube shell. Hafnium and zirconium gettering materials were used in connection with a pre-purified distilled lithium charge to ensure internal chemical compatibility. After initial performance verification, the 14.1-mm i.d. heat pipe was operated at 15 kW throughput at 1500 0 K for 100 hours. No performance degradation was observed during the test

Evaluation report on CCTF Core-II reflood Test C2-15 (Run 75)

International Nuclear Information System (INIS)

Okubo, Tsutomu; Iguchi, Tadashi; Akimoto, Hajime; Murao, Yoshio

1992-01-01

This report presents an evaluation on the CCTF Core-II Test C2-15 (Run 75). The purpose of the test is to investigate whether the thermo-hydrodynamic behavior is different between the CCTF and the FLECHT-SET reflooding tests. For this purpose, test conditions of the present test were set as close as possible to those of concerned FLECHT-SET 2714B experiment, taking account of differences in facility design. Investigating results of both the tests, the following conclusions are obtained: (1) Some discrepancies were observed in the measured test conditions between the two tests. Out of them, difference in the Acc injection duration was large and affected test results, such as the water accumulation in the downcomer and the core and the core cooling, during the initial period. However, this effect was found to become small with time. (2) Taking account of this difference and the difference in the broken cold leg pressure loss coefficient between the two facilities, the overall reflooding behavior is judged to be similar in the two facilities. (3) The CCTF test results showed the core heat transfer enhancement in the higher power region due to its steep radial power distribution, whereas the FLECHT-SET did not due to its rather flat radial power distribution. This enhancement was observed significantly at 1.83 m but was smaller at the higher elevation. (4) The heat transfer was nearly identical between the two tests and an existing correlation could well predict the heat transfer coefficients of both the tests at the location where the heat transfer enhancement mentioned above (3) were small, during the time period when the effect of the difference in the Acc injection mentioned above (1) were small. (5) Therefore, the core cooling is expected to be almost the same in the CCTF and the FLECHT-SET under the same core boundary conditions and core radial power distribution. (author)

Some factors affecting radiative heat transport in PWR cores

International Nuclear Information System (INIS)

Hall, A.N.

1989-04-01

This report discusses radiative heat transport in Pressurized Water Reactor cores, using simple models to illustrate basic features of the transport process. Heat transport by conduction and convection is ignored in order to focus attention on the restrictions on radiative heat transport imposed by the geometry of the heat emitting and absorbing structures. The importance of the spacing of the emitting and absorbing structures is emphasised. Steady state temperature distributions are found for models of cores which are uniformly heated by fission product decay. In all of the models, a steady state temperature distribution can only be obtained if the central core temperature is in excess of the melting point of UO 2 . It has recently been reported that the MIMAS computer code, which takes into account radiative heat transport, has been used to model the heat-up of the Three Mile Island-2 reactor core, and the computations indicate that the core could not have reached the melting point of UO 2 at any time or any place. We discuss this result in the light of the calculations presented in this paper. It appears that the predicted stabilisation of the core temperatures at ∼ 2200 0 C may be a consequence of the artificially large spacing between the radial rings employed in the MIMAS code, rather than a result of physical significance. (author)

Water table tests of proposed heat transfer tunnels for small turbine vanes

Science.gov (United States)

Meitner, P. L.

1974-01-01

Water-table flow tests were conducted for proposed heat-transfer tunnels which were designed to provide uniform flow into their respective test sections of a single core engine turbine vane and a full annular ring of helicopter turbine vanes. Water-table tests were also performed for the single-vane test section of the core engine tunnel. The flow in the heat-transfer tunnels was shown to be acceptable.

Two-dimensional thermal-hydraulic behavior in core in SCTF Core-II forced feed reflood tests

International Nuclear Information System (INIS)

Iwamura, Takamichi; Sobajima, Makoto; Okubo, Tsutomu; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi

1987-01-01

Major purpose of the Slab Core Test Program is to investigate the two-dimensional thermal-hydraulic behavior in the core during the reflood phase of a PWR-LOCA. It was revealed in the previous Slab Core Test Facility (SCTF) Core-II test results that the heat transfer was enhanced in the higher power bundles and degraded in the lower power bundles in the non-uniform radial power profile tests. In order to separately evaluate the effect of the radial power (Q) distribution itself and the effect of the radial temperature (T) distribution, four tests were performed with steep Q and T, flat Q and T, steep Q and flat T, and flat Q and steep T. Based on the test results, it was concluded that the radial temperature distribution which accompanied the radial power distribution was the dominant factor of the two-dimensional thermal-hydraulic behavior in the core during the initial period. Selected data from these four tests are also presented in this report. Some data from Test S2-12 (steep Q, T) were compared with TRAC post-test calculations performed by the Los Alamos National Laboratory. (author)

Critical evaluation of molybdenum and its alloys for use in space reactor core heat pipes

International Nuclear Information System (INIS)

Lundberg, L.B.

1981-01-01

The choice of pure molybdenum as the prime candidate material for space reactor core heat pipes is examined, and the advantages and disadvantages of this material are brought into focus. Even though pure molybdenum heat pipes have been built and tested, this metal's high ductile-brittle transition temperature and modest creep strength place significant design restrictions on a core heat pipe made from it. Molybdenum alloys are examined with regard to their promise as potential replacements for pure molybdenum. The properties of TZM and molybdenum-rhenium alloys are examined, and it appears that Mo-Re alloys with 10 to 15 wt % rhenium offer the most advantage as an alternative to pure molybdenum in space reactor core heat pipes

Thermal Hydraulic Tests for Reactor Core Safety

Energy Technology Data Exchange (ETDEWEB)

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

2007-06-15

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

Radiant heating tests of several liquid metal heat-pipe sandwich panels

International Nuclear Information System (INIS)

Camarda, C.J.; Basiulis, A.

1983-08-01

Integral heat pipe sandwich panels, which synergistically combine the thermal efficiency of heat pipes and the structural efficiency of honeycomb sandwich construction, were conceived as a means of alleviating thermal stress problems in the Langley Scramjet Engine. Test panels which utilized two different wickable honeycomb cores, facesheets with screen mesh sintered to the internal surfaces, and a liquid metal working fluid (either sodium or potassium) were tested by radiant heating at various heat load levels. The heat pipe panels reduced maximum temperature differences by 31 percent with sodium working fluid and 45 percent with potassium working fluid. Results indicate that a heat pipe sandwich panel is a potential, simple solution to the engine thermal stress problem. Other interesting applications of the concept include: cold plates for electronic component and circuit card cooling, radiators for large space platforms, low distortion large area structures (e.g., space antennas) and laser mirrors

Laser Heating of the Core-Shell Nanowires

Science.gov (United States)

Astefanoaei, Iordana; Dumitru, Ioan; Stancu, Alexandru

2016-12-01

The induced thermal stress in a heating process is an important parameter to be known and controlled in the magnetization process of core-shell nanowires. This paper analyses the stress produced by a laser heating source placed at one end of a core-shell type structure. The thermal field was computed with the non-Fourier heat transport equation using a finite element method (FEM) implemented in Comsol Multiphysics. The internal stresses are essentially due to thermal gradients and different expansion characteristics of core and shell materials. The stress values were computed using the thermo elastic formalism and are depending on the laser beam parameters (spot size, power etc.) and system characteristics (dimensions, thermal characteristics). Stresses in the GPa range were estimated and consequently we find that the magnetic state of the system can be influenced significantly. A shell material as the glass which is a good thermal insulator induces in the magnetic core, the smaller stresses and consequently the smaller magnetoelastic energy. These results lead to a better understanding of the switching process in the magnetic materials.

DENSE MOLECULAR CORES BEING EXTERNALLY HEATED

Energy Technology Data Exchange (ETDEWEB)

Kim, Gwanjeong; Lee, Chang Won; Kim, Mi-Ryang [Radio Astronomy division, Korea Astronomy and Space Science Institute, 776 Daedeokdae-ro, Yuseong-gu, Daejeon, 34055 (Korea, Republic of); Gopinathan, Maheswar [Aryabhatta Research Institute of Observational Sciences, Manora Peak, Nainital 263129 (India); Jeong, Woong-Seob, E-mail: archer81@kasi.re.kr [Department of Astronomy and Space Science, University of Science and Technology, 217 Gajungro, Yuseong-gu, Daejeon, 34113 (Korea, Republic of)

2016-06-20

We present results of our study of eight dense cores, previously classified as starless, using infrared (3–160 μ m) imaging observations with the AKARI telescope and molecular line (HCN and N{sub 2}H{sup +}) mapping observations with the KVN telescope. Combining our results with the archival IR to millimeter continuum data, we examined the starless nature of these eight cores. Two of the eight cores are found to harbor faint protostars having luminosities of ∼0.3–4.4 L {sub ⊙}. The other six cores are found to remain starless and probably are in a dynamically transitional state. The temperature maps produced using multi-wavelength images show an enhancement of about 3–6 K toward the outer boundary of these cores, suggesting that they are most likely being heated externally by nearby stars and/or interstellar radiation fields. Large virial parameters and an overdominance of red asymmetric line profiles over the cores may indicate that the cores are set into either an expansion or an oscillatory motion, probably due to the external heating. Most of the starless cores show a coreshine effect due to the scattering of light by the micron-sized dust grains. This may imply that the age of the cores is of the order of ∼10{sup 5} years, which is consistent with the timescale required for the cores to evolve into an oscillatory stage due to external perturbation. Our observational results support the idea that the external feedback from nearby stars and/or interstellar radiation fields may play an important role in the dynamical evolution of the cores.

Numerical Analysis of Heat Storage and Heat Conductivity in the Concrete Hollow Core Deck Element

DEFF Research Database (Denmark)

Pomianowski, Michal Zbigniew; Heiselberg, Per; Jensen, Rasmus Lund

2011-01-01

extent these simplified models estimate the heat storage potential of precast hollow-core concrete decks correctly. This study investigates various approaches on how to model the heat transfer within the air void in the deck. Furthermore, it is analysed how different heat transfer models influence...... the overall heat transfer and heat storage in the hollow-core decks. The presented results allow comparison between detailed results from 2D-COMSOL simulations and simple 1D calculations from the whole building simulation tool such as BSim program and moreover, it is possible to validate the calculation...... method in BSim for the concrete deck element with air voids. Finally, this paper presents a comparison of the calculated heat conductivity of the hollow-core concrete deck and the measured heat conductivity for the same deck by using hot box apparatus....

Thermohydraulic behaviour and heat transfer in the molten core

International Nuclear Information System (INIS)

Reineke, H.H.

1977-01-01

Increasing the application of nuclear reactors to produce electrical power extremely unprobable accidents should be investigated too. In the Federal Republic of Germany, a research program is performed for some years engaged in accidents at light water reactors in which the melting of the reactor core is presumed. A part of this program is to investigate the thermohydraulic and the heat transfer behavior in an accumulation of molten core material. The knowledge of these events is necessary to analyse the accident exactly. Further on the results of this work are of great importance to build a catcher for the molten core material. As a result of the decay heat the molten material is heated up and the density differences induce a free convection motion. In this work the thermohydraulic behavior and the distribution of the escaping heat fluxes for several accumulations of molten core material were determined. The numerical methods for solving the system of partial differential equation were used to develop computer codes, able to compute the average and local heat fluxes at the walls enclosing the molten core material and the inside increase of the temperature. The numerical computations were confirmed and verified by experimental investigations. In these investigations the molten core material was always assumed as a homogeneous fluid. In this case, the results could be reproduced by simple power laws

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

Cause and countermeasure for heat up of HTTR core support plate at power rise tests

Energy Technology Data Exchange (ETDEWEB)

Fujimoto, Nozomu; Takada, Eiji; Nakagawa, Shigeaki; Tachibana, Yukio; Kawasaki, Kozo; Saikusa, Akio; Kojima, Takao; Iyoku, Tatuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

2002-01-01

HTTR has carried out many kinds of tests as power rise tests in which reactor power rises step by step after attained the first criticality. In the tests, temperature of a core support plate reached higher than expected at each power level, the temperature was expected to be higher than the maximum working temperature at 100% power level. Therefore, tests under the high temperature test operation mode, in which the core flow rate was different, were carried out to predict the temperature at 100% power precisely, and investigate the cause of the temperature rise. From the investigation, it was clear that the cause was gap flow in the core support structure. Furthermore, it was estimated that the temperature of the core support plate rose locally due to change in gap width between the core support plate and a seal plate due to change in core pressure drop. The maximum working temperature of the core support plate was revised. The integrity of core support plate under the revised maximum working temperature condition was confirmed by stress analyses. (author)

Fission product retention in TRISO coated UO2 particle fuels subjected to HTR simulated core heating tests

International Nuclear Information System (INIS)

Baldwin, C.A.; Kania, M.J.

1991-01-01

Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbounded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600 deg. C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800 deg. C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800 deg. C and above may exist. (author). 6 refs, 6 figs, 4 tabs

Fission product retention in TRISCO coated UO2 particle fuels subjected to HTR simulated core heating tests

International Nuclear Information System (INIS)

Baldwin, C.A.; Kania, M.J.

1990-11-01

Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbonded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600 degree C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800 degree C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800 degree C and above may exist. 6 refs., 6 figs., 4 tabs

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.

Experiment data report for semiscale MOD-1 tests S-03-A, S-03-B, S-03-C, and S-03-D (reflood heat transfer tests)

International Nuclear Information System (INIS)

1976-05-01

Recorded test data are presented for Tests S-03-A, S-03-B, S-03-C, and S-03-D of the Semiscale Mod-1 reflood heat transfer series (Test Series 3). The tests conducted in this series are separate effects core reflood tests performed to determine the reflood heat transfer characteristics of the 5.5 foot Mod-1 rod bundle. Tests S-03-A through S-03-D were forced-feed reflood tests in which the reflood rate was held constant during each test. The tests were conducted to investigate the effects on system response resulting from variations in operating conditions of pressure, temperature, core power, reflood coolant, subcooling, and peak heater rod thermocouple temperature at reflood initiation. Test S-03-A was conducted from an initial system temperature of about 230 0 F at a pressure of 20 psia. Tests S-03-B through S-03-D were conducted from an initial system temperature of about 290 0 F at a pressure of 60 psia. In all four tests, reflood coolant was injected directly into the core barrel by means of a specially designed core inlet manifold. The electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core during reflood. All four tests were conducted with a flat radial power profile. During reflood, core power was reduced from the initial level according to the American Nuclear Society (ANS) decay heat curve plus 20 percent for pressurized water reactor (PWR) core decay heat. The cold leg broken loop piping was open to the pressure suppression system (PSS). A separate steam supply system connected to the PSS was controlled to maintain constant pressure during the tests

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-15

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

System pressure effects on reflooding phenomena observed in the SCTF Core-I forced flooding tests

International Nuclear Information System (INIS)

Adachi, Hiromichi; Sudo, Yukio; Sobajima, Makoto; Iwamura, Takamichi; Osakabe, Masahiro; Ohnuki, Akira; Abe, Yutaka

1983-06-01

The Slab Core Test Facility was constructed to investigate two-dimensional thermo-hydrodynamics in the core and the interaction in fluid behavior between the core and the upper plenum during the last part of blowdown, refill and reflood phases of a posturated loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). The present report described the analytical results on the effects of system pressure on reflooding phenomena observed in Tests Sl-SH2, Sl-01 and Sl-02 which are belonging to the SCTF Core-I forced-feed reflooding test series. Nominal system pressures in these tests are 0.4, 0.2 and 0.15 MPa, respectively. By comparison among the data of these three tests, the effects of system pressure on thermo-hydrodynamic behavior in the pressure vessel including the core and the primary coolant loops of the SCTF can be clarified under the forced flooding condition. Major items investigated in the present report are (1) overall temperature behaviors in the core, (2) change of heat transfer coefficient and heat flux at the rod surface before the quench, (3) two-dimensional thermo-hydrodynamic behaviors in the core and upper plenum and (4) hot leg carryover. (author)

Elevated-Temperature Tests Under Static and Aerodynamic Conditions on Honeycomb-Core Sandwich Panels

Science.gov (United States)

Groen, Joseph M.; Johnson, Aldie E., Jr.

1959-01-01

Stainless-steel honeycomb-core sandwich panels which differed primarily in skin thicknesses were tested at elevated temperatures under static and aerodynamic conditions. The results of these tests were evaluated to determine the insulating effectiveness and structural integrity of the panels. The static radiant-heating tests were performed in front of a quartz-tube radiant heater at panel skin temperatures up to 1,5000 F. The aerodynamic tests were made in a Mach 1.4 heated blowdown wind tunnel. The tunnel temperature was augmented by additional heat supplied by a radiant heater which raised the panel surface temperature above 8000 F during air flow. Static radiant-heating tests of 2 minutes duration showed that all the panels protected the load-carrying structure about equally well. Thin-skin panels showed an advantage for this short-time test over thick-skin panels from a standpoint of weight against insulation. Permanent inelastic strains in the form of local buckles over each cell of the honeycomb core caused an increase in surface roughness. During the aero- dynamic tests all of the panels survived with little or no damage, and panel flutter did not occur.

HYDRATE CORE DRILLING TESTS

Energy Technology Data Exchange (ETDEWEB)

John H. Cohen; Thomas E. Williams; Ali G. Kadaster; Bill V. Liddell

2002-11-01

The ''Methane Hydrate Production from Alaskan Permafrost'' project is a three-year endeavor being conducted by Maurer Technology Inc. (MTI), Noble, and Anadarko Petroleum, in partnership with the U.S. DOE National Energy Technology Laboratory (NETL). The project's goal is to build on previous and ongoing R&D in the area of onshore hydrate deposition. The project team plans to design and implement a program to safely and economically drill, core and produce gas from arctic hydrates. The current work scope includes drilling and coring one well on Anadarko leases in FY 2003 during the winter drilling season. A specially built on-site core analysis laboratory will be used to determine some of the physical characteristics of the hydrates and surrounding rock. Prior to going to the field, the project team designed and conducted a controlled series of coring tests for simulating coring of hydrate formations. A variety of equipment and procedures were tested and modified to develop a practical solution for this special application. This Topical Report summarizes these coring tests. A special facility was designed and installed at MTI's Drilling Research Center (DRC) in Houston and used to conduct coring tests. Equipment and procedures were tested by cutting cores from frozen mixtures of sand and water supported by casing and designed to simulate hydrate formations. Tests were conducted with chilled drilling fluids. Tests showed that frozen core can be washed out and reduced in size by the action of the drilling fluid. Washing of the core by the drilling fluid caused a reduction in core diameter, making core recovery very difficult (if not impossible). One successful solution was to drill the last 6 inches of core dry (without fluid circulation). These tests demonstrated that it will be difficult to capture core when drilling in permafrost or hydrates without implementing certain safeguards. Among the coring tests was a simulated hydrate

Evaluation report on CCTF CORE-I REFLOOD TEST Cl-1, (Run 010)

International Nuclear Information System (INIS)

Sudoh, Takashi; Murao, Yoshio.

1983-09-01

This report describes the effects of the loop flow resistance on the thermohydraulic behavior in the primary system during the reflood phase. The investigation is based on the results of the test Cl-1 which was performed with increased loop flow resistance in the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute. The loop flow resistance was about 40% higher in the present test than in the reference test Cl-5. The results of two tests were compared and the following conclusions were obtained: 1) The total loop flow rate and the core flooding rate were reduced by about 20% with the increased loop flow resistance 2) The core heat transfer was also lowered, then, the turnaround and the quench times extended at the locations above the core midplane. 3) The measured maximum temperature in the core was 50 K higher for the present test than for the reference test. (author)

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

Energy Technology Data Exchange (ETDEWEB)

Ehud Greenspan

2008-09-30

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

Experiment data report for semiscale Mod-1, test S-02-7. Blowndown heat transfer test

International Nuclear Information System (INIS)

Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

1975-11-01

Recorded test data are presented for Test S-02-7 of the Semiscale Mod-1 blowdown heat transfer test series conducted to investigate the thermal and hydraulic phenomena accompanying an hypothesized loss-of-coolant accident (LOCA) in a water-cooled nuclear reactor system and to provide data for the assessment of the Loss-of-Fluid Test (LOFT) design basis. Test S-02-7 was conducted from an initial cold leg fluid temperature of 543 0 F and an initial pressure of 2,263 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient with full design core power (1.6 MW). An electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core with power set to provide a flat radial power profile. System flow was set to achieve the full design core temperature differential of 66 0 F. Blowdown to the pressure suppression system was accomplished without simulated emergency core cooling injection or pressure suppression system coolant spray. The uninterpreted data from Test S-02-7 are presented for future data analysis and test results reporting activities. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to assure that they are reasonable and consistent. Also included as an appendix are selected data from a test identified as Test S-02-7C. This test was an initial attempt at Test S-02-7 in which an inadvertent power trip occurred at 2.3 seconds after rupture. Selected data comparisons of the results from Test S-02-7 and S-02-7C are presented to indicate the repeatability of system behavior

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

International Nuclear Information System (INIS)

Setiyanto

2014-01-01

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

Demonstration of Efficient Core Heating of Magnetized Fast Ignition in FIREX project

Science.gov (United States)

Johzaki, Tomoyuki

2017-10-01

Extensive theoretical and experimental research in the FIREX ``I project over the past decade revealed that the large angular divergence of the laser generated electron beam is one of the most critical problems inhibiting efficient core heating in electron-driven fast ignition. To solve this problem, beam guiding using externally applied kilo-tesla class magnetic field was proposed, and its feasibility has recently been numerically demonstrated. In 2016, integrated experiments at ILE Osaka University demonstrated core heating efficiencies reaching > 5 % and heated core temperatures of 1.7 keV. In these experiments, a kilo-tesla class magnetic field was applied to a cone-attached Cu(II) oleate spherical solid target by using a laser-driven capacitor-coil. The target was then imploded by G-XII laser and heated by the PW-class LFEX laser. The heating efficiency was evaluated by measuring the number of Cu-K- α photons emitted. The heated core temperature was estimated by the X-ray intensity ratio of Cu Li-like and He-like emission lines. To understand the detailed dynamics of the core heating process, we carried out integrated simulations using the FI3 code system. Effects of magnetic fields on the implosion and electron beam transport, detailed core heating dynamics, and the resultant heating efficiency and core temperature will be presented. I will also discuss the prospect for an ignition-scale design of magnetized fast ignition using a solid ball target. This work is partially supported by JSPA KAKENHI Grant Number JP16H02245, JP26400532, JP15K21767, JP26400532, JP16K05638 and is performed with the support and the auspices of the NIFS Collaboration Research program (NIFS12KUGK057, NIFS15KUGK087).

Experiment on heat transfer in simulated molten core/concrete interaction

International Nuclear Information System (INIS)

Katsumura, Yukihiro; Hashizume, Hidetoshi; Toda, Saburo; Kawaguchi, Takahiro.

1993-01-01

In order to investigate heat transfer between molten core and concrete in LWR severe accidents, experiments were performed using water as the molten core, paraffin as the concrete, and air as gases from the decomposition of concrete. It was found that the heat transfer on the interface between paraffin and water were promoted strongly by the air gas. (author)

Solid-Core Heat-Pipe Nuclear Batterly Type Reactor

International Nuclear Information System (INIS)

Ehud Greenspan

2008-01-01

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

3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

International Nuclear Information System (INIS)

Chvetsov, I.; Volkov, A.

2000-01-01

For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)

Evaluation report on CCTF Core-II reflood test second shakedown test, C2-SH2 (Run 54)

International Nuclear Information System (INIS)

Iguchi, Tadashi; Sugimoto, Jun; Akimoto, Hajime; Okubo, Tsutomu; Murao, Yoshio

1985-03-01

A low power test (the initial averaged linear power density = 1.18 kW/m) and the base case test (1.4 kW/m) were performed with the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute, in order to study the effect of the power on the reflood phenomena. The former linear power density corresponds nearly to the scaled linear power density based on the current safety evaluation criterio. During the early period of the reflood ( 200s) the heat transfer coefficient became higher and resultantly the quench front advanced faster in the low power test. The core flooding rate was nearly identical between both tests, independently of the different power. The insensitiveness of the power to the core flooding rate was also observed in FLECHT-SET performed in the USA. A significatn large differential pressure oscillation at ECC ports was experienced in the low power test, and it may be important for the long term core cooling although it has not been taken note on the previous studies. (author)

Evaluation report on SCTF Core-II test S2-08

International Nuclear Information System (INIS)

Ohnuki, Akira; Iwamura, Takamichi; Abe, Yutaka; Murao, Yoshio; Adachi, Hiromichi.

1991-01-01

The present report investigates the effects of the difference of the core inlet subcooling during reflood in a PWR-LOCA on the thermal-hydraulic behaviors including two-dimensional behaviors in the pressure vessel in the Slab Core Test Facility (SCTF) Core-II tests under gravity feed mode. The following test results are examined: Tests S2-02 (Reference test) and Test S2-08 (High subcooling test). The degree of the difference of the subcooling between the two tests was about 20 to 35 K in the LPCI period. The following conclusions were obtained from this study: (1) Higher the subcooling gave larger amount of water accumulation in the core and gave better core cooling. These tendencies were also recognized in comparisons under the same distance from the quench front. Since the same tendencies can be predicted in the analyses with REFLA code because of the lower steam generation rate below quench front in the high subcooling test, the differences in the tests are supposed to be caused by the same reason. (2) Higher the subcooling gave larger amount of water accumulation in upper plenum. The carry-over liquid mass into hot leg became smaller in the later period in the higher subcooling test. These differences for carry-over and de-entrainment characteristics can be explained by the differences of quench velocity and of steam mass flow rate generated in the core. (3) No significant influence of the different degree of the subcooling was observed on the two-dimensional thermal-hydraulic behaviors in the pressure vessel. Namely, radial differences of sectional void fraction, heat transfer coefficient and the pressure among bundles at the same elevation were almost the same amount for the two tests. Radial differences of liquid levels in the upper plenum was also almost the same amount for the two tests. (J.P.N.)

Diversified emergency core cooling in CANDU with a passive moderator heat rejection system

Energy Technology Data Exchange (ETDEWEB)

Spinks, N [AECL Research, Chalk River Labs., Chalk River, ON (Canada)

1996-12-01

A passive moderator heat rejection system is being developed for CANDU reactors which, combined with a conventional emergency-coolant injection system, provides the diversity to reduce core-melt frequency to order 10{sup -7} per unit-year. This is similar to the approach used in the design of contemporary CANDU shutdown systems which leads to a frequency of order 10{sup -8} per unit-year for events leading to loss of shutdown. Testing of a full height 1/60 power-and-volume-scaled loop has demonstrated the feasibility of the passive system for removal of moderator heat during normal operation and during accidents. With the frequency of core-melt reduced, by these measures, to order 10{sup -7} per unit year, no need should exist for further mitigation. (author). 3 refs, 2 figs.

A solar-thermal energy harvesting scheme: enhanced heat capacity of molten HITEC salt mixed with Sn/SiO(x) core-shell nanoparticles.

Science.gov (United States)

Lai, Chih-Chung; Chang, Wen-Chih; Hu, Wen-Liang; Wang, Zhiming M; Lu, Ming-Chang; Chueh, Yu-Lun

2014-05-07

We demonstrated enhanced solar-thermal storage by releasing the latent heat of Sn/SiO(x) core-shell nanoparticles (NPs) embedded in a eutectic salt. The microstructures and chemical compositions of Sn/SiO(x) core-shell NPs were characterized. In situ heating XRD provides dynamic crystalline information about the Sn/SiO(x) core-shell NPs during cyclic heating processes. The latent heat of ∼29 J g(-1) for Sn/SiO(x) core-shell NPs was measured, and 30% enhanced heat capacity was achieved from 1.57 to 2.03 J g(-1) K(-1) for the HITEC solar salt without and with, respectively, a mixture of 5% Sn/SiO(x) core-shell NPs. In addition, an endurance cycle test was performed to prove a stable operation in practical applications. The approach provides a method to enhance energy storage in solar-thermal power plants.

Convective heating of the inner core of red giants prior to the peak of the core helium flash

International Nuclear Information System (INIS)

Cole, P.W.; Demarque, P.; Deupree, R.G.

1985-01-01

The effects of convective overshooting across the temperature inversion in the cores of red giants are investigated from the onset of the core convection zone to the peak of the core helium flash using a model for overshooting in stellar evolution, based on two-dimensional and three-dimensional hydrodynamic simulations of the core helium flash. A major effect of the overshooting is the substantial heating of the material interior to the temperature inversion, producing a smoother temperature profile. This interior heating is thus unimportant until approximately 1 week preceding the time of maximum temperature, but then produces temperature changes on a time scale short with respect to the evolution time scale. Interior heating (1) alters the standard relation of the maximum temperature and the density at the point of maximum temperature, (2) makes the maximum temperature occur at a smaller mass fraction, (3) causes the time of maximum temperature to occur hundreds of years earlier in the red giant evolution, and (4) redistributes the mass from the location of maximum temperature. Since the degree of degeneracy is known to affect the violence of the flash in the hydrodynamic phase, internal heating may play an important role in determining the subsequent evolution of the core

Core-Sheath Paraffin-Wax-Loaded Nanofibers by Electrospinning for Heat Storage.

Science.gov (United States)

Lu, Yuan; Xiao, Xiudi; Zhan, Yongjun; Huan, Changmeng; Qi, Shuai; Cheng, Haoliang; Xu, Gang

2018-04-18

Paraffin wax (PW) is widely used for smart thermoregulation materials due to its good thermal performance. However, the leakage and low thermal conductivity of PW hinder its application in the heat storage field. Accordingly, developing effective methods to address these issues is of great importance. In this study, we explored a facile approach to obtain PW-loaded core-sheath structured flexible nanofibers films via coaxial electrospinning technique. The PW as the core layer was successfully encapsulated by the sheath-layer poly(methyl methacrylate). The diameter of the fiber core increased from 395 to 848 nm as the core solution speed rate increased from 0.1 to 0.5 mL/h. In addition, it can be seen that higher core solution speed rate could lead to higher PW encapsulation efficiency according to the transmission electron microscopy results. The core-sheath nanofiber films, moreover, possessed the highest latent heat of 58.25 J/g and solidifying enthalpy of -56.49 J/g. In addition, we found that after 200 thermal cycles, there was little change in latent heat, which demonstrated that it is beneficial for the PW-loaded core-sheath structure to overcome the leakage issue and enhance thermal stability properties for the thermoregulation film.

Precessing vortex core in a swirling wake with heat release

International Nuclear Information System (INIS)

Gorbunova, A.; Klimov, A.; Molevich, N.; Moralev, I.; Porfiriev, D.; Sugak, S.; Zavershinskii, I.

2016-01-01

Highlights: • Precessing vortex core is left-handed co-rotated bending single-vortex structure. • The precession frequency grows with the heat-source power. • Growth of the heat-source power decreases vortex core oscillations. • The left-handed bending mode is the most unstable mode in the low-density wake. - Abstract: Numerical simulation of the non-stationary three-dimensional swirling flow is presented for an open tube with a paraxial heat source. In the considered type of swirling flows, it is shown that a precessing vortex core (PVC) appears. The obtained PVC is a left-handed co-rotated bending single-vortex structure. The influence of the heat release enhancement on parameters of PVC is investigated. Using various turbulence models (the Spalart–Allmaras, k–ω and SST models), it is shown that an increase in the heat-source power leads to an increase in the PVC frequency and to a decrease in the amplitude of PVC oscillations. Moreover, we conduct the linear stability analysis of the simplified flow model with paraxial heating (the Rankine vortex with the piecewise axial flow and density) and demonstrate that its results correspond to the results of numerical simulations rather well. In particular, we prove that the left-handed bending mode (m = +1) is the most unstable one in the low-density wake and its frequency increases with a decrease of density ratio that is similar to the behavior of precession frequency with an increase of heat-source power.

Spectral Effects on Fast Wave Core Heating and Current Drive

International Nuclear Information System (INIS)

Phillips, C.K.; Bell, R.E.; Berry, L.A.; Bonoli, P.T.; Harvey, R.W.; Hosea, J.C.; Jaeger, E.F.; LeBlanc, B.P.; Ryan, P.M.; Taylor, G.; Valeo, E.J.; Wilson, J.R.; Wright, J.C.; Yuh, H. and the NSTX Team

2009-01-01

Recent results obtained with high harmonic fast wave (HHFW) heating and current drive (CD) on NSTX strongly support the hypothesis that the onset of perpendicular fast wave propagation right at or very near the launcher is a primary cause for a reduction in core heating efficiency at long wavelengths that is also observed in ICRF heating experiments in numerous tokamaks. A dramatic increase in core heating efficiency was first achieved in NSTX L-mode helium majority plasmas when the onset for perpendicular wave propagation was moved away from the antenna and nearby vessel structures. Efficient core heating in deuterium majority L mode and H mode discharges, in which the edge density is typically higher than in comparable helium majority plasmas, was then accomplished by reducing the edge density in front of the launcher with lithium conditioning and avoiding operational points prone to instabilities. These results indicate that careful tailoring of the edge density profiles in ITER should be considered to limit rf power losses to the antenna and plasma facing materials. Finally, in plasmas with reduced rf power losses in the edge regions, the first direct measurements of high harmonic fast wave current drive were obtained with the motional Stark effect (MSE) diagnostic. The location and radial dependence of HHFW CD measured by MSE are in reasonable agreement with predictions from both full wave and ray tracing simulations

Interpretation of the results of the CORA-33 dry core BWR test

International Nuclear Information System (INIS)

Ott, L.J.; Hagen, S.

1993-01-01

All BWR degraded core experiments performed prior to CORA-33 were conducted under ''wet'' core degradation conditions for which water remains within the core and continuous steaming feeds metal/steam oxidation reactions on the in-core metallic surfaces. However, one dominant set of accident scenarios would occur with reduced metal oxidation under ''dry'' core degradation conditions and, prior to CORA-33, this set had been neglected experimentally. The CORA-33 experiment was designed specifically to address this dominant set of BWR ''dry'' core severe accident scenarios and to partially resolve phenomenological uncertainties concerning the behavior of relocating metallic melts draining into the lower regions of a ''dry'' BWR core. CORA-33 was conducted on October 1, 1992, in the CORA tests facility at KfK. Review of the CORA-33 data indicates that the test objectives were achieved; that is, core degradation occurred at a core heatup rate and a test section axial temperature profile that are prototypic of full-core nuclear power plant (NPP) simulations at ''dry'' core conditions. Simulations of the CORA-33 test at ORNL have required modification of existing control blade/canister materials interaction models to include the eutectic melting of the stainless steel/Zircaloy interaction products and the heat of mixing of stainless steel and Zircaloy. The timing and location of canister failure and melt intrusion into the fuel assembly appear to be adequately simulated by the ORNL models. This paper will present the results of the posttest analyses carried out at ORNL based upon the experimental data and the posttest examination of the test bundle at KfK. The implications of these results with respect to degraded core modeling and the associated safety issues are also discussed

Heat loss and fluid leakage tests of the ROSA-III facility

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Tasaka, Kanji; Shiba, Masayoshi

1981-12-01

The report presents characteristic test results about the steady state heat loss, one of the inherent characteristics of the ROSA-III test facility. The steady state heat loss tests were conducted at five different temperature conditions between 111 0 C and 290 0 C . Net heat loss rates were obtained by estimating the electric power supplied to the core, heat input from the recirculation pumps and steam leakage rate. The heat loss characteristics have important contribution to analyses of the ROSA-III small break tests. A following simple relation was obtained between the net heat loss rate Q*sub(HL) (kJ/s) (*: radical) of the ROSA-III facility and the temperature difference ΔT ( 0 C) between the fluid temperature of the system and the room temperature, Q*sub(HL) = 0.56 x ΔT. (*: radical) And the steam leak flow at normal operating condition of the ROSA-III test, (P = 7.2 MPa) was obtained as 8.9 x 10 -3 kg/s and corresponding steam leakage energy as 10.5 kJ/s. The heat input from the recirculation pumps was indirectly estimated under a constant speed by assuming the heat input was equal to the brake horce power of the pumps. (author)

Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

International Nuclear Information System (INIS)

Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki

2003-01-01

The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 degree sign C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 degree sign C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open

In-reactor testing of the closed cycle gas core reactor---the nuclear light bulb concept

International Nuclear Information System (INIS)

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

1993-01-01

The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (>1800 s) and thrust (>445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (∼4000 K). The following paper describes analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented here include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRR. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRR for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept

Evaluation report on SCTF Core-III tests S3-14, S3-15 and S3-16

International Nuclear Information System (INIS)

Iwamura, Takamichi; Iguchi, Tadashi; Akimoto, Hajime; Okubo, Tsutomu; Ohnuki, Akira; Sakaki, Isao; Adachi, Hiromichi; Murao, Yoshio

1988-03-01

It was revealed from previous reflood tests using Slab Core Test Facility (SCTF) that the heat transfer was enhanced in high power bundles and degraded in low power bundles due to the radial power distribution. In the present study, the effects of local power ratio itself and the shape of radial power distribution were separately investigated by comparing three tests: S3-14 (flat power distribution), S3-15 (inclined power distribution) and S3-16 (steep power distribution). Those three tests were performed under the same boundary conditions, total power and initial stored energy. The emergency core cooling (ECC) water was injected into the lower plenum. The test results indicated that the heat transfer enhancement and degradation was governed mainly by the bundlewise radial power ratio and less dependent on the shape of radial power distribution when the difference in power ratios between adjacent bundles is less than 0.44. The degree of heat transfer enhancement at high power bundles was increased with the radial peak power ratio. Even at an average power bundle, the heat transfer was enhanced in the non-uniform radial power distribution tests. (author)

Contribution to the modelling of flows and heat transfers during the reflooding phase of a PWR core

International Nuclear Information System (INIS)

Colas, D.

1984-01-01

This thesis contributes to modelise thermohydraulic phenomena occuring in a pressurized water nuclear reactor core during the reflood phase of a LOCA. The reference accident and phenomena occuring during reflooding are described as well as flow regime and heat transfer proposed models. With these models, we developed a code to compute fluid conditions and fuel rods temperatures in a reactor core chanel. In order to test this code, results of computation are compared with experiments (FLECHT Skewed Tests) and a conclusion is drawn [fr

Turbulence-induced heat transfer in PBMR core using LES and RANS

International Nuclear Information System (INIS)

Lee, Jung-Jae; Yoon, Su-Jong; Park, Goon-Cherl; Lee, Won-Jae

2007-01-01

This paper introduces the results of numerical simulations on flow fields and relevant heat transfer in the pebble bed reactor (PBR) core, since the coolant passes a highly complicated random flow path with a high Reynolds number, an appropriate treatment of the turbulence is required. A set of simple experiments for the flow over a circular cylinder with heat transfer was conducted to finally select the large eddy simulation (LES) and k-ω model among the considering Reynolds-averaged Navier-Stokes (RANS) models for PBR application. Using these models, the PBR cores, whose geometries were simplified to the body-centered cubical (BCC) and face-centered cubical (FCC) structures, were simulated. A larger pressure drop, a more random flow field, a higher vorticity magnitude and a higher temperature at the local hot spots on the pebble surface were found in the results of the LES than in those of RANS for both geometries. In cases of the LES, the flow structures were resolved up to the grid scales. Irregular distributions of the flow and local heat transfer were found in the BCC core, while relatively regular distributions for the FCC core. The turbulent nature of the coolant flow in the pebble core evidently affected the fuel surface temperature distribution. (author)

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

International Nuclear Information System (INIS)

Kotas, J.F.; Stroh, K.R.

1983-01-01

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident that simulates a control-rod withdrawal at full power

Rates of chemical reaction and atmospheric heating during core debris expulsion from a pressurized vessel

International Nuclear Information System (INIS)

Powers, D.A.; Tarbell, W.W.; Brockman, J.E.; Pilch, M.

1986-01-01

Core debris may be expelled from a pressurized reactor vessel during a severe nuclear reactor accident. Experimental studies of core debris expulsion from pressurized vessels have established that the expelled material can be lofted into the atmosphere of the reactor containment as particulate 0.4 to 2 mm in diameter. These particles will vigorously react with steam and oxygen in the containment atmosphere. Data on such reactions during tests with 80 kg of expelled melt will be reported. A model of the reaction rates based on gas phase mass transport will be described and shown to account for atmospheric heating and aerosol generation observed in the tests

Transient core-debris bed heat-removal experiments and analysis

International Nuclear Information System (INIS)

Ginsberg, T.; Klein, J.; Klages, J.; Schwarz, C.E.; Chen, J.C.

1982-08-01

An experimental investigation is reported of the thermal interaction between superheated core debris and water during postulated light-water reactor degraded core accidents. Data are presented for the heat transfer characteristics of packed beds of 3 mm spheres which are cooled by overlying pools of water. Results of transient bed temperature and steam flow rate measurements are presented for bed heights in the range 218 mm-433 mm and initial particle bed temperatures between 530K and 972K. Results display a two-part sequential quench process. Initial frontal cooling leaves pockets or channels of unquenched spheres. Data suggest that heat transfer process is limited by a mechanism of countercurrent two-phase flow. An analytical model which combines a bed energy equation with either a quasisteady version of the Lipinski debris bed model or a critical heat flux model reasonably well predicts the characteristic features of the bed quench process. Implications with respect to reactor safety are discussed

Analysis of core plasma heating and ignition by relativistic electrons

International Nuclear Information System (INIS)

Nakao, Y.

2002-01-01

Clarification of the pre-compressed plasma heating by fast electrons produced by relativistic laser-plasma interaction is one of the most important issues of the fast ignition scheme in ICF. On the basis of overall calculations including the heating process, both by relativistic hot electrons and alpha-particles, and the hydrodynamic evolution of bulk plasma, we examine the feature of core plasma heating and the possibility of ignition. The deposition of the electron energy via long-range collective mode, i.e. Langmuir wave excitation, is shown to be comparable to that through binary electron-electron collisions; the calculation neglecting the wave excitation considerably underestimates the core plasma heating. The ignition condition is also shown in terms of the intensity I(h) and temperature T(h) of hot electrons. It is found that I(h) required for ignition increases in proportion to T(h). For efficiently achieving the fast ignition, electron beams with relatively 'low' energy (e.g.T(h) below 1 MeV) are desirable. (author)

Multiobjective optimization for design of multifunctional sandwich panel heat pipes with micro-architected truss cores

International Nuclear Information System (INIS)

Roper, Christopher S.

2011-01-01

A micro-architected multifunctional structure, a sandwich panel heat pipe with a micro-scale truss core and arterial wick, is modeled and optimized. To characterize multiple functionalities, objective equations are formulated for density, compressive modulus, compressive strength, and maximum heat flux. Multiobjective optimization is used to determine the Pareto-optimal design surfaces, which consist of hundreds of individually optimized designs. The Pareto-optimal surfaces for different working fluids (water, ethanol, and perfluoro(methylcyclohexane)) as well as different micro-scale truss core materials (metal, ceramic, and polymer) are determined and compared. Examination of the Pareto fronts allows comparison of the trade-offs between density, compressive stiffness, compressive strength, and maximum heat flux in the design of multifunctional sandwich panel heat pipes with micro-scale truss cores. Heat fluxes up to 3.0 MW/m 2 are predicted for silicon carbide truss core heat pipes with water as the working fluid.

Sodium heat pipe module test for the SAFE-30 reactor prototype

International Nuclear Information System (INIS)

Reid, Robert S.; Sena, J. Tom; Martinez, Adam L.

2001-01-01

Reliable, long-life, low-cost heat pipes can enable safe, affordable space fission power and propulsion systems. Advanced versions of these systems can in turn allow rapid access to any point in the solar system. Twelve stainless steel-sodium heat pipe modules were built and tested at Los Alamos for use in a non-nuclear thermohydraulic simulation of the SAFE-30 reactor (Poston et al., 2000). SAFE-30 is a near-term, low-cost space fission system demonstration. The heat pipes were designed to remove thermal power from the SAFE-30 core, and transfer this power to an electrical power conversion system. These heat pipe modules were delivered to NASA Marshall Space Flight Center in August 2000 and were assembled and tested in a prototypical configuration during September and October 2000. The construction and test of one of the SAFE-30 modules is described

Numerical Analysis on Heat Flux Distribution through the Steel Liner of the Ex-vessel Core Catcher

Energy Technology Data Exchange (ETDEWEB)

Oh, Se Hong; Choi, Choeng Ryul [ELSOLTEC, Yongin (Korea, Republic of); Kim, Byung Jo; Lee, Kyu Bok [KEPCO, Gimcheon (Korea, Republic of); Hwang, Do Hyun [KHNP-CRI, Daejeon (Korea, Republic of)

2016-05-15

In order to prevent material failure of steel container of the core catcher system due to high temperatures, heat flux through the steel liner wall must be kept below the critical heat flux (CHF), and vapor dry-out of the cooling channel must be avoided. In this study, CFD methodology has been developed to simulate the heat flux distribution in the core catcher system, involving following physical phenomena: natural convection in the corium pool, boiling heat transfer and solidification/melting of the corium. A CFD methodology has been developed to simulate the thermal/hydraulic phenomena in the core catcher system, and a numerical analysis has been carried out to estimate the heat flux through the steel liner of the core catcher. High heat flux values are formed at the free surface of the corium pool. However, the heat flux through the steel liner is maintained below the critical heat flux.

Simulation of heat and mass transfer processes in molten core debris-concrete systems. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Felde, D K

1979-01-01

The heat and mass transport phenomena taking place in volumetrically-heated fluids have become of interest in recent years due to their significance in assessments of fast reactor safety and post-accident heat removal (PAHR). Following a hypothetical core disruptive accident (HCDA), the core and reactor internals may melt down. The core debis melting through the reactor vessel and guard vessel may eventually contact the concrete of the reactor cell floor. The interaction of the core debris with the concrete as well as the melting of the debris pool into the concrete will significantly affect efforts to prevent breaching of the containment and the resultant release of radioactive effluents to the environment.

Demonstration testing and evaluation of in situ heating of soil

International Nuclear Information System (INIS)

1995-01-01

This document describes the Quality Assurance Project Plan (QAPP) for IITRI Project C06787 entitled open-quotes Demonstration Testing and Evaluation of In Situ Heating of Soilclose quotes. A work plan for the above mentioned work was previously submitted. This QAPP describes the sampling and analysis of soil core-samples obtained from the K-25 Site (Oak Ridge Gaseous Diffusion Plant) where an in-situ heating and soil decontamination demonstration experiment will be performed. Soil samples taken before and after the experiment will be analyzed for selected volatile organic compounds. The Work Plan mentioned above provides a complete description of the demonstration site, the soil sampling plan, test plan, etc

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

International Nuclear Information System (INIS)

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

1992-07-01

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

An Operators View of Reliability Testing and Decay Heat Rejection Systems

International Nuclear Information System (INIS)

Henderson, J.D.C.

1975-01-01

The object of this paper is to review the in-situ testing of DHR systems, and to convey policy rather than to indicate a definitive test programme. The test policy is aimed primarily at commissioning the plant and secondly at providing such support for reliability predictions as is practical. Provisions for removal of decay heat from the core and from the reactor tank are described in papers by Broadley and Davies

Heat transfer analysis to investigate the core catcher plate assembly in SFR

International Nuclear Information System (INIS)

Patil, Swapnil; Sharma, Anil Kumar; Velusamy, K.; Nashine, B.K.; Selvaraj, P.

2015-01-01

Severe accident scenario in Sodium Cooled Fast Reactor (SFR) is the major concern for public acceptance. After severe accident, the molten core continuously generates substantial decay heat. However, an in-vessel core catcher plate is provided to remove the decay heat passively. The numerical investigation of pool hydraulics phenomena in sodium pool of typical Indian SFR has been carried out. The debris may form a heap with different angle over the core catcher plate due to molten fuel density and interaction force. Therefore, the debris bed with different heap angle has been analyzed for steady and transient state conditions. The governing equation of fluid flow and heat transfer are solved by finite volume method based solver with the k-ε turbulent model. The time period Δ for which temperature is exceeding above safety limit with different debris heap angle have been established. (author)

Test plan for core drilling ignitability testing

International Nuclear Information System (INIS)

Witwer, K.S.

1996-01-01

The objective of this testing is to determine if ignition occurs while core drilling in a flammable gas environment. Drilling parameters are chosen so as to provide bounding conditions for the core sampling environment. If ignition does not occur under the conditions set forth in this test, then a satisfactory level of confidence will be obtained which would allow field operations under the normal drilling conditions

Large scale reflood test with cylindrical core test facility (CCTF). Core I. FY 1979 tests

International Nuclear Information System (INIS)

Murao, Yoshio; Akimoto, Hajime; Okubo, Tsutomu; Sudoh, Takashi; Hirano, Kenmei

1982-03-01

This report presents the results of analysis of the data obtained in the CCTF Core I test series (19 tests) in FY. 1979 as an interim report. The Analysis of the test results showed that: (1) The present safety evaluation model on the reflood phenomena during LOCA conservatively represents the phenomena observed in the tests except for the downcomer thermohydrodynamic behavior. (2) The downcomer liquid level rose slowly and it took long time for the water to reach a terminal level or the spill-over level. It was presume that such a results was due to an overly conservative selection of the ECC flow rate. This presumption will be checked against a future test result for an increased flow rate. The loop-seal-water filling test was unsuccessful due to a premature power shutdown by the core protection circuit. The test will be conducted again. The tests to be performed in the future are summerized. Tests for investigation of the refill phenomena were also proposed. (author)

Testing of a Stirling engine for heat + power cogeneration; Test eines Stirlingmotors zur Kraft-Waerme-Kopplung

Energy Technology Data Exchange (ETDEWEB)

Kramer, M.; Heinen, J. [RWE Energy AG, Essen (Germany)

2007-01-15

As part of a technology evaluation of distributed energy generators, RWE Energy AG extensively tested a micro combined heat and power appliance, powered by a Stirling engine developed by the British firm Microgen Energy Limited. Microgen Energy Limited is a specialist in micro combined heat and power (microCHP) based on unique Free-Piston Stirling generator technology Microgen is working with leading appliance manufacturers to integrate its core technology into a range of innovative microCHP products. The investigations concentrated on the determination of capacity, efficiency and emissions, the grid connection and behaviour at start-up and under varying loads. This article summarises the results of the tests and gives an overview of micro-CHP technologies (CHP=combined heat and power) and their possible significance to the market in the future. (orig.)

Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core

Energy Technology Data Exchange (ETDEWEB)

Kawaji, Masahiro [City College of New York, NY (United States); Valentin, Francisco I. [City College of New York, NY (United States); Artoun, Narbeh [City College of New York, NY (United States); Banerjee, Sanjoy [City College of New York, NY (United States); Sohal, Manohar [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schultz, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); McEligot, Donald M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-12-21

The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.

Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core

International Nuclear Information System (INIS)

Kawaji, Masahiro; Valentin, Francisco I.; Artoun, Narbeh; Banerjee, Sanjoy; Sohal, Manohar; Schultz, Richard; McEligot, Donald M.

2015-01-01

The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.

CCTF CORE I test results

International Nuclear Information System (INIS)

Murao, Yoshio; Sudoh, Takashi; Akimoto, Hajime; Iguchi, Tadashi; Sugimoto, Jun; Fujiki, Kazuo; Hirano, Kenmei

1982-07-01

This report presents the results of the following CCTF CORE I tests conducted in FY. 1980. (1) Multi-dimensional effect test, (2) Evaluation model test, (3) FLECHT coupling test. On the first test, one-dimensional treatment of the core thermohydrodynamics was discussed. On the second and third tests, the test results were compared with the results calculated by the evaluation model codes and the results of the corresponding FLECHT-SET test (Run 2714B), respectively. The work was performed under contracts with the Atomic Energy Bureau of Science and Technology Agency of Japan. (author)

Numerical simulation of passive heat removal under severe core meltdown scenario in a sodium cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

David, Dijo K.; Mangarjuna Rao, P., E-mail: pmr@igcar.gov.in; Nashine, B.K.; Selvaraj, P.; Chellapandi, P.

2015-09-15

Highlights: • PAHR in SFR under large core relocation to in-vessel core catcher is numerically analyzed. • A 1-D thermal conduction model and a 2-D axisymmetric CFD model are developed for turbulent natural convection phenomenon. • The side pool (cold pool) was found out to be instrumental in storing heat and dissipating it to the heat sink. • Single tray type in-vessel core catcher is found to be thermally effective under one-fourth core relocation. - Abstract: A sequence of highly unlikely events leading to significant meltdown of the Sodium cooled Fast Reactor (SFR) core can cause the failure of reactor vessel if the molten fuel debris settles at the bottom of the reactor main vessel. To prevent this, pool type SFRs are usually provided with an in-vessel core catcher above the bottom wall of the main vessel. The core catcher should collect, retain and passively cool these debris by facilitating decay heat removal by natural convection. In the present work, the heat removal capability of the existing single tray core catcher design has been evaluated numerically by analyzing the transient development of natural convection loops inside SFR pool. A 1-D heat diffusion model and a simplified 2-D axi-symmetric CFD model are developed for the same. Maximum temperature of the core catcher plate evaluated for different core meltdown scenarios using these models showed that there is much higher heat removal potential for single tray in-vessel SFR core catcher compared to the design basis case of melting of 7 subassemblies under total instantaneous blockage of a subassembly. The study also revealed that the side pool of cold sodium plays a significant role in decay heat removal. The maximum debris bed temperature attained during the initial hours of PAHR does not depend much on when the Decay Heat Exchanger (DHX) gets operational, and it substantiates the inherent safety of the system. The present study paves the way for better understanding of the thermal

Integrated CFD investigation of heat transfer enhancement using multi-tray core catcher in SFR

International Nuclear Information System (INIS)

Rakhi; Sharma, Anil Kumar; Velusamy, K.

2017-01-01

Highlights: • Heat transfer enhancement using multi-tray core catcher for SFR is investigated. • The capability of a single core collector tray is estimated. • Double and triple collector trays with innovative designs is discussed. • Provision of openings in the trays contributed to enhanced natural circulation. - Abstract: To render future SFR more robust and safe, certain BDBE have been considered in the recent years. A Core Disruptive Accident leading to a whole core meltdown scenario has gained the interest of researchers. Various design concepts and safety measures have been suggested and incorporated in design to address such a low probability scenario. A core catcher concept, in particular, has proved to be inevitable as an in-vessel core retention device in SFR for safe retention of core debris arising out after the severe accident. This study aims to analyse the cooling capability of the innovative design concept of core catcher to remove decay heat of degraded core after the accident. First, the capability of single collection tray is established and then the study is extended to two and three collection trays with different design concepts. Transient forms of governing equations of mass, momentum and energy conservations along with k-ε turbulence model are solved by finite volume based CFD solver. Boussinesq approximation is invoked to model buoyancy in sodium. The study shows that a single collection tray is capable of removing up to 20 MW decay heat load in a typical 500 MWe pool type SFR. Further, studies are carried out to improve the natural circulation of sodium around the source, in the lower plenum and to distribute core debris of the whole core to multiple collection trays. It is found that the double and triple collection trays can accommodate decay loads up to 29 MW. Provision of openings in the collection trays has proved to be effective in improving the heat transfer and sodium flow as well as in distributing the core debris to the

Evaluation report on CCTF Core-II reflood test C2-16 (Run 76)

International Nuclear Information System (INIS)

Iguchi, Tadashi; Akimoto, Hajime; Okubo, Tsutomu; Hojo, Tsuneyuki; Murao, Yoshio; Sugimoto, Jun.

1987-03-01

This report presents the result of the upper plenum injection (UPI) test C2-16 (Run 76), which was conducted on October 23, 1984, with the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute (JAERI). The CCTF is a 1/21.4 scale model of a 1,100 MWe PWR with four loop active components to provide information on the system and core thermo-hydrodynamics during reflood. The objectives of the test are to investigate the reflood phenomena with single failure UPI condition and to investigate the effect of the asymmetry of UPI on the reflood phenomena. The test was performed with an asymmetric UPI condition at the injection rate simulating single failure of LPCI pumps. It was observed that, (1) a UPI test simulating no LPCI pump failure gave the slightly lower peak clad temperature than a UPI test simulating single LPCI pump failure, indicating that single LPCI pump failure assumption is conserrative for UPI condition, and (2) an asymmetric UPI lead to a higher core water accumulation and then a higher heat transfer coefficient, resultantly a lower peak clad temperature than a symmetric UPI, indicating that asymmetric UPI does not lead to a poorer core cooling than symmetric UPI. (author)

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

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.)

Evaluation report on SCTF Core-III test S3-06

International Nuclear Information System (INIS)

Iwamura, Takamichi; Iguchi, Tadashi; Akimoto, Hajime; Okubo, Tsutomu; Ohnuki, Akira; Adachi, Hiromichi; Murao, Yoshio; Minato, Akihiko; Sakaki, Isao.

1988-10-01

In order to investigate the effect of radial power distribution on the thermal-hydraulic characteristics during the reflood phase of a PWR-LOCA with a combined injection type ECCS, a core cooling separate effect test S3-06 and a combined injection test S3-16-Phase 2 were performed using the Slab Core Test Facility (SCTF) Core-III. The radial power distributions in these two tests simulated a reference distribution for a PWR with a combined injection type ECCS and a steep distribution for a PWR with a cold leg injection type ECCS, respectively. Under the radial power distribution of a PWR with a combined injection type ECCS, the radial power distribution had little effect on the thermal-hydraulic behavior in the two-phase up-flow region due to the approximately flat power distribution in this region (power ratio = 1.04 ∼ 1.08). The overall fluid behavior in the pressure vessel was also little affected by the radial power distribution. On the other hand, under the steep radial power distribution (peak power ratio = 1.36), the degree of heat transfer enhancement in high power bundles in the two-phase up-flow region was dominated by the bundlewise radial power ratio as in the case of a PWR with a cold leg injection type ECCS. (author)

Heat Transfer in Pebble-Bed Nuclear Reactor Cores Cooled by Fluoride Salts

Science.gov (United States)

Huddar, Lakshana Ravindranath

underlining key distortions between the experimental and the prototypical conditions. This dissertation is broadly split into four parts. Firstly, the heat transfer phenomenology in the PB-FHR core was outlined. Although the viscous dissipation term and the thermal diffusion term (including thermal dispersion) were similar in magnitude, they were overshadowed by the advection term which was about 104 times bigger during normal operation and 105 times bigger during accident transients in which natural circulation becomes the main mode of fluid flow. Thus it is safe to neglect the viscous dissipation and the thermal diffusion terms in the PB-FHR core without a significant loss of accuracy. Secondly, separate effects tests (SET) were performed using simulant oils, and the results were compared to the prototypical conditions using flinak as the fluoride salt. The main purpose of these experiments was to study natural convection heat transfer and identify any distortions between the two cases. An isolated copper sphere was immersed in flinak and a parallel experiment was performed using simulant oil. A large discrepancy between the flinak and the oil was noted, due to distortions from assuming quasi-steady state conditions. A steady state experiment using a cylindrical heater immersed in oil was also performed, and the results compared to a similar experiment done at Oak Ridge National Laboratory (ORNL) using flinak. The Nusselt numbers matched within 10% for laminar flows. This supports the conclusion that natural convection similitude does exist for oils used in scaled experiments, allowing natural convection data to be used for for FHR and MSR modeling. This is important, due to the lack of significant experimental data showing natural convection in fluoride salts, so these SETs add to the overall understanding of their heat transfer properties. With the knowledge of the distortions between the oil and the salt, an experiment to measure heat transfer coefficients within a

Computer-Aided Test Flow in Core-Based Design

OpenAIRE

Zivkovic, V.; Tangelder, R.J.W.T.; Kerkhoff, Hans G.

2000-01-01

This paper copes with the test-pattern generation and fault coverage determination in the core based design. The basic core-test strategy that one has to apply in the core-based design is stated in this work. A Computer-Aided Test (CAT) flow is proposed resulting in accurate fault coverage of embedded cores. The CAT now is applied to a few cores within the Philips Core Test Pilot IC project

In-core flow rate distribution measurement test of the JOYO irradiation core

International Nuclear Information System (INIS)

Suzuki, Toshihiro; Isozaki, Kazunori; Suzuki, Soju

1996-01-01

A flow rate distribution measurement test was carried out for the JOYO irradiation core (the MK-II core) after the 29th duty cycle operation. The main object of the test is to confirm the proper flow rate distribution at the final phase of the MK-II core. The each flow rate at the outlet of subassemblies was measured by the permanent magnetic flowmeter inserted avail of fuel exchange hole in the rotating plug. This is third test in the MK-II core, after 10 years absence from the final test (1985). Total of 550 subassemblies were exchanged and accumulated reactor operation time reached up to 38,000 hours from the previous test. As a conclusion, it confirmed that the flow rate distribution has been kept suitable in the final phase of the MK-II core. (author)

Heating analysis of cobalt adjusters in reactor core

International Nuclear Information System (INIS)

Mei Qiliang; Li Kang; Fu Yaru

2011-01-01

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

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.

Calculation and analysis of generator limiting regimes with respect to stator end core heating

Directory of Open Access Journals (Sweden)

Kostić Miloje

2015-01-01

Full Text Available A new simplified procedure for defining the limiting operating regimes on the generator capability curve, with respect to stator end core heating, is proposed and described in this paper. First of all, a simplified analysis of axial flux leakage that penetrates into the end plates of the stator is carried out and the corresponding power losses are calculated. Then the analysis of measured point temperature increases over the stator end core, and a qualitative and quantitative overview of the effects, are presented. A simplified procedure for defining the limiting regime with regard to the heating stator end core, which is illustrated for the case of an operating diagram for a given generator of apparent power of 727 MVA (B2 is also described. The given limiting line constructed using this method is similar to the appropriate line constructed on the basis of complex and lengthy factory and on-site tests performed by the manufacturer and the user. According to the results and the check, the proposed method has been proved and the application of the simplified procedure can be recommended for use along with other procedures, at least when it comes to similar synchronous generators in Serbia's Electric Power Industry.

Possible generation of heat from nuclear fusion in Earth's inner core.

Science.gov (United States)

Fukuhara, Mikio

2016-11-23

The cause and source of the heat released from Earth's interior have not yet been determined. Some research groups have proposed that the heat is supplied by radioactive decay or by a nuclear georeactor. Here we postulate that the generation of heat is the result of three-body nuclear fusion of deuterons confined in hexagonal FeDx core-centre crystals; the reaction rate is enhanced by the combined attraction effects of high-pressure (~364 GPa) and high-temperature (~5700 K) and by the physical catalysis of neutral pions: 2 D +  2 D +  2 D → 2 1 H +  4 He + 2  + 20.85 MeV. The possible heat generation rate can be calculated as 8.12 × 10 12  J/m 3 , based on the assumption that Earth's primitive heat supply has already been exhausted. The H and He atoms produced and the anti-neutrino are incorporated as Fe-H based alloys in the H-rich portion of inner core, are released from Earth's interior to the universe, and pass through Earth, respectively.

Dynamo Tests for Stratification Below the Core-Mantle Boundary

Science.gov (United States)

Olson, P.; Landeau, M.

2017-12-01

Evidence from seismology, mineral physics, and core dynamics points to a layer with an overall stable stratification in the Earth's outer core, possibly thermal in origin, extending below the core-mantle boundary (CMB) for several hundred kilometers. In contrast, energetic deep mantle convection with elevated heat flux implies locally unstable thermal stratification below the CMB in places, consistent with interpretations of non-dipole geomagnetic field behavior that favor upwelling flows below the CMB. Here, we model the structure of convection and magnetic fields in the core using numerical dynamos with laterally heterogeneous boundary heat flux in order to rationalize this conflicting evidence. Strongly heterogeneous boundary heat flux generates localized convection beneath the CMB that coexists with an overall stable stratification there. Partially stratified dynamos have distinctive time average magnetic field structures. Without stratification or with stratification confined to a thin layer, the octupole component is small and the CMB magnetic field structure includes polar intensity minima. With more extensive stratification, the octupole component is large and the magnetic field structure includes intense patches or high intensity lobes in the polar regions. Comparisons with the time-averaged geomagnetic field are generally favorable for partial stratification in a thin layer but unfavorable for stratification in a thick layer beneath the CMB.

Core dynamics analysis for reactivity insertion and loss of coolant flow tests using the HTTR

International Nuclear Information System (INIS)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

2007-01-01

The High Temperature engineering Test Reactor (HTTR) is a graphite-moderated and a gas-cooled reactor with a thermal power of 30 MW and a reactor outlet coolant temperature of 950degC (SAITO, 1994). Safety demonstration tests using the HTTR are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs) (TACHIBANA 2002) (NAKAGAWA 2004). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named ACCORD (TAKAMATSU 2006), was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We used a conventional method, namely, a one-dimensional flow channel model and reactor kinetics model with a single temperature coefficient, taking into account the temperature changes in the core. However, a slight difference between the analytical and experimental results was observed. Therefore, we have modified this code to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the loss of coolant flow tests by tripping one or two out of three gas circulators. Finally, the pre-analytical result of

Study of heat removal by natural convection from the internal core catcher in PFBR using water model experiments

International Nuclear Information System (INIS)

Jasmin Sudha, A.; Punitha, G.; Das, S.K.; Lydia, G.; Murthy, S.S.; Malarvizhi, B.; Harvey, J.; Kannan, S.E.

2005-01-01

Full text of publication follows: In the event of a core meltdown accident in a Fast Breeder Reactor, the molten core material settling on the bottom of the main vessel can endanger the structural integrity of the main vessel. In the design of Prototype Fast Breeder Reactor in India, the construction of which is about to commence, a core catcher is provided as the internal core retention device to collect and retain the core debris in a coolable configuration. Heat transfer by natural convection above and below the core catcher plate, in the zone beneath the core support structure is evaluated from water mockup experiments in the 1:4 geometrically scaled setup. These studies were undertaken towards comparison of experimentally measured temperatures at different locations with the numerical results. The core catcher assembly consists of a core catcher plate, a heat shield plate and a chimney. Decay heat from the core debris is simulated by electrical heating of the heat shield plate. An opening is provided in the cover plate to reproduce the situation in the actual accident where the core debris would have breached a part of the core support structure. Experiments were carried out with different heat flux levels prevailing upon the heat shield plate. Temperature monitoring was done at more than 100 locations, distributed both on the solid components and in water. The temperature data was analysed to get the temperature profile at different steady state conditions. Flow visualisation was also carried out using water soluble dye to establish the direction of the convective currents. The captured images show that water flows through the slots provided in the top portion of the chimney in the upward direction as evidenced from the diffusion of dye injected inside the chimney. Both the temperature data and flow visualisation confirm mixing of water through the opening in the core support structure which indicates that natural convection is set up in that zone

Under sodium reliability tests on core components and in-core instrumentation

International Nuclear Information System (INIS)

Ruppert, E.; Stehle, H.; Vinzens, K.

1977-01-01

A sodium test facility for fast breeder core components (AKB), built by INTERATOM at Bensberg, has been operating since 1971 to test fuel dummies and blanket elements as well as absorber elements under simulated normal and extreme reactor conditions. Individual full-scale fuel or blanket elements and arrays of seven elements, modelling a section of the SNR-300 reactor core, have been tested under a wide range of sodium mass flow and isothermal test conditions up to 925K as well as under cyclic changed temperature transients. Besides endurance testing of the core components a special sodium and high-temperature instrumentation is provided to investigate thermohydraulic and vibrational behaviour of the test objects. During all test periods the main subassembly characteristics could be reproduced and the reliability of the instrumentation could be proven. (orig.) [de

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.)

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

Visualization of Heat Transfer and Core Damage With RGUI 1.5

International Nuclear Information System (INIS)

Mesina, George L.

2002-01-01

Graphical User Interfaces (GUI) have become an integral and essential part of computer software. In the ever-changing world of computing, they provide the user with a valuable means to learn, understand, and use the application software while also helping applications adapt to and span different computing paradigms, such as different operating systems. For these reasons, GUI development for nuclear plant analysis programs has been ongoing for a decade and a half and much progress has been made. With the development of codes such as RELAP5-3D [1] and SCDAP/RELAP5 that have multi-dimensional modeling capability, it has become necessary to represent three-dimensional, calculated data. The RELAP5-3D Graphical User Interface (RGUI) [4] was designed specifically for this purpose. It reduces the difficulty of analyzing complex three-dimensional models and enhances the analysts' ability to recognize plant behavior visually. Previous versions of RGUI [5] focused on visualizing reactor coolant behavior during a simulated transient or accident. Recent work has extended RGUI to display two other phenomena, heat transfer and core damage. Heat transfer is depicted through the visualization of RELAP5-3D heat structures. Core damage is visualized by displaying fuel rods and other core structures in a reactor vessel screen. Conditions within the core are displayed via numerical results and color maps. These new features of RGUI 1.5 are described and illustrated. (authors)

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.

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.

Report on material and fabrication tests of the KUHFR core vessel

International Nuclear Information System (INIS)

Yoshida, H.; Kozuka, T.; Achiwa, N.; Mitani, S.; Kawano, S.; Araki, Y.; Shibata, T.

1983-01-01

For the material of the cylindrical reactor core vessel of the Kyoto University High Flux Reactor (KUHFR), A6061 alloy is selected because the aged state of the alloy is known to show the highest resistance against void swelling due to high-dose irradiation. The fabrication possibility of the large-scale tubes is also tested because the sizes (40 cmdiameter and 43 cmdiameter x 960 cm with a thickness of 10 mm for the inner- and outer-tubes, respectively) are just over the largest limit of the conventional factory fabrication. The results are summarized as follows. (1) From an ingot of A6061 alloy a raw inner-tube is hot-extruded by the 3,000 ton press machine. The shape of the extruded tubes is effectively corrected by stretch forming and other special methods. (2) The real scale tubes are heat-treated under the various conditions (T1, T4 and T6) and their size changes are measured just after the every heat-treatment. (3) The hydropressure for a pipe prepared by welding from an aged-tube shows a fairly uniform strain distribution and the breaking initiation at the reasonable pressure in the welded part. (4) Each of the welded specimens prepared using three kinds of welding rods shows sufficient strength in both of bending and tensile test for the JIS standard. Their microstructures correspond to the result of the mechanical tests for each welded specimen. The confidence for the fabrication possibility of the real core vessel has been given through the present tests. (author)

Reflooding phenomena of German PWR estimated from CCTF [Cylindrical Core Test Facility], SCTF [Slab Core Test Facility] and UPTF [Upper Plenum Test Facility] results

International Nuclear Information System (INIS)

Murao, Y.; Iguchi, T.; Sugimoto, J.

1988-09-01

The reflooding behavior in a PWR with a combined injection type ECCS was studied by comparing the test results from Cylindrical Core Test Facility (CCTF), Slab Core Test Facility (SCTF) and Upper Plenum Test Facility (UPTF). Core thermal-hydraulics is discussed mainly based on SCTF test data. In addition, the water accumulation behavior in hot legs and the break-through characteristics at tie plate are discussed

The nuclear heating calculation scheme for material testing in the future Jules Horowitz Reactor

International Nuclear Information System (INIS)

Huot, N.; Aggery, A.; Blanchet, D.; Courcelle, A.; Czernecki, S.; Di-Salvo, J.; Doederlein, C.; Serviere, H.; Willermoz, G.

2004-01-01

An innovative nuclear heating calculation scheme for materials testing carried out in in the future Jules Horowitz reactor (JHR) is described. A heterogeneous gamma source calculation is first performed at assembly level using the deterministic code APOLLO2. This is followed by a Monte Carlo gamma transport calculation in the whole core using the TRIPOLI4 code. The calculated gamma sources at the assembly level are applied in the whole core simulation using a weighting based on power distribution obtained from the neutronic core calculation. (authors)

LMFBR post accident heat removal testing needs and conceptual design of a test facility

International Nuclear Information System (INIS)

Kleefeldt, K.; Kuechle, M.; Royl, P.; Werle, H.; Boenisch, G.; Heinzel, V.; Mueller, R.A.; Schramm, K.; Smidt, D.

1977-03-01

A study has been carried out in which the needs and requirements for a test facility were derived, enabling detailed investigation of key phenomena anticipated during the post accident heat removal (PAHR) phase as a consequence of a postulated LMFBR whole core accident. Part I of the study concentrates on demonstrating the PAHR phenomena and related testing needs. Three types of experiments were identified which require in-pile testing, ranging from 10 to 70 cm test bed diameter and correspondingly, 30 to 5 W/g minimum power density in the test fuel. In part II a conceptual design for a test facility is presented, emphasizing the capability for accomodating large test beds. This is achieved by a below-reactor-vessel testing device, neutronically coupled to a 100 MWt sodium cooled fast reactor. (orig.) [de

A dynamic model for helium core heat exchangers

International Nuclear Information System (INIS)

Schiesser, W.E.; Shih, H.J.; Hartozog, D.G.; Herron, D.M.; Nahmias, D.; Stuber, W.G.; Hindmarsh, A.C.

1990-04-01

To meet the helium (He) requirements of the superconducting supercollider (SSC), the cryogenic plants must be able to respond to time-varying loads. Thus the design and simulation of the cryogenic plants requires dynamic models of their principal components, and in particular, the core heat exchangers. In this paper, we detail the derivation and computer implementation of a model for core heat exchangers consisting of three partial differential equations (PDES) for each fluid stream (the continuity, energy and momentum balances for the He), and one PDE for each parting sheet (the energy balance for the parting sheet metal); the PDEs have time and axial position along the exchanger as independent variables. The computer code can accommodate any number of fluid streams and parting sheets in an adiabatic group. Features of the code include: rigorous or approximate thermodynamic properties for He, upwind and downwind approximation of the PDE spatial derivatives, and sparse matrix time integration. The outputs from the code include the time-dependent axial profiles of the fluid He mass flux, density, pressure, temperature, internal energy and enthalpy. The code is written in transportable Fortran 77, and can therefore be executed on essentially any computer

A dynamic model for helium core heat exchangers

International Nuclear Information System (INIS)

Schiesser, W.E.; Shih, H.J.; Hartzog, D.G.; Herron, D.M.; Nahmias, D.; Stuber, W.G.; Hindmarsh, A.C.

1990-01-01

To meet the helium (He) requirements of the superconducting supercollider (SSC), the cryogenic plants must be able to respond to time-varying loads. Thus the design and simulation of the cryogenic plants requires dynamic models of their principal components, and in particular, the core heat exchangers. In this paper, we detail the derivation and computer implementation of a model for core heat exchangers consisting of three partial differential equations (PDEs) for each fluid stream (the continuity, energy and momentum balances for the He), and one PDE for each parting sheet (the energy balance for the parting sheet metal); the PDEs have time and axial position along the exchanger as independent variables. The computer code can accommodate any number of fluid streams and parting sheets in an adiabatic group. Features of the code include: rigorous or approximate thermodynamic properties for He, upwind and downwind approximation of the PDE spatial derivatives, and sparse matrix time integration. The outputs from the code include the time-dependent axial profiles of the fluid He mass flux, density, pressure, temperature, internal energy and enthalpy. The code is written in transportable Fortran 77, and can therefore be executed on essentially any computer. 10 refs., 10 figs

Two-phase flow pattern and heat transfer during core uncovery

International Nuclear Information System (INIS)

Osakabe, Masahiro; Koizumi, Yasuo; Tasaka, Kanji

1987-01-01

The low and high power core uncovery patterns were observed in the high-pressure quasi-steady core uncovery experiments in a 25-rod bundle. The boundary between the two patterns was obtained in the experiments. The difference of two patterns was considered to be due to the slug-annular transition below the dryout points. The Osakabe's slug-annular transition model was the good boundary between the two patterns. The small break loss-of-coolant accident (LOCA) experiments were conducted by using the integral experimental facility with the 1,168-rod core. The transient core uncovery pattern was expected as the low power core uncovery pattern based on the quasisteady experiments mentioned above. The transient core uncovery patterns were classified into the boiloff and hydraulic core uncovery. In the boiloff core uncovery, the dryout points were controlled with the mixture level like the quasi-steady state. In the hydraulic core uncovery, the dryout points were not controlled with the mixture level alone, and the multi-dimensional dryout process in the core and the relatively high heat transfer above the dryout points were observed. It was considered that a part of water was remained above the dryout points due to the rapid depression of core liquid level. (author)

NEPTUN/5052, PWR LOCA Cooling Heat Transfer Tests for Loft, Reflood Test

International Nuclear Information System (INIS)

Richner, M.; Analytis, G.Th.; Aksan, S.N.

1993-01-01

1 - Description of test facility: NEPTUN is designed to perform PWR LOCA simulation experiments, which provide the full length emergency cooling heat transfer tests for LOFT. Therefore the NEPTUN heater bundle with 33 electrical heater elements and 4 guide tubes simulates a section of the LOFT nuclear core. The main test loop also contains measuring systems for the carry-over rate and for the steam expelled, and a back-pressure control system. A water loop brings the water to the initial reflooding conditions. In addition, auxiliary systems maintain normal operating conditions. 2 - Description of test: Test 5052 is one of a series of 40 reflood tests performed in NEPTUN. Before the start of the test, the flooding water in its circuit is brought to the following conditions: pressure = 4.1 bar; velocity = 2.5 cm/sec; subcooling temperature = 78 C; single rod power = 2.45 kW; maximal initial cladding temperature = 867 C. 3 - Status: CSNI1013/01, 21-Jul-1993 Arrived at NEADB

Foam Core Particleboards with Intumescent FRT Veneer: Cone Calorimeter Testing With Varying Adhesives, Surface Layer Thicknesses, and Processing Conditions

Science.gov (United States)

Mark A. Dietenberger; Johannes Welling; Ali Shalbafan

2014-01-01

Intumescent FRT Veneers adhered to the surface of foam core particleboard to provide adequate fire protection were evaluated by means of cone calorimeter tests (ASTM E1354). The foam core particleboards were prepared with variations in surface layer treatment, adhesives, surface layer thicknesses, and processing conditions. Ignitability, heat release rate profile, peak...

Acceptance test procedure for core sample trucks

International Nuclear Information System (INIS)

Smalley, J.L.

1995-01-01

The purpose of this Acceptance Test Procedure is to provide instruction and documentation for acceptance testing of the rotary mode core sample trucks, HO-68K-4600 and HO-68K-4647. The rotary mode core sample trucks were based upon the design of the second core sample truck (HO-68K-4345) which was constructed to implement rotary mode sampling of the waste tanks at Hanford. Acceptance testing of the rotary mode core sample trucks will verify that the design requirements have been met. All testing will be non-radioactive and stand-in materials shall be used to simulate waste tank conditions. Compressed air will be substituted for nitrogen during the majority of testing, with nitrogen being used only for flow characterization

Core optimization studies for a small heating reactor

International Nuclear Information System (INIS)

Galperin, A.

1986-11-01

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

Evaluation report on CCTF core-I reflood test C1-19 (RUN 38)

International Nuclear Information System (INIS)

Murao, Yoshio; Fujiki, Kazuo; Akimoto, Hajime

1983-02-01

A test named the Evaluation Model (EM) test was performed, whose test conditions were simulated the reflood phase predicted with the safety evaluation analysis. The test results were compared with the blindfold results predicted by Evaluation Model (EM) codes. The main conclusions are as follows: (1) The core heat transfer model built in the EM codes gives conservative results. (2) The system models in the present EM codes are found to be well balanced integrally over the system. (3) Conservative items and items to be improved are pointed out. The downcomer slow water accumulation observed in the lower flow rate test was not appeared in the EM test. (author)

Post shut-down decay heat removal from nuclear reactor core by natural convection loops in sodium pool

Energy Technology Data Exchange (ETDEWEB)

Rajamani, A. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Sundararajan, T., E-mail: tsundar@iitm.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Parthasarathy, U.; Velusamy, K. [Nuclear Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

2016-05-15

Highlights: • Transient simulations are performed for a worst case scenario of station black-out. • Inter-wrapper flow between various sub-assemblies reduces peak core temperature. • Various natural convection paths limits fuel clad temperatures below critical level. - Abstract: The 500 MWe Indian pool type Prototype Fast Breeder Reactor (PFBR) has a passive core cooling system, known as the Safety Grade Decay Heat Removal System (SGDHRS) which aids to remove decay heat after shut down phase. Immediately after reactor shut down the fission products in the core continue to generate heat due to beta decay which exponentially decreases with time. In the event of a complete station blackout, the coolant pump system may not be available and the safety grade decay heat removal system transports the decay heat from the core and dissipates it safely to the atmosphere. Apart from SGDHRS, various natural convection loops in the sodium pool carry the heat away from the core and deposit it temporarily in the sodium pool. The buoyancy driven flow through the small inter-wrapper gaps (known as inter-wrapper flow) between fuel subassemblies plays an important role in carrying the decay heat from the sub-assemblies to the hot sodium pool, immediately after reactor shut down. This paper presents the transient prediction of flow and temperature evolution in the reactor subassemblies and the sodium pool, coupled with the safety grade decay heat removal system. It is shown that with a properly sized decay heat exchanger based on liquid sodium and air chimney stacks, the post shutdown decay heat can be safely dissipated to atmospheric air passively.

Multiregional coupled conduction--convection model for heat transfer in an HTGR core

International Nuclear Information System (INIS)

Giles, G.E. Jr.; Childs, K.W.; Sanders, J.P.

1978-01-01

HEXEREI is a three-dimensional, coupled conduction-convection heat transfer and multichannel fluid dynamic analysis computer code with both steady-state and transient capabilities. The program was developed to provide thermal-fluid dynamic analysis of a core following the general design for high-temperature gas-cooled reactors (HTGRs); its purpose was to provide licensing evaluations for the U.S. Nuclear Regulatory Commission. In order to efficiently model the HTGR core, the nodal geometry of HEXEREI was chosen as a regular hexagonal array perpendicular to the axis of and bounded by a right circular cylinder. The cylindrical nodal geometry surrounds the hexagonal center portion of the mesh; these two different types of nodal geometries must be connected by interface nodes to complete the accurate modeling of the HTGR core. HEXEREI will automatically generate a nodal geometry that will accurately model a complex assembly of hexagonal and irregular prisms. The accuracy of the model was proven by a comparison of computed values with analytical results for steady-state and transient heat transfer problems. HEXEREI incorporates convective heat transfer to the coolant in many parallel axial flow channels. Forced and natural convection (which permits different flow directions in parallel channels) is included in the heat transfer and fluid dynamic models. HEXEREI incorporates a variety of steady-state and transient solution techniques that can be matched with a particular problem to minimize the computational time. HEXEREI was compared with a code of similar capabilities that was based on a Cartesian mesh. This code modeled only one specific core design, and the mesh spacing was closer than that generated by HEXEREI. Good agreement was obtained with the detail provided by the representations

Surveillance test of the JMTR core components

International Nuclear Information System (INIS)

Takeda, Takashi; Amezawa, Hiroo; Tobita, Kenji

1986-02-01

Surveillance test for the core components of Japan Materials Testing Reactor (JMTR) was started in 1966, and completed in 1985 without one capsule. Most of capsules in the program, except one beryllium specimens, were removed from the core, and carred out the post-irradiation tests at the JMTR Hot Laboratory. The data is applied to review of JMTR core components management plan. JMTR surveillance test was carried out with several kind of materials of JMTR core components, Berylium as the reflector, Hafnium as the neutron absorber of control rod, 17-4PH stainless steel as a roller spring of the control rod, and 304 stainless steel as the grid plate. Results are described in this report. (author)

Operation manual for the core flow test loop zone power-supply controller

Energy Technology Data Exchange (ETDEWEB)

Harper, R.E.

1981-11-01

The core flow test loop, which is part of the Gas-Cooled Fast Breeder Reactor Program (GCFR) at ORNL, is a high-pressure, high-temperature, out-of-reactor helium circulation system that is being constructed to permit study of the performance at steady-state and transient conditions of simulated segments of core assemblies for a GCFR demonstration plant. The simulated core segments, which are divided into zones, contain electrical heating elements to simulate the heat generated by fission. To control the power which is applied to a zone, a novel multitapped transformer and zone power control system have been designed and built which satisfy stringent design criteria. The controller can match power output to demand to within better than +-1% over a 900:1 dynamic range and perform full-power transients within 1 s. The power is applied in such a way as to minimize the electromagnetic interference at the bandwidth of the loop instrumentation, and the controller incorporates several error detection techniques, making it inherently fail-safe. The operation manual describes the specifications, operating instructions, error detection capabilities, error recovery, troubleshooting, calibration and QA procedures, and maintenance requirements. Also included are sections on the theory of operation, circuitry description, and a complete set of schematics.

Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

2017-03-15

Highlights: • Decay heat removal from degraded core of a typical SFR is highlighted. • Influence of number of DHXs in operation on PAHR is analyzed. • Investigations on structural integrity of the inner vessel and core catcher. • Feasibility study for retention of a part of debris in upper pool of SFR. - Abstract: Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also

Computer-Aided Test Flow in Core-Based Design

NARCIS (Netherlands)

Zivkovic, V.; Tangelder, R.J.W.T.; Kerkhoff, Hans G.

2000-01-01

This paper copes with the efficient test-pattern generation in a core-based design. A consistent Computer-Aided Test (CAT) flow is proposed based on the required core-test strategy. It generates a test-pattern set for the embedded cores with high fault coverage and low DfT area overhead. The CAT

Core-concrete molten pool dynamics and interfacial heat transfer

International Nuclear Information System (INIS)

Benjamin, A.S.

1980-01-01

Theoretical models are derived for the heat transfer from molten oxide pools to an underlying concrete surface and from molten steel pools to a general concrete containment. To accomplish this, two separate effects models are first developed, one emphasizing the vigorous agitation of the molten pool by gases evolving from the concrete and the other considering the insulating effect of a slag layer produced by concrete melting. The resulting algebraic expressions, combined into a general core-concrete heat transfer representation, are shown to provide very good agreement with experiments involving molten steel pours into concrete crucibles

Transitions to improved core electron heat confinement in JT-II plasmas

International Nuclear Information System (INIS)

Estrada, T.; Medina, F.; Ascasibar, E.; Balbin, R.; Castejon, F.; Hidalgo, C.; Lopez-Bruna, D.; Petrov, S.

2008-01-01

Transitions to improved core electron heat confinement are triggered by low order rational magnetic surfaces in TJ-II ECH plasmas. Transitions triggered by the rational surface n=4/m=2 show an increase in the ion temperature synchronized with the increase in the electron temperature. SXR measurements demonstrate that, under certain circumstances, the rational surface positioned inside the plasma core region precedes and provides a trigger for the transition. (author)

Grimsel Test Site: heat test, final report

International Nuclear Information System (INIS)

Schneefuss, J.; Glaess, F.; Gommlich, G.; Schmidt, M.

1989-05-01

The Swiss concept for the storage of radioactive waste consists in placing it in compact, dense rock formations. An experiment 'Heat Test' carried out by the 'Gesellschaft fuer Strahlen- und Umweltforschung' in Nagra's Grimsel rock laboratory simulated the heat production of stored radioactive waste. The aim was to evaluate processes for the demonstration of the suitability of a final repository for heat-producing radioactive waste in cristalline rock, to investigate the thermic, mechanic and hydraulic reactions to an artificial heat source, and to develop corresponding calculating models. The duration of the tests was about 3 years. In this report the measured thermic, mechanic and hydraulic reactions are documented and discussed in detail. A simple, rotation symmetrical FEM-model was used for the preparatory and experiment-accompanying modelling of the thermomechanical conditions in the heat test. The test showed that suitable measuring methods for the surveillance of the geomechanics of a final repository are available and that the reactions of the crystalline host rock to the heat source remain locally limited and can be modelled with relatively small effort. 29 refs., 33 figs., 10 tabs

Development of heat transfer package for core thermal-hydraulic design and analysis of upgraded JRR-3

International Nuclear Information System (INIS)

Sudo, Yukio; Ikawa, Hiromasa; Kaminaga, Masanori

1985-01-01

A heat transfer package was developed for the core thermal-hydraulic design and analysis of the Japan Research Reactor-3 (JRR-3) which is to be remodeled to a 20 MWt pool-type, light water-cooled reactor with 20 % low enriched uranium (LEU) plate-type fuel. This paper presents the constitution of the developed heat transfer package and the applicability of the heat transfer correlations adopted in it, based on the heat transfer experiments in which thermal-hydraulic features of the new JRR-3 core were properly reflected. (author)

Computer-Aided Test Flow in Core-Based Design

NARCIS (Netherlands)

Zivkovic, V.; Tangelder, R.J.W.T.; Kerkhoff, Hans G.

2000-01-01

This paper copes with the test-pattern generation and fault coverage determination in the core based design. The basic core-test strategy that one has to apply in the core-based design is stated in this work. A Computer-Aided Test (CAT) flow is proposed resulting in accurate fault coverage of

Evaluation of linear heat rates for the power-to-melt tests on 'JOYO' using the Monte-Carlo code 'MVP'

International Nuclear Information System (INIS)

Yokoyama, Kenji; Ishikawa, Makoto

2000-04-01

The linear heat rates of the power-to-melt (PTM) tests, performed with B5D-1 and B5D-2 subassemblies on the Experimental Fast Reactor 'JOYO', are evaluated with the continuous energy Monte-Carlo code, MVP. We can apply a whole core model to MVP, but it takes very long time for the calculation. Therefore, judging from the structure of B5D subassembly, we used the MVP code to calculate the radial distribution of linear heat rate and used the deterministic method to calculate the axial distribution. We also derived the formulas for this method. Furthermore, we evaluated the error of the linear heat rate, by evaluating the experimental error of the reactor power, the statistical error of Monte-Carlo method, the calculational model error of the deterministic method and so on. On the other hand, we also evaluated the burnup rate of the B5D assembly and compared with the measured value in the post-irradiation test. The main results are following: B5D-1 (B5101, F613632, core center). Linear heat rate: 600 W/cm±2.2%. Burnup rate: 0.977. B5D-2 (B5214, G80124, core center). Linear heat rate: 641 W/cm±2.2%. Burnup rate: 0.886. (author)

ISOTROPIC HEATING OF GALAXY CLUSTER CORES VIA RAPIDLY REORIENTING ACTIVE GALACTIC NUCLEUS JETS

International Nuclear Information System (INIS)

Babul, Arif; Sharma, Prateek; Reynolds, Christopher S.

2013-01-01

Active galactic nucleus (AGN) jets carry more than sufficient energy to stave off catastrophic cooling of the intracluster medium (ICM) in the cores of cool-core clusters. However, in order to prevent catastrophic cooling, the ICM must be heated in a near-isotropic fashion and narrow bipolar jets with P jet = 10 44–45 erg s –1 , typical of radio AGNs at cluster centers, are inefficient in heating the gas in the transverse direction to the jets. We argue that due to existent conditions in cluster cores, the supermassive black holes (SMBHs) will, in addition to accreting gas via radiatively inefficient flows, experience short stochastic episodes of enhanced accretion via thin disks. In general, the orientation of these accretion disks will be misaligned with the spin axis of the black holes (BHs) and the ensuing torques will cause the BH's spin axis (and therefore the jet axis) to slew and rapidly change direction. This model not only explains recent observations showing successive generations of jet-lobes-bubbles in individual cool-core clusters that are offset from each other in the angular direction with respect to the cluster center, but also shows that AGN jets can heat the cluster core nearly isotropically on the gas cooling timescale. Our model does require that the SMBHs at the centers of cool-core clusters be spinning relatively slowly. Torques from individual misaligned disks are ineffective at tilting rapidly spinning BHs by more than a few degrees. Additionally, since SMBHs that host thin accretion disks will manifest as quasars, we predict that roughly 1-2 rich clusters within z < 0.5 should have quasars at their centers.

Effect of upper plenum water accumuration on reflooding phenomena under forced-feed flooding in SCTF Core-I tests

International Nuclear Information System (INIS)

Sudo, Yukio; Sobajima, Makoto; Iwamura, Takamichi; Osakabe, Masahiro; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi

1983-07-01

Large Scale Reflood Test Program has been performed under contract with the Atomic Energy Bureau of Science and Technology Agency of Japan since 1976. The Slab Core Test Program is a part of the Large Scale Reflood Test Program along with the Cylindrical Core Test Program. Major purpose of the Slab Core Test Program is to investigate two-dimensional, thermo-hydrodynamic behavior in the core and the effect of fluid communication between the core and the upper plenum on the reflood phenomena in a postulated loss-of-coolant accident of a PWR. A significant upper plenum water accumulation was observed in the Base Case Test Sl-01 which was carried out under forced-feed flooding condition. To investigate the effects of upper plenum water accumulation on reflooding phenomena, accumulated water is extracted out of the upper plenum in Test Sl-03 by full opening of valves for extraction lines located just above the upper core support plate. This report presents this effect of upper plenum water accumulation on reflooding phenomena through the comparison of Tests Sl-01 and Sl-03. In spite of full opening of valves for upper plenum water extraction in Test Sl-03, a little water accumulation was observed which is of the same magnitude as in Test Sl-01 for about 200 s after the beginning of reflood. From 200 s after the beginning of reflood, however, the upper plenum water accumulation is much less in Test Sl-03 than in Test Sl-01, showing the following effects of upper plenum water accumulation. In Test Sl-03, (1) the two-dimensionality of horizontal fluid distribution is much less both above and in the core, (2) water carryover through hot leg and water accumulation in the core are less, (3) quench time is rather delayed in the upper part of the core by less water fall back from the upper plenum, and (4) difference in the core thermal behavior and core heat transfer are not significant in the middle and lower part of the core. (author)

Test-Access Planning and Test Scheduling for Embedded Core-Based System Chips

OpenAIRE

Goel, Sandeep Kumar

2005-01-01

Advances in the semiconductor process technology enable the creation of a complete system on one single die, the so-called system chip or SOC. To reduce time-to-market for large SOCs, reuse of pre-designed and pre-veried blocks called cores is employed. Like the design style, testing of SOCs can be best approached in a core-based fashion. In order to enable core-based test development, an embedded core should be isolated from its surrounding circuitry and electrical test access from chip pins...

Critical heat flux predictions for the Sandia Annular Core Research Reactor

International Nuclear Information System (INIS)

Rao, D.V.; El-Genk, M.S.

1994-08-01

This study provides best estimate predictions of the Critical Heat Flux (CHF) and the Critical Heat Flux Ratio (CHFR) to support the proposed upgrade of the Annual Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) from its present value of 2 MWt to 4 MWt. These predictions are based on the University of New Mexico (UNM) - CHF correlation, originally developed for uniformly heated vertical annuli. The UNM-CHF correlation is applicable to low-flow and low-pressure conditions, which are typical of those in the ACRR. The three hypotheses that examined the effect of the nonuniform axial heat flux distribution in the ACRR core are (1) the local conditions hypotheses, (2) the total power hypothesis, and (3) the global conditions hypothesis. These hypotheses, in conjunction with the UNM-CHF correlation, are used to estimate the CHF and CHFR in the ACRR. Because the total power hypothesis predictions of power per rod at CHF are approximately 15%-20% lower than those corresponding to saturation exit conditions, it can be concluded that the total power hypothesis considerably underestimates the CHF for nonuniformly heated geometries. This conclusion is in agreement with previous experimental results. The global conditions hypothesis, which is more conservative and more accurate of the other two, provides the most reliable predictions of CHF/CHFR for the ACRR. The global conditions hypothesis predictions of CHFR varied between 2.1 and 3.9, with the higher value corresponding to the lower water inlet temperature of 20 degrees C

Core heat transfer experiment for JRR-3 to be upgraded at 20 MWt, 2

International Nuclear Information System (INIS)

Sudo, Yukio; Miyata, Keiichi; Ikawa, Hiromasa; Ohgawara, Masami; Kaminaga, Masanori

1985-09-01

Experiments were carried out to investigate the condition of onset of nucleate boiling (ONB) and the departure from nucleate boiling (DNB) heat flux under forced convection in a vertical rectangular channel, both of which take important roles in the core thermal-hydraulic design of the upgraded JRR-3. This report presents the validity and applicability of the correlations proposed for ONB condition and DNB heat flux, based on the analysis of the experimental results. The upgraded JRR-3 is a low-pressure, low-temperature research reactor and the core heat generation is removed by two cooling modes, one is natural circulation under upflow up to 200 kW and the other is forced circulation under downflow up to 20 MW. Therefore, the difference in heat transfer characteristics between upflow and downflow were investigated in the experiments, which were carried out by using a heated channel properly simulating a subchannel of fuel element because the heat transfer characteristics are considered to be strongly dependent on the configuration of flow channel. (author)

Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model

Directory of Open Access Journals (Sweden)

Reza Akbari

2017-08-01

Full Text Available Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very fast model for predicting thermal hydraulics behavior of pressurized water reactor core has been developed. For verifying the results of this model, we used RELAP5 code as US nuclear regulatory approved thermal hydraulics code. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core.

A scaling study of the natural circulation flow of the ex-vessel core catcher cooling system of a 1400MW PWR for designing a scale-down test facility

International Nuclear Information System (INIS)

Rhee, Bo. W.; Ha, K. S.; Park, R. J.; Song, J. H.

2012-01-01

A scaling study on the steady state natural circulation flow along the flow path of the ex-vessel core catcher cooling system of 1400MWe PWR is described. The scaling criteria for reproducing the same thermalhydraulic characteristics of the natural circulation flow as the prototype core catcher cooling system in the scale-down test facility is derived and the resulting natural circulation flow characteristics of the prototype and scale-down facility analyzed and compared. The purpose of this study is to apply the similarity law to the prototype EU-APR1400 core catcher cooling system and the model test facility of this prototype system and derive a relationship between the heating channel characteristics and the down-comer piping characteristics so as to determine the down-comer pipe size and the orifice size of the model test facility. As the geometry and the heating wall heat flux of the heating channel of the model test facility will be the same as those of the prototype core catcher cooling system except the width of the heating channel is reduced, the axial distribution of the coolant quality (or void fraction) is expected to resemble each other between the prototype and model facility. Thus using this fact, the down-comer piping design characteristics of the model facility can be determined from the relationship derived from the similarity law

Evaluation report on CCTF Core-II reflood tests C2-AC1 (run 51) and C2-4 (run 62)

International Nuclear Information System (INIS)

Sugimoto, Jun; Iguchi, Tadashi; Murao, Yoshio

1984-02-01

A reflood test program has been conducted at Japan Atomic Energy Research Institute (JAERI) using large scale test facilities named Cylindrical Core Test Facility (CCTF) and Slab Core Test Facility (SCTF). The present report describes the effect of the initial clad temperature i.e., the initial stored energy on reflood phenomena observed in CCTF Core-II tests C2-ACl and C2-4. The peak clad temperatures of tests C2-ACl and C2-4 were 863 K and 1069 K, respectively at reflood initiation. With higher initial clad temperature, obtained were lower water accumulation in the core and upper plenum, and higher loop mass flow rate in an early reflood transient due to larger heat release of the stored energy in the core. Core inlet flow conditions were only affected shortly after the reflood initiation, causing the suppressed flooding rate and the larger U-tube flow oscillation between the core and the downcomer. In the core, with higher initial clad temperature, slower quench front propagation and higher turnaround temperature were observed. Responses to a higher initial clad temperature were similar to those observed in CCTF Core-I and FLECHT tests. Thus, the lower temperature rise with higher initial clad temperature was experimentally confirmed. The importance of higher flooding rate at initial period was analytically shown for further decreasing the temperature rise. (author)

Analysis of gamma heating at TRIGA mark reactor core Bandung using plate type fuel

International Nuclear Information System (INIS)

Setiyanto; Tukiran Surbakti

2016-01-01

In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities) and central irradiation position (CIP), especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0.87 W/g), but very low value for Lazy Susan position (lest then 0.11 W/g). Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. (author)

Evaluation report on CCTF CORE-I REFLOOD TEST Cl-4 (Run 13) and Cl-15 (Run 24)

International Nuclear Information System (INIS)

Sudoh, Takashi; Murao, Yoshio.

1983-08-01

The tests Cl-4 and Cl-15 were performed with the Cylindrical Core Test Facility (CCTF) to investigate the effects of the depressurization process to simulate the refill phase, and the effects of the nitrogen to be injected after the end of the accumulator injection on the thermo-hydraulic behavior in the core and primary loop system during refill and reflood phases. In these tests, after the lower plenum was filled to 0.9m level with saturated water at 0.6 MPa, the accumulator water was injected into three intact cold legs in the depressurization period from 0.6 MPa to 0.2 MPa. The water in the lower plenum voided during the depressurization and the significant steam condensation occurred in or near the intact cold legs. The condensation caused high steam flow rate in the intact loops and the lower plenum flashing resulted in suppressed core water accumulation. The slightly lower core heat transfer coefficient due to the less core water caused the higher turnaround temperature and the longer quench time than those of the normal reflood test without the depressurization process. The nitrogen injection followed the accumulator injection was allowed in the test Cl-15. However, significant effects of the nitrogen injection was not observed. (author)

Results of out-of-pile experiments to investigate the possibilities of cooling a core melt with internal heat production

International Nuclear Information System (INIS)

Fieg, G.

1976-01-01

After serious hypothetical reactor accidents, melted core materials with internal heat production can occur in large quantities. A retention of these molten core masses within the containment must be ensured. The knowledge of the heat transport from volume-heated layers is necessary to clarify this matter. (orig./LH) [de

Modification of heating system on HeaTiNG-02 test section of beta test loop

International Nuclear Information System (INIS)

Sagino; Dedy Haryanto; Riswan Djambiar; Edy Sumarno

2013-01-01

Modifications have been carried out on the heating test section heating-02 on the integration strand Beta Test (UUB). The activities carried out to overcome the obstacles that arise in the test section when used. Constraint that often arises is the fall of the heating source super chantal when it reaches a certain temperature. To mitigate the super chantal is initially converted into a horizontal vertical position. Change from vertical to horizontal position on super chantal aims to stabilize the position of super chantal, so it needs to be modified in the heating system. Modification activities include manufacturing, installation and testing of super chantal and refractory stone as super chantal support. Manufacturing refractory stone formation and assembly into the heater in accordance with design modifications that have been done in electromechanical workshop obtained using some machine tools. Testing results of fabrication has been done by providing voltage 110 volts until it reaches operating temperature 400°C. Test results obtained super chantal stable position when it reaches operating temperature, and heater of heating-02 test section feasible to be used for experiments. (author)

Heat and fission product transport in molten core material pool with crust

International Nuclear Information System (INIS)

Yun, J.I.; Suh, K.Y.; Kang, C.S.

2005-01-01

Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the reactor vessel during a severe accident. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool is estimated by product of the mass concentration and energy conversion factor of each fission product. Twenty-nine elements are chosen and classified by their chemical properties to calculate heat generation rate in the pool. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis is performed for heat and fission product transport in a molten core material pool during the Three Mile Island Unit 2 (TMI-2) accident. The pool is assumed to be a partially filled hemisphere, whose change in geometry is neglected during the numerical calculation. Calculated results indicate that the peak temperature in the molten pool is significantly lowered, since a substantial amount of the volatile fission products is released from the molten pool during progression of the accident. The results may directly be applied to the existing severe accident analysis codes to more mechanistically determine the thermal load to the reactor vessel lower head during the in-vessel retention

Comparison of heat-testing methodology.

Science.gov (United States)

Bierma, Mark M; McClanahan, Scott; Baisden, Michael K; Bowles, Walter R

2012-08-01

Patients with irreversible pulpitis occasionally present with a chief complaint of sensitivity to heat. To appropriately diagnose the offending tooth, a variety of techniques have been developed to reproduce this chief complaint. Such techniques cause temperature increases that are potentially damaging to the pulp. Newer electronic instruments control the temperature of a heat-testing tip that is placed directly against a tooth. The aim of this study was to determine which method produced the most consistent and safe temperature increase within the pulp. This consistency facilitates the clinician's ability to differentiate between a normal pulp and irreversible pulpitis. Four operators applied the following methods to each of 4 extracted maxillary premolars (for a total of 16 trials per method): heated gutta-percha, heated ball burnisher, hot water, and a System B unit or Elements unit with a heat-testing tip. Each test was performed for 60 seconds, and the temperatures were recorded via a thermocouple in the pulp chamber. Analysis of the data was performed by using the intraclass correlation coefficient. The least consistent warming was found with hot water. The heat-testing tip also demonstrated greater consistency between operators compared with the other methods. Hot water and the heated ball burnisher caused temperature increases high enough to damage pulp tissue. The Elements unit with a heat-testing tip provides the most consistent warming of the dental pulp. Copyright © 2012 American Association of Endodontists. Published by Elsevier Inc. All rights reserved.

Nuclear start-up, testing and core management of the Fast Test Reactor (FTR)

International Nuclear Information System (INIS)

Bennett, R.A.; Daughtry, J.W.; Harris, R.A.; Jones, D.H.; Nelson, J.V.; Rawlins, J.A.; Rothrock, R.B.; Sevenich, R.A.; Zimmerman, B.D.

1980-01-01

Plans for the nuclear start-up, low and high power physics testing, and core management of the Fast Test Reactor (FTR) are described. Owing to the arrangement of the fuel-handling system, which permits continuous instrument lead access to experiments during refuelling, it is most efficient to load the reactor in an asymmetric fashion, filling one-third core sectors at a time. The core neutron level will be monitored during this process using both in-core and ex-core detectors. A variety of physics tests are planned following the core loading. Because of the experimental purpose of the reactor, these tests will include a comprehensive characterization programme involving both active and passive neutron and gamma measurements. Following start-up tests, the FTR will be operated as a fast neutron irradiation facility, to test a wide variety of fast reactor core components and materials. Nuclear analyses will be made prior to each irradiation cycle to confirm that the planned arrangement of standard and experimental components satisfies all safety and operational constraints, and that all experiments are located so as to achieve their desired irradiation environment. (author)

Heat Transfer Characteristics of SiC-coated Heat Pipe for Passive Decay Heat Removal

International Nuclear Information System (INIS)

Kim, Kyung Mo; Kim, In Guk; Jeong, Yeong Shin; Bang, In Cheol

2014-01-01

The main concern with the Fukushima accident was the failure of active and passive core cooling systems. The main function of existing passive decay heat removal systems is feeding additional coolant to the reactor core. Thus, an established emergency core cooling system (ECCS) cannot operate properly because of impossible depressurization under the station blackout (SBO) condition. Therefore, a new concept for passive decay heat removal system is required. In this study, an innovative hybrid control rod concept is considered for passive in-core decay heat removal that differs from the existing direct vessel injection core cooling system and passive auxiliary feedwater system (PAFS). The heat transfer between the evaporator and condenser sections occurs by phase change of the working fluid and capillary action induced by wick structures installed on the inner wall of the heat pipe. In this study, a hybrid control rod is developed to take the roles of both neutron absorption and heat removal by combining the functions of a heat pipe and control rod. Previous studies on enhancing the heat removal capacity of heat pipes used nanofluids, self-rewetting fluids, various wick structures and condensers. Many studies have examined the thermal performances of heat pipes using various nanofluids. They concluded that the enhanced thermal performance of the heat pipe using nanofluids is due to nanoparticle deposition on the wick structures. Thus, the wick structure of heat pipes has been modified by nanoparticle deposition to enhance the heat removal capacity. However, previous studies used relatively small heat pipes and narrow ranges of heat loads. The environment of a nuclear reactor is very specific, and the decay heat produced by fission products after shutdown is relatively large. Thus, this study tested a large-scale heat pipe over a wide range of power. The concept of a hybrid heat pipe for an advanced in-core decay heat removal system was introduced for complete

Heat Transfer Characteristics of SiC-coated Heat Pipe for Passive Decay Heat Removal

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung Mo; Kim, In Guk; Jeong, Yeong Shin; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-10-15

The main concern with the Fukushima accident was the failure of active and passive core cooling systems. The main function of existing passive decay heat removal systems is feeding additional coolant to the reactor core. Thus, an established emergency core cooling system (ECCS) cannot operate properly because of impossible depressurization under the station blackout (SBO) condition. Therefore, a new concept for passive decay heat removal system is required. In this study, an innovative hybrid control rod concept is considered for passive in-core decay heat removal that differs from the existing direct vessel injection core cooling system and passive auxiliary feedwater system (PAFS). The heat transfer between the evaporator and condenser sections occurs by phase change of the working fluid and capillary action induced by wick structures installed on the inner wall of the heat pipe. In this study, a hybrid control rod is developed to take the roles of both neutron absorption and heat removal by combining the functions of a heat pipe and control rod. Previous studies on enhancing the heat removal capacity of heat pipes used nanofluids, self-rewetting fluids, various wick structures and condensers. Many studies have examined the thermal performances of heat pipes using various nanofluids. They concluded that the enhanced thermal performance of the heat pipe using nanofluids is due to nanoparticle deposition on the wick structures. Thus, the wick structure of heat pipes has been modified by nanoparticle deposition to enhance the heat removal capacity. However, previous studies used relatively small heat pipes and narrow ranges of heat loads. The environment of a nuclear reactor is very specific, and the decay heat produced by fission products after shutdown is relatively large. Thus, this study tested a large-scale heat pipe over a wide range of power. The concept of a hybrid heat pipe for an advanced in-core decay heat removal system was introduced for complete

LOCA simulation tests in the RD-12 loop with multiple heat channels

International Nuclear Information System (INIS)

Ardron, K.H.; McGee, G.R.; Hawley, E.H.

1985-11-01

A series of tests has been performed in the RD-12 loop to study the bahaviour of a CANDU-type, primary heat transport system (PHTS) during the blowdown and injection phases of a loss-of-coolant accident (LOCA). Specifically, the tests were used to investigate flow stagnation and refilling of the core following a LOCA. RD-12 is a pressurized water loop with the basic geometry of a CANDU reactor PHTS, but at approximately 1/125 volume scale. The loop consists of U-tube steam generators, pumps, headers, feeders, and heated channels arranged in the symmetrical figure-of-eight configuration of the CANDU PHTS. In the LOCA simulation tests, the loop contained four horizontal heated channels, each containing a seven-element assembly of indirectly heated, fuel-rod simulators. The channels were nominally identical, and were arranged in parallel pairs between the headers in each half-circuit. Tests were carried out using various restricting orifices to represent pipe breaks of different sizes. The break sizes were specifically chosen such that stagnation conditions in the heated channels would be likely to occur. In some tests, the primary pumps were programmed to run down over a 100-s period to simulate a LOCA with simultaneous loss of pump power. Test results showed that, for certain break sizes, periods of low flow occurred in the channels in one half of the loop, leading to flow stratification and sheath temperature excursions. This report reviews the results of two of the tests, and discusses possible mechanisms that may have led to the low channel flow conditions observed in some cases. Plans for future experiments in the larger scale RD-14 facility are outlined. 5 refs

Design heating test section HeaTiNG-02

International Nuclear Information System (INIS)

Riswan Djambiar; Sagino; Dedy Haryanto; Joko Prasetio Witoko

2013-01-01

HeaTiNG-02 is a component test loop BETA which serves as a heater in conducting experimental heat transfer processes in two-phase flow in narrow slit-shaped plate, considering this phenomenon is one of the conditions postulated accident scenarios a NPP type PWR. To produce heat for the heating component takes the AC power the source voltage can be set from 0 Volts to 220 Volts with no more than a maximum power of 25 KVA. To obtain the thermal conditions on HeaTiNG-02 heating wire dimensions need to be determined and the corresponding voltage so that it will an expected power. Determination of the dimensions of the heater wire through calculations using electricity formulations. Retrieved draft heater test BETA (UUB) HeaTiNG-02 use material super kanthal (FeCuAl) in diameter (Ø) = 2 mm and wire length 3770 mm. Voltage regulators with a maximum power of 25 kVA with a minimum voltage of 0 volts and a maximum of 220 volts. Heater is used as the base to form refractory stone trench. (author)

Influence of composition and rate heating on formation of black core in bodies obtained with red ceramic

International Nuclear Information System (INIS)

Santana, L.N.L.; Goncalves, W.P.; Silva, B.J. da; Macedo, R.S.; Santos, R.C.; Lisboa, D.

2011-01-01

In the heating of pieces of red pottery can the defect known as black core, this may deteriorate the technical and aesthetic characteristics of the final product. This study evaluated the influence of chemical composition and heating rate on the formation of black core in bodies red ceramic. The masses were treated and samples were extruded, dried, sintered at 900 °C, with heating rates of 5, 10, 15, 20 and 30 °C / min. and determined the following properties: water absorption, linear shrinkage and flexural strength. The pieces made with the mass containing lower content of iron oxide showed better resistance to bending when subjected to rapid heating. The presence of the black core was identified through visual analysis of the pieces after the break, being more apparent in parts subject to rates above 5 °C / min. (author)

Simulation of nonlinear dynamics of a PWR core by an improved lumped formulation for fuel heat transfer

International Nuclear Information System (INIS)

Su, Jian; Cotta, Renato M.

2000-01-01

In this work, thermohydraulic behaviour of PWR, during reactivity insertion and partial loss-of-flow, is simulated by using a simplified mathematical model of reactor core and primary coolant. An improved lumped parameter formulation for transient heat conduction in fuel rod is used for core heat transfer modelling. Transient temperature response of fuel, cladding and coolant is analysed. (author)

Pressurized Hybrid Heat Pipe for Passive IN-Core Cooling System (PINCs) in Advanced Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung Mo; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2016-05-15

The representative operating limit of the thermosyphon heat pipe is flooding limit that arises from the countercurrent flow of vapor and liquid. The effect of difference between wetted perimeter and heated perimeter on the flooding limit of the thermosyphons has not been studied; despite the effect of cross-sectional area of the vapor path on the heat transfer characteristics of thermosyphons have been studied. Additionally, the hybrid heat pipe must operate at the high temperature and high pressure environment because it will be inserted to the active core to remove the decay heat. However, the previously studied heat pipes operated below the atmospheric pressure. Therefore, the effect of the unique geometry for hybrid heat pipe and operating pressure on the heat transfer characteristics including the flooding limit of hybrid heat pipe was experimentally measured. Hybrid heat pipe as a new conceptual decay heat removal device was proposed. For the development of hybrid heat pipe operating at high temperature and high pressure conditions, the pressurized hybrid heat pipe was prepared and the thermal performances including operation limits of hybrid heat pipe were experimentally measured. Followings were obtained: (1) As operating pressure of the heat pipe increases, the evaporation heat transfer coefficient increases due to heat transfer with convective pool boiling mode. (2) Non-condensable gas charged in the test section for the pressurization lowered the condensation heat transfer by impeding the vapor flow to the condenser. (3) The deviations between experimentally measured flooding limits for hybrid heat pipes and the values from correlation for annular thermosyphon were observed.

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

Tuned Chamber Core Panel Acoustic Test Results

Science.gov (United States)

Schiller, Noah H.; Allen, Albert R.

2016-01-01

This report documents acoustic testing of tuned chamber core panels, which can be used to supplement the low-frequency performance of conventional acoustic treatment. The tuned chamber core concept incorporates low-frequency noise control directly within the primary structure and is applicable to sandwich constructions with a directional core, including corrugated-, truss-, and fluted-core designs. These types of sandwich structures have long, hollow channels (or chambers) in the core. By adding small holes through one of the facesheets, the hollow chambers can be utilized as an array of low-frequency acoustic resonators. These resonators can then be used to attenuate low-frequency noise (below 400 Hz) inside a vehicle compartment without increasing the weight or size of the structure. The results of this test program demonstrate that the tuned chamber core concept is effective when used in isolation or combined with acoustic foam treatments. Specifically, an array of acoustic resonators integrated within the core of the panels was shown to improve both the low-frequency absorption and transmission loss of the structure in targeted one-third octave bands.

A concept of passive safety pressurized water reactor system with inherent matching nature of core heat generation and heat removal

International Nuclear Information System (INIS)

Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke

1995-01-01

The reduction of manpower in operation and maintenance by simplification of the system are essential to improve the safety and the economy of future light water reactors. At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for this purpose and in the concept minimization of developing work and conservation of scale-up capability in design were considered. The inherent matching nature of core heat generation and heat removal rate is introduced by the core with high reactivity coefficient for moderator density and low reactivity coefficient for fuel temperature (Doppler effect) and once-through steam generators (SGs). This nature makes the nuclear steam supply system physically-slave for the steam and energy conversion system by controlling feed water mass flow rate. The nature can be obtained by eliminating chemical shim and adopting in-vessel control rod drive mechanism (CRDM) units and a low power density core. In order to simplify the system, a large pressurizer, canned pumps, passive residual heat removal systems with air coolers as a final heat sink and passive coolant injection system are adopted and the functions of volume and boron concentration control and seal water supply are eliminated from the chemical and volume control system (CVCS). The emergency diesel generators and auxiliary component cooling system of 'safety class' for transferring heat to sea water as a final heat sink in emergency are also eliminated. All of systems are built in the containment except for the air coolers of the passive residual heat removal system. The analysis of the system revealed that the primary coolant expansion in 100% load reduction in 60 s can be mitigated in the pressurizer without actuating the pressure relief valves and the pressure in 50% load change in 30 s does not exceed the maximum allowable pressure in accidental conditions in regardless of pressure regulation. (author)

Heat transfer in reactor cavity during core-concrete interaction

International Nuclear Information System (INIS)

Adroguer, B.; Cenerino, G.

1989-08-01

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

Thermal-Hydraulic Tests for Reactor Core Safety

International Nuclear Information System (INIS)

Chun, Se Young; Chung, Moon Ki; Baek, Won Pil and others

2005-04-01

The reflood experiments for single rod annulus geometry have been performed to investigate the effect of spacer grid on thermal-hydraulics under reflood conditions. The reflood experimental loop for 6x6 rod bundle with a spacer grid developed in Korea has been provided. About 8000 data points for Post-CHF heat transfer have been obtained from the experiments About 1400 CHF data points for 3x3 Water and 5x5 Freon rod bundles have been obtained. The existing evaluation methodology for core safety under return-to-power conditions has been investigated using KAERI low flow CHF database. The hydraulic tests for turbulence mixing characteristics in subchannel of 5x5 rod bundle have been carried out using advanced measurement technique, LVD and the database for various spacer grids have been provided. In order to measure the turbulence mixing characteristics in details, the hydraulic loop with a magnified 5x5 rod bundle has been prepared. The database which was constructed through a systematic thermal hydraulic tests for the reflood phenomenon, CHF, Post-CHF is surely to be useful to the industry field, the regulation body and the development of thermal-hydraulic analysis code

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Concept Design of a Gravity Core Cooling Tank as a Passive Residual Heat Removal System for a Research Reactor

International Nuclear Information System (INIS)

Lee, Kwonyeong; Chi, Daeyoung; Kim, Seong Hoon; Seo, Kyoungwoo; Yoon, Juhyeon

2014-01-01

A core downward flow is considered to use a plate type fuel because it is benefit to install the fuel in the core. If a flow inversion from a downward to upward flow in the core by a natural circulation is introduced within a high heat flux region of residual heat, the fuel fails instantly due to zero flow. Therefore, the core downward flow should be sufficiently maintained until the residual heat is in a low heat flux region. In a small power research reactor, inertia generated by a flywheel of the PCP can maintain a downward flow shortly and resolve the problem of a flow inversion. However, a high power research reactor more than 10 MW should have an additional method to have a longer downward flow until a low heat flux. Usually, other research reactors have selected an active residual heat removal system as a safety class. But, an active safety system is difficult to design and expensive to construct. A Gravity Core Cooling Tank (GCCT) beside the reactor pool with a Residual Heat Removal Pipe connecting two pools was developed and designed preliminarily as a passive residual heat removal system for an open-pool type research reactor. It is very simple to design and cheap to construct. Additionally, a non-safety, but active residual heat removal system is applied with the GCCT. It is a Pool Water Cooling and Purification System. It can improve the usability of the research reactor by removing the thermal waves, and purify the reactor pool, the Primary Cooling System, and the GCCT. Moreover, it can reduce the pool top radiation level

Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

International Nuclear Information System (INIS)

Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol

2015-01-01

As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

Modeling and simulation of loss of the ultimate heat sink in a typical material testing reactor

International Nuclear Information System (INIS)

El-Khatib, Hisham; El-Morshedy, Salah El-Din; Higazy, Maher G.; El-Shazly, Karam

2013-01-01

Highlights: ► A thermal–hydraulic model has been developed to simulate loss of the ultimate heat sink in MTR. ► The model involves three coupled sub-models for core, heat exchanger and cooling tower. ► The model is validated against PARET for steady-state and verified by operation data for transients. ► The model is used to simulate the behavior of the reactor under a loss of the ultimate heat sink. ► The model results are analyzed and discussed. -- Abstract: A thermal–hydraulic model has been developed to simulate loss of the ultimate heat sink in a typical material testing reactor (MTR). The model involves three interactively coupled sub-models for reactor core, heat exchanger and cooling tower. The model is validated against PARET code for steady-state operation and verified by the reactor operation records for transients. Then, the model is used to simulate the thermal–hydraulic behavior of the reactor under a loss of the ultimate heat sink event. The simulation is performed for two operation regimes: regime I representing 11 MW power and three cooling tower cells operated, and regime II representing 22 MW power and six cooling tower cells operated. In regime I, the simulation is performed for 1, 2 and 3 cooling tower cells failed while in regime II, it is performed for 1, 2, 3, 4, 5 and 6 cooling tower cells failed. The simulation is performed under protected conditions where the safety action called power reduction is triggered by reactor protection system to decrease the reactor power by 20% when the coolant inlet temperature to the core reaches 43 °C and scram is triggered if the core inlet temperature reaches 44 °C. The model results are analyzed and discussed.

Results of recent reactor-material tests on dispersal of oxide fuel from a disrupted core

International Nuclear Information System (INIS)

Spencer, B.W.; Wilson, R.J.; Vetter, D.L.; Erickson, E.G.; Dewey, G.

1985-01-01

The results of experimental investigations and related analyses are reported addressing the dispersal of molten oxide fuel from a disrupted core via various available pathways for the CRBR system. These investigations included the GAPFLOW tests in which pressure-driven and gravity drainage tests were performed using dispersal pathways mocking up the intersubassembly gaps, the CAMEL C6 and C7 tests in which molten fuel entered sodium-filled control assembly ducts under prototypic thermal-hydraulic conditions, and the Lower Internals Drainage (LID) tests in which molten fuel drained downward through simulated below-core structure (orifice plate stacks) as the bottom of control assembly ducts. The results of SHOTGUN tests addressing basic freezing of molten UO 2 and UO 2 /metal mixtures flowing through circular tubes are also reported. Test results have invariably shown the existance of stable UO 2 crusts on the inside surfaces of the flow paths. Appreciable removal of fuel was indicated prior to freezing-induced immobilization. Application of heat transfer models based upon the presence of stable, insulating fuel crusts tends to overpredict the removal process

Test Anxiety, Computer-Adaptive Testing and the Common Core

Science.gov (United States)

Colwell, Nicole Makas

2013-01-01

This paper highlights the current findings and issues regarding the role of computer-adaptive testing in test anxiety. The computer-adaptive test (CAT) proposed by one of the Common Core consortia brings these issues to the forefront. Research has long indicated that test anxiety impairs student performance. More recent research indicates that…

Heat transfer test in a vertical tube using CO2 at supercritical pressures

International Nuclear Information System (INIS)

Kim, Hwan Yeol; Kim, Hyungrae; Song, Jin Ho; Cho, Bong Hyun; Bae, Yoon Yeong

2007-01-01

Heat transfer test facility, SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt Generation), was constructed at KAERI (Korea Atomic Energy Research Institute) for an investigation of the thermal-hydraulic behaviors of supercritical CO 2 at the various geometries of the test section. The test data will be used for the reactor core design of the SCWR (SuperCritical Water-cooled Reactor). As a working fluid, CO 2 was selected to make use of the low critical pressure and temperature of CO 2 compared with water. An experimental study was carried out in the SPHINX to investigate the characteristics of heat transfer and pressure drop at a vertical single tube with an inside diameter of 4.4 mm in case of an upward flow of supercritical CO 2 . The heat and mass fluxes were varied at a given pressure. The mass flux was in the range of 400-1,200 kg/m 2 s and the heat flux was chosen up to 150 kW/m 2 . The selected pressures were 7.75, 8.12, and 8.85 MPa. A heat transfer deterioration occurred at the lower mass fluxes. The experimental heat transfer coefficients were compared with the ones predicted by several existing correlations. The standard deviation was about 20% for each correlation and an apparent discrepancy was not found among the correlations. The major components of the pressure drop were a gravitational pressure drop and a frictional pressure drop. The frictional pressure drop increases as the mass flux and heat flux increase. (author)

The effect of humidified heated breathing circuit on core body temperature in perioperative hypothermia during thyroid surgery.

Science.gov (United States)

Park, Hue Jung; Moon, Ho Sik; Moon, Se Ho; Do Jeong, Hyeon; Jeon, Young Jae; Do Han, Keung; Koh, Hyun Jung

2017-01-01

Purpose: During general anesthesia, human body easily reaches a hypothermic state, which is mainly caused by heat redistribution. Most studies suggested that humidified heated breathing circuits (HHBC) have little influence on maintenance of the core temperature during early phase of anesthesia. This study was aimed at examining heat preservation effect with HHBC in case of undergoing surgery with less exposure of surgical fields and short surgical duration. Methods: Patients aged 19 to 70 yr - old, ASA-PS I or II who were scheduled for elective thyroidectomy were assigned and divided to the group using HHBC (G1) and the group using conventional circuit (G2) by random allocation. During operation, core, skin, and room temperatures were measured every 5minutes by specific thermometer. Results: G1 was decreased by a lesser extent than G2 in core temperature, apparently higher at 30 and 60 minutes after induction. Skin and room temperatures showed no differences between the two groups (p>0.05). Consequently, we confirmed HHBC efficiently prevented a decrease in core temperature during early period in small operation which has difficulty in preparing warming devices or environments were not usually considered. Conclusions: This study showed that HHBC influences heat redistribution in early period of operation and can lessen the magnitude of the decrease in core body temperature. Therefore, it can be applied efficiently for other active warming devices in mild hypothermia.

Characteristic test of initial HTTR core

International Nuclear Information System (INIS)

Nojiri, Naoki; Shimakawa, Satoshi; Fujimoto, Nozomu; Goto, Minoru

2004-01-01

This paper describes the results of core physics test in start-up and power-up of the HTTR. The tests were conducted in order to ensure performance and safety of the high temperature gas cooled reactor, and was carried out to measure the critical approach, the excess reactivity, the shutdown margin, the control rod worth, the reactivity coefficient, the neutron flux distribution and the power distribution. The expected core performance and the required reactor safety characteristics were verified from the results of measurements and calculations

Technical specification improvements to containment heat removal and emergency core cooling systems: Final report

International Nuclear Information System (INIS)

Sullivan, W.P.; Ha, C.; Pentzien, D.C.; Visweswaran, S.

1988-07-01

This report presents the results of an analysis for technical specification improvements to the emergency core cooling systems (ECCS) and containment heat removal systems (EPRI Research Project 2142-3). The objective of this project is to further develop a reliability- and risk-based methodology to provide improvements by considering groups of surveillance test intervals and allowed out-of-service times jointly. This was done for the technical specifications for the ECCS, containment heat removal equipment, and supporting systems of a boiling water reactor plant. The project (1) developed a methodology for optimizing groups of surveillance test intervals and allowed out-of-service times jointly, (2) applied the methodology in a case study of a specific operating plant, Hatch-2, and (3) evaluated benefits of the application. The results of the case study demonstrate that beneficial technical specification improvements can be realized with application of the methodology. By tightening a small group of sensitive surveillance test intervals (STIs) and allowed out-of-service times (AOTs), a larger group of less sensitive STIs and AOTs can be extended resulting in an overall plant operating cost improvement without reducing the plant safety. The reliability- and risk-based methodology and results from this project can be effectively applied for technical specification improvements at other operating plants

Flow and heat transfer thermohydraulic modelisation during the reflooding phase of a P.W.R.'s core

International Nuclear Information System (INIS)

Raymond, Patrick

1978-04-01

Some generalities about L.O.C.A. are first recalled. The French experimental studies about Emergency Core Cooling System are briefly described. The different heat transfer mechanisms to take into account, according to the flow pattern in the dry zone, and the correlations or methods to calculate them, are defined. Then the Thermohydraulic code computer: FLIRA, which describe the reflooding phase, and a modelisation taking into account the different flow patterns are setted. A first interpretation of ERSEC experiments with a tubular test section shows that it is possible, with this modelisation and some classical heat transfer correlations, to describe the reflooding phase. [fr

Development and testing of passive autocatalytic recombiners cooled by heat pipes

International Nuclear Information System (INIS)

Granzow, Christoph

2012-01-01

A severe accident in a nuclear power plant (NPP) can lead to core damage in conjunction with the release of large amounts of hydrogen. As hydrogen mitigation measure, passive autocatalytic recombiners (PARs) are used in today's pressurized water reactors. PARs recombine hydrogen and oxygen contained in the air to steam. The heat from this exothermic reaction causes the catalyst and its surroundings to heat up. If parts of the PAR heat up above the ignition temperature of the gas mixture, a spontaneous deflagration or detonation can occur. The aim of this work is the prevention of such high temperatures by means of passive cooling of the catalyst with heat pipes. Heat pipes are completely passive heat exchanger with a very high effective thermal conductivity. For a deeper understanding of the reaction kinetics at lower temperatures, single catalytic coated heat pipes are studied in a flow reactor. The development of a modular small-scale PAR model is then based on a test series with cooled catalyst sheets. Finally, the PAR model is tested inside a pressure vessel under boundary conditions similar to a real NPP. The experiments show, that the temperatures of the cooled catalytic sheets stay significantly below the temperature of the uncooled sheets and below the ignition temperature of the gas mixture under any set boundary conditions, although no significant reduction of the conversion efficiency can be observed. As a last point, a mathematical model of the reaction kinetics of the recombination process as well as a model of the fluid dynamic and thermohydraulic processes in a heat pipe are developed with the data obtained from the experiments.

Xenon oscillation tests in four-loop PWR cores

International Nuclear Information System (INIS)

Aoki, Norihiko; Osaka, Kenichi; Shimada, Shoichiro; Tochihara, Hiroshi; Machii, Seigo

1980-01-01

The Kansai Electric Power Co.'s OHI Unit 1 and 2 are the first 4-loop PWRs in Japan which use 17 x 17 fuel assemblies and have essentially the same plant parameters. A 4-loop core has larger core radius and higher power density than those of 2- or 3-loop cores, and is less stable for Xe oscillation. It is therefore important to confirm that Xe oscillations in radial direction are sufficiently stable in a 4-loop core. Radial and axial Xe oscillation tests were performed during the startup physics tests of OHI Unit 1 and 2; Xe oscillation was induced by perturbation of control rods and the Xe effect on power distribution observed periodically. The test results show that the transverse Xe oscillation in the 4-loop core is sufficiently stable and that the agreement between the measurement and the calculated prediction is good. (author)

Ky. Road-Tests Common Core

Science.gov (United States)

Ujifusa, Andrew

2012-01-01

Results from new state tests in Kentucky--the first in the nation explicitly tied to the Common Core State Standards--show that the share of students scoring "proficient" or better in reading and math dropped by roughly a third or more in both elementary and middle school the first year the tests were given. Kentucky in 2010 was the…

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

Test report for core drilling ignitability testing

International Nuclear Information System (INIS)

Witwer, K.S.

1996-01-01

Testing was carried out with the cooperation of Westinghouse Hanford Company and the United States Bureau of Mines at the Pittsburgh Research Center in Pennsylvania under the Memorandum of Agreement 14- 09-0050-3666. Several core drilling equipment items, specifically those which can come in contact with flammable gasses while drilling into some waste tanks, were tested under conditions similar to actual field sampling conditions. Rotary drilling against steel and rock as well as drop testing of several different pieces of equipment in a flammable gas environment were the specific items addressed. The test items completed either caused no ignition of the gas mixture, or, after having hardware changes or drilling parameters modified, produced no ignition in repeat testing

Cylindrical core reflood test facility (CCTF) and slab core reflood test facility (SCTF) for Japan Atomic Energy Research Institute (JAERI)

International Nuclear Information System (INIS)

1981-01-01

IHI has designed and constructed the CCTF at JAERI to be used in the safety analysis research on the loss of coolant accident in a PWR plant. This test facility is planned so that reflood phenomenon in the PWR plant (a phenomenon is that the bared and overheated core is reflooded by the emergency core cooling system when the coolant loss accident occurred) is simulated under various test conditions. The CCTF is the largest-scale test plant in the world, composed of approximately 2000 simulated fuel rods (electric heaters), 1 simulated pressure vessel, 4 primary cooling loops, 2 simulated steam generators, emergency core cooling system, and so on. The test conditions are controlled, and the test steps are sequentially progressed by the computing system, and test data are collected by the data acquisition system. Furthermore, IHI is now designing and constructing the SCTF in accordance with the JAERI research plan. The SCTF is similar to the CCTF in scale. Main feature of the SCTF is the form of the simulated core and the simulated pressure vessel, which is of slab construction to be representative of the radial section of the PWR reactor. Reliable and various data for safety analysis are expected by the CCTF and the SCTF. (author)

Heat-transfer analysis of the existing HEU and proposed LEU cores of Pakistan research reactor

International Nuclear Information System (INIS)

Khan, L.A.; Nabbi, R.

1987-02-01

In connection with conversion of Pakistan Research Reactor (PARR) from the use of Highly Enriched Uranium (HEU) fuel to the use of Low Enriched Uranium (LEU) fuel, steady-state thermal hydraulic analysis of both existing HEU and proposed LEU cores has been carried out. Keeping in mind the possibility of power upgrading, the performance of proposed LEU core, under 10 MW operating conditions, has also been evaluated. Computer code HEATHYD has been used for this purpose. In order to verify the reliability of the code, IAEA benchmark 2 MW reactor was analyzed. The cooling parameters evaluated include: coolant velocity, critical velocity, pressure drop, temperature distribution in the core, heat fluxes at onset of nucleate boiling, flow instability and burnout and corresponding safety margins. From the results of the study it can be concluded that the conversion of the core to LEU fuel will result in higher safety margins, as compared to existing HEU core, mainly because the increased number of fuel plates in the proposed design will reduce the average heat flux significantly. Anyhow upgrading of the reactor power to 10 MW will need the flow rate to be adjusted between 850 to 900 m 3 /hr, to achieve reasonable safety margins, at least, comparable with the existing HEU core. (orig.)

Auxiliary Heat Exchanger Flow Distribution Test

International Nuclear Information System (INIS)

Kaufman, J.S.; Bressler, M.M.

1983-01-01

The Auxiliary Heat Exchanger Flow Distribution Test was the first part of a test program to develop a water-cooled (tube-side), compact heat exchanger for removing heat from the circulating gas in a high-temperature gas-cooled reactor (HTGR). Measurements of velocity and pressure were made with various shell side inlet and outlet configurations. A flow configuration was developed which provides acceptable velocity distribution throughout the heat exchanger without adding excessive pressure drop

Experiment data report for semiscale Mod-1 Test S-01-5 (isothermal blowdown with core resistance simulator)

International Nuclear Information System (INIS)

Zender, S.N.; Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

1975-04-01

Recorded test data are presented for Test S-01-5 of the semiscale Mod-1 isothermal blowdown test series. Test S-01-5 is one of several semiscale Mod-1 experiments which are counterparts of the LOFT nonnuclear experiments. System hardware is representative of LOFT with the design based on volumetric scaling methods and with initial conditions duplicating those identified for LOFT nonnuclear tests. Test S-01-5 was conducted with the secondary side of the steam generator pressurized with nitrogen gas in order to effectively eliminate heat transfer from the steam generator during blowdown and thereby to investigate the effect on overall system behavior of heat transfer from the steam generator. An orificed structure was used in the pressure vessel to simulate the LOFT core simulator. The test was initiated at isothermal conditions of 2270 psig and 540 0 F by a simulated offset shear of the cold leg broken loop piping. During system depressurization, coolant was injected into the cold leg of the operating loop to simulate emergency core cooling (ECC). Following the blowdown portion of the test, coolant spray was introduced into the pressure suppression tank to determine the response of the pressure suppression system. The uninterpreted data from Test S-01-5 and the reference material needed for future data analysis and test results reporting activities are presented. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to assure that they are reasonable and consistent. (U.S.)

Studies of Deteriorated Heat Transfer in Prismatic Cores Stemming from Irradiation-Induced Geometry Distortion

International Nuclear Information System (INIS)

Williams, Brian G.; Schultz, Richard R.; McEligot, Don M.; McCreery, Glenn

2015-01-01

A reference design for the Next Generation Nuclear Plant (NGNP) is to use General Atomics Modular High Temperature Gas-cooled Reactor (MHTGR). For such a configuration in normal operation, the helium coolant flow proceeds from the upper plenum to the lower plenum principally through the core coolant channels and the interstitial gaps (bypass flow) that separate the prismatic blocks from one another. Only the core prismatic blocks have coolant channels. The interstitial gaps are present throughout the core, the inner reflector region, and the out reflector region. The bypass flows in a prismatic gas-cooled reactor (GCR) are of potential concern because they reduce the desired flow rates in the coolant channels and, thereby, can increase outlet gas temperatures and maximum fuel temperatures. Consequently, it is appropriate to account for bypass flows in reactor thermal gas dynamic analyses. The objectives of this project include the following: fundamentally understand bypass flow and heat transfer at scaled, undistorted conditions and with geometry distortions; develop improved estimates of associated loss coefficients, surface friction and heat transfer for systems and network codes; and obtain related data for validation of CFD (computational fluid dynamic) or system (e.g., RELAP5) codes which can be employed in predictions for a GCR for normal power, reduced power, and residual heat removal operations.

Studies of Deteriorated Heat Transfer in Prismatic Cores Stemming from Irradiation-Induced Geometry Distortion

Energy Technology Data Exchange (ETDEWEB)

Williams, Brian G. [Idaho State Univ., Pocatello, ID (United States); Schultz, Richard R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); McEligot, Don M. [Univ. of Idaho, Moscow, ID (United States); McCreery, Glenn [Battelle Energy Alliance, LLC, Idaho Falls, ID (United States)

2015-08-31

A reference design for the Next Generation Nuclear Plant (NGNP) is to use General Atomics Modular High Temperature Gas-cooled Reactor (MHTGR). For such a configuration in normal operation, the helium coolant flow proceeds from the upper plenum to the lower plenum principally through the core coolant channels and the interstitial gaps (bypass flow) that separate the prismatic blocks from one another. Only the core prismatic blocks have coolant channels. The interstitial gaps are present throughout the core, the inner reflector region, and the out reflector region. The bypass flows in a prismatic gas-cooled reactor (GCR) are of potential concern because they reduce the desired flow rates in the coolant channels and, thereby, can increase outlet gas temperatures and maximum fuel temperatures. Consequently, it is appropriate to account for bypass flows in reactor thermal gas dynamic analyses. The objectives of this project include the following: fundamentally understand bypass flow and heat transfer at scaled, undistorted conditions and with geometry distortions; develop improved estimates of associated loss coefficients, surface friction and heat transfer for systems and network codes; and obtain related data for validation of CFD (computational fluid dynamic) or system (e.g., RELAP5) codes which can be employed in predictions for a GCR for normal power, reduced power, and residual heat removal operations.

Experimental and numerical investigations of high temperature gas heat transfer and flow in a VHTR reactor core

Science.gov (United States)

Valentin Rodriguez, Francisco Ivan

High pressure/high temperature forced and natural convection experiments have been conducted in support of the development of a Very High Temperature Reactor (VHTR) with a prismatic core. VHTRs are designed with the capability to withstand accidents by preventing nuclear fuel meltdown, using passive safety mechanisms; a product of advanced reactor designs including the implementation of inert gases like helium as coolants. The present experiments utilize a high temperature/high pressure gas flow test facility constructed for forced and natural circulation experiments. This work examines fundamental aspects of high temperature gas heat transfer applied to VHTR operational and accident scenarios. Two different types of experiments, forced convection and natural circulation, were conducted under high pressure and high temperature conditions using three different gases: air, nitrogen and helium. The experimental data were analyzed to obtain heat transfer coefficient data in the form of Nusselt numbers as a function of Reynolds, Grashof and Prandtl numbers. This work also examines the flow laminarization phenomenon (turbulent flows displaying much lower heat transfer parameters than expected due to intense heating conditions) in detail for a full range of Reynolds numbers including: laminar, transition and turbulent flows under forced convection and its impact on heat transfer. This phenomenon could give rise to deterioration in convection heat transfer and occurrence of hot spots in the reactor core. Forced and mixed convection data analyzed indicated the occurrence of flow laminarization phenomenon due to the buoyancy and acceleration effects induced by strong heating. Turbulence parameters were also measured using a hot wire anemometer in forced convection experiments to confirm the existence of the flow laminarization phenomenon. In particular, these results demonstrated the influence of pressure on delayed transition between laminar and turbulent flow. The heat

Development of whole core thermal-hydraulic analysis program ACT. 3. Coupling core module with primary heat transport system module

International Nuclear Information System (INIS)

Ohtaka, Masahiko; Ohshima, Hiroyuki

1998-10-01

A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including inter-wrapper flow under various reactor operation conditions. In this work, the core module as a main part of the ACT developed last year, which simulates thermal-hydraulics in the subassemblies and the inter-subassembly gaps, was coupled with an one dimensional plant system thermal-hydraulic analysis code LEDHER to simulate transients in the primary heat transport system and to give appropriate boundary conditions to the core model. The effective algorithm to couple these two calculation modules was developed, which required minimum modification of them. In order to couple these two calculation modules on the computing system, parallel computing technique using PVM (Parallel Virtual Machine) programming environment was applied. The code system was applied to analyze an out-of-pile sodium experiment simulating core with 7 subassemblies under transient condition for code verification. It was confirmed that the analytical results show a similar tendency of experimental results. (author)

Results of fission product release from intermediate-scale MCCI [molten core-concrete interaction] tests

International Nuclear Information System (INIS)

Spencer, B.W.; Thompson, D.H.; Fink, J.K.; Gunther, W.H.; Sehgal, B.R.

1988-01-01

A program of reactor-material molten core-concrete interaction (MCCI) tests and related analyses are under way at Argonne National Laboratory under sponsorship of the Electric Power Research Institute (EPRI). The particular objective of these tests is to provide data pertaining to the release of nonvolatile fission products such as La, Ba, and Sr, plus other aerosol materials, from the coupled thermal-hydraulic and chemical processes of the MCCI. The first stages of the program involving small and intermediate-scale tests have been completed. Three small-scale tests (/approximately/5 kg corium) and nine intermediate-scale tests (/approximately/30 kg corium) were performed between September 1985 and September 1987. Real reactor materials were used in these tests. Sustained internal heat generation at nominally 1 kW per kg of melt was provided by direct electrical heating of the corium mixture. MCCI tests were performed with both fully and partially oxidized corium mixtures that contained a variety of nonradioactive materials such as La 2 O 3 , BaO, and SrO to represent fission products. Both limestone/common sand and basaltic concrete basemats were used. The system was instrumented for characterization of the thermal hydraulic, chemical, gas release, and aerosol release processes

Space Launch System Base Heating Test: Sub-Scale Rocket Engine/Motor Design, Development and Performance Analysis

Science.gov (United States)

Mehta, Manish; Seaford, Mark; Kovarik, Brian; Dufrene, Aaron; Solly, Nathan; Kirchner, Robert; Engel, Carl D.

2014-01-01

The Space Launch System (SLS) base heating test is broken down into two test programs: (1) Pathfinder and (2) Main Test. The Pathfinder Test Program focuses on the design, development, hot-fire test and performance analyses of the 2% sub-scale SLS core-stage and booster element propulsion systems. The core-stage propulsion system is composed of four gaseous oxygen/hydrogen RS-25D model engines and the booster element is composed of two aluminum-based model solid rocket motors (SRMs). The first section of the paper discusses the motivation and test facility specifications for the test program. The second section briefly investigates the internal flow path of the design. The third section briefly shows the performance of the model RS-25D engines and SRMs for the conducted short duration hot-fire tests. Good agreement is observed based on design prediction analysis and test data. This program is a challenging research and development effort that has not been attempted in 40+ years for a NASA vehicle.

Construction of In-core Structure Test Section in HENDEL, (1)

International Nuclear Information System (INIS)

Kunitomi, Kazuhiko; Inagaki, Yoshiyuki; Ioka, Ikuo; Kondoh, Yasuo; Nekoya, Shinichi; Miyamoto, Yoshiaki; Akisada, Toshihiro; Yamaguchi, Shigeru.

1988-01-01

An In-core Structure Test Section (T 2 ) in Helium Engineering Demonstration Loop (HENDEL) simulates a part of the core bottom structure with the same scale as that of a high temperature engineering test reactor (HTTR) designed in Japan Atomic Energy Research Institute. The design and construction of T 2 test section were started in March 1983, and completed in June 1986. The main objectives of the T 2 test section are to verify thermal-hydraulic performance and integrity of the core bottom structure. The report describes the general outline of T 2 test section, and experience gained from construction and preliminary test with regard to the simulated core bottom structure. (author)

Multi-bundle sodium experiments for thermohydraulics in core subassemblies during natural circulation decay heat removal operation

International Nuclear Information System (INIS)

Kamide, H.; Ieda, Y.; Toda, S.; Isozaki, T.; Sugawara, S.

1993-01-01

Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor core during natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)

Natural convection and radiation heat transfer in a vertical porous layer with a hexagonal honeycomb core. 2nd Report. Experiment on heat transfer; Honeycomb core de shikirareta enchoku takoshitsu sonai no shizen tairyu - fukusha fukugo netsu dentatsu. 2. Dennetsu jikken

Energy Technology Data Exchange (ETDEWEB)

Yamaguchi, Y; Asako, Y [Tokyo Metropolitan Univ., Tokyo (Japan). Faculty of Technology

1997-06-25

The combined natural convection and radiation heat transfer characteristics in a vertical porous layer with a hexagonal honeycomb core were investigate experimentally. The temperature distributions on the honeycomb core wall and the combined heat transfer rates through the porous layer were measured. The measurements of the heat transfer were accomplished using the guarded hot plate (GHP) method. The honeycomb core wall was made of paper and large mesh foamed resins were inserted into the honeycomb enclosures. The measurements were performed while varying the radiation parameters between 0.5 to 0.65, varying the temperature ratios between 0.01 to 0.1 and varying the Darcy-Rayleigh numbers between 5 to 80, and for a fixed aspect ratio of H/L=1. The experimental results for Nusselt numbers agreed well with our available numerical results. 9 refs., 8 figs.

The Yucca Mountain Project prototype air-coring test, U12g tunnel, Nevada test site

International Nuclear Information System (INIS)

Ray, J.M.; Newsom, J.C.

1994-12-01

The Prototype Air-Coring Test was conducted at the Nevada Test Site (NTS) G-Tunnel facility to evaluate standard coring techniques, modified slightly for air circulation, for use in testing at a prospective nuclear waste repository at Yucca Mountain, Nevada. Air-coring technology allows sampling of subsurface lithology with minimal perturbation to ambient characteristic such as that required for exploratory holes near aquifers, environmental applications, and site characterization work. Two horizontal holes were cored, one 50 ft long and the other 150 ft long, in densely welded fractured tuff to simulate the difficult drilling conditions anticipated at Yucca Mountain. Drilling data from seven holes on three other prototype tests in nonwelded tuff were also collected for comparison. The test was used to establish preliminary standards of performance for drilling and dust collection equipment and to assess procedural efficiencies. The Longyear-38 drill achieved 97% recovery for HQ-size core (-2.5 in.), and the Atlas Copco dust collector (DCT-90) captured 1500 lb of fugitive dust in a mine environment with only minor modifications. Average hole production rates were 6-8 ft per 6-h shift in welded tuff and almost 20 ft per shift on deeper holes in nonwelded tuff. Lexan liners were successfully used to encapsulate core samples during the coring process and protect core properties effectively. The Prototype Air-Coring Test demonstrated that horizontal air coring in fractured welded tuff (to at least 150 ft) can be safely accomplished by proper selection, integration, and minor modification of standard drilling equipment, using appropriate procedures and engineering controls. The test also indicated that rig logistics, equipment, and methods need improvement before attempting a large-scale dry drilling program at Yucca Mountain

Critical heat flux evaluation

International Nuclear Information System (INIS)

Banner, D.

1995-01-01

Critical heat flux (CHF) is of importance for nuclear safety and represents the major limiting factors for reactor cores. Critical heat flux is caused by a sharp reduction in the heat transfer coefficient located at the outer surface of fuel rods. Safety requires that this phenomenon also called the boiling crisis should be precluded under nominal or incidental conditions (Class I and II events). CHF evaluation in reactor cores is basically a two-step approach. Fuel assemblies are first tested in experimental loops in order to determine CHF limits under various flow conditions. Then, core thermal-hydraulic calculations are performed for safety evaluation. The paper will go into more details about the boiling crisis in order to pinpoint complexity and lack of fundamental understanding in many areas. Experimental test sections needed to collect data over wide thermal-hydraulic and geometric ranges are described CHF safety margin evaluation in reactors cores is discussed by presenting how uncertainties are mentioned. From basic considerations to current concerns, the following topics are discussed; knowledge of the boiling crisis, CHF predictors, and advances thermal-hydraulic codes. (authors). 15 refs., 4 figs

Effect of heat transfer in the fog region during core reflooding

International Nuclear Information System (INIS)

Rouai, N. M.; El-sawy, H. M.

1993-01-01

Core reflooding following a loss of coolant accident (LOCA) in a pressurized water reactor (PWR) received considerable attention during the past thirty years. In this paper a one dimensional model is used to study the effect of the heat transfer in the fog region ahead of the wet front reflooding rate of a cylindrical fuel element following a LOCA in a PWR. The heat conduction equation in the cladding is solved in coordinate system moving with the wet front under a variety of condition to investigate the effects of such parameters as the initial cladding surface temperature, the decay heat generation rate in the fuel and the mode of heat transfer prevailing. The cladding surface is divided into three axial regions according to the mechanism of heat transfer, namely, a boiling region behind the wet front, a fog region ahead of the wet front and a dry region further downstream of the wet front. The effect of changing the heat transfer coefficient in the fog region on the rewetting rate and on the fog length is investigated. The results of this simple model show that increasing the heat transfer in the fog region increases the rewetting velocity and consequently decreases the fog length. The results are in general agreement with a more accurate two-dimensional model and experimental data. (author)

Natural convection as the way of heat removal from fast reactor core at cooldown regimes

International Nuclear Information System (INIS)

Zhukov, A.V.; Kuzina, J.A.; Uhov, V.A.; Sorokin, G.A.

2000-01-01

The problems of thermohydraulics in fast reactors at cooldown regimes at heat removal by natural convection are considered The results of experiments and calculations obtained in various countries in this area are presented. The special attention is given to heat removal through inter-assembly space in the core and also to problems of thermohydraulics in the upper plenum. (author)

Arc Heated Scramjet Test Facility

Data.gov (United States)

Federal Laboratory Consortium — The Arc Heated Scramjet Test Facility is an arc heated facility which simulates the true enthalpy of flight over the Mach number range of about 4.7 to 8 for free-jet...

Space Launch System Base Heating Test: Environments and Base Flow Physics

Science.gov (United States)

Mehta, Manish; Knox, Kyle S.; Seaford, C. Mark; Dufrene, Aaron T.

2016-01-01

The NASA Space Launch System (SLS) vehicle is composed of four RS-25 liquid oxygen- hydrogen rocket engines in the core-stage and two 5-segment solid rocket boosters and as a result six hot supersonic plumes interact within the aft section of the vehicle during ight. Due to the complex nature of rocket plume-induced ows within the launch vehicle base during ascent and a new vehicle con guration, sub-scale wind tunnel testing is required to reduce SLS base convective environment uncertainty and design risk levels. This hot- re test program was conducted at the CUBRC Large Energy National Shock (LENS) II short-duration test facility to simulate ight from altitudes of 50 kft to 210 kft. The test program is a challenging and innovative e ort that has not been attempted in 40+ years for a NASA vehicle. This presentation discusses the various trends of base convective heat ux and pressure as a function of altitude at various locations within the core-stage and booster base regions of the two-percent SLS wind tunnel model. In-depth understanding of the base ow physics is presented using the test data, infrared high-speed imaging and theory. The normalized test design environments are compared to various NASA semi- empirical numerical models to determine exceedance and conservatism of the ight scaled test-derived base design environments. Brief discussion of thermal impact to the launch vehicle base components is also presented.

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.

Influence of heat treatment on hole transfer dynamics in core-shell quantum dot/organic hole conductor hybrid films

Science.gov (United States)

Sun, Mingye; Zheng, Youjin; Zhang, Lei; Zhao, Liping; Zhang, Bing

2017-08-01

The influence of heat treatment on hole transfer (HT) processes from the CdSe/ZnS and CdSe/CdS/ZnS quantum dots (QDs) to 4,4‧,4″-Tris(carbazol-9-yl)-triphenylamine (TCTA) in QD/TCTA hybrid films has been researched with time-resolved photoluminescence (PL) spectroscopy. The PL dynamic results demonstrated a heat-treatment-temperature-dependent HT process from the core-shell CdSe QDs to TCTA. The HT rates and efficiencies can be effectively increased due to reduced distance between core-shell CdSe QDs and TCTA after heat treatment. The CdS shell exhibited a more obvious effect on HT from the core-shell CdSe QDs to TCTA than on electron transfer to TiO2, due to higher barrier for holes to tunnel through CdS shell and larger effective mass of holes in CdS than electrons. These results indicate that heat treatment would be an effective means to further optimize solid-state QD sensitized solar cells and rational design of CdS shell is significant.

Using the Oscillation Test Method to test for Delay Faults in Embedded Cores

NARCIS (Netherlands)

Vermaak, H.J.; Kerkhoff, Hans G.

2004-01-01

Continual advances In the manufacturing processes of integrated circuits provide designers the ability to create more complex and denser architectures and increased functionality on a single chip. The increased usage of embedded cores necessitates a core-based test strategy in which cores are also

Heat flow at the proposed Appalachian Ultradeep Core Hole (ADCOH) Site: Tectonic implications

Science.gov (United States)

Costain, John K.; Decker, Edward R.

The heat flow in northwestern South Carolina at the Appalachian Ultradeep Core Hole (ADCOH) site area is approximately 55 mW/m². This data supplements other data to the east in the Piedmont and Atlantic Coastal Plain provinces where heat flows > 55 mW/m² are characteristic of post- and late-synmetamorphic granitoids. Piedmont heat flow and heat generation data for granites, metagranites, and one Slate Belt site, in a zone approximately parallel to major structural Appalachian trends, define a linear relation. Tectonic truncation of heat-producing crust at a depth of about 8 km (a depth equal to the slope of the heat flow-heat production line) is proposed to explain the linear relation. Using the value of reduced heat flow estimated from this empirical relation, and assuming thicknesses of heat-producing crust defined by new ADCOH seismic data, the heat flow and heat production at the ADCOH site are consistent with a depth to the base of the Inner Piedmont crystalline allochthon of about 5.5 km. Seismic data at the ADCOH site confirm that the Inner Piedmont is tectonically truncated at about 5.5 km by the Blue Ridge master decollement. Temperatures at 10 km at the ADCOH site are predicted to be less than 200 °C.

Evaluation report on SCTF Core-III Test S3-22

International Nuclear Information System (INIS)

Okubo, Tsutomu; Iguchi, Tadashi; Iwamura, Takamichi; Akimoto, Hajime; Ohnuki, Akira; Abe, Yutaka; Murao, Yoshio; Adachi, Hiromichi.

1991-07-01

Two tests (Tests S3-20 and S3-22) were conducted with JAERI's Slab Core Test Facility (SCTF) Core-III in order to investigate water break-through and core cooling behaviors under the intermittent ECC water delivery from the hot legs to one location in the upper plenum and the alternate ECC water delivery to two locations in the upper plenum during reflooding, respectively. This report presents an analysis on Test S3-22 (the alternate case). Subcooled ECC water was injected alternately just above the upper core support plate above Bundles 7 and 8 and Bundles 3 and 4. The total injection rate from both injection ports was the same as that in SCTF Test S3-20 and Test S3-13. Analyzing the test data together with those of Tests S3-13 and S3-20 the following has been found: (1) Alternate break-through occurred immediately corresponding to the alternate ECC water injection except for one period, during which no break-through was observed. However, there observed a difference in break-through behavior that break-through was strong above the low power region, whereas weak above the high power region. (2) Although its break-through behavior was different, nearly the same core cooling as in the continuous or intermittent ECC water delivery case was observed except for the period around quench. (3) Around quench time, degraded core cooling comparing to the continuous or intermittent ECC water delivery case was observed. That is, quench time at the midplane level of the present test was 35 s later than in the continuous case. This is considered to result from decrease in core water inventory caused by water sealing at the cross-over leg. (J.P.N.)

The Influence of Heat Flux Boundary Heterogeneity on Heat Transport in Earth's Core

Science.gov (United States)

Davies, C. J.; Mound, J. E.

2017-12-01

Rotating convection in planetary systems can be subjected to large lateral variations in heat flux from above; for example, due to the interaction between the metallic cores of terrestrial planets and their overlying silicate mantles. The boundary anomalies can significantly reorganise the pattern of convection and influence global diagnostics such as the Nusselt number. We have conducted a suite of numerical simulations of rotating convection in a spherical shell geometry comparing convection with homogeneous boundary conditions to that with two patterns of heat flux variation at the outer boundary: one hemispheric pattern, and one derived from seismic tomographic imaging of Earth's lower mantle. We consider Ekman numbers down to 10-6 and flux-based Rayleigh numbers up to 800 times critical. The heterogeneous boundary conditions tend to increase the Nusselt number relative to the equivalent homogeneous case by altering both the flow and temperature fields, particularly near the top of the convecting region. The enhancement in Nusselt number tends to increase as the amplitude and wavelength of the boundary heterogeneity is increased and as the system becomes more supercritical. In our suite of models, the increase in Nusselt number can be as large as 25%. The slope of the Nusselt-Rayleigh scaling also changes when boundary heterogeneity is included, which has implications when extrapolating to planetary conditions. Additionally, regions of effective thermal stratification can develop when strongly heterogeneous heat flux conditions are applied at the outer boundary.

Heated wire humidification circuit attenuates the decrease of core temperature during general anesthesia in patients undergoing arthroscopic hip surgery.

Science.gov (United States)

Park, Sooyong; Yoon, Seok-Hwa; Youn, Ann Misun; Song, Seung Hyun; Hwang, Ja Gyung

2017-12-01

Intraoperative hypothermia is common in patients undergoing general anesthesia during arthroscopic hip surgery. In the present study, we assessed the effect of heating and humidifying the airway with a heated wire humidification circuit (HHC) to attenuate the decrease of core temperature and prevent hypothermia in patients undergoing arthroscopic hip surgery under general anesthesia. Fifty-six patients scheduled for arthroscopic hip surgery were randomly assigned to either a control group using a breathing circuit connected with a heat and moisture exchanger (HME) (n = 28) or an HHC group using a heated wire humidification circuit (n = 28). The decrease in core temperature was measured from anesthetic induction and every 15 minutes thereafter using an esophageal stethoscope. Decrease in core temperature from anesthetic induction to 120 minutes after induction was lower in the HHC group (-0.60 ± 0.27℃) compared to the control group (-0.86 ± 0.29℃) (P = 0.001). However, there was no statistically significant difference in the incidence of intraoperative hypothermia or the incidence of shivering in the postanesthetic care unit. The use of HHC may be considered as a method to attenuate intraoperative decrease in core temperature during arthroscopic hip surgery performed under general anesthesia and exceeding 2 hours in duration.

Fast Flux Test Facility core system

International Nuclear Information System (INIS)

Ethridge, J.L.; Baker, R.B.; Leggett, R.D.; Pitner, A.L.; Waltar, A.E.

1990-11-01

A review of Liquid Metal Reactor (LMR) core system accomplishments provides an excellent road map through the maze of issues that faced reactor designers 10 years ago. At that time relatively large uncertainties were associated with fuel pin and fuel assembly performance, irradiation of structural materials, and performance of absorber assemblies. The extensive core systems irradiation program at the US Department of Energy's Fast Flux Test Facility (FFTF) has addressed each of these principal issues. As a result of the progress made, the attention of long-range LMR planners and designers can shift away from improving core systems and focus on reducing capital costs to ensure the LMR can compete economically in the 21st century with other nuclear reactor concepts. 3 refs., 6 figs., 1 tab

Analysis of SCTF/CCTF counterpart test results

International Nuclear Information System (INIS)

Okubo, Tsutomu; Sobajima, Makoto; Iwamura, Takamichi; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi; Murao, Yoshio

1990-06-01

Slab Core Test Facility (SCTF) and Cylindrical Core Test Facility (CCTF) are large scale experimental facilities of Japan Atomic Energy Research Institute (JAERI) for the investigation of reflooding behavior during a postulated loss-of-coolant accident (LOCA) in PWRs. Although the flow area scaling ratios of both facilities to a 1,000 MWe class PWR are the same and 1/21.4, the SCTF has the same core width as the radius of the reference PWR while the CCTF has a 1/4.5 times shorter core radius. Therefore, a few SCTF/CCTF counterpart tests were conducted in order to investigate the difference in core reflooding behavior between in the SCTF and CCTF tests as well as the effect of core radial length on core two-dimensional thermo-hydrodynamic behavior. This report present the test results and an analysis on them. Major results obtained are: (1) Taking account of the differences in test conditions and facility design, core reflooding behavior is considered to be similar between the SCTF and the CCTF test. Main difference of the facility design is in the effective core flow area and this is considered to result in the difference in core water accumulation behavior. (2) The effect of core radial length on core two-dimensional thermo-hydrodynamic behavior has been observed to be significant and heat transfer enhancement or degradation in radial direction is more significant for the longer radius core. (3) In addition, where the core power varies significantly in the radial direction, significant heat transfer enhancement has been observed in the higher power bundle during the LPCI period. Also, in the peripheral region, heat transfer degradation has been observed more significantly in the outer bundle even they have the same bundle power. (4) Magnitude of these heat transfer enhancement or degradation was larger at the higher elevation than the midplane level in the SCTF test, whereas smaller in the CCTF test. (author)

DARK MATTER HEATING AND EARLY CORE FORMATION IN DWARF GALAXIES

International Nuclear Information System (INIS)

Madau, Piero; Shen, Sijing; Governato, Fabio

2014-01-01

We present more results from a fully cosmological ΛCDM simulation of a group of isolated dwarf galaxies that has been shown to reproduce the observed stellar mass and cold gas content, resolved star formation histories, and metallicities of dwarfs in the Local Volume. Here we investigate the energetics and timetable of the cusp-core transformation. As suggested by previous work, supernova-driven gas outflows remove dark matter (DM) cusps and create kiloparsec-size cores in all systems having a stellar mass M * > 10 6 M ☉ . The D M core mass removal efficiency — dark mass ejected per unit stellar mass—ranges today from a few to a dozen, and increases with decreasing host mass. Because dwarfs form the bulk of their stars prior to redshift 1 and the amount of work required for DM heating and core formation scales approximately as M vir 5/3 , the unbinding of the DM cusp starts early and the formation of cored profiles is not as energetically onerous as previously claimed. DM particles in the cusp typically migrate to 2-3 core radii after absorbing a few percent of the energy released by supernovae. The present-day slopes of the inner DM mass profiles, Γ ≡ dlog M/dlog R ≅ 2.5-3, of the simulated ''Bashful'' and ''Doc'' dwarfs are similar to those measured in the luminous Fornax and Sculptor dwarf spheroidals. None of the simulated galaxies has a circular velocity profile exceeding 20 km s –1 in the inner 1 kpc, implying that supernova feedback is key to solve the ''too-big-to-fail'' problem for Milky Way subhalos

Core Cutting Test with Vertical Rock Cutting Rig (VRCR)

Science.gov (United States)

Yasar, Serdar; Osman Yilmaz, Ali

2017-12-01

Roadheaders are frequently used machines in mining and tunnelling, and performance prediction of roadheaders is important for project economics and stability. Several methods were proposed so far for this purpose and, rock cutting tests are the best choice. Rock cutting tests are generally divided into two groups which are namely, full scale rock cutting tests and small scale rock cutting tests. These two tests have some superiorities and deficiencies over themselves. However, in many cases, where rock sampling becomes problematic, small scale rock cutting test (core cutting test) is preferred for performance prediction, since small block samples and core samples can be conducted to rock cutting testing. Common problem for rock cutting tests are that they can be found in very limited research centres. In this study, a new mobile rock cutting testing equipment, vertical rock cutting rig (VRCR) was introduced. Standard testing procedure was conducted on seven rock samples which were the part of a former study on cutting rocks with another small scale rock cutting test. Results showed that core cutting test can be realized successfully with VRCR with the validation of paired samples t-test.

Plutonium cores of zenith

Energy Technology Data Exchange (ETDEWEB)

Barclay, F R; Cameron, I R; Drageset, A; Freemantle, R G; Wilson, D J

1965-03-15

The report describes a series of experiments carried out with plutonium fuel in the heated zero power reactor ZENITH, with the aim of testing current theoretical methods, with particular reference to excess reactivity, temperature coefficients, differential spectrum and reaction rate distributions. Two cores of widely different fissile/moderator atom ratios were loaded in order to test the theory under significantly varied spectrum conditions.

Tank 101-SY Window E core sample: Interpretation of results

International Nuclear Information System (INIS)

Reynolds, D.A.

1993-02-01

A full depth core sample was taken for tank 241-SY-101 in December 1991 during a time period called ''Window E.'' This was the second full depth core sample from this tank during the year. The core had two major portions that are known as the convective zone and the nonconvective zone. A crust was on the top of tank but as poorly sampled. The analysis of the Window E core sample stressed segment composite chemical analysis instead of segment by segment as in Window C. Adiabatic calorimetry on samples from both cores showed a slow self heating reaction above 150 degrees C on dried samples. The exothermic events were milder than similar synthetic samples. The chemical and physical properties complemented the information from Window C. The Window E material from the convective zone was more viscous than the Window C convective zone material. The nonconvective zone viscosities were similar for both cores. Heating and dilution tests were made to test mitigation concepts

Tank 101-SY Window E core sample: Interpretation of results

Energy Technology Data Exchange (ETDEWEB)

Reynolds, D.A.

1993-02-01

A full depth core sample was taken for tank 241-SY-101 in December 1991 during a time period called ``Window E.`` This was the second full depth core sample from this tank during the year. The core had two major portions that are known as the convective zone and the nonconvective zone. A crust was on the top of tank but as poorly sampled. The analysis of the Window E core sample stressed segment composite chemical analysis instead of segment by segment as in Window C. Adiabatic calorimetry on samples from both cores showed a slow self heating reaction above 150{degrees}C on dried samples. The exothermic events were milder than similar synthetic samples. The chemical and physical properties complemented the information from Window C. The Window E material from the convective zone was more viscous than the Window C convective zone material. The nonconvective zone viscosities were similar for both cores. Heating and dilution tests were made to test mitigation concepts.

Evaluation report on CCTF Core-I reflood tests C1-5 (Run 14), C1-7 (Run 16) and C1-14 (Run 23)

International Nuclear Information System (INIS)

Sugimoto, Jun; Muurao, Yoshio

1983-02-01

The present report describes the effects of the initial clad temperature on the reflood phenomena observed in the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute. The evaluation is based on the data of tests C1-5, C1-7 and C1-14 of the CCTF-Core I test series. Nominal initial peak clad temperatures in these tests are 600 0 C, 700 0 C and 800 0 C, respectively. With the higher initial clad temperature, the higher loop mass flow rate and the lower water accumulation in the core and the upper plenum were obtained in an early reflood transient. However, the core inlet flow conditions, which is sensitive to the core cooling, were not much affected by the higher initial clad temperature. The slower quench front propagation was observed with the higher initial clad temperature. However, the heat transfer coefficient was almost identical with each other before the turnaround time, which resulted in the lower temperature rise with the highest initial clad temperature. This qualitatively agreed with the results of the forced feed FLECHT experiment. (author)

Niobium 1 percent zirconium/potassium and titanium/potassium life-test heat pipe design and testing

Science.gov (United States)

Sena, J. Tom; Merrigan, Michael A.

Experimental lifetime performance studies currently in progress use Niobium 1 percent Zirconium (Nb-1Zr) and Titanium (Ti) heat pipes with potassium (K) as the working fluid. A heat pipe life test matrix was developed for testing the heat pipes. Because the corrosion rates in alkali metal heat pipes are affected by temperature and working fluid evaporation flux, the variable parameters of the experimental matrix are established as steady operating temperature and input heat flux density. Total impurity inventory is a factor in corrosion rate so impurity levels are being evaluated in the heat pipe materials before and after testing. Eight Nb-1Zr/K heat pipes were designed, fabricated, and tested. Two of the heat pipes have completed testing whereas the other six are currently in test. These are gravity assist heat pipes operating in a reflux mode. The heat pipes are tested by sets, one set of two and two sets of three heat pipes. Three Ti/K heat pipes are also in life test. These heat pipes are tested as a set in a horizontal position in a capillary pumped annular flow mode. Each of the heat pipes is encapsulated in a quartz vacuum container with a water calorimeter over the vacuum container for power throughput measurements. Thermocouples are attached to the heat pipes for measuring temperature. Heat input to the heat pipes is via an RF coil. The heat pipes are operating at between 800 and 900 K, with heat input fluxes of 13.8 to 30 W/sq cm. Of the Nb-1Zr/K heat pipes, two of the heat pipes have been in operation for 14,000 hours, three over 10,000 hours, and three over 7,000 hours. The Ti/K heat pipes have been in operation for 1,266 hours.

Shivering heat production and core cooling during head-in and head-out immersion in 17 degrees C water.

Science.gov (United States)

Pretorius, Thea; Cahill, Farrell; Kocay, Sheila; Giesbrecht, Gordon G

2008-05-01

Many cold-water scenarios cause the head to be partially or fully immersed (e.g., ship wreck survival, scuba diving, cold-water adventure swim racing, cold-water drowning, etc.). However, the specific effects of head cold exposure are minimally understood. This study isolated the effect of whole-head submersion in cold water on surface heat loss and body core cooling when the protective shivering mechanism was intact. Eight healthy men were studied in 17 degrees C water under four conditions: the body was either insulated or exposed, with the head either out of the water or completely submersed under the water within each insulated/exposed subcondition. Submersion of the head (7% of the body surface area) in the body-exposed condition increased total heat loss by 11% (P < 0.05). After 45 min, head-submersion increased core cooling by 343% in the body-insulated subcondition (head-out: 0.13 +/- 0.2 degree C, head-in: 0.47 +/- 0.3 degree C; P < 0.05) and by 56% in the body-exposed subcondition (head-out: 0.40 +/- 0.3 degree C and head-in: 0.73 +/- 0.6 degree C; P < 0.05). In both body-exposed and body-insulated subconditions, head submersion increased the rate of core cooling disproportionally more than the relative increase in total heat loss. This exaggerated core-cooling effect is consistent with a head cooling induced reduction of the thermal core, which could be stimulated by cooling of thermosensitive and/or trigeminal receptors in the scalp, neck, and face. These cooling effects of head submersion are not prevented by shivering heat production.

ac power control in the Core Flow Test Loop

International Nuclear Information System (INIS)

McDonald, D.W.

1980-01-01

This work represents a status report on a development effort to design an ac power controller for the Core Flow Test Loop. The Core Flow Test Loop will be an engineering test facility which will simulate the thermal environment of a gas-cooled fast-breeder reactor. The problems and limitations of using sinusoidal ac power to simulate the power generated within a nuclear reactor are addressed. The transformer-thyristor configuration chosen for the Core Flow Test Loop power supply is presented. The initial considerations, design, and analysis of a closed-loop controller prototype are detailed. The design is then analyzed for improved performance possibilities and failure modes are investigated at length. A summary of the work completed to date and a proposed outline for continued development completes the report

Test plan: Potash Core Test. WIPP experimental program borehole plugging

International Nuclear Information System (INIS)

Christensen, C.L.

1979-09-01

The Potash Core Test will utilize a WIPP emplaced plug to obtain samples of an in-situ cured plug of known mix constituents for bench scale testing. An earlier effort involved recovery at the salt horizon of Plug 217, a 17 year old plug in a potash exploration hole for bond testing, but the lack of particulars in the emplacement precluded significant determination of plug performance

Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

2009-06-15

Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

High Temperature Test Possibility at the HANARO Out-core Region through a Thermal Analysis

International Nuclear Information System (INIS)

Kang, Young-Hwan; Choi, Myung-Hwan; Cho, Man-Soon; Choo, Kee-Nam; Kim, Bong-Goo

2007-01-01

The development of an advanced reactor system such as a next generation nuclear plant and other generation IV systems require new fuels, claddings, and structural materials. To characterize the performance of these new materials, it is necessary for us to have a leading-edge technology to satisfy the specific test requirements such as the conditions of high neutron exposures and high operating temperatures. Thus, nuclear data on HANARO's vertical test holes have been gathered and reviewed to evaluate the usability of the test holes located at the out-core zone of HANARO. In 2007, neutron flux levels of the concerned test holes and the gamma heat of the specimens and two different specimen holder materials of Al and Mo at the concerned test hole were obtained to enhance the utilization of the HANARO reactor and to develop new design concepts for high temperature irradiation tests. Based on the data, a series of thermal analyses was implemented to provide a reasonable demonstration and guidance on limitations or application

Application of startup/core management code system to YGN 3 startup testing

International Nuclear Information System (INIS)

Chi, Sung Goo; Hah, Yung Joon; Doo, Jin Yong; Kim, Dae Kyum

1995-01-01

YGN 3 is the first nuclear power plant in Korea to use the fixed incore detector system for startup testing and core management. The startup/core management code system was developed from existing ABB-C-E codes and applied for YGN 3 startup testing, especially for physics and CPC(Core Protection Calculator)/COLSS (Core Operating Limit Supervisory System) related testing. The startup/core management code system consists of startup codes which include the CEBASE, CECOR, CEFAST and CEDOPS, and startup data reduction codes which include FLOWRATE, COREPERF, CALMET, and VARTAV. These codes were implemented on an HP/Apollo model 9000 series 400 workstation at the YGN 3 site and successfully applied to startup testing and core management. The startup codes made a great contribution in upgrading the reliability of test results and reducing the test period by taking and analyzing core data automatically. The data reduction code saved the manpower and time for test data reduction and decreased the chance for error in the analysis. It is expected that this code system will make similar contributions for reducing the startup testing duration of YGN 4 and UCN3,4

Core-adjacent instrumentation systems for pebble bed reactors for process heat application - state of planning

International Nuclear Information System (INIS)

Benninghofen, G.; Serafin, N.; Spillekothen, H.G.; Hecker, R.; Brixy, H.; Serpekian, T.

1982-06-01

Planning and theoretical/experimental development work for core surveillance instrumentation systems is being performed to meet requirements of pebble bed reactors for process heat application. Detailed and proved instrumentation concepts are now available for the core-adjacent instrumentation systems. The current work and the results of neutron flux measurements at high temperatures are described. Operation devices for long-term accurate gas outlet temperature measurements up to approximately 1423 deg. K will also be discussed. (author)

Simulation tests for temperature mixing in a core bottom model of the HTR-module

International Nuclear Information System (INIS)

Damm, G.; Wehrlein, R.

1992-01-01

Interatom and Siemens are developing a helium-cooled Modular High Temperature Reactor. Under nominal operating conditions temperature differences of up to 120deg C will occur in the 700deg C hot helium flow leaving the core. In addition, cold gas leakages into the hot gas header can produce even higher temperature differences in the coolant flow. At the outlet of the reactor only a very low temperature difference of maximum ± 15deg C is allowed in order to avoid damages at the heat exchanging components due to alternating thermal loads. Since it is not possible to calculate the complex flow behaviour, experimental investigations of the temperature mixing in the core bottom had to be carried out in order to guarantee the necessary reduction of temperature differences in the helium. The presented air simulation tests in a 1:2.9 scaled plexiglas model of the core bottom showed an extremely high mixing rate of the hot gas header and the hot gas duct of the reactor. The temperature mixing of the simulated coolant flow as well as the leakage flows was larger than 95%. Transfered to reactor conditions this means a temperature difference of only ± 3deg C for the main flow at a quite resonable pressure drop. For the cold gas leakages temperature differences in the hot gas up to 400deg C proved to be permissible. The results of the simulation experiments in the Aerodynamic Test Facility of Interatom permitted to design a shorter bottom reflector of the core. (orig.)

High-heat-flux testing of helium-cooled heat exchangers for fusion applications

International Nuclear Information System (INIS)

Youchison, D.L.; Izenson, M.G.; Baxi, C.B.; Rosenfeld, J.H.

1996-01-01

High-heat-flux experiments on three types of helium-cooled divertor mock-ups were performed on the 30-kW electron beam test system and its associated helium flow loop at Sandia National Laboratories. A dispersion-strengthened copper alloy (DSCu) was used in the manufacture of all the mock-ups. The first heat exchanger provides for enhanced heat transfer at relatively low flow rates and much reduced pumping requirements. The Creare sample was tested to a maximum absorbed heat flux of 5.8 MW/m 2 . The second used low pressure drops and high mass flow rates to achieve good heat removal. The GA specimen was tested to a maximum absorbed heat flux of 9 MW/m 2 while maintaining a surface temperature below 400 degree C. A second experiment resulted in a maximum absorbed heat flux of 34 MW/m 2 and surface temperatures near 533 degree C. The third specimen was a DSCu, axial flow, helium-cooled divertor mock-up filled with a porous metal wick which effectively increases the available heat transfer area. Low mass flow and high pressure drop operation at 4.0 MPa were characteristic of this divertor module. It survived a maximum absorbed heat flux of 16 MW/m 2 and reached a surface temperature of 740 degree C. Thermacore also manufactured a follow-on, dual channel porous metal-type heat exchanger, which survived a maximum absorbed heat flux of 14 MW/m 2 and reached a maximum surface temperature of 690 degree C. 11refs., 20 figs., 3 tabs

ROSA-V/LSTF vessel top head LOCA tests SB-PV-07 and SB-PV-08 with break sizes of 1.0 and 0.1% and operator recovery actions for core cooling

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

2010-02-01

A series of break size parameter tests (SB-PV-07 and SB-PV-08) were conducted at the Large Scale Test Facility (LSTF) of ROSA-V Program by simulating a vessel top small break loss-of-coolant accident (SBLOCA) at a pressurized water reactor (PWR). Typical phenomena to the vessel top break LOCA and effectiveness of operator recovery actions on core cooling were studied under an assumption of total failure of high pressure injection (HPI) system. The LSTF simulates a 4-loop 3423 MWt PWR by a full-height, full-pressure and 1/48 volume scaling two-loop system. Typical phenomena of vessel top break LOCA are clarified for the cases with break sizes of 1.0 and 0.1% cold leg break equivalent. The results from a 0.5% top break LOCA test (SB-PV-02) in the early ROSA-IV Program was referred during discussion. Operator actions of HPI recovery in the 1.0% top break test and steam generator (SG) depressurization in the 0.1% top break test were initiated when temperature at core exit thermocouple (CET) reached 623 K during core boil-off. Both operator actions resulted in immediate recovery of core cooling. Based on the obtained data, several thermal-hydraulic phenomena were discussed further such as relations between vessel top head water level and steam discharge at the break, and between coolant mass inventory transient and core heat-up and quench behavior, and CET performances to detect core heat-up under influences of three-dimensional (3D) steam flows in the core and core exit. (author)

OECD MCCI 2-D Core Concrete Interaction (CCI) tests : CCI-2 test data report-thermalhydraulic results, Rev. 0 October 15, 2004.

Energy Technology Data Exchange (ETDEWEB)

Farmer, M. T.; Lomperski, S.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

2011-05-23

The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of two long-term 2-D Core-Concrete Interaction (CCI) experiments designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-2 experiment, which was conducted on August 24, 2004. Test specifications for CCI-2 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 400 kg

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.)

Shock Compression and Melting of an Fe-Ni-Si Alloy: Implications for the Temperature Profile of the Earth's Core and the Heat Flux Across the Core-Mantle Boundary

Science.gov (United States)

Zhang, Youjun; Sekine, Toshimori; Lin, Jung-Fu; He, Hongliang; Liu, Fusheng; Zhang, Mingjian; Sato, Tomoko; Zhu, Wenjun; Yu, Yin

2018-02-01

Understanding the melting behavior and the thermal equation of state of Fe-Ni alloyed with candidate light elements at conditions of the Earth's core is critical for our knowledge of the region's thermal structure and chemical composition and the heat flow across the liquid outer core into the lowermost mantle. Here we studied the shock equation of state and melting curve of an Fe-8 wt% Ni-10 wt% Si alloy up to 250 GPa by hypervelocity impacts with direct velocity and reliable temperature measurements. Our results show that the addition of 10 wt% Si to Fe-8 wt% Ni alloy slightly depresses the melting temperature of iron by 200-300 (±200) K at the core-mantle boundary ( 136 GPa) and by 600-800 (±500) K at the inner core-outer core boundary ( 330 GPa), respectively. Our results indicate that Si has a relatively mild effect on the melting temperature of iron compared with S and O. Our thermodynamic modeling shows that Fe-5 wt% Ni alloyed with 6 wt% Si and 2 wt% S (which has a density-velocity profile that matches the outer core's seismic profile well) exhibits an adiabatic profile with temperatures of 3900 K and 5300 K at the top and bottom of the outer core, respectively. If Si is a major light element in the core, a geotherm modeled for the outer core indicates a thermal gradient of 5.8-6.8 (±1.6) K/km in the D″ region and a high heat flow of 13-19 TW across the core-mantle boundary.

Feasibility study for a postaccident heat removal facility

International Nuclear Information System (INIS)

Barts, E.W.; Apperson, C.E. Jr.; Dunwoody, W.E.; Bennett, J.G.

1978-01-01

An initial feasibility investigation for PAHRTEF, a Postaccident Heat Removal Test Facility, is presented. The facility would provide an experimental capability for PAHR experiments beyond that available in any currently existing or proposed U.S. safety test facility. The facility design presented in this report is based upon the technology developed for the ROVER nuclear rocket propulsion program. The core is a graphite-moderated, helium-cooled, epithermal core with radial reflector control. The PAHR experiments are located just below the reactor containment vessel, very near the bottom of the core. The experiments (up to 55% enriched) are driven and controlled by neutrons leaking axially from the core such that the PAHRTEF core and the experiment form a coupled reactor system. The experiment can be designed so that it is extremely unlikely that the test fuel by itself could form a critical system. The investigation indicates that adequate fission heating of large PAHR experiments could be provided at low driver core power levels. Both the reactor and the experiment handling and examination equipment can use available technology and, whenever possible, existing equipment and buildings

Analysis of excess reactivity of JOYO MK-III performance test core

International Nuclear Information System (INIS)

Maeda, Shigetaka; Yokoyama, Kenji

2003-10-01

JOYO is currently being upgraded to the high performance irradiation bed JOYO MK-III core'. The MK-III core is divided into two fuel regions with different plutonium contents. To obtain a higher neutron flux, the active core height was reduced from 55 cm to 50 cm. The reflector subassemblies were replaced by shielding subassemblies in the outer two rows. Twenty of the MK-III outer core fuel subassemblies in the performance test core were partially burned in the transition core. Four irradiation test rigs, which do not contain any fuel material, were loaded in the center of the performance test core. In order to evaluate the excess reactivity of MK-III performance test core accurately, we evaluated it by applying not only the JOYO MK-II core management code system MAGI, but also the MK-III core management code system HESTIA, the JUPITER standard analysis method and the Monte Carlo method with JFS-3-J3.2R content set. The excess reactivity evaluations obtained by the JUPITER standard analysis method were corrected to results based on transport theory with zero mesh-size in space and angle. A bias factor based on the MK-II 35th core, which sensitivity was similar to MK-III performance test core's, was also applied, except in the case where an adjusted nuclear cross-section library was used. Exact three-dimensional, pin-by-pin geometry and continuous-energy cross sections were used in the Monte Carlo calculation. The estimated error components associated with cross-sections, methods correction factors and the bias factor were combined based on Takeda's theory. Those independently calculated values agree well and range from 2.8 to 3.4%Δk/kk'. The calculation result of the MK-III core management code system HESTLA was 3.13% Δk/kk'. The estimated errors for bias method range from 0.1 to 0.2%Δk/kk'. The error in the case using adjusted cross-section was 0.3%Δk/kk'. (author)

Evaluation of upward heat flux in ex-vessel molten core heat transfer using MELCOR

International Nuclear Information System (INIS)

Park, S.Y.; Park, J.H.; Kim, S.D.; Kim, D.H.; Kim, H.D.

2000-01-01

The purpose of this study is to share experiences of MELCOR application to resolve the molten corium-concrete interaction (MCCI) issue in the Korea Next Generation Reactor (KNGR). In the evaluation of concrete erosion, the heat transfer modeling from the molten corium internal to the corium pool surface is very important and uncertain. MELCOR employs Kutateladze or Greene's bubble-enhanced heat transfer model for the internal heat transfer. The phenomenological uncertainty is so large that the model provides several model parameters in addition to the phenomenological model for user flexibility. However, the model parameters do not work on Kutateladze correlation at the top of the molten layer. From our experience, a code modification is suggested to match the upward heat flux with the experimental results. In this analysis, minor modification was carried out to calculate heat flux from the top molten layer to corium surface, and efforts were made to find out the best value of the model parameter based on upward heat flux of MACE test M1B. Discussion also includes its application to KNGR. (author)

Application of advanced model of radiative heat transfer in a rod geometry to QUENCH and PARAMETER tests

International Nuclear Information System (INIS)

Vasiliev, A.D.; Kobelev, G.V.; Astafieva, V.O.

2007-01-01

Radiative heat transfer is very important in different fields of mechanical engineering and related technologies including nuclear reactors, heat transfer in furnaces, aerospace, different high-temperature assemblies. In particular, in the course of a hypothetical severe accident at PWR-type nuclear reactor the temperatures inside the reactor vessel reach high values at which taking into account of radiative heat exchange between the structures of reactor (including core and other reactor vessel elements) gets important. Radiative heat transfer dominates the late phase of severe accident because radiative heat fluxes (proportional to T4, where T is the temperature) are generally considerably higher than convective and conductive heat fluxes in a system. In particular, heat transfer due to radiation determines the heating and degradation of the core and surrounding steel in-vessel structures and finally influences the composition, temperature and mass of materials pouring out of the reactor vessel after its loss of integrity. Existing models of radiative heat exchange use many limitations and approximations: approximate estimation of view factors and beam lengths; the geometry change in the course of the accident is neglected; the database for emissivities of materials is not complete; absorption/emission by steam-noncondensable medium is taken into account approximately. The module MRAD was developed in this paper to model the radiative heat exchange in rod-like geometry typical of PWR-type reactor. Radiative heat exchange is computed using dividing on zones (zonal method) as in existing radiation models implemented to severe accident numerical codes such as ICARE, SCDAP/RELAP, MELCOR but improved in following aspects: new approach to evaluation of view factors and mean beam length; detailed evaluation of gas absorptivity and emissivity; account of effective radiative thermal conductivity for the large core; account of geometry modification in the course of severe

Constraints on The Coupled Thermal Evolution of the Earth's Core and Mantle, The Age of The Inner Core, And The Origin of the 186Os/188Os Core(?) Signal in Plume-Derived Lavas

Science.gov (United States)

Lassiter, J. C.

2005-12-01

Thermal and chemical interaction between the core and mantle has played a critical role in the thermal and chemical evolution of the Earth's interior. Outer core convection is driven by core cooling and inner core crystallization. Core/mantle heat transfer also buffers mantle potential temperature, resulting in slower rates of mantle cooling (~50-100 K/Ga) than would be predicted from the discrepancy between current rates of surface heat loss (~44 TW) and internal radioactive heat production (~20 TW). Core/mantle heat transfer may also generate thermal mantle plumes responsible for ocean island volcanic chains such as the Hawaiian Islands. Several studies suggest that mantle plumes, in addition to transporting heat from the core/mantle boundary, also carry a chemical signature of core/mantle interaction. Elevated 186Os/188Os ratios in lavas from Hawaii, Gorgona, and in the 2.8 Ga Kostomuksha komatiites have been interpreted as reflecting incorporation of an outer core component with high time-integrated Pt/Os and Re/Os ( Brandon et al., 1999, 2003; Puchtel et al., 2005). Preferential partitioning of Os relative to Re and Pt into the inner core during inner core growth may generate elevated Re/Os and Pt/Os ratios in the residual outer core. Because of the long half-life of 190Pt (the parent of 186Os, t1/2 = 489 Ga), an elevated 186Os/188Os outer core signature in plume lavas requires that inner core crystallization began early in Earth history, most likely prior to 3.5 Ga. This in turn requires low time-averaged core/mantle heat flow (<~2.5 TW) or large quantities of heat-producing elements in the core. Core/mantle heat flow may be estimated using boundary-layer theory, by measuring the heat transported in mantle plumes, by estimating the heat transported along the outer core adiabat, or by comparing the rates of heat production, surface heat loss, and secular cooling of the mantle. All of these independent methods suggest time-averaged core/mantle heat flow of ~5

Residential gas-fired sorption heat pumps. Test and technology evaluation

Energy Technology Data Exchange (ETDEWEB)

Naeslund, M.

2008-12-15

Heat pumps may be the next step in gas-fired residential space heating. Together with solar energy it is an option to combine natural gas and renewable energy. Heat pumps for residential space heating are likely to be based on the absorption or adsorption process, i.e. sorption heat pumps. Manufacturers claim that the efficiency could reach 140-160%. The annual efficiency will be lower but it is clear that gas-fired heat pumps can involve an efficiency and technology step equal to the transition from non-condensing gas boilers with atmospheric burners to condensing boilers. This report contains a review of the current sorption gas-fired heat pumps for residential space heating and also the visible development trends. A prototype heat pump has been laboratory tested. Field test results from Germany and the Netherlands are also used for a technology evaluation. The tested heat pump unit combines a small heat pump and a supplementary condensing gas boiler. Field tests show an average annual efficiency of 120% for this prototype design. The manufacturer abandoned the tested design during the project period and the current development concentrates on a heat pump design only comprising the heat pump, although larger. The heat pump development at three manufacturers in Germany indicates a commercial stage around 2010-2011. A fairly high electricity consumption compared to traditional condensing boilers was observed in the tested heat pump. Based on current prices for natural gas and electricity the cost savings were estimated to 12% and 27% for heat pumps with 120% and 150% annual efficiency respectively. There is currently no widespread performance testing procedure useful for annual efficiency calculations of gas-fired heat pumps. The situation seems to be clearer for electric compression heat pumps regarding proposed testing and calculation procedures. A German environmental label exists and gasfired sorption heat pumps are also slightly treated in the Eco-design work

The Effects of the Heat and Moisture Exchanger on Humidity, Airway Temperature, and Core Body Temperature

National Research Council Canada - National Science Library

Delventhal, Mary

1999-01-01

Findings from several studies have demonstrated that the use of a heat and moisture exchanger increases airway humidity, which in turn increases mean airway temperature and prevents decreases in core body temperature...

Analysis and testing of W-DHR system for decay heat removal in the lead-cooled ELSY reactor

International Nuclear Information System (INIS)

Bandini, Giacomino; Meloni, Paride; Polidori, Massimiliano; Gaggini, Piero; Labanti, Valerio; Tarantino, Mariano; Cinotti, Luciano; Presciuttini, Leonardo

2009-01-01

An innovative LFR system that complies with GEN IV goals is under design in the frame of ELSY European project. ELSY is a lead-cooled pool-type reactor of about 1500 MW thermal power which normally relies on the secondary system for decay heat removal. Since the secondary system is not safety-grade and must be fully depressurized in case of detection of a steam generator tube rupture, an independent and much reliable decay heat removal (DHR) system is foreseen on the primary side. Owing to the limited capability of the Reactor Vessel Air Cooling System (RVACS) in this large power reactor, additional safety-grade loops equipped with coolers immersed in the primary coolant are necessary for an efficient removal of decay heat. Some of these loops (W-DHR) are of innovative design and may operate with water at atmospheric pressure. In the frame of the ICE program to be performed on the integral facility CIRCE at ENEA/Brasimone research centre within the EUROTRANS European project, integral circulation experiments with core heat transport and heat removal by steam generator will be conducted in a reactor pool-type configuration. Taking advantage from this experimental program, a mock-up of W-DHR heat exchanger will be tested in order to investigate its functional behavior for decay heat removal. Some pre-test calculations of W-DHR heat exchanger operation in CIRCE have been performed with the RELAP5 thermal-hydraulic code in order to support the heat exchanger design and test conduct. In this paper the experimental activity to be conducted in CIRCE and main results from W-DHR pre-test calculations are presented, along with a preliminary investigation of the W-DHR system efficiency in ELSY configuration. (author)

Heat Removal Performance of Hybrid Control Rod for Passive In-Core Cooling System

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung Mo; Jeong, Yeong Shin; Kim, In Guk; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

2015-10-15

The two-phase closed heat transfer device can be divided by thermosyphon heat pipe and capillary wicked heat pipe which uses gravitational force or capillary pumping pressure as a driving force of the convection of working fluid. If there is a temperature difference between reactor core and ultimate heat sink, the decay heat removal and reactor shutdown is possible at any accident conditions without external power sources. To apply the hybrid control rod to the commercial nuclear power plants, its modelling about various parameters is the most important work. Also, its unique geometry is coexistence of neutron absorber material and working fluid in a cladding material having annular vapor path. Although thermosyphon heat pipe (THP) or wicked heat pipe (WHP) shows high heat transfer coefficients for limited space, the maximum heat removal capacity is restricted by several phenomena due to their unique heat transfer mechanism. Validation of the existing correlations on the annular vapor path thermosyphon (ATHP) which has different wetted perimeter and heated diameter must be conducted. The effect of inner structure, and fill ratio of the working fluid on the thermal performance of heat pipe has not been investigated. As a first step of the development of hybrid heat pipe, the ATHP which contains neutron absorber in the concentric thermosyphon (CTHP) was prepared and the thermal performance of the annular thermosyphon was experimentally studied. The heat transfer characteristics and flooding limit of the annular vapor path thermosyphon was studied experimentally to model the performance of hybrid control rod. The following results were obtained: (1) The annular vapor path thermosyphon showed better evaporation heat transfer due to the enhanced convection between adiabatic and condenser section. (2) Effect of fill ratio on the heat transfer characteristics was negligible. (3) Existing correlations about flooding limit of thermosyphon could not reflect the annular vapor

Test of In-core Flux Detectors in KNK II

CERN Document Server

Hoppe, P

1979-01-01

The development of in-core detectors for Liquid Metal Fast Breeder Reactors (LMFBRs) is still in an early stage, and little operation experience is available. Therefore self-powered neutron and gamma detectors and neutron sensitive ionization chambers -especially developed for LMFBRs- have been tested in the Fast Sodium Cooled Test Reactor KNK II. Seven flux detectors have been installed in the core of KNK II by means of a special test rig. Five of them failed already within the first week during operation in the reactor. Due to measurements of electrical resistances and capacities, sodium penetrating into the detectors or cables probably seems to be the cause. As tests prior to the installation in the core proved the tightness of all detectors, it is suspected that small cracks have developed in the detector casings or in the outer cable sheaths during their exposure to the hot coolant. Two ionization chambers did not show these faults. However, one of them failed because the saturation current plateau disap...

Effects of Re-heating Tissue Samples to Core Body Temperature on High-Velocity Ballistic Projectile-tissue Interactions.

Science.gov (United States)

Humphrey, Caitlin; Henneberg, Maciej; Wachsberger, Christian; Maiden, Nicholas; Kumaratilake, Jaliya

2017-11-01

Damage produced by high-speed projectiles on organic tissue will depend on the physical properties of the tissues. Conditioning organic tissue samples to human core body temperature (37°C) prior to conducting ballistic experiments enables their behavior to closely mimic that of living tissues. To minimize autolytic changes after death, the tissues are refrigerated soon after their removal from the body and re-heated to 37°C prior to testing. This research investigates whether heating 50-mm-cube samples of porcine liver, kidney, and heart to 37°C for varying durations (maximum 7 h) can affect the penetration response of a high-speed, steel sphere projectile. Longer conditioning times for heart and liver resulted in a slight loss of velocity/energy of the projectile, but the reverse effect occurred for the kidney. Possible reasons for these trends include autolytic changes causing softening (heart and liver) and dehydration causing an increase in density (kidney). © 2017 American Academy of Forensic Sciences.

Core radial power profile effect on system and core cooling behavior during reflood phase of PWR-LOCA with CCTF data

International Nuclear Information System (INIS)

Akimoto, Hajime; Iguchi, Tadashi; Murao, Yoshio

1985-01-01

In the reactor safety assessment during reflood phase of a PWR-LOCA, it is assumed implicitly that the core thermal hydraulic behavior is evaluated by the one-dimensional model with an average power rod. In order to assess the applicability of the one-dimensional treatment, integral tests were performed with various core radial power profiles using the Cylindrical Core Test Facility (CCTF) whose core includes about 2,000 heater rods. The CCTF results confirm that the core radial power profile has weak effect on the thermal hydraulic behavior in the primary system except core. It is also confirmed that the core differential pressure in the axial direction is predicted by the one-dimensional core model with an average power rod even in the case with a steep radial power profile in the core. Even though the core heat transfer coefficient is dependent on the core radial power profile, it is found that the error of the peak clad surface temperature calculation is less than 15 K using the one-dimensional model in the CCTF tests. The CCTF results support the one-dimensional treatment assumed in the reactor safety assessment. (author)

Experiment data report for Semiscale Mod-1 Test S-29-1 (integral test with asymmetrical break)

International Nuclear Information System (INIS)

Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

1976-07-01

Recorded test data are presented for Test S-29-1 of the Semiscale Mod-1 special heat transfer test series. This test is among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident (LOCA) in a pressurized-water reactor system. Test S-29-1 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,260 psia. An asymmetrical offset shear cold leg break was used to investigate the system response to a depressurization transient with a flow distribution different from that associated with a symmetrical cold leg break. System flow was set to achieve a core fluid temperature differential of 66 0 F at full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling (DNB) might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core cooling injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown

Concepts for passive heat removal and filtration systems under core meltdown conditions

International Nuclear Information System (INIS)

Wilhelm, J.G.; Neitzel, H.-J.

1993-01-01

The objective of the new containment concept being developed by KfK is the complete passive enclosure of a power reactor after a core meltdown accident by means of a solid containment structure and passive removal of the decay heat. This is to be accomplished by cooling the containment walls with ambient air, with thermoconvection as the driving force. The concept of the containment is described. Data are given of the heat removal and the requirements for filtration of the exhaust air, which is contaminated due to the leak rate assumed for the inner containment. The concept for the filter system is described. Various solutions for reduction of the large volumetric flow to be filtered are discussed. 3 refs., 8 figs

Hanford coring bit temperature monitor development testing results report

International Nuclear Information System (INIS)

Rey, D.

1995-05-01

Instrumentation which directly monitors the temperature of a coring bit used to retrieve core samples of high level nuclear waste stored in tanks at Hanford was developed at Sandia National Laboratories. Monitoring the temperature of the coring bit is desired to enhance the safety of the coring operations. A unique application of mature technologies was used to accomplish the measurement. This report documents the results of development testing performed at Sandia to assure the instrumentation will withstand the severe environments present in the waste tanks

Evaluation of CCTF Core-II second acceptance Test C2-AC2 (Run 052)

International Nuclear Information System (INIS)

Okubo, Tsutomu; Murao, Yoshio

1984-03-01

In order to investigate the thermo-hydrodynamic behavior in a PWR during the reflood phase of the LOCA, large scale reflooding tests have been conducted at JAERI using the CCTF Core-I and Core-II facilities. This report presents the investigation on the difference in the thermo-hydrodynamic behavior observed between in the CCTF Core-I and Core-II facilities. For this purpose the test data of the second CCTF Core-II acceptance test C2-AC2 (Run 052) were evaluated by using the data of the Test CL-21 (Run 040) in the Core-I test series. The experimental conditions for these two tests were almost identical. Comparing the data of those two tests, the following is obtained. 1. The system behavior observed in the Core-II facility was nearly identical to that observed in the Core-I facility. 2. The core behavior observed in the Core-II facility was also nearly identical to that observed in the Core-I facility except for the top quenching behavior. 3. The differences in the top quenching behavior between the two facilities were as follows: (1) The selective occurrence of top quenching below the open holes of the upper core support plate observed in the Core-I facility was not observed in the Core-II facility. (2) Top quenching tended to occur less in the Core-II facility in the region where the initial average linear power density was over 1.69 kW/m. (author)

BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses

International Nuclear Information System (INIS)

Wiles, L.E.; Lombardo, N.J.; Heeb, C.M.; Jenquin, U.P.; Michener, T.E.; Wheeler, C.L.; Creer, J.M.; McCann, R.A.

1986-06-01

This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions

Experimental investigation of natural convection heat transfer in volumetrically heated spherical segments. Final report

International Nuclear Information System (INIS)

Asfia, F.; Dhir, V.

1998-03-01

One strategy for preventing the failure of lower head of a nuclear reactor vessel is to flood the concrete cavity with subcooled water in accidents in which relocation of core material into the vessel lower head occurs. After the core material relocates into the vessel, a crust of solid material forms on the inner wall of the vessel, however, most of the pool remains molten and natural convection exists in the pool. At present, uncertainty exists with respect to natural convection heat transfer coefficients between the pool of molten core material and the reactor vessel wall. In the present work, experiments were conducted to examine natural convection heat transfer in internally heated partially filled spherical pools with external cooling. In the experiments, Freon-113 contained in a Pyrex bell jar was used as a test liquid. The pool was bounded with a spherical segment at the bottom, and was heated with magnetrons taken from a conventional microwave oven. The vessel was cooled from the outside with natural convection of water or with nucleate boiling of liquid nitrogen

Startup testing of Romania dual-core test reactor

International Nuclear Information System (INIS)

Whittemore, W.L.

1980-01-01

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

Heat transfer performance test of PDHRS heat exchangers of PGSFR using STELLA-1 facility

Energy Technology Data Exchange (ETDEWEB)

Hong, Jonggan, E-mail: hong@kaeri.re.kr; Yeom, Sujin; Eoh, Jae-Hyuk; Lee, Tae-Ho; Jeong, Ji-Young

2017-03-15

Highlights: • Heat transfer performance test of heat exchangers of PGSFR PDHRS is conducted using STELLA-1 facility. • Steady-state test results of DHX and AHX show good agreement with theoretical results of design codes. • Design codes for DHX and AHX are validated by STELLA-1 experimental results. • Heat transport capability of DHX and AHX is turned out to be satisfactory for reliable plant operation. - Abstract: The STELLA-1 facility was designed and constructed to carry out separate effect tests of the decay heat exchanger (DHX) and natural draft sodium-to-air heat exchanger (AHX), which are key components of the safety-grade decay heat removal system in PGSFR. The DHX is a sodium-to-sodium heat exchanger with a straight tube arrangement, and the AHX is a sodium-to-air heat exchanger with a helically coiled tube arrangement. The model heat exchangers in STELLA-1 have been designed to meet their own similitude conditions from the prototype ones, of which scale ratios were set to be unity in height (or length) and 1/2.5 in heat transfer rate. Consequently, the overall heat transfer coefficients and log-mean temperature differences of the prototypes have been preserved as well. The steady-state test results for each model heat exchanger obtained from STELLA-1 showed good agreement with the theoretical results of the computer design codes for thermal-sizing and a performance analysis of the DHX and AHX. In the DHX result comparison, the discrepancies in the heat transfer rate ranged from −4.4% to 2.0%, and in the AHX result comparison, they ranged from −11.1% to 12.6%. Therefore, the first step in thermal design codes validation for sodium heat exchangers, e.g., DHX and AHX, has been successfully completed with the experimental database obtained from STELLA-1. In addition, the heat transfer performance of the DHX and AHX was found to be satisfactory enough to secure a reliable decay heat removal performance.

Double-Lap Shear Test For Honeycomb Core

Science.gov (United States)

Nettles, Alan T.; Hodge, Andrew J.

1992-01-01

Double-lap test measures shear strength of panel made of honeycomb core with 8-ply carbon-fiber/epoxy face sheets. Developed to overcome three principal disadvantages of prior standard single-lap shear test: specimen had to be more than 17 in. long; metal face sheets had to be used; and test introduced torque, with consequent bending and peeling of face sheets and spurious tensile or compressive loading of honeycomb.

Linking lowermost mantle structure, core-mantle boundary heat flux and mantle plume formation

Science.gov (United States)

Li, Mingming; Zhong, Shijie; Olson, Peter

2018-04-01

The dynamics of Earth's lowermost mantle exert significant control on the formation of mantle plumes and the core-mantle boundary (CMB) heat flux. However, it is not clear if and how the variation of CMB heat flux and mantle plume activity are related. Here, we perform geodynamic model experiments that show how temporal variations in CMB heat flux and pulses of mantle plumes are related to morphologic changes of the thermochemical piles of large-scale compositional heterogeneities in Earth's lowermost mantle, represented by the large low shear velocity provinces (LLSVPs). We find good correlation between the morphologic changes of the thermochemical piles and the time variation of CMB heat flux. The morphology of the thermochemical piles is significantly altered during the initiation and ascent of strong mantle plumes, and the changes in pile morphology cause variations in the local and the total CMB heat flux. Our modeling results indicate that plume-induced episodic variations of CMB heat flux link geomagnetic superchrons to pulses of surface volcanism, although the relative timing of these two phenomena remains problematic. We also find that the density distribution in thermochemical piles is heterogeneous, and that the piles are denser on average than the surrounding mantle when both thermal and chemical effects are included.

OECD MCCI project long-term 2-D molten core concrete interaction test design report, Rev. 0. September 30, 2002

International Nuclear Information System (INIS)

Farmer, M.T.; Kilsdonk, D.J.; Lomperski, S.; Aeschliman, R.W.; Basu, S.

2011-01-01

The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following two technical objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of the first program objective, the Small-Scale Water Ingression and Crust Strength (SSWICS) test series has been initiated to provide fundamental information on the ability of water to ingress into cracks and fissures that form in the debris during quench, thereby augmenting the otherwise conduction-limited heat transfer process. A test plan for Melt Eruption Separate Effects Tests (MESET) has also been developed to provide information on the extent of crust growth and melt eruptions as a function of gas sparging rate under well-controlled experiment conditions. In terms of the second program objective, the project Management Board (MB) has approved startup activities required to carry out

Proposed heat transfer model for the gas-liquid heat transfer effects observed in the Stanford Research Institute scaled tests

International Nuclear Information System (INIS)

Corradini, M.; Sonin, A.A.; Todreas, N.

1976-12-01

In 1971-72, the Stanford Research Institute conducted a series of scaled experiments which simulated a sodium-vapor expansion in a hypothetical core disruptive accident (HCDA) for the Fast Flux Test Facility. A non-condensible explosive source was used to model the pressure-volume expansion characteristics of sodium vapor as predicted by computer code calculations. Rigid piston-cylinder experiments ( 1 / 10 and 1 / 30 scale) were undertaken to determine these expansion characteristics. The results showed that the pressure-volume characteristics depend significantly on the presence of water in the cylinder reducing the work output by about 50 percent when a sufficient water depth was present. The study presented proposes that the mechanism of heat transfer between the water and high temperature gas was due to area enhancement by Taylor instabilities at the gas-liquid interface. A simple heat transfer model is proposed which describes this energy transport process and agrees well with the experimental data from both scaled experiments. The consequences of this analysis suggest that an estimate of the heat transfer to the cold slug during a full-scale HCDA due to sodium vapor expansion and the accompanying reduction in mechanical work energy warrants further investigation. The implication of this analysis is that for either sodium or fuel vapor expansion in an HCDA, there is an inherent heat transfer mechanism which significantly reduces the work output of the expanding bubble

Efficient Test Application for Core-Based Systems Using Twisted-Ring Counters

OpenAIRE

Anshuman Chandra; Krishnendu Chakrabarty; Mark C. Hansen

2001-01-01

We present novel test set encoding and pattern decompression methods for core-based systems. These are based on the use of twisted-ring counters and offer a number of important advantages–significant test compression (over 10X in many cases), less tester memory and reduced testing time, the ability to use a slow tester without compromising test quality or testing time, and no performance degradation for the core under test. Surprisingly, the encoded test sets obtained from partially-specified...

First in-core measurement results obtained with the innovative mobile calorimeter CALMOS inside the OSIRIS material testing reactor

International Nuclear Information System (INIS)

Carcreff, Hubert; Salmon, Laurent; Courtaux, Cedric

2014-01-01

Nuclear heating rate inside an MTR has to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry. An innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70 MWth OSIRIS reactor operated by CEA. Thanks to a new calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. Development of the calorimetric probe required manufacturing and irradiation of mock-ups in the ex-core area, where nuclear heating rate does not exceed 2 W.g -1 . The calorimeter working mode, the different measurement procedures, main modeling and ex-core experimental results have been already presented in previous papers. In this paper, we present in-core results obtained from 2011 to 2013 with the final device. For the first time, this new experimental measurement system was operated in several experimental locations, with nominal in-core thermal hydraulic conditions, nominal neutron flux and nuclear heating rate up to 6 W.g -1 (in graphite). After a brief presentation of the displacement system specificities, first nuclear heating distributions are presented and discussed. The Finite Element model of the calorimeter was upgraded in order to match calculated temperatures with measured ones. This 'validated' model allowed to estimate a Kc factor which tends to correct small nonlinearities when heating rate is calculated from the 'calibration method'. A comparison is made between nuclear heating rates determined from 'calibration' and 'zero methods'. In addition, an evaluation of the global uncertainty associated to the measurements is detailed. Finally, a comparison is made with available measurements obtained from previous calorimeters. (authors)

Performance testing of a hydrogen heat pipe

International Nuclear Information System (INIS)

Alario, J.; Kosson, R.

1980-01-01

Test results are presented for a reentrant groove heat pipe with hydrogen working fluid. The heat pipe became operational between 20 and 30 K after a cooldown from 77 K without any difficulty. Steady-state performance data taken over a 19 to 23 K temperature range indicated the following: (1) maximum heat transport capacity 5.4 W-m (2) static wicking height 1.42 cm and (3) overall heat pipe conductance 1.7 W/C. These data agreed remarkably well with extrapolations made from comparable ammonia test results. The maximum heat transport capacity is 9.5% larger than the extrapolated value, but the static wicking height is the same. The overall conductance is 29% of the ammonia value, which is close to the ratio of liquid thermal conductivities (24%). Also, recovery from a completely frozen condition was accomplished within 5 min by simply applying an evaporater heat load of 1.8 W

Evaluation report on CCTF Core-II reflood test C2-4 (Run 62)

International Nuclear Information System (INIS)

Okubo, Tsutomu; Iguchi, Tadashi; Sugimoto, Jun; Akimoto, Hajime; Murao, Yoshio; Okabe, Kazuharu.

1985-03-01

This report presents a data evaluation of the CCTF Core-II test C2-4 (Run 62), which was conducted on May 12, 1983. This test was conducted to investigate the reproducibility of tests in the CCTF Core-II test series. Therefore, the initial and boundary conditions of the present test were determined to be the same as those for the previously performed base case test (Test C2-SH1). Comparing the data of the present test with those of Test C2-SH1, the following results are obtained. (1) The initial and boundary conditions for the two tests were nearly identical except the temperature of the core barrel and the lower plenum fluid. The difference in the latter is considered to result in the difference in the core inlet subcooling of about 6 K at most. (2) The system behavior was almost identical. (3) The core cooling behavior was also nearly identical except a little difference in the rod surface temperature in the upper part of the high power region. (4) Taking account that the difference mentioned above in item (3) is small and can be explained qualitatively to be caused by the difference in the core inlet subcooling mentioned above in item (1), it is considered practically that there is the reproducibility of the thermo-hydrodynamic behavior in the CCTF Core-II tests. (author)

Evaluation report on CCTF core-I reflood tests Cl-5 (Run 14), Cl-10 (Run 19) and Cl-12 (Run 21)

International Nuclear Information System (INIS)

Akimoto, Hajime; Murao, Yoshio

1983-06-01

Three tests Cl-5 (Run 14), Cl-10 (Run 19) and Cl-12 (Run 21) were performed using the Cylindrical Core Test Facility to study the effect of the containment pressure on the core cooling and the system behaviors during the reflood phase of a PWR-LOCA. The containment pressures of these tests were 0.15, 0.20 and 0.30 MPa for the tests Cl-10, Cl-5 and Cl-12, respectively. Through the comparison of the test results from these three tests, the following results were obtained. (1) The higher containment pressure gave the higher heat transfer coefficient in the core. This resulted in the lower turnaround temperature, the shorter turnaround time and the shorter quench time at the higher containment pressure. (2) In the higher containment pressure test, the higher core water head, the higher upper plenum water head, the higher downcomer water head in the early period and the lower downcomer water head in the later period were observed than those in the lower containment pressure test. This resulted in the higher pressure drop through the intact loop in the early period of the tests and the lower pressure drop in the later period of the test with the containment pressure. (3) The pressure drop through the broken cold leg pressurized the primary system. The pressure drop through the broken cold leg was decreased with the containment pressure. (4) The core inlet mass flow rate was increased with the containment pressure as observed in the FLECHT-SET phase B1 test. In quantity, however, the effect of the containment pressure on the increase of the core inlet mass flow rate was less in the CCTF than that in the FLECHT-SET. The less sensitivity in the CCTF was attributed mainly to the great pressure drop through the broken cold leg, which was not observed in the FLECHT-SET with big broken cold leg. (5) The system effect of the containment pressure was explained quantitatively. (author)

ANALYSIS OF GAMMA HEATING AT TRIGA MARK REACTOR CORE BANDUNG USING PLATE TYPE FUEL

Directory of Open Access Journals (Sweden)

Setiyanto Setiyanto

2016-10-01

Full Text Available ABSTRACT In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities and central irradiation position (CIP, especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0,87 W/g, but very low value for Lazy Susan position (lest then 0,11 W/g. Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. Keywords: gamma heating, nuclear reactor, research reactor, reactor safety.   ABSTRAK Dengan dihentikannya produksi elemen bakar reaktor jenis Triga oleh produsen, maka semua reaktor TRIGA di dunia terganggu operasinya, termasuk juga reaktor TRIGA 2000 di Bandung. Untuk mendukung pengoperasian reaktor TRIGA Bandung

Heating facility for blanket and performance test

Energy Technology Data Exchange (ETDEWEB)

Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Sato, Satoshi; Hatano, Toshihisa; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Hara, Shigemitsu

1999-03-01

A design and a fabrication of heating test facility for a mock-up of the blanket module to be installed in International Thermonuclear Experimental Reactor (ITER) have been conducted to evaluate/demonstrate its heat removal performance and structural soundness under cyclic heat loads. To simulate surface heat flux to the blanket module, infrared heating method is adopted so as to heat large surface area uniformly. The infrared heater is used in vacuum environment (10{sup -4} Torr{approx}), and the lamps are cooled by air flowing through an annulus between the lamp and a cover tube made of quartz glass. Elastomer O rings (available to be used up to {approx}300degC) and used for vacuum seal at outer surface of the cover tube. To prevent excessive heating of the O ring, the end part of the cover tube is specially designed including the tube shape, flow path of air and gold coating on the surface of the cover tube to protect the O ring against thermal radiation from glowing tungsten filament. To examine the performance of the facility, steady state and cyclic operation of the infrared heater were conducted using a small-scaled shielding blanket mock-up as a test specimen. The important results are as follows: (1) Heat flux at the surface of the small-scaled mock-up measured by a calorimeter was {approx}0.2 MW/m{sup 2}. (2) A comparison of thermal analysis results and measured temperature responses showed that the small-scaled mock-up had good heat removal performance. (3) Steady state operation and cyclic operation with step response between the rated and zero powers of the infrared heater were successfully performed, and it was confirmed that this heating facility was well-prepared and available for the thermal cyclic test of a blanket module. (author)

Strength and permeability tests on ultra-large Stripa granite core

International Nuclear Information System (INIS)

Thorpe, R.; Watkins, D.J.; Ralph, W.E.; Hsu, R.; Flexser, S.

1980-09-01

This report presents the results of laboratory tests on a 1 meter diameter by 2 meters high sample of granitic (quartz monzonite) rock from the Stripa mine in Sweden. The tests were designed to study the mechanical and hydraulic properties of the rock. Injection and withdrawal permeability tests were performed at several levels of axial stress using a borehole through the long axis of the core. The sample was pervasively fractured and its behavior under uniaxial compressive stress was very complicated. Its stress-strain behavior at low stresses was generally similar to that of small cores containing single healed fractures. However, this large core failed at a peak stress of 7.55 MPa, much less than the typical strength measured in small cores. The complex failure mechanism included a significant creep component. The sample was highly permeable, with flows-per-unit head ranging from 0.11 to 1.55 cm 2 /sec. Initial application of axial load caused a decrease in permeability, but this was followed by rapid increase in conductivity coincident with the failure of the core. The hydraulic regime in the fracture system was too intricate to be satisfactorily modeled by simple analogs based on the observed closure of the principal fractures. The test results contribute to the data base being compiled for the rock mass at the Stripa site, but their proper application will require synthesis of results from several laboratory and in situ test programs

LMFR core and heat exchanger thermohydraulic design: former USSR and present Russian approaches

International Nuclear Information System (INIS)

1999-01-01

The information presented in this report is dealing with liquid metal cooled fast reactors some of which are in operation (France, Japan, Russian federation) or under construction. Comprehensive thermal hydraulic research both experimental and numeric applied to such reactors was carried out in the Institute of Physics and Power Engineering (IPPE), Obninsk, Russian Federation. The IAEA Working Group on fast Reactors (IWGFR) recommended that IPPE should generalize its thermal hydraulic studies as well as results of other countries published previously in the field of liquid metal flow distribution and heat transfer in fuel pin and heat exchanger rod bundles (France, Germany, Japan, India, Russian Federation, United Kingdom and United States). The validity of computer codes and design approaches was proven by comparison of calculated results with measured values of velocity, pressure, temperature distributions in rod bundles cooled/heated by liquid metal, usually sodium. The report includes the methodology and philosophy of the analytical and experimental investigations when applied to core and heat exchanger thermal hydraulic design of Light Water Moderated Fast Reactors (LMFRs)

Test of In-core Flux Detectors in KNK II

International Nuclear Information System (INIS)

Hoppe, P.; Mitzel, F.

1979-10-01

The development of in-core detectors for Liquid Metal Fast Breeder Reactors (LMFBRs) is still in an early stage, and little operation experience is available. Therefore self-powered neutron and gamma detectors and neutron sensitive ionization chambers -especially developed for LMFBRs- have been tested in the Fast Sodium Cooled Test Reactor KNK II. Seven flux detectors have been installed in the core of KNK II by means of a special test rig. Five of them failed already within the first week during operation in the reactor. Due to measurements of electrical resistances and capacities, sodium penetrating into the detectors or cables probably seems to be the cause. As tests prior to the installation in the core proved the tightness of all detectors, it is suspected that small cracks have developed in the detector casings or in the outer cable sheaths during their exposure to the hot coolant. Two ionization chambers did not show these faults. However, one of them failed because the saturation current plateau disappeared and the other one's sensitivity decreased by a factor of five during the test period. It is suspected that in both cases changes of the filling gas might be involved

Gamma ray heating rates due to chromium isotopes in stellar core during late stages of high mass stars (>10M⊙

Directory of Open Access Journals (Sweden)

Nabi Jameel-Un

2017-01-01

Full Text Available Gamma ray heating rates are thought to play a crucial role during the pre-supernova stage of high mass stars. Gamma ray heating rates, due to β±-decay and electron (positron capture on chromium isotopes, are calculated using proton-neutron quasiparticle random phase approximation theory. The electron capture significantly affects the lepton fraction (Ye and accelerates the core contraction. The gamma rays emitted as a result of weak processes heat the core and tend to hinder the cooling and contraction due to electron capture and neutrino emission. The emitted gamma rays tend to produce enormous entropy and set the convection to play its role at this stage. The gamma heating rates, on 50-60Cr, are calculated for the density range 10 < ρ (g.cm-3 < 1011 and temperature range 107 < T (K < 3.0×1010.

Study on HANARO core conversion using U-Mo fuel

International Nuclear Information System (INIS)

Lee, K.H.; Lee, C.S.; Seo, C.G.; Park, S.J.; Kim, H.; Kim, C.K.

2002-01-01

Two types of fuel rods with different fuel meat diameter and uranium density are considered for HANARO core conversion with high density U-Mo fuel. Arranging standard fuels of 5.0 g U/cc and 6.35 mm in diameter at the inner ring of an assembly and reduced fuels of 4.3 g U/cc and 5.49 mm in diameter at the outer ring of an assembly flattens the assembly power distribution and avoids the increase of linear heat generation rate due to using higher uranium density and less number of fuel rods. The maximum linear heat generation rate is similar with the current reference core and four fuel sites at the outer core in the reflector tank is converted to the irradiation sites to suit more demand on fuel tests and radioisotope production at outer core sites. This new core has 32% longer fuel cycle than the current reference core. (author)

Estimation of groundwater flow from temperature monitoring in a borehole heat exchanger during a thermal response test

Science.gov (United States)

Yoshioka, Mayumi; Takakura, Shinichi; Uchida, Youhei

2018-05-01

To estimate the groundwater flow around a borehole heat exchanger (BHE), thermal properties of geological core samples were measured and a thermal response test (TRT) was performed in the Tsukuba upland, Japan. The thermal properties were measured at 57 points along a 50-m-long geological core, consisting predominantly of sand, silt, and clay, drilled near the BHE. In this TRT, the vertical temperature in the BHE was also monitored during and after the test. Results for the thermal properties of the core samples and from the monitoring indicated that groundwater flow enhanced thermal transfers, especially at shallow depths. The groundwater velocities around the BHE were estimated using a two-dimensional numerical model with monitoring data on temperature changes. According to the results, the estimated groundwater velocity was generally consistent with hydrogeological data from previous studies, except for the data collected at shallow depths consisting of a clay layer. The reasons for this discrepancy at shallow depths were predicted to be preferential flow and the occurrence of vertical flow through the BHE grout, induced by the hydrogeological conditions.

Reliability of core test – Critical assessment and proposed new approach

OpenAIRE

Shafik Khoury; Ali Abdel-Hakam Aliabdo; Ahmed Ghazy

2014-01-01

Core test is commonly required in the area of concrete industry to evaluate the concrete strength and sometimes it becomes the unique tool for safety assessment of existing concrete structures. Core test is therefore introduced in most codes. An extensive literature survey on different international codes’ provisions; including the Egyptian, British, European and ACI Codes, for core analysis is presented. All studied codes’ provisions seem to be unreliable for predicting the in-situ concrete ...

Current Status and Performance Tests of Korea Heat Load Test Facility KoHLT-EB

Energy Technology Data Exchange (ETDEWEB)

Kim, Sukkwon; Jin, Hyunggon; Shin, Kyuin; Choi, Boguen; Lee, Eohwak; Yoon, Jaesung; Lee, Dongwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Duckhoi; Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

2013-05-15

A commissioning test has been scheduled to establish the installation and preliminary performance experiments of the copper hypervapotron mockups. And a qualification test will be performed to evaluate the CuCrZr duct liner in the ITER neutral beam injection facility and the ITER first wall small-scale mockups of the semi-prototype, at up to 1.5 and 5 MW/m{sup 2} high heat flux. Also, this system will be used to test other PFCs for ITER and materials for tokamak reactors. Korean high heat flux test facility(KoHLT-EB; Korea Heat Load Test facility - Electron Beam) by using an electron beam system has been constructed in KAERI to perform the qualification test for ITER blanket FW semi-prototype mockups, hypervapotron cooling devices in fusion devices, and other ITER plasma facing components. The commissioning and performance tests with the supplier of e-gun system have been performed on November 2012. The high heat flux test for hypervapotron cooling device and calorimetry were performed to measure the surface heat flux, the temperature profile and cooling performance. Korean high heat flux test facility for the plasma facing components of nuclear fusion machines will be constructed to evaluate the performance of each component. This facility for the plasma facing materials will be equipped with an electron beam system with a 60 kV acceleration gun.

OECD MMCI 2-D Core Concrete Interaction (CCI) tests : CCCI-1 test data report-thermalhydraulic results. Rev 0 January 31, 2004.

Energy Technology Data Exchange (ETDEWEB)

Farmer, M. T.; Lomperski, S.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division)

2011-05-23

The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten coreconcrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of two long-term 2-D Core-Concrete Interaction (CCI) experiments designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-1 experiment, which was conducted on December 19, 2003. Test specifications for CCI-1 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 400 kg

Acceptance test report for core sample trucks 3 and 4

International Nuclear Information System (INIS)

Corbett, J.E.

1996-01-01

The purpose of this Acceptance Test Report is to provide documentation for the acceptance testing of the rotary mode core sample trucks 3 and 4, designated as HO-68K-4600 and HO-68K-4647, respectively. This report conforms to the guidelines established in WHC-IP-1026, ''Engineering Practice Guidelines,'' Appendix M, ''Acceptance Test Procedures and Reports.'' Rotary mode core sample trucks 3 and 4 were based upon the design of the second core sample truck (HO-68K-4345) which was constructed to implement rotary mode sampling of the waste tanks at Hanford. Successful completion of acceptance testing on June 30, 1995 verified that all design requirements were met. This report is divided into four sections, beginning with general information. Acceptance testing was performed on trucks 3 and 4 during the months of March through June, 1995. All testing was performed at the ''Rock Slinger'' test site in the 200 West area. The sequence of testing was determined by equipment availability, and the initial revision of the Acceptance Test Procedure (ATP) was used for both trucks. Testing was directed by ICF-KH, with the support of WHC Characterization Equipment Engineering and Characterization Project Operations. Testing was completed per the ATP without discrepancies or deviations, except as noted

An Equation Governing Ultralow-Velocity Zones: Implications for Holes in the ULVZ, Lateral Chemical Reactions at the Core-Mantle Boundary, and Damping of Heat Flux Variations in the Core

Science.gov (United States)

Hernlund, J. W.; Matsui, H.

2017-12-01

Ultralow-velocity zones (ULVZ) are increasingly illuminated by seismology, revealing surprising diversity in size, shape, and physical characteristics. The only viable hypotheses are that ULVZs are a compositionally distinct FeO-enriched dense material, which could have formed by fractional crystallization of a basal magma ocean, segregation of subducted banded iron formations, precipitation of solids from the outer core, partial melting and segregation of iron-rich melts from subducted basalts, or most likely a combination of many different processes. But many questions remain: Are ULVZ partially molten in some places, and not in others? Are ULVZ simply the thicker portions of an otherwise global thin layer, covering the entire CMB and thus blocking or moderating chemical interactions between the core and overlying mantle? Is such a layer inter-connected and able to conduct electrical currents that allow electro-magnetic coupling of core and mantle angular momentum? Are they being eroded and shrinking in size due to viscous entrainment, or is more material being added to ULVZ over time? Here we derive an advection-diffusion-like equation that governs the dynamical evolution of a chemically distinct ULVZ. Analysis of this equation shows that ULVZ should become readily swept aside by viscous mantle flows at the CMB, exposing "ordinary mantle" to the top of the core, thus inducing chemical heterogeneity that drives lateral CMB chemical reactions. These reactions are correlated with heat flux, thus maintaining large-scale pressure variations atop the core that induce cyclone-like flows centered around ULVZ and ponded subducted slabs. We suggest that turbulent diffusion across adjacent cyclone streams inside a stratified region atop the core readily accommodates lateral transport and re-distribution of components such as O and Si, in addition to heat. Our model implies that the deeper core is at least partly shielded from the influence of strong heat flux variations at

Advanced Marketing Core Curriculum. Test Items and Assessment Techniques.

Science.gov (United States)

Smith, Clifton L.; And Others

This document contains duties and tasks, multiple-choice test items, and other assessment techniques for Missouri's advanced marketing core curriculum. The core curriculum begins with a list of 13 suggested textbook resources. Next, nine duties with their associated tasks are given. Under each task appears one or more citations to appropriate…

Evaluation report on CCTF Core-II reflood test C2-9 (Run 68)

International Nuclear Information System (INIS)

Akimoto, Hajime; Iguchi, Tadashi; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu; Sugimoto, Jun.

1987-02-01

In order to study the LPCI flow rate effect on the core cooling and system behavior, a test was performed with the LPCI flow rate of 0.025 m 3 /s, which corresponds to the flow rate in case of no pump failure in a PWR system. Through the comparisons of test results with those from the reference test with the LPCI flow rate of 0.011 m 3 /s, the following conclusions were obtained: (1) The higher LPCI flow rate resulted in the worse core-cooling in these two tests. The test results show that the lower LPCI flow rate is not necessarily a conservative assumption for the evaluation of the core cooling during the reflood phase of a PWR LOCA. (2) The worse core-cooling in the high LPCI flow rate test is attributed to the lower core-pressure than in the reference test. It is found that the lower core-pressure results from the lower pressure drop through the broken cold leg. (3) It is expected that the current evaluation model(EM) code is still conservative because it usually predicts the low pressure drop through the broken cold leg. (4) The flow oscillation in the cold leg was not significant even in the high LPCI flow rate test before the whole core quench. (author)

Experimental investigation of thermal-hydraulic performance of PCCS with horizontal tube heat exchangers: single U-tube test

International Nuclear Information System (INIS)

Nakamura, Hideo; Anoda, Yoshinari; Arai, Kenji; Kurita, Tomohisa

2000-01-01

JAERI and JAPC started a cooperative study to verify performance of a PCCS (Passive Containment Cooling System) using horizontal heat exchanger for next-generation BWR in 1998. A test facility with a horizontal single U-tube was constructed in JAERI in 1999 to investigate fundamental condensation behavior under influences of non-condensable gas. Preliminary pre-test analyses were performed using RELAP5/ MOD3.2.1.2 code to expect the experimental outcomes by incorporating a correlation for condensation degradation because of non-condensable gas by Ueno et al. for better prediction. Preliminary results from both experiments (shakedown) and pre-test analyses indicated that the PCCS using horizontal U-tube heat exchanger is promising. Steam generated under assumed severe accident conditions; steam generation rate approx. = 1% core power, non-condensable gas concentration of 1% and simulated containment vessel pressure of 0.7 MPa, was totally condensed with a small differential pressure across inlet and outlet plenum. Experimental data will be accumulated to develop models and correlations for a better prediction of responses of the PCCS using horizontal heat exchanger during postulated severe accidents. (author)

Fast Flux Test Facility core restraint system performance

International Nuclear Information System (INIS)

Hecht, S.L.; Trenchard, R.G.

1990-02-01

Characterizing Fast Flux Test Facility (FFTF) core restraint system performance has been ongoing since the first operating cycle. Characterization consists of prerun analysis for each core load, in-reactor and postirradiation measurements of subassembly withdrawal loads and deformations, and using measurement data to fine tune predictive models. Monitoring FFTF operations and performing trend analysis has made it possible to gain insight into core restraint system performance and head off refueling difficulties while maximizing component lifetimes. Additionally, valuable information for improved designs and operating methods has been obtained. Focus is on past operating experience, emphasizing performance improvements and avoidance of potential problems. 4 refs., 12 figs., 2 tabs

Advance of core design method for ATR

International Nuclear Information System (INIS)

Maeda, Seiichirou; Ihara, Toshiteru; Iijima, Takashi; Seino, Hideaki; Kobayashi, Tetsurou; Takeuchi, Michio; Sugawara, Satoru; Matsumoto, Mitsuo.

1995-01-01

Core characteristics of ATR demonstration plant has been revised such as increasing the fuel burnup and the channel power, which is achieved by changing the number of fuel rod per fuel assembly from 28 to 36. The research and development concerning the core design method for ATR have been continued. The calculational errors of core analysis code have been evaluated using the operational data of FUGEN and the full scale simulated test results in DCA (Deuterium Critical Assembly) and HTL (Heat Transfer Loop) at O-arai engineering center. It is confirmed that the calculational error of power distribution is smaller than the design value of ATR demonstration plant. Critical heat flux correlation curve for 36 fuel rod cluster has been developed and the probability evaluation method based on its curve, which is more rational to evaluate the fuel dryout, has been adopted. (author)

Warm-Core Intensification of a Hurricane Through Horizontal Eddy Heat Transports Inside the Eye

Science.gov (United States)

Braun, Scott A.; Montgomery, Michael T.; Fulton, John; Nolan, David S.

2001-01-01

A simulation of Hurricane Bob (1991) using the PSU/NCAR MM5 mesoscale model with a finest mesh spacing of 1.3 km is used to diagnose the heat budget of the hurricane. Heat budget terms, including latent and radiative heating, boundary layer forcing, and advection terms were output directly from the model for a 6-h period with 2-min frequency. Previous studies of warm core formation have emphasized the warming associated with gentle subsidence within the eye. The simulation of Hurricane Bob also identifies subsidence warming as a major factor for eye warming, but also shows a significant contribution from horizontal advective terms. When averaged over the area of the eye, excluding the eyewall (at least in an azimuthal mean sense), subsidence is found to strongly warm the mid-troposphere (2-9 km) while horizontal advection warms the mid to upper troposphere (5-13 km) with about equal magnitude. Partitioning of the horizontal advective terms into azimuthal mean and eddy components shows that the mean radial circulation cannot, as expected, generally contribute to this warming, but that it is produced almost entirely by the horizontal eddy transport of heat into the eye. A further breakdown of the eddy components into azimuthal wave numbers 1, 2, and higher indicates that the warming is dominated by wave number 1 asymmetries, with smaller contributions coming from higher wave numbers. Warming by horizontal eddy transport is consistent with idealized modeling of vortex Rossby waves and work is in progress to identify and clarify the role of vortex Rossby waves in warm-core intensification in both the full-physics model and idealized models.

Qualification test for ITER HCCR-TBS mockups with high heat flux test facility

Energy Technology Data Exchange (ETDEWEB)

Kim, Suk-Kwon, E-mail: skkim93@kaeri.re.kr [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Seong Dae; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

2016-11-01

Highlights: • The test mockups for ITER HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) in Korea were designed and fabricated. • A thermo-hydraulic analysis was performed using a high heat flux test facility by using electron beam. • The plan for qualification tests was developed to evaluate the thermo-hydraulic efficiency in accordance with the requirements of the ITER Organization. - Abstract: The test mockups for ITER HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) in Korea were designed and fabricated, and an integrity and thermo-hydraulic performance test should be completed under the same or similar operation conditions of ITER. The test plan for a thermo-hydraulic analysis was developed by using a high heat flux test facility, called the Korean heat load test facility by using electron beam (KoHLT-EB). This facility is utilized for a qualification test of the plasma facing component (PFC) for the ITER first wall and DEMO divertor, and for the thermo-hydraulic experiments. In this work, KoHLT-EB will be used for the plan of the performance qualification test of the ITER HCCR-TBS mockups. This qualification tests should be performed to evaluate the thermo-hydraulic efficiency in accordance with the requirements of the ITER Organization (IO), which describe the specifications and qualifications of the heat flux test facility and test procedure for ITER PFC.

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

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)

Summary of CCTF test results

International Nuclear Information System (INIS)

Iguchi, T.; Murao, Y.; Sugimoto, J.; Akimoto, H.; Okubo, T.; Hojo, T.

1987-01-01

Conservatism of current safety analysis was assessed by comparing the predicted result with cylindrical core test facility (CCTF) test result performed at Japan Atomic Energy Research Institute. WREM code was selected for the assessment. The overall conservatism of the WREM code on the peak clad temperature prediction was confirmed against CCTF evaluation model (EM) test which simulated the typical initial and boundary conditions in the safety evaluation analysis. WREM code predicted the reasonable core boundary conditions and the conservatism of the code came mainly from core calculation. The conservatism of the WREM code against CCTF data could be attributed to the following three points: (1) no horizontal mixing assumption between subchannels at each elevation; (2) no modeling on heat transfer enhancement caused by the radial core power profile; and (3) conservative heat transfer correlations in the code

DART Core/Combustor-Noise Initial Test Results

Science.gov (United States)

Boyle, Devin K.; Henderson, Brenda S.; Hultgren, Lennart S.

2017-01-01

Contributions from the combustor to the overall propulsion noise of civilian transport aircraft are starting to become important due to turbofan design trends and advances in mitigation of other noise sources. Future propulsion systems for ultra-efficient commercial air vehicles are projected to be of increasingly higher bypass ratio from larger fans combined with much smaller cores, with ultra-clean burning fuel-flexible combustors. Unless effective noise-reduction strategies are developed, combustor noise is likely to become a prominent contributor to overall airport community noise in the future. The new NASA DGEN Aero0propulsion Research Turbofan (DART) is a cost-efficient testbed for the study of core-noise physics and mitigation. This presentation gives a brief description of the recently completed DART core combustor-noise baseline test in the NASA GRC Aero-Acoustic Propulsion Laboratory (AAPL). Acoustic data was simultaneously acquired using the AAPL overhead microphone array in the engine aft quadrant far field, a single midfield microphone, and two semi-infinite-tube unsteady pressure sensors at the core-nozzle exit. An initial assessment shows that the data is of high quality and compares well with results from a quick 2014 feasibility test. Combustor noise components of measured total-noise signatures were educed using a two-signal source-separation method an dare found to occur in the expected frequency range. The research described herein is aligned with the NASA Ultra-Efficient Commercial Transport strategic thrust and is supported by the NASA Advanced Air Vehicle Program, Advanced Air Transport Technology Project, under the Aircraft Noise Reduction Subproject.

The validity and reliability of a dynamic neuromuscular stabilization-heel sliding test for core stability.

Science.gov (United States)

Cha, Young Joo; Lee, Jae Jin; Kim, Do Hyun; You, Joshua Sung H

2017-10-23

Core stabilization plays an important role in the regulation of postural stability. To overcome shortcomings associated with pain and severe core instability during conventional core stabilization tests, we recently developed the dynamic neuromuscular stabilization-based heel sliding (DNS-HS) test. The purpose of this study was to establish the criterion validity and test-retest reliability of the novel DNS-HS test. Twenty young adults with core instability completed both the bilateral straight leg lowering test (BSLLT) and DNS-HS test for the criterion validity study and repeated the DNS-HS test for the test-retest reliability study. Criterion validity was determined by comparing hip joint angle data that were obtained from BSLLT and DNS-HS measures. The test-retest reliability was determined by comparing hip joint angle data. Criterion validity was (ICC2,3) = 0.700 (preliability was (ICC3,3) = 0.953 (pvalidity data demonstrated a good relationship between the gold standard BSLLT and DNS-HS core stability measures. Test-retest reliability data suggests that DNS-HS core stability was a reliable test for core stability. Clinically, the DNS-HS test is useful to objectively quantify core instability and allow early detection and evaluation.

Transitions to improved core electron heat confinement triggered by low order rational magnetic surfaces in the stellarator TJ-II

International Nuclear Information System (INIS)

Estrada, T.; Medina, F.; Lopez-Bruna, D.; AscasIbar, E.; BalbIn, R.; Cappa, A.; Castejon, F.; Eguilior, S.; Fernandez, A.; Guasp, J.; Hidalgo, C.; Petrov, S.

2007-01-01

Transitions to improved core electron heat confinement are triggered by low order rational magnetic surfaces in TJ-II electron cyclotron heated (ECH) plasmas. Experiments are performed changing the magnetic shear around the rational surface n = 3/m = 2 to study its influence on the transition; ECH power modulation is used to look at transport properties. The improvement in the electron heat confinement shows no obvious dependence on the magnetic shear. Transitions triggered by the rational surface n = 4/m = 2 show, in addition, an increase in the ion temperature synchronized with the increase in the electron temperature. Ion temperature changes had not been previously observed either in TJ-II or in any other helical device. SXR measurements demonstrate that, under certain circumstances, the rational surface positioned inside the plasma core region precedes and provides a trigger for the transition

Sport-specific endurance plank test for evaluation of global core muscle function.

Science.gov (United States)

Tong, Tom K; Wu, Shing; Nie, Jinlei

2014-02-01

To examine the validity and reliability of a sports-specific endurance plank test for the evaluation of global core muscle function. Repeated-measures study. Laboratory environment. Twenty-eight male and eight female young athletes. Surface electromyography (sEMG) of selected trunk flexors and extensors, and an intervention of pre-fatigue core workout were applied for test validation. Intraclass correlation coefficient (ICC), coefficient of variation (CV), and the measurement bias ratio */÷ ratio limits of agreement (LOA) were calculated to assess reliability and measurement error. Test validity was shown by the sEMG of selected core muscles, which indicated >50% increase in muscle activation during the test; and the definite discrimination of the ∼30% reduction in global core muscle endurance subsequent to a pre-fatigue core workout. For test-retest reliability, when the first attempt of three repeated trials was considered as familiarisation, the ICC was 0.99 (95% CI: 0.98-0.99), CV was 2.0 ± 1.56% and the measurement bias ratio */÷ ratio LOA was 0.99 */÷ 1.07. The findings suggest that the sport-specific endurance plank test is a valid, reliable and practical method for assessing global core muscle endurance in athletes given that at least one familiarisation trial takes place prior to measurement. Copyright © 2013 Elsevier Ltd. All rights reserved.

OECD MCCI project 2-D Core Concrete Interaction (CCI) tests : CCI-3 test data report-thermalhydraulic results. Rev. 0 October 15, 2005.

Energy Technology Data Exchange (ETDEWEB)

Farmer, M. T.; Lomperski, S.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

2011-05-23

The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of a third long-term 2-D Core-Concrete Interaction (CCI) experiment designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-3 experiment, which was conducted on September 22, 2005. Test specifications for CCI-3 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 375

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...

Preliminary scoping safety analyses of the limiting design basis protected accidents for the Fast Flux Test Facility tritium production core

International Nuclear Information System (INIS)

Heard, F.J.

1997-01-01

The SAS4A/SASSYS-l computer code is used to perform a series of analyses for the limiting protected design basis transient events given a representative tritium and medical isotope production core design proposed for the Fast Flux Test Facility. The FFTF tritium and isotope production mission will require a different core loading which features higher enrichment fuel, tritium targets, and medical isotope production assemblies. Changes in several key core parameters, such as the Doppler coefficient and delayed neutron fraction will affect the transient response of the reactor. Both reactivity insertion and reduction of heat removal events were analyzed. The analysis methods and modeling assumptions are described. Results of the analyses and comparison against fuel pin performance criteria are presented to provide quantification that the plant protection system is adequate to maintain the necessary safety margins and assure cladding integrity

Development of JOYO MK-II core characteristics database

International Nuclear Information System (INIS)

Tabuchi, Shiro; Aoyama, Takafumi

2000-01-01

The MK-II core of the experimental fast reactor JOYO served as the irradiation bed for testing fuels and materials for FBR development since 1982 for 15 years. During the MK-II operation, extensive data were accumulated from the core management calculations and characteristics tests conducted in thirty-one duty operations and thirteen special test operations. These core management data and core characteristics data were compiled into a database recorded on CD-ROM for user convenience. The calculated core management data are the text style data. The 'Configuration Data' include the history of the fuel exchange and core arrangement for each cycle. The Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of about 300 fuel subassemblies, and 60 irradiation subassemblies. The 'Output Data' include the neutron fluxes, gamma fluxes, power density, linear heat rates, coolant and fuel temperature distributions of each core position at the beginning and end of each cycle. The measured core characteristics data, such as the excess reactivity, control rod worths, temperature coefficient, power coefficient, and burn-up coefficient are also included along with the measurement conditions. (J.P.N.)

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

Modelling guidelines for core exit temperature simulations with system codes

Energy Technology Data Exchange (ETDEWEB)

Freixa, J., E-mail: jordi.freixa-terradas@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Martínez-Quiroga, V., E-mail: victor.martinez@nortuen.com [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Zerkak, O., E-mail: omar.zerkak@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Reventós, F., E-mail: francesc.reventos@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain)

2015-05-15

Highlights: • Core exit temperature is used in PWRs as an indication of core heat up. • Modelling guidelines of CET response with system codes. • Modelling of heat transfer processes in the core and UP regions. - Abstract: Core exit temperature (CET) measurements play an important role in the sequence of actions under accidental conditions in pressurized water reactors (PWR). Given the difficulties in placing measurements in the core region, CET readings are used as criterion for the initiation of accident management (AM) procedures because they can indicate a core heat up scenario. However, the CET responses have some limitation in detecting inadequate core cooling and core uncovery simply because the measurement is not placed inside the core. Therefore, it is of main importance in the field of nuclear safety for PWR power plants to assess the capabilities of system codes for simulating the relation between the CET and the peak cladding temperature (PCT). The work presented in this paper intends to address this open question by making use of experimental work at integral test facilities (ITF) where experiments related to the evolution of the CET and the PCT during transient conditions have been carried out. In particular, simulations of two experiments performed at the ROSA/LSTF and PKL facilities are presented. The two experiments are part of a counterpart exercise between the OECD/NEA ROSA-2 and OECD/NEA PKL-2 projects. The simulations are used to derive guidelines in how to correctly reproduce the CET response during a core heat up scenario. Three aspects have been identified to be of main importance: (1) the need for a 3-dimensional representation of the core and Upper Plenum (UP) regions in order to model the heterogeneity of the power zones and axial areas, (2) the detailed representation of the active and passive heat structures, and (3) the use of simulated thermocouples instead of steam temperatures to represent the CET readings.

Thermal performance of Egypt's research reactor core (ET-RR-1)

International Nuclear Information System (INIS)

Khattab, M.; Mariy, A.

1986-01-01

The steady state thermal performance of the ET-RR-1 core system is theoretically investigated by different models describing the heat flux and the coolant mass flow rate. The magnitude of the heat generated by a fuel element depends upon its position in the core. Normal and uniform distributions for heat flux and coolant mass flow rate are considered. The clad and coolant temperatures at different core positions are evaluated and compared with the experimental measurements at different operating conditions. The results indicated large discrepancy between the predicted and the experimental results. Therefore, the previous models and the experimental results are evaluated in order to develop the best model that describes the thermal performance of the ET-RR-1 core. The adapted model gives 99.5% significant confidence limit. The effect of increasing the heat flux or decreasing the mass flow rate by 20% from its maximum recommended operating condition is tested and discussed. Also, the thermal behaviour towards increasing the reactor power more than its maximum operating condition is discussed. The present work could also be used in extending the investigation to other PWR reactor operating conditions

Experiment data report for Semiscale Mod-1 Test S-05-2 (alternate ECC injection test)

International Nuclear Information System (INIS)

Feldman, E.M.; Collins, B.L.; Sackett, K.E.

1977-02-01

Recorded test data are presented for Test S-05-2 of the Semiscale Mod-1 alternate emergency core coolant (ECC) injection test series. This test is one of several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-2 was conducted from an initial cold leg fluid temperature of 545 0 F and an initial pressure of 2263 psia. A simulated double-ended offset shear cold leg break was used to investigate core and system response to a depressurization and reflood transient with ECC injection at the intact loop pump suction and broken loop cold leg. A reduced lower plenum volume was used for this test to more accurately represent the lower plenum of a PWR, based on system volume scaling. System flow was set to achieve a core fluid temperature differential of 65 0 F at a core power level of 1.44 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a slightly peaked radial power profile was used in the pressure vessel to simulate the predicted surface heat flux of nuclear fuel rods during a loss-of-coolant accident

Thermal-hydraulic analysis of the semiscale Mod-1 blowdown heat transfer test series

International Nuclear Information System (INIS)

Cozzuol, J.M.

1976-06-01

Selected experimental thermal-hydraulic data from the recent Semiscale Mod-1 blowdown heat transfer test series are analyzed from an experimental viewpoint with emphasis on explaining those phenomena which influence core fluid behavior. Comparisons are made between the trends measured by the system instrumentation and the trends predicted by the RELAP4 computer code to aid in obtaining an understanding of the interactions between phenomena occurring in different parts of the system. The analyses presented in this report are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict the system response of a pressurized water reactor during a postulated loss-of-coolant accident

Analysis of select Mod-1 semiscale blowdown heat transfer tests. Final report

International Nuclear Information System (INIS)

Irani, A.A.; Fujita, N.; Mecham, D.C.; Ching, J.T.; Gose, G.C.; Hentzen, R.D.; Sawtelle, G.R.; Moore, K.V.

1976-10-01

The report contains the RELAP4 analysis and sensitivity studies of Semiscale Tests S-02-2 and S-02-7. The Semiscale System is an electrically heated experiment designed to produce data on system performance typical of PWR thermal-hydraulic behavior. The RELAP4 program used for these analyses is a digital computer program developed to predict the thermal-hydraulic behavior of experimental systems and water-cooled nuclear reactors subjected to postulated transients. The results of the analysis for Test S-02-2 were in very good agreement with the data. Two parameters which required improvement were identified. These were the lower plenum density and the mass flow on the vessel side of the break. Subsequently, before analyzing Test S-02-7, the lower plenum was renodalized and the critical flow model at the vessel side break was modified. The results of the analysis of Test S-02-7 compared more favorably with the data than those of S-02-2. Additional sensitivity studies included time step studies, steam generator and downcomer modeling, and core nodalization

Space Launch System Base Heating Test: Experimental Operations & Results

Science.gov (United States)

Dufrene, Aaron; Mehta, Manish; MacLean, Matthew; Seaford, Mark; Holden, Michael

2016-01-01

NASA's Space Launch System (SLS) uses four clustered liquid rocket engines along with two solid rocket boosters. The interaction between all six rocket exhaust plumes will produce a complex and severe thermal environment in the base of the vehicle. This work focuses on a recent 2% scale, hot-fire SLS base heating test. These base heating tests are short-duration tests executed with chamber pressures near the full-scale values with gaseous hydrogen/oxygen engines and RSRMV analogous solid propellant motors. The LENS II shock tunnel/Ludwieg tube tunnel was used at or near flight duplicated conditions up to Mach 5. Model development was based on the Space Shuttle base heating tests with several improvements including doubling of the maximum chamber pressures and duplication of freestream conditions. Test methodology and conditions are presented, and base heating results from 76 runs are reported in non-dimensional form. Regions of high heating are identified and comparisons of various configuration and conditions are highlighted. Base pressure and radiometer results are also reported.

Summary of experimental core turbulence characteristics in ohmic and electron cyclotron resonance heated discharges in T-10 tokamak plasmas

International Nuclear Information System (INIS)

Vershkov, V.A.; Shelukhin, D.A.; Soldatov, S.V.; Urazbaev, A.O.; Grashin, S.A.; Eliseev, L.G.; Melnikov, A.V.

2005-01-01

This report summarizes the results of experimental turbulence investigations carried out at T-10 for more than 10 years. The turbulence characteristics were investigated using correlation reflectometry, multipin Langmuir probe (MLP) and heavy ion beam probe diagnostics. The reflectometry capabilities were analysed using 2D full-wave simulations and verified by direct comparison using a MLP. The ohmic and electron cyclotron resonance heated discharges show the distinct transition from the core turbulence, having complex spectral structure, to the unstructured one in the scrape-off layer. The core turbulence includes 'broad band, quasi-coherent' features, arising due to the excitation of rational surfaces with high poloidal m-numbers, with a low frequency near zero and specific oscillations at 15-30 kHz. All experimentally measured properties of low frequency and high frequency quasi-coherent oscillations are in good agreement with predictions of linear theory for the ion temperature gradient/dissipative trapped electron mode instabilities. Significant local changes in the turbulence characteristics were observed at the edge velocity shear layer and in the core near q = 1 radius after switching off the electron cyclotron resonance heating (ECRH). The local decrease in the electron heat conductivity and decrease in the turbulence level could be evidence of the formation of an electron internal transport barrier. The dynamic behaviour of the core turbulence was also investigated for the case of fast edge cooling and the beginning phase of ECRH

Natural Convection Heat Transfer of Oxide Pool During In-Vessel Retention of Core Melts

Energy Technology Data Exchange (ETDEWEB)

Park, Hae-Kyun; Chung, Bum-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The integrity of reactor vessel may be threatened by the heat generation at the oxide pool and to the natural convection heat transfer to the reactor vessel by those two layers. Therefore, External Reactor Vessel Cooling (ERVC) is performed in order to secure the integrity of the reactor vessel. Whether the IVR(In-Vessel Retention) Strategy can be applicable to a larger reactor is the technical concern, which nourished the research interest for the natural convection heat transfer of metal and oxide pool and ERVC performance. Especially, it is hard to simulate oxide pool by experimentally due to the high level of buoyancy. Moreover, the volumetrically exothermic working fluid should be adopted to simulate the behavior of the core melts. Therefore, the volumetric heat sources that immersed in the working fluid have been adopted to simulate oxide pool by experiment. We investigated oxide pool with two different designs of the volumetric heat sources that adopted previous experiments. The investigation was performed by mass transfer experiment using analogy between heat and mass transfers. The results were compared to previous studies. We simulated the natural convection heat transfer of the oxide pool by mass transfer experiment. The isothermally cooled condition was established by limiting current technique firstly. The results were compared to previous studies under identical design of the volumetric heat sources. The average Nu's of the curvature and the top plate were close to the previous studies.

Testing and performance analysis of a hollow fiber-based core for evaporative cooling and liquid desiccant dehumidification

DEFF Research Database (Denmark)

Jradi, Muhyiddine; Riffat, Saffa

2016-01-01

In this study, an innovative heat and mass transfer core is proposed to provide thermal comfort and humidity control using a hollow fiber contactor with multiple bundles of micro-porous hollow fibers. The hollow fiberbased core utilizes 12 bundles aligned vertically, each with 1,000 packed...

Overview report of RAMONA-NEPTUN program on passive decay heat removal

International Nuclear Information System (INIS)

Weinberg, D.; Rust, K.; Hoffmann, H.

1996-03-01

The design of the advanced sodium-cooled European Fast Reactor provides a safety graded decay heat removal concept which ensures the coolability of the primary system by natural convection when forced cooling is lost. The findings of the RAMONA and NEPTUN experiments indicate that the decay heat can be safely removed by natural convection. The operation of the decay heat exchangers being installed in the upper plenum causes the formation of a thermal stratification associated with a pronounced temperature gradient. The vertical extent of the stratification and the qualitity of the gradient are depending on the fact whether a permeable or an impermeable shell covers the above core structure. A delayed startup time of the decay heat exchangers leads only to a slight increase of the temperatures in the upper plenum. A complete failure of half of the decay heat exchangers causes a higher temperature level in the primary system, but does not alter the global temperature distribution. The transient development of the temperatures is faster going on in a three-loop model than in a four-loop model due to the lower amount of heat stored in the compacter primary vessel. If no coolant reaches the core inlet side via the intermediate heat exchangers, the core remains coolable. In this case, cold water of the upper plenum penetrates into the subassemblies (thermosyphon effects) and the interwrapper spaces existing in the NEPTUN core. The core coolability from above is feasible without any difficulty though the temperatures increase to a minor degree at the top end of the core. The thermal hydraulic computer code FLUTAN was applied for the 3D numerical simulation of the majority of the steady state RAMONA and NEPTUN tests as well as for selected transient RAMONA tests. (orig./HP) [de

Two-dimensional horizontal model seismic test and analysis for HTGR core

International Nuclear Information System (INIS)

Ikushima, Takeshi; Honma, Toshiaki.

1988-05-01

The resistance against earthquakes of high-temperature gas-cooled reactor (HTGR) core with block-type fuels is not fully ascertained yet. Seismic studies must be made if such a reactor plant is to be installed in areas with frequent earthquakes. The paper presented the test results of seismic behavior of a half scale two-dimensional horizontal slice core model and analysis. The following is a summary of the more important results. (1) When the core is subjected to the single axis excitation and simultaneous two-axis excitations to the core across-corners, it has elliptical motion. The core stays lumped motion at the low excitation frequencies. (2) When the load is placed on side fixed reflector blocks from outside to the core center, the core displacement and reflector impact reaction force decrease. (3) The maximum displacement occurs at simultaneous two-axis excitations. The maximum displacement occurs at the single axis excitation to the core across-flats. (4) The results of two-dimensional horizontal slice core model was compared with the results of two-dimensional vertical one. It is clarified that the seismic response of actual core can be predicted from the results of two-dimensional vertical slice core model. (5) The maximum reflector impact reaction force for seismic waves was below 60 percent of that for sinusoidal waves. (6) Vibration behavior and impact response are in good agreement between test and analysis. (author)

Simulation of a hypothetical core disruptive accident in the mars test-facility

International Nuclear Information System (INIS)

Robbe, M.F.; Lepareux, M.

2001-01-01

In France, a large experimental programme MARA/MARS was undertaken in the 80's to estimate the mechanical consequences of an HCDA (Hypothetical Core Disruptive Accident) and to validate the SIRIUS computer code used at that time for the numerical simulations. At the end of the 80's, it was preferred to add a HCDA sodium-bubble-argon tri-component constitutive law to the general ALE fast dynamics finite element CASTEM-PLEXUS code rather than going on developing and using the specialized SIRIUS code. The experimental results of the MARA programme were used in the 90's to validate and qualify the CASTEM-PLEXUS code. A first series of computations of the tests MARA 8, MARA 10 and MARS was realised. The simulations showed a rather good agreement between the experimental and computed results for the MARA 8 and MARA 10 tests - even if there were some discrepancies - but the prediction of the MARS structure displacements and strains was overestimated. This conservatism was supposed to come from the fact that several MARS non axisymmetric structures like core elements, pumps and heat exchangers were not represented in the CASTEM-PLEXUS model. These structures, acting as porous barriers, had a protective effect on the mock-up containment by absorbing energy and slowing down the fluid impacting the containment. For these reasons, we developed in CASTEM-PLEXUS a new HCDA constitutive law taking into account the presence of the internal structures (without meshing them) by means of an equivalent porosity method. In other respects, the process used for dealing with the fluid-structure coupling in CASTEM-PLEXUS was improved. Thus a second series of simulations of the tests MARA8 and MARA10 was realised. A simulation of the test MARS was carried out too with the same simplified representation of the peripheral structures as in order to estimate the improvement provided by the new fluid-structure coupling. This paper presents a third numerical simulation of the MARS test with the

Assessment of Two-Phase Flow Heat Transfer Correlations for Molten Core-Concrete Interaction Study

International Nuclear Information System (INIS)

Tourniaire, B.; Varo, O.

2006-01-01

The prediction of heat transfer between corium pool and concrete basemat is of particular significance in the framework of the study of PWR's severe accident. Heat transfer directly governs the ablation velocity of concrete in case of molten core-concrete interaction (MCCI) and, consequently, the time delay when the reactor cavity may fail. From a restricted hydrodynamic point of view, this issue is related to heat transfer between a heated bubbling pool and a porous wall with gas injection. Several experimental studies have been performed with simulant materials and many correlations have been provided to address this issue. The main purpose of this paper is to assess these correlations from comparisons against the available experimental data. After a review of these data, the different correlations are presented. Attention focuses here on the correlations generally used in MCCI study: Kutateladze-Malenkov, Konsetov and BALI correlations. The Deckwer's correlation is also included in this review. The comparisons between the results of these correlations and the experimental data are then discussed. (authors)

Evaluation method for core thermohydraulics during natural circulation in fast reactors numerical predictions of inter-wrapper flow

International Nuclear Information System (INIS)

Kamide, H.; Kimura, N.; Miyakoshi, H.; Nagasawa, K.

2001-01-01

Decay heat removal using natural circulation is one of the important functions for the safety of fast reactors. As a decay heat removal system, direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this design, dumped heat exchanger provides cold sodium and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such natural circulation phenomena in a reactor core has been developed, which models each subassembly as a rectangular duct with gap region between the subassemblies and also the upper plenum in a reactor vessel. This numerical simulation method was verified based on experimental data of a sodium test using 7- subassembly core model and also a water test which simulates IWF using the 1/12 sector model of a reactor core. We applied the estimation method to the natural circulation in a 600 MW class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer. It is desired for prediction methods on the natural circulation to simulate these phenomena. (author)

Flow visualization study of inverted annular flow of post dryout heat transfer region

International Nuclear Information System (INIS)

Ishii, M.; De Jarlais, G.

1985-01-01

The inverted annular flow is important in the area of LWR accident analysis in terms of the maximum cladding temperature and effectiveness of the emergency core cooling. However, the inverted annular flow thermal-hydraulics is not well understood due to its special heat transfer condition of film boiling. In view of this, the inverted flow is studied in detail experimentally. A new experimental apparatus has been constructed in which film boiling heat transfer can be established in a transparent test section. Data on liquid core stability, core break-up mechanism, and dispersed-core liquid slug and droplet sizes are obtained using F 113 as a test fluid. Both high speed movies and flash photographs are used

The validation of science virtual test to assess 7th grade students’ critical thinking on matter and heat topic (SVT-MH)

Science.gov (United States)

Sya’bandari, Y.; Firman, H.; Rusyati, L.

2018-05-01

The method used in this research was descriptive research for profiling the validation of SVT-MH to measure students’ critical thinking on matter and heat topic in junior high school. The subject is junior high school students of 7th grade (13 years old) while science teacher and expert as the validators. The instruments that used as a tool to obtain the data are rubric expert judgment (content, media, education) and rubric of readability test. There are four steps to validate SVT-MH in 7th grade Junior High School. These steps are analysis of core competence and basic competence based on Curriculum 2013, expert judgment (content, media, education), readability test and trial test (limited and larger trial test). The instrument validation resulted 30 items that represent 8 elements and 21 sub-elements to measure students’ critical thinking based on Inch in matter and heat topic. The alpha Cronbach (α) is 0.642 which means that the instrument is sufficient to measure students’ critical thinking matter and heat topic.

Experiment data report for semiscale Mod-1 test S-04-1 (baseline ECC test)

International Nuclear Information System (INIS)

Crapo, H.S.; Collins, B.L.; Sackett, K.E.

1976-09-01

Recorded test data are presented for Test S-04-1 of the Semiscale Mod-1 Baseline ECC Test Series. This test is among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor system. Test S-04-1 was conducted from an initial cold leg fluid temperature of 542 0 F and an initial pressure of 2,263 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization and reflood transient using system volume scaled coolant injection parameters. System flow was set to achieve a core fluid temperature differential of 66 0 F at a full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW in such a manner as to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core cooling injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown

Heat-pipe thermionic reactor concept

DEFF Research Database (Denmark)

Storm Pedersen, E.

1967-01-01

Main components are reactor core, heat pipe, thermionic converter, secondary cooling system, and waste heat radiator; thermal power generated in reactor core is transported by heat pipes to thermionic converters located outside reactor core behind radiation shield; thermionic emitters are in direct...

Heat, hydrogen peroxide, and UV resistance of Bacillus subtilis spores with increased core water content and with or without major DNA-binding proteins

International Nuclear Information System (INIS)

Popham, D.L.; Sengupta, S.; Setlow, P.

1995-01-01

Spores of a Bacillus subtilis strain with an insertion mutation in the dacB gene, which codes for an enzyme involved in spore cortex biosynthesis, have a higher core water content than wild-type spores. Spores lacking the two major α/β-type small, acid-soluble proteins (SASP) (termed a α - β - spores) have the same core water content as do wild-type spores, but α - β - dacB spores had more core water than did dacB spores. The resistance of α - β - , α - β - dacB, dacB, and wild-type spores to dry and moist heat, hydrogen peroxide, and UV radiation has been determined, as has the role of DNA damage in spore killing by moist heat and hydrogen peroxide. These data (1) suggest that core water content has little if any role in spore UV resistance and are consistent with binding of α/β-type SASP to DNA being the major mechanism providing protection to spores from UV radiation; (2) suggest that binding of αβ-type SASP to DNA is the major mechanism unique to spores providing protection from dry heat; (3) suggest that spore resistance to moist heat and hydrogen peroxide is affected to a large degree by the core water content, as increased core water resulted in large decreases in spore resistance to these agents; and (4) indicate that since this decreased resistance (i.e., in dacB spores) is not associated with increased spore killing by DNA damage, spore DNA must normally be extremely well protected against such damage, presumably by the saturation of spore DNA by α/β-type SASP. 19 refs., 2 figs., 5 tabs

Evaluation report on CCTF Core-II reflood test C2-1 (Run 55)

International Nuclear Information System (INIS)

Iguchi, Tadashi; Sugimoto, Jun; Akimoto, Hajime; Okubo, Tsutomu; Murao, Yoshio

1991-10-01

A high pressure test (0.42 MPa) on the reflood phenomena was performed with the CCTF. The result of the test was compared with the experimental result of the base case test (0.2 MPa). (1) The overall flow characteristics in the high pressure test were qualitatively similar to that of the base case test. Any qualitatively different phenomena were not recognized during reflood phase. This indicates that it is reasonable to utilize the physical reflood model developed from the result of the base case test to the high pressure condition at least up to 0.42 MPa for prediction of reflood behavior of PWRs. (2) On the other hand, following quantitative influence of high pressure on reflood phenomena was observed. The core cooling was better, and the mass flow rate of the steam generated in the core was larger. However, the steam velocity was smaller due to higher density of the steam. Therefore, the steam discharge through loops was easier and hence the so-called steam binding effect was weaker. And, the water accumulation rate in the core was larger. Consequently the core flooding mass flow rate was larger. Since the core cooling was better, the maximum core temperature was lower and the last quenching was earlier. This result was the same as that previously observed in CCTF tests in the scope of the pressure upto 0.3 MPa. (3) The higher pressure leads to the better core cooling, and hence the safety margin increases with the increase in the pressure. (author)

Flow visualization study of inverted annular flow of post dryout heat transfer region

International Nuclear Information System (INIS)

Ishii, M.; De Jarlais, G.

1985-01-01

The inverted annular flow is important in the area of LWR accident analysis in terms of the maximum cladding temperature and effectiveness of the emergency core cooling. However, the inverted annular flow thermal-hydraulics is not well understood due to its special heat transfer condition of film boiling. The review of existing data indicates further research is needed in the areas of basic hydrodynamics related to liquid core disintegration mechanisms, slug and droplet formation, entrainment, and droplet size distributions. In view of this, the inverted flow is studied in detail experimentally. A new experimental apparatus has been constructed in which film boiling heat transfer can be established in a transparent test section. The test section consists of two coaxial quartz tubes. The annular gap between these two tubes is filled with a hot, clear fluid (syltherm 800) so as to maintain film boiling temperatures and heat transfer rates at the inner quartz tube wall. Data on liquid core stability, core break-up mechanism, and dispersed-core liquid slug and droplet sizes are obtained using F 113 as a test fluid. Both high speed movies and flash photographs (3 μsec) are used

Verification and Validation of Heat Transfer Model of AGREE Code

Energy Technology Data Exchange (ETDEWEB)

Tak, N. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Seker, V.; Drzewiecki, T. J.; Downar, T. J. [Department of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Michigan (United States); Kelly, J. M. [US Nuclear Regulatory Commission, Washington (United States)

2013-05-15

The AGREE code was originally developed as a multi physics simulation code to perform design and safety analysis of Pebble Bed Reactors (PBR). Currently, additional capability for the analysis of Prismatic Modular Reactor (PMR) core is in progress. Newly implemented fluid model for a PMR core is based on a subchannel approach which has been widely used in the analyses of light water reactor (LWR) cores. A hexagonal fuel (or graphite block) is discretized into triangular prism nodes having effective conductivities. Then, a meso-scale heat transfer model is applied to the unit cell geometry of a prismatic fuel block. Both unit cell geometries of multi-hole and pin-in-hole types of prismatic fuel blocks are considered in AGREE. The main objective of this work is to verify and validate the heat transfer model newly implemented for a PMR core in the AGREE code. The measured data in the HENDEL experiment were used for the validation of the heat transfer model for a pin-in-hole fuel block. However, the HENDEL tests were limited to only steady-state conditions of pin-in-hole fuel blocks. There exist no available experimental data regarding a heat transfer in multi-hole fuel blocks. Therefore, numerical benchmarks using conceptual problems are considered to verify the heat transfer model of AGREE for multi-hole fuel blocks as well as transient conditions. The CORONA and GAMMA+ codes were used to compare the numerical results. In this work, the verification and validation study were performed for the heat transfer model of the AGREE code using the HENDEL experiment and the numerical benchmarks of selected conceptual problems. The results of the present work show that the heat transfer model of AGREE is accurate and reliable for prismatic fuel blocks. Further validation of AGREE is in progress for a whole reactor problem using the HTTR safety test data such as control rod withdrawal tests and loss-of-forced convection tests.

Space Launch System, Core Stage, Structural Test Design and Implementation

Science.gov (United States)

Shaughnessy, Ray

2017-01-01

As part of the National Aeronautics and Space Administration's (NASA) Space Launch System (SLS) Program, engineers at NASA's Marshall Space Flight Center (MSFC) in Huntsville, Alabama are working to design, develop and implement the SLS Core Stage structural testing. The SLS will have the capability to return humans to the Moon and beyond and its first launch is scheduled for December of 2017. The SLS Core Stage consist of five major elements; Forward Skirt, Liquid Oxygen (LOX) tank, Intertank (IT), Liquid Hydrogen (LH2) tank and the Engine Section (ES). Structural Test Articles (STA) for each of these elements are being designed and produced by Boeing at Michoud Assembly Facility located in New Orleans, La. The structural test for the Core Stage STAs (LH2, LOX, IT and ES) are to be conducted by the MSFC Test Laboratory. Additionally, the MSFC Test Laboratory manages the Structural Test Equipment (STE) design and development to support the STAs. It was decided early (April 2012) in the project life that the LH2 and LOX tank STAs would require new test stands and the Engine Section and Intertank would be tested in existing facilities. This decision impacted schedules immediately because the new facilities would require Construction of Facilities (C of F) funds that require congressional approval and long lead times. The Engine Section and Intertank structural test are to be conducted in existing facilities which will limit lead times required to support the first launch of SLS. With a SLS launch date of December, 2017 Boeing had a need date for testing to be complete by September of 2017 to support flight certification requirements. The test facilities were required to be ready by October of 2016 to support test article delivery. The race was on to get the stands ready before Test Article delivery and meet the test complete date of September 2017. This paper documents the past and current design and development phases and the supporting processes, tools, and

Data on test results of vessel cooling system of high temperature engineering test reactor

International Nuclear Information System (INIS)

Saikusa, Akio; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

2003-02-01

High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

Control of the ASTRA decay heat removal system

International Nuclear Information System (INIS)

Nedelik, A.

1982-11-01

To ensure a minimum of core cooling even under severest accident conditions (loss of reactor pool water) a core spray system for decay heat removal has been installed at the ASTRA-reactor. The automatic and manual control of the system, its power supply and test procedures are shortly described. (Author)

Heat Storage Performance of the Prefabricated Hollow Core Concrete Deck Element with Integrated Microencapsulated Phase Change Material

DEFF Research Database (Denmark)

Pomianowski, Michal Zbigniew; Heiselberg, Per; Jensen, Rasmus Lund

2012-01-01

The paper presents the numerically calculated dynamic heat storage capacity of the prefabricated hollow core concrete deck element with and without microencapsulated phase change material (PCM). The reference deck is the ordinary deck made of standard concrete material and that is broadly used...

Static analysis of material testing reactor cores:critical core calculations

International Nuclear Information System (INIS)

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

1999-01-01

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

Verification of in-core thermal and hydraulic analysis code FLOWNET/TRUMP for the high temperature engineering test reactor (HTTR) at JAERI

International Nuclear Information System (INIS)

Maruyama, Soh; Sudo, Yukio; Saito, Shinzo; Kiso, Yoshihiro; Hayakawa, Hitoshi

1989-01-01

The FLOWNET/TRUMP code consists of a flow network analysis code 'FLOWNET' for calculations of coolant flow distribution and coolant temperature distribution in the core with a thermal conduction analysis code 'TRUMP' for calculation of temperature distribution in solid structures. The verification of FLOWNET/TRUMP was made by the comparison of the analytical results with the results of steady state experiments by the HENDEL multichannel test rig, T1-M, which consisted of twelve simulated fuel rods heated electrically and eleven hexagonal graphite fuel blocks. The T1-M simulated the one fuel column in the core. The analytical results agreed well with the results of the experiment in which the HTTR operating conditions were simulated. (orig.)

Ice-sheet flow conditions deduced from mechanical tests of ice core

DEFF Research Database (Denmark)

Miyamoto, Atsushi; Narita, Hideki; Hondoh, Takeo

1999-01-01

Uniaxial compression tests were performed on samples of the Greenland Ice Core Project (GRIP) deep ice core, both in the field and later in a cold-room laboratory, in order to understand the ice-flow behavior of large ice sheets. Experiments were conducted under conditions of constant strain rate....... It was revealed that cloudy bands affect ice-deformation processes, but the details remain unclear. Udgivelsesdato: June......Uniaxial compression tests were performed on samples of the Greenland Ice Core Project (GRIP) deep ice core, both in the field and later in a cold-room laboratory, in order to understand the ice-flow behavior of large ice sheets. Experiments were conducted under conditions of constant strain rate......-core samples with basal planes parallel to the horizontal plane of the ice sheet. The ice-flow enhancement factors show a gradual increase with depth down to approximately 2000 m. These results can be interpreted in terms of an increase in the fourth-order Schmid factor. Below 2000 m depth, the flow...

Methodology for Life Testing of Refractory Metal / Sodium Heat Pipes

International Nuclear Information System (INIS)

Martin, James J.; Reid, Robert S.

2006-01-01

This work establishes an approach to generate carefully controlled data to find heat pipe operating life with material-fluid combinations capable of extended operation. To accomplish this goal acceleration is required to compress 10 years of operational life into 3 years of laboratory testing through a combination of increased temperature and mass fluence. Specific test series have been identified, based on American Society for Testing and Materials (ASTM) specifications, to investigate long-term corrosion rates. The refractory metal selected for demonstration purposes is a molybdenum-44.5% rhenium alloy formed by powder metallurgy. The heat pipes each have an annular crescent wick formed by hot isostatic pressing of molybdenum-rhenium wire mesh. The heat pipes are filled by vacuum distillation with purity sampling of the completed assembly. Round-the-clock heat pipe tests with 6-month destructive and non-destructive inspection intervals are conducted to identify the onset and level of corrosion. Non-contact techniques are employed to provide power to the evaporator (radio frequency induction heating at 1 to 5 kW per heat pipe) and calorimetry at the condenser (static gas gap coupled water cooled calorimeter). The planned operating temperature range extends from 1123 to 1323 K. Accomplishments before project cancellation included successful development of the heat pipe wick fabrication technique, establishment of all engineering designs, baseline operational test requirements, and procurement/assembly of supporting test hardware systems. (authors)

Test results from a helium gas-cooled porous metal heat exchanger

International Nuclear Information System (INIS)

North, M.T.; Rosenfeld, J.H.; Youchison, D.L.

1996-01-01

A helium-cooled porous metal heat exchanger was built and tested, which successfully absorbed heat fluxes exceeding all previously tested gas-cooled designs. Helium-cooled plasma-facing components are being evaluated for fusion applications. Helium is a favorable coolant for fusion devices because it is not a plasma contaminant, it is not easily activated, and it is easily removed from the device in the event of a leak. The main drawback of gas coolants is their relatively poor thermal transport properties. This limitation can be removed through use of a highly efficient heat exchanger design. A low flow resistance porous metal heat exchanger design was developed, based on the requirements for the Faraday shield for the International Thermonuclear Experimental Reactor (ITER) device. High heat flux tests were conducted on two representative test articles at the Plasma Materials Test Facility (PMTF) at Sandia National Laboratories. Absorbed heat fluxes as high as 40 MW/m 2 were successfully removed during these tests without failure of the devices. Commercial applications for electronics cooling and other high heat flux applications are being identified

Formation of core transport barrier and CH-Mode by ion Bernstein wave heating in PBX-M

International Nuclear Information System (INIS)

Ono, M.; Bell, R.; Bernabei, S.; Gettelfinger, G.; Hatcher, R.; Kaita, R.; Kaye, S.; Kugel, H.; LeBlanc, B.; Manickam, J.

1995-01-01

Observation of core transport barrier formation (for particles, ion and electron energies, and toroidal momentum) by ion Bernstein wave heating (IBWH) in PBX-M plasma is reported. The formation of a transport barrier leads to a strong peaking and significant increase of the core pressure (70%) and toroidal momentum (20%), and has been termed the core-high confinement mode (CH-Mode). This formation of a transport barrier is consistent, in terms of the expected barrier location as well as the required threshold power, with a theoretical model based on the poloidal sheared flow generation by the ion Bernstein wave power. The use of ion Bernstein wave (IBW) induced sheared flow as a tool to control plasma pressure and bootstrap current profiles shows a favorable scaling for the use in future reactor grade tokamak plasmas

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...

Study on practical of eddy current testing of core and core support graphite components in HTTR

International Nuclear Information System (INIS)

Ishihara, Masahiro; Iyoku, Tatsuo; Ooka, Norikazu; Shindo, Yoshihisa; Kawae, Hidetoshi; Hayashi, Motomitsu; Kambe, Mamoru; Takahashi, Masaaki; Ide, Akira.

1994-01-01

Core and core support graphite components in the HTTR (High Temperature Engineering Test Reactor) are mainly made of nuclear-grade IG-110 and PGX graphites. Nondestructive inspection with Eddy Current Testing (ECT) is planned to be applied to these components. The method of ECT has been already established for metallic components, however, cannot be applied directly to the graphite ones, because the characteristics of graphite are quite different in micro-structure from those of metals. Therefore, ECT method and condition were studied for the application of the ECT to the graphite components. This paper describes the study on practical method and conditions of ECT for above mentioned graphite structures. (author)

A study of HANARO core conversion using high density U-Mo fuel

International Nuclear Information System (INIS)

Lee, K.H.; Lee, C.S.; Lee, B.C.; Park, S.J.; Kim, H.; Kim, C.K.

2002-01-01

Currently, HANARO is using 3.15gU/cc U3Si/Al as a driver fuel. HANARO has seven vertical irradiation holes in the core region. Three of them including a central trap are located in the inner region of the core and mainly being used for material irradiation tests. Four of them are located in the reflector tank but cooled by primary coolant. They are used for fuel irradiation tests or radioisotope development tests. For minimum core modification using high density U-Mo fuels, no dimension change is assumed in the current fuel rods and the cladding thickness remains the same in this study. The high density U-Mo fuel will have up to about twice the linear uranium loading of a current HANARO driver fuel. Using this high density fuel 8 fuel sites can be replaced with irradiation sites. Three kinds of conceptual cores are considered using 5 gU/cc U-7Mo/Al and 16 gU/cc U-7Mo. The increase of the linear heat generation rate due to the decrease of total fuel length can be overcome by more uniform radial and axial power distribution using different uranium densities and different fuel meat diameters are introduced into those cores. The new core has 4.54 times larger surface-to-volume ratio than the reference core. The core uranium loading, linear heat generation rate, excess reactivity, and control rod worth as well as the neutron spectra are analysed for each core. (author)

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.

Research and development of selected components of the high-temperature reactor for process heat generation - results and their application

International Nuclear Information System (INIS)

Theymann, W.; Lange, G.

1989-01-01

For the process heat supplying high-temperature reactor (PNP) a comprehensive research and development program was performed. Investigations in three fields of the program are reported: heat transfer, gas flow guidance components, and seismic properties of the core structure. Results are presented for the statistics of heat transfer in the core and for heat transfer under operational conditions of a PNP-plant. Further topics are cooling of the side reflector, hot gas mixing in the core bottom region, optimization of inlet flow into the steam generator, and flow tests on a large diameter shut-off valve. Performance tests on hot gas insulations in a special test facility are described as well as tests on connecting elements for coaxial ducts. The measured data on dynamic excitation of the pebble bed with the SAMSON test facility allow an analytical description of the pebble bed core with respect to seismic behaviour. The results of experiments and calculations, using the computer codes CRUNCH-1D and -2D, for seismic excitation of the suspended top reflector are discussed. The seismic tests will be completed in 1989 with the side reflector investigations. A comprehensive seismic verification will then be available. (orig.)

A Simplified Method for Stationary Heat Transfer of a Hollow Core Concrete Slab Used for TABS

DEFF Research Database (Denmark)

Yu, Tao; Heiselberg, Per Kvols; Lei, Bo

2014-01-01

Thermally activated building systems (TABS) have been an energy efficient way to improve the indoor thermal comfort. Due to the complicated structure, heat transfer prediction for a hollow core concrete used for TABS is difficult. This paper proposes a simplified method using equivalent thermal...... resistance for the stationary heat transfer of this kind of system. Numerical simulations are carried out to validate this method, and this method shows very small deviations from the numerical simulations. Meanwhile, this method is used to investigate the influence of the thickness of insulation on the heat...... transfer. The insulation with a thickness of more than 0.06 m can keep over 95 % of the heat transferred from the lower surface, which is beneficial to the radiant ceiling cooling. Finally, this method is extended to involve the effect of the pipe, and the numerical comparison results show that this method...

Heat pumps in field test; Waermepumpen im Feldtest

Energy Technology Data Exchange (ETDEWEB)

Becker, R. [Fraunhofer-Institut fuer Solare Energiesysteme (ISE), Freiburg im Breisgau (Germany); Miara, M.; Russ, C.

2007-09-15

The Fraunhofer ISE has launched two field tests of newly installed heat pumps in 2006. Both deal with the measurement of a high number of heat pump units under real conditions in small houses. Values of volume flows, temperatures, heat quantity and electricity consumption are collected and daily saved and analysed at the Fraunhofer ISE. (orig.)

Heat transfer between immiscible liquids enhanced by gas bubbling

International Nuclear Information System (INIS)

Greene, G.A.; Schwarz, C.E.; Klages, J.; Klein, J.

1982-08-01

The phenomena of core-concrete interactions impact upon containment integrity of light water reactors (LWR) following postulated complete meltdown of the core by containment pressurization, production of combustible gases, and basemat penetration. Experiments have been performed with non-reactor materials to investigate one aspect of this problem, heat transfer between overlying immiscible liquids whose interface is disturbed by a transverse non-condensable gas flux emanating from below. Hydrodynamic studies have been performed to test a criterion for onset of entrainment due to bubbling through the interface and subsequent heat transfer studies were performed to assess the effect of bubbling on interfacial heat transfer rates, both with and without bubble induced entrainment. Non-entraining interfacial heat transfer data with mercury-water/oil fluid pairs were observed to be bounded from below within a factor of two to three by the Szekeley surface renewal heat transfer model. However heat transfer data for fluid pairs which are found to entrain (water-oil), believed to be characteristic of molten reactor core-concrete conditions, were measured to be up to two orders of magnitude greater than surface renewal predictions and are calculated by a simple entrainment heat transfer model

Evaluation report on SCTF Core-III tests S3-7 and S3-8

International Nuclear Information System (INIS)

Okubo, Tsutomu; Iguchi, Tadashi; Iwamura, Takamichi

1990-03-01

It has been said that the Emergency Core Cooling (ECC) water injected into the hot legs flows into the upper plenum and then falls back to the core (i.e. break-through) during reflood phase in a German type Pressurized Water Reactor (GPWR) with the combined-injection-type ECCS, and that the break-through occurs where the water temperature at the tie plate area is lower and subcooled. Based on this information two tests were conducted with the Slab Core Test Facility (SCTF) Core-III in order to investigate the effects of the water temperature distribution at the tie plate area on the break-through and the core cooling. In these tests, the subcooled ECC water was injected just above the Upper Core Support Plate (UCSP) in order to establish the desired water temperature distribution at the tie plate area. In one test (Test S3-7) the ECC water injection above the UCSP was performed above Bundles 3 and 4, and in the other test (Test S3-8) above Bundles 7 and 8 during initial 60 s a and then was changed to above Bundles 3 and 4. The test data were compared with those of Test S3-SH1, in which the injection was performed above Bundles 7 and 8 and the other test conditions were the same as in Tests S3-7 and S3-8. Analyzing these test data, the following has been found: The break-through occurs where the water temperature at the tie plate area is subcooled and the core cooling is enhanced significantly in the break-through region. The break-through location changes, with some time lag, following the change of the water temperature distribution at the tie plate area. Furthermore, the core cooling in the non-break-through regions is almost the same regardless of the location of the break-through. (author)

Testing, development and demonstration of large scale solar district heating systems

DEFF Research Database (Denmark)

Furbo, Simon; Fan, Jianhua; Perers, Bengt

2015-01-01

In 2013-2014 the project “Testing, development and demonstration of large scale solar district heating systems” was carried out within the Sino-Danish Renewable Energy Development Programme, the so called RED programme jointly developed by the Chinese and Danish governments. In the project Danish...... know how on solar heating plants and solar heating test technology have been transferred from Denmark to China, large solar heating systems have been promoted in China, test capabilities on solar collectors and large scale solar heating systems have been improved in China and Danish-Chinese cooperation...

ORNL instrumentation performance for Slab Core Test Facility (SCTF)-Core I Reflood Test Facility

International Nuclear Information System (INIS)

Hardy, J.E.; Hess, R.A.; Hylton, J.O.

1983-11-01

Instrumentation was developed for making measurements in experimental refill-reflood test facilities. These unique instrumentation systems were designed to survive the severe environmental conditions that exist during a simulated pressurized water reactor loss-of-coolant accident (LOCA). Measurement of in-vessel fluid phenomena such as two-phase flow velocity and void fraction and film thickness and film velocity are required for better understanding of reactor behavior during LOCAs. The Advanced Instrumentation for Reflood Studies (AIRS) Program fabricated and delivered instrumentation systems and data reduction software algorithms that allowed the above measurements to be made. Data produced by AIRS sensors during three experimental runs in the Japanese Slab Core Test Facility are presented. Although many of the sensors failed before any useful data could be obtained, the remaining probes gave encouraging and useful results. These results are the first of their kind produced during simulated refill-reflood stage of a LOCA near actual thermohydrodynamic conditions

Induction Heating on Dynamic Tensile Tests in CEA Saclay

International Nuclear Information System (INIS)

Averty, X.; Yvon, P.; Duguay, C.; Pizzanelli, J. P.; Basini, V.

2001-01-01

The LCMI (Laboratory for characterization of irradiated materials), located in CEA from Saclay, is in charge of the mechanical tests on irradiated materials. The dynamic tensile testing machine, in a hot cell equipped with two remote handling, has been first improved in 1995, to fulfill the French safety programs on Reactivity Initiated Accident (RIA). One objective of this machine is to obtain mechanical property data on current Zircaloy cladding types needed to quality the cladding's response under RIA or LOCA transient loading and thermal conditions. For the RIA, this means testing at strain rates up to 5 s' and heating rates up to 200 degree centigree-s''-1, while for Loss of Coolant Accidents (LOCA) testing at strain rates of 10''-3 s''-1 and heating rates of 20 degree centigree s''-1 would be appropriate. The tensile samples are machined with a spark erosion machine, directly from pieces of cladding previously de fueled. Two kinds of samples can be machined in the cladding. Axial samples in order to test axial mechanical characteristics Ring samples in order to test transverse mechanical characteristics, more representative of RIA conditions. On one hand, the axial tensile tests were performed using the Joule effect, and heating rates up to about 500 degree centigree .s''-1 were obtained. This enabled us to perform the axial tests in a satisfactory manner. On the other hand, the tensile ring were first performed in a vertical furnace with a heating rate about 0.2 degree centigree.s''-1 and a thermal stability about 1 degree centigree. For temperatures above 480 degree centigree, the mechanical characteristics showed a sharp drop which could be attributed to irradiation defect annealing. Therefore we have recently developed an induction heating system to reach heating rates high enough (200 degree centigree.s''-1) to prevent any significant annealing before performing the ring tensile tests. To apply a uniaxial tangential tension, two matching half

First In-Core Measurement Results Obtained with the Innovative Mobile Calorimeter CALMOS inside the OSIRIS Material Testing Reactor

International Nuclear Information System (INIS)

Carcreff, Hubert; Salmon, Laurent; Courtaux, Cedric

2013-06-01

Nuclear heating rate inside an MTR has to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry [1, 2]. An innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70 MWth OSIRIS reactor operated by CEA. Thanks to a new calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. The development of the calorimetric probe required the manufacturing and the irradiation of mock-ups in the ex-core area, where nuclear heating rate does not exceed 2 W.g -1 . The calorimeter working mode, the different measurement procedures allowed with such a new probe and main modeling and experimental results have been already presented [3, 4]. In this paper, we present the first results obtained during several measurement campaigns carried out in 2012 and 2013 inside the OSIRIS core with the final device. For the first time, this new experimental measurement system was operated in nominal in-core thermo hydraulic conditions with nominal neutron and gamma fluxes (up to 6 W.g -1 ) in several experimental locations. After a brief presentation of the displacement system specificities, first nuclear heating distributions are presented and discussed. Experimental data were also used to upgrade the Finite Element model of the calorimeter in order to match measured temperatures with calculated ones. This model allowed to estimate a Kc correction factor which takes into account small nonlinearities when the heating rate is deduced from the calibration method. A comparison is made between nuclear heating rates determined from the probe calibration and from the zero method. In addition, an evaluation of the global uncertainty associated to the measurements is detailed. Finally, a global comparison is made with available measurements obtained from previous calorimeters. (authors)

JOYO MK-II core characteristics database

International Nuclear Information System (INIS)

Ohkawachi, Yasushi; Maeda, Shigetaka; Sekine, Takashi; Aoyama, Takafumi

2003-04-01

The 'JOYO' MK-II core characteristics database was compiled and published in 1998. Comments and requests from many users led to the creation of a revised edition. The revisions include changes to the MAGI calculation code system to use the 70 group JFS-3-J3.2 constant set processed from the JENDL-3.2 library. Total control rod worth, reactor kinetic parameters and the MK-II core performance test results were included per user's requests. The core characteristics obtained from the 32 nd to 35 th operational cycles, which were conducted in the MK-III transition core, were newly added in this revised version. The MK-II core management data and core characteristics data were recorded to CD-ROM for user convenience. The Configuration Data' include the core arrangement and refueling record for each operational cycle. The 'Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of 362 driver fuel subassemblies and 69 irradiation test subassemblies. The 'Output Data' contain the calculated neutron flux, gamma flux, power density, linear heat rate, coolant and fuel temperature distribution of all the fuel subassemblies at the beginning and end of each operational cycle. The 'Core Characteristics Data' include the measured excess reactivity, control rod worth calibration curve, and reactivity coefficients of temperature, power and burn-up. (author)

Technical management on commissioning test of nuclear heating reactor

International Nuclear Information System (INIS)

Zhang Yajun; Su Qingshan

1999-01-01

The commissioning is the last construction stage of a nuclear heating project. The commissioning quality will directly affect on the safe operation and availability of the heating reactor. The author presents the whole test process until the completion of the test report from the point of test documents, including the preparation and execution of the test, the management of the various unexpected events during the test. And it will be emphatically discussed that the managing procedures of the various unexpected events during the test, including temporary control change, setpoint change, unexpected events and design change

Experiment data report for semiscale Mod-1 test S-04-2 (baseline ECC test)

International Nuclear Information System (INIS)

Crapo, H.S.; Collins, B.L.; Sackett, K.E.

1976-09-01

Recorded test data are presented for Test S-04-2 of the Semiscale Mod-1 Baseline ECC test series. This test is among Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor system. Test S-04-2 was conducted from an initial cold leg fluid temperature of 542 0 F and an initial pressure of 2,263 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization and reflood transient using emergency core coolant injection parameters based on downcomer volume scaling. System flow was set to achieve a core fluid temperature differential of 66 0 F at a full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW to simulate the surface heat flux response of nuclear fuel rods until such sime that departure from nucleate boiling might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core coolant injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown. The purpose of the report is to make available the uninterpreted data from Test S-04-2 for future data analysis and test results reporting activities. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to assure that they are reasonable and consistent

Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

International Nuclear Information System (INIS)

2014-08-01

The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-08-15

The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

International Nuclear Information System (INIS)

Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won; Cho, Seungyon

2014-01-01

The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity

Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

2014-10-15

The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity.

Thermodynamic and heat transfer analysis of heat recovery from engine test cell by Organic Rankine Cycle

Science.gov (United States)

Shokati, Naser; Mohammadkhani, Farzad; Farrokhi, Navid; Ranjbar, Faramarz

2014-12-01

During manufacture of engines, evaluation of engine performance is essential. This is accomplished in test cells. During the test, a significant portion of heat energy released by the fuel is wasted. In this study, in order to recover these heat losses, Organic Rankine Cycle (ORC) is recommended. The study has been conducted assuming the diesel oil to be composed of a single hydrocarbon such as C12H26. The composition of exhaust gases (products of combustion) have been computed (and not determined experimentally) from the stoichiometric equation representing the combustion reaction. The test cell heat losses are recovered in three separate heat exchangers (preheater, evaporator and superheater). These heat exchangers are separately designed, and the whole system is analyzed from energy and exergy viewpoints. Finally, a parametric study is performed to investigate the effect of different variables on the system performance characteristics such as the ORC net power, heat exchangers effectiveness, the first law efficiency, exergy destruction and heat transfer surfaces. The results of the study show that by utilizing ORC, heat recovery equivalent to 8.85 % of the engine power is possible. The evaporator has the highest exergy destruction rate, while the pump has the lowest among the system components. Heat transfer surfaces are calculated to be 173.6, 58.7, and 11.87 m2 for the preheater, evaporator and superheater, respectively.

Critical heat flux tests for a 12 finned-element assembly