Effect of the design change of the LSSBP on core flow distribution of APR+ Reactor

International Nuclear Information System (INIS)

Kim, Kihwan; Euh, Dong-Jin; Choi, Hae-Seob; Kwon, Tae-Soon

2014-01-01

The uniform core inlet flow distribution of an Advanced Power Reactor Plus (APR+) is required to prevent the failure rate of the HIPER fuel assembly and improve the core thermal margin. KEPCO-E and C and KAERI proposed a design change of the Lower Support Structure Bottom Plate (LSSBP), since the core flow rates were intense near the outer region of the intact LSSBP in a previous study. In this study, an experiment was carried out to evaluate the effect of the design change of the LSSBP on the core flow distribution using the APR+ Core Flow and Pressure (ACOP) test facility. The results showed great improvement on the core flow distribution under a 4-pump balanced flow condition. Under the 4-pump balanced flow condition, fifteen tests were repeated using the ACOP test facility to verify the effect of the 50% blocked flow area at the outer region of the LSSBP on the core inlet flow distribution. The profiles of the core inlet mass flow rates were analyzed using ensemble averaged values, and compared with that of the intact LSSBP. The results showed great improvement for the overall core region. The change in design of the LSSBP is expected to improve the hydraulic performance of an APR+ reactor

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

Reflood behavior at low initial clad temperature in Slab Core Test Facility Core-II

International Nuclear Information System (INIS)

Akimoto, Hajime; Sobajima, Makoto; Abe, Yutaka; Iwamura, Takamichi; Ohnuki, Akira; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu; Adachi, Hiromichi.

1990-07-01

In order to study the reflood behavior with low initial clad temperature, a reflood test was performed using the Slab Core Test Facility (SCTF) with initial clad temperature of 573 K. The test conditions of the test are identical with those of SCTF base case test S2-SH1 (initial clad temperature 1073 K) except the initial clad temperature. Through the comparison of results from these two tests, the following conclusions were obtained. (1) The low initial clad temperature resulted in the low differential pressures through the primary loops due to smaller steam generation in the core. (2) The low initial clad temperature caused the accumulated mass in the core to be increased and the accumulated mass in the downcomer to be decreased in the period of the lower plenum injection with accumulator (before 50s). In the later period of the cold leg injection with LPCI (after 100s), the water accumulation rates in the core and the downcomer were almost the same between both tests. (3) The low initial clad temperature resulted in the increase of the core inlet mass flow rate in the lower plenum injection period. However, the core inlet mass flow rate was almost the same regardless of the initial clad temperature in the later period of the cold leg injection period. (4) The low initial clad temperature resulted in the low turnaround temperature, high temperature rise and fast bottom quench front propagation. (5) In the region apart from the quench front, low initial clad temperature resulted in the lower heat transfer. In the region near the quench front, almost the same heat transfer coefficient was observed between both tests. (6) No flow oscillation with a long period was observed in the SCTF test with low initial clad temperature of 573 K, while it was remarkable in the Cylindrical Core Test Facility (CCTF) test which was performed with the same initial clad temperature. (J.P.N.)

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

CFD analysis of flow distribution at the core inlet of SMART

Energy Technology Data Exchange (ETDEWEB)

Bae, Youngmin, E-mail: ybae@kaeri.re.kr [Korea Atomic Energy Research Institute, 150 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, Young In; Park, Cheon Tae [Korea Atomic Energy Research Institute, 150 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

2013-05-15

Highlights: ► Core inlet flow distribution of system-integrated modular advanced reactor (SMART) is numerically investigated. ► Effects of mesh resolution, turbulence model, Reynolds number, and inflow condition are examined. ► Numerical results show that lower core support plate effectively distributes the flow at the core inlet of SMART. -- Abstract: This study numerically investigates the flow distribution at the core inlet region of the system-integrated modular advanced reactor (SMART). The single-phase turbulent flow is computed by the commercial CFD code, Fluent 12.0 on the computational domain consisting of three main parts: fuel assemblies, a lower core support plate, and a flow skirt. Simulations are carried out with different mesh resolutions, turbulence models, and upstream boundary conditions. The CFD results show that the flow distribution at the core inlet is almost identical for the two Reynolds numbers and turbulence models tested here, and the effect of mesh refinement on the flow distribution at the core inlet is negligible. It is also found that under a uniform upstream boundary condition, the maximum difference in mass flow rate between the fuel assemblies is less than 2%, while it slightly increases to 2.3% under a non-uniform condition. These results consequently indicate that the present design of the lower core support plate effectively distributes the flow at the core inlet of SMART, even when the flow discharged from the upstream has a certain degree of non-uniformity.

Investigation on the Core Bypass Flow in a Very High Temperature Reactor

Energy Technology Data Exchange (ETDEWEB)

Hassan, Yassin

2013-10-22

Uncertainties associated with the core bypass flow are some of the key issues that directly influence the coolant mass flow distribution and magnitude, and thus the operational core temperature profiles, in the very high-temperature reactor (VHTR). Designers will attempt to configure the core geometry so the core cooling flow rate magnitude and distribution conform to the design values. The objective of this project is to study the bypass flow both experimentally and computationally. Researchers will develop experimental data using state-of-the-art particle image velocimetry in a small test facility. The team will attempt to obtain full field temperature distribution using racks of thermocouples. The experimental data are intended to benchmark computational fluid dynamics (CFD) codes by providing detailed information. These experimental data are urgently needed for validation of the CFD codes. The following are the project tasks: • Construct a small-scale bench-top experiment to resemble the bypass flow between the graphite blocks, varying parameters to address their impact on bypass flow. Wall roughness of the graphite block walls, spacing between the blocks, and temperature of the blocks are some of the parameters to be tested. • Perform CFD to evaluate pre- and post-test calculations and turbulence models, including sensitivity studies to achieve high accuracy. • Develop the state-of-the art large eddy simulation (LES) using appropriate subgrid modeling. • Develop models to be used in systems thermal hydraulics codes to account and estimate the bypass flows. These computer programs include, among others, RELAP3D, MELCOR, GAMMA, and GAS-NET. Actual core bypass flow rate may vary considerably from the design value. Although the uncertainty of the bypass flow rate is not known, some sources have stated that the bypass flow rates in the Fort St. Vrain reactor were between 8 and 25 percent of the total reactor mass flow rate. If bypass flow rates are on the

Accuracy of small diameter sheathed thermocouples for the core flow test loop

International Nuclear Information System (INIS)

Anderson, R.L.; Kollie, T.G.

1979-04-01

This report summarizes the research and development on 0.5-mm-diameter, compacted, metal sheathed thermocouples. The objectives of this research effort have been: to identify and analyze the sources of temperature measurement errors in the use of 0.5-mm-diameter sheathed thermocouples to measure the surface temperature of the cladding of fuel-rod simulators in the Core Flow Test Loop (CFTL) at ORNL; to devise methods for reducing or correcting for these temperature measurement errors; to estimate the overall temperature measurement uncertainties; and to recommend modifications in the manufacture, installation, or materials used to minimize temperature measurement uncertainties in the CFTL experiments

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.

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 of core modeling effect on transients for multi-flow zone design of SFR

International Nuclear Information System (INIS)

Shin, Andong; Choi, Yong Won

2016-01-01

SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

Evaluation of core modeling effect on transients for multi-flow zone design of SFR

Energy Technology Data Exchange (ETDEWEB)

Shin, Andong; Choi, Yong Won [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-10-15

SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

Effects of Core Cavity on a Flow Distribution

Energy Technology Data Exchange (ETDEWEB)

Kwon, Tae-Soon; Kim, Kihwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The axial pressure drop is removed in the free core condition, But the actual core has lots of fuel bundles and mixing vanes to the flow direction. The axial pressure drop induces flow uniformity. In a uniform flow having no shear stress, the cross flow or cross flow mixing decreases. The mixing factor is important in the reactor safety during a Steam Line Break (SLB) or Main Steam Line Break (MSLB) transients. And the effect of core cavity is needed to evaluate the realistic core mixing factor quantification. The multi-dimensional flow mixing phenomena in a core cavity has been studied using a CFD code. The 1/5-scale model was applied for the reactor flow analysis. A single phase water flow conditions were considered for the 4-cold leg and DVI flows. To quantify the mixing intensity, a boron scalar was introduced to the ECC injection water at cold legs and DVI nozzles. The present CFD pre-study was performed to quantify the effects of core structure on the mixing phenomena. The quantified boron mixing scalar in the core simulator model represented the effect of core cavity on the core mixing phenomena. This simulation results also give the information for sensor resolution to measure the boron concentration in the experiments and response time to detect mixing phenomena at the core and reactor vessel.

Study and analysis on the flow induced vibration of the core barrel of PWR

International Nuclear Information System (INIS)

Yao Weida; Shi Guolin; Jiang Nanyan; Peng YongYong; Zhang Huijun; Wang Yufen; Xie Yongcheng; Guo Chunhua; Shen Qinping

1989-01-01

The deduction of the resemblance criterion and the design of the test model by applying flow-solid coupling theory are described. The model analysis of a core barrel both in the air and stationary water were performed in a 1:10 model, thus obtaining the dynamic characteristic. In a 1:5 reactor model with a hydraulic closed loop, the inner structure and support were modeled for performing hydraulic closed loop, the inner structure and support were modeled for performing hydraulic vibration test of the core barrel. The flow induced pulse pressure of the core barrel and corresponding response were obtained by using miniature pressure capsule, strain gauge and accelerometer. Power spectrum, correlation functions, transfer function and amplitudes under different flow velocities were calculated. The hydraulic vibration test shows that the core barrel will be in safety during its 30-year life time

Study and analysis for the flow-induced vibration of the core barrel of a PWR

International Nuclear Information System (INIS)

Yao Weida; Shi Guolin; Jiang Nanyan

1989-01-01

The resemblance criteria are derived and a test model is designed by applying the flow-soild coupling theory. After having completed the model analysis of the pressurized water reactor (PWR) core barrel in an 1:10 model, the dynamic characteristics are obtained. In an 1:5 reactor model with a hydraulic closed loop, the hydraulic vibration tests of the core barrel are performed, and the relations between the flow rate and the flow-induced pulse pressure on core barrel, acceleration and strain signals have been measured. The corresponding responses and a group of computational equations for hydraulic vibration are derived from these two experiments. The computational hydraulic vibration responses for core barrel in Qinshan Nuclear Power Plant are in good agreement with the test results, and it shows that the core barrel is safe within its lifetime of 30 years

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

International Nuclear Information System (INIS)

Hudson, T.L.

1981-09-01

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

CFD Analysis on a Core Outlet Flow through the Fuel Alignment Plant of SMART

International Nuclear Information System (INIS)

Kim, Y. I.; Bae, Y. M.; Kim, K. K.

2014-01-01

CFD (Computational Fluid Dynamics) simulations were performed to confirm the core flow distribution for SMART, which acquired standard design approval in 2012. In this paper, CFD simulation is also used to calculate the pressure distribution of a core outlet, a Fuel Alignment Plate (FAP), for SMART. In SMART, the fluid discharged from the Steam Generator comes into a Flow Mixing Header Assembly (FMHA), and is rearranged and split into a very fine size. The FMHA is greatly important for enhancing the flow distribution of a downcomer during a normal operation, transient, and even accidents. Then, the fluid discharged from the FMHA flows into the core upstream through flow skirt holes. The Low Core Support Plate (LCSP) reallocates the flow introducing into the inlet core from the core upstream. The deviation of flow distribution becomes smaller or almost disappears by LCSP holes having relatively large loss coefficient compared to the downstream flow deviation. In an open core, the flow deviation at the core inlet region is diminished by cross flow as it goes upward. Near the core outlet, the flow distribution can be distorted by the influence of a Fuel Alignment Plate (FAP) installed above the fuels. In this paper, the effect of the core outlet flow structure such as the FAP holes of SMART is investigated. Before the calculation, the influences of mesh size and turbulence models are inspected. CFD simulations were performed to investigate the effect of FAP flow holes on the core outlet flow of SMART. As a preliminary study, the dependency of the mesh size and turbulence models was tested; a fine grid was applied, the effect of which is negligible, and the core outlet flow is not sensitive to the turbulence models. In brief, the flow resistance of FAP is less than 15% of that of the fuel assemblies. The flow resistance deviation between two flow path patterns is less than 1% of that of active core. Even two flow path patterns located at the downstream location of the

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)

Effects of non-uniform core flow on peak cladding temperature: MOXY/SCORE sensitivity calculations

Energy Technology Data Exchange (ETDEWEB)

Chang, S.C.

1979-08-15

The MOXY/SCORE computer program is used to evaluate the potential effect on peak cladding temperature of selective cooling that may result from a nonuniform mass flux at the core boundaries during the blowdown phase of the LOFT L2-4 test. The results of this study indicate that the effect of the flow nonuniformity at the core boundaries will be neutralized by a strong radial flow redistribution in the neighborhood of core boundaries. The implication is that the flow nonuniformity at the core boundaries has no significant effect on the thermal-hydraulic behavior and cladding temperature at the hot plane.

Effects of non-uniform core flow on peak cladding temperature: MOXY/SCORE sensitivity calculations

International Nuclear Information System (INIS)

Chang, S.C.

1979-01-01

The MOXY/SCORE computer program is used to evaluate the potential effect on peak cladding temperature of selective cooling that may result from a nonuniform mass flux at the core boundaries during the blowdown phase of the LOFT L2-4 test. The results of this study indicate that the effect of the flow nonuniformity at the core boundaries will be neutralized by a strong radial flow redistribution in the neighborhood of core boundaries. The implication is that the flow nonuniformity at the core boundaries has no significant effect on the thermal-hydraulic behavior and cladding temperature at the hot plane

Evaluation report on CCTF Core-I reflood tests Cl-2 (Run 11) and Cl-3 (Run 12)

International Nuclear Information System (INIS)

Akimoto, Hajime; Murao, Yoshio

1983-06-01

In order to clarify the effect of the initial superheat of the downcomer wall on the system and the core cooling behaviors during the reflood phase of a PWR-LOCA, two tests were performed with the Cylindrical Core Test Facility (CCTF). One is the superheated wall test (test Cl-2) with the initial superheat of 79K, as in the actual PWR, and the other is the saturated wall test (test Cl-3) without any initial superheat. Through the comparisons of the test results from these two tests, the following conclusions were obtained. (1) The initial superheat of the downcomer wall resulted in the lower downcomer water head as observed in our separate-effect tests for the downcomer water head. (2) The superheat also caused the core inlet subcooling to be decreased, and led to the lower core water head. (3) The mass flow rate through the intact loop was reduced only by 4% by the initial superheat of the downcomer wall because the core water head was reduced as well as the downcomer water head. Whereas the mass flow rate through the broken loop was increased because of the increased pressure drop through the broken cold leg. (4) The difference of the core inlet mass flow rate was small between the superheated and the saturated wall tests. It can be considered that small difference of the core inlet mass flow rate results from the compensation of the decreased mass flow rate through the intact loops by the increased mass flow rate through the broken loop. (5) The main discrepancies of the core cooling and the carry-over behaviors between two CCTF tests, were consistent with those observed in the parametric tests for the core inlet subcooling of the FLECHT LOW FLOODING TEST series. (author)

Flow distribution in ET-RR-1 core

International Nuclear Information System (INIS)

Khattab, M.; Mina, A.R.

1989-01-01

In nuclear reactors the flow may be arranged through individual bundles by orifices to achieve better thermal performance. A model based on constant pressure drop across different core regions is developed to determine the flow distribution in reactor core. The friction and grids in the bundles as well as the orifices diameters have an influence on modifying the flow distribution. The application of the proposed model on ET-RR-1 gives reasonable prediction of flow distribution

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)

Rapidly changing flows in the Earth's core

DEFF Research Database (Denmark)

Olsen, Nils; Mandea, M.

2008-01-01

A large part of the Earth's magnetic field is generated by fluid motion in the molten outer core(1). As a result of continuous satellite measurements since 1999, the core magnetic field and its recent variations can now be described with a high resolution in space and time(2). These data have...... field occurring over only a few months, indicative of fluid flow at the top of the core, can in fact be resolved. Using nine years of magnetic field data obtained by satellites as well as Earth-based observatories, we determine the temporal changes in the core magnetic field and flow in the core. We...

Core surface flow modelling from high-resolution secular variation

DEFF Research Database (Denmark)

Holme, R.; Olsen, Nils

2006-01-01

-flux hypothesis, but the spectrum of the SV implies that a conclusive test of frozen-flux is not possible. We parametrize the effects of diffusion as an expected misfit in the flow prediction due to departure from the frozen-flux hypothesis; at low spherical harmonic degrees, this contribution dominates...... the expected departure of the SV predictions from flow to the observed SV, while at high degrees the SV model uncertainty is dominant. We construct fine-scale core surface flows to model the SV. Flow non-uniqueness is a serious problem because the flows are sufficiently small scale to allow flow around non......-series of magnetic data and better parametrization of the external magnetic field....

Reactor core flow rate control system

International Nuclear Information System (INIS)

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

1996-01-01

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

Evaluation report on CCTF core-II reflood test C2 - 8 (Run 67)

International Nuclear Information System (INIS)

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

1987-01-01

In order to study the system pressure effect of the core cooling and flow behavior during the reflood phase of a PWR LOCA, a test was performed with CCTF under the system pressure pf 0.15 MPa as a counterpart test of the CCTF test C2-1 (system pressure 0.42 MPa) and the CCTF test C2-4 (system pressure 0.20 MPa). Through the comparisons of results from these three tests, the following conclusions were obtained: (1) The higher system pressure resulted in the lower temperature rise, the shorter turnaround time and the shorter quench time as observed in the CCTF Core-I system pressure effect tests. (2) The higher system pressure resulted in higher core water head, higher upper plenum water head, higher mass flow rate through the primary loops. On the other hand, the higher system pressure resulted in lower downcomer water head and lower pressure drop through the primary loops and the broken cold leg. These system pressure effects on the flow behavior in the primary system are almost the same as observed in the system pressure effect tests in the CCTF Core-I test series. (3) Before the mixture level in the upper plenum reached the level of the hot leg nozzle, the loop flow resistance coefficient of the intact loops was nearly constant regardless of the system pressure. After the mixture level reached the level of the hot leg nozzle, the loop flow resistance coefficient was increased due to the water accumulation in the hot leg piping and the inlet plenum of the steam generator in these tests. (J.P.N.)

Core-annular flow through a horizontal pipe : Hydrodynamic counterbalancing of buoyancy force on core

NARCIS (Netherlands)

Ooms, G.; Vuik, C.; Poesio, P.

2007-01-01

A theoretical investigation has been made of core-annular flow: the flow of a high-viscosity liquid core surrounded by a low-viscosity liquid annular layer through a horizontal pipe. Special attention is paid to the question of how the buoyancy force on the core, caused by a density difference

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

CFD Analysis for Predicting Flow Resistance of the Cross Flow Gap in Prismatic VHTR Core

International Nuclear Information System (INIS)

Lee, Jeong Hun; Yoon, Su Jong; Park, Goon Cherl; Park, Jong Woon

2011-01-01

The core of Very High Temperature Reactor (VHTR) consists of assemblies of hexagonal graphite blocks and its height and across-flats width are 800 mm and 360 mm respectively. They are equipped with 108 coolant holes 16 mm in diameter. Up to ten fuel blocks arranged in vertical order form a fuel element column and the neutron flux varies over the cross section of the core. It makes different axial shrinkage of fuel element and this leads to make wedge-shaped gaps between the base and top surfaces of stacked blocks. The cross flow is defined as the core flow that passes through this cross gaps. The cross flow complicates the flow distribution of reactor core. Moreover, the cross flow could lead to uneven coolant distribution and consequently to superheating of individual fuel element zones with increased fission product release. Since the core cross flow has a negative impact on safety and efficiency of VHTR, core cross flow phenomena have to be investigated to improve the core thermal margin of VHTR. In particular, to predict amount of flow at the cross flow gap obtaining accurate flow loss coefficient is important. Nevertheless, there has not been much effort in domestic. The experiment of cross flow was carried out by H. G. Groehn in 1981 Germany. For the study of cross flow the applicability of CFD code should be validated. In this paper a commercial CFD code CFX-12 validation will be carried out with this cross flow experiment. Validated data can be used for validation of other thermal-hydraulic analysis codes

Investigation of two-phase flow instability under SMART-P core conditions

International Nuclear Information System (INIS)

Hwang, Dae Hyun; Lee, Chung Chan

2005-01-01

An integral-type advanced light water reactor, named SMART-P, is being continuously studied at KAERI. The reactor core consists of hundreds of closed-channel type fuel assemblies with vertical upward flows. The upper and lower parts of the fuel assembly channels are connected to the common heads. The constant pressure drop imposed on the channel is responsible for the occurrence of density wave oscillations under local boiling and/or natural circulation conditions. The fuel assembly channel with oscillatory flow is highly susceptible to experience the CHF which may cause the fuel failure due to a sudden increase of the cladding temperature. Thus, prevention of the flow instability is an important criterion for the SMART-P core design. Experimental and analytical studies have been conducted in order to investigate the onset of flow instability (OFI) under SMART core conditions. The parallel channel oscillations were observed in a high pressure water-loop test facility. A linear stability analysis model in the frequency-domain was developed for the prediction of the marginal stability boundary (MSB) in the parallel boiling channels

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

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

Two-dimensional thermal-hydraulic behavior in core in SCTF Core-II cold leg injection tests

International Nuclear Information System (INIS)

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

1985-07-01

Major purpose of the Slab Core Test Program is to investigate the two-dimensional thermal-hydraulic behavior in the core during the reflood phase in a PWR-LOCA. In order to investigate the effects of radial power profile, three cold leg injection tests with different radial power profiles under the same total heating power and core stored energy were performed by using the Slab Core Test Facility (SCTF) Core-II. It was revealed by comparing these three tests 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. The turnaround temperature in the high power bundles were evaluated to be reduced by about 40 to 120 K. On the other hand, a two-dimensional flow in the core was also induced by the non-uniform water accumulation in the upper plenum and the quench was delayed resultantly in the bundles corresponding to the peripheral bundles of a PWR. However, the effect of the non-uniform upper plenum water accumulation on the turnaround temperature was small because the effect dominated after the turnaround of the cladding temperature. Selected data from Tests S2-SH1, S2-SH2 and S2-O6 are also presented in this report. Some data from Tests S2-SH1 and S2-SH2 were compared with TRAC post-test calculations performed by the Los Alamos National Laboratory. (author)

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)

Preliminary Calculations of Bypass Flow Distribution in a Multi-Block Air Test

International Nuclear Information System (INIS)

Kim, Min Hwan; Tak, Nam Il

2011-01-01

The development of a methodology for the bypass flow assessment in a prismatic VHTR (Very High Temperature Reactor) core has been conducted at KAERI. A preliminary estimation of variation of local bypass flow gap size between graphite blocks in the NHDD core were carried out. With the predicted gap sizes, their influence on the bypass flow distribution and the core hot spot was assessed. Due to the complexity of gap distributions, a system thermo-fluid analysis code is suggested as a tool for the core thermo-fluid analysis, the model and correlations of which should be validated. In order to generate data for validating the bypass flow analysis model, an experimental facility for a multi-block air test was constructed at Seoul National University (SNU). This study is focused on the preliminary evaluation of flow distribution in the test section to understand how the flow is distributed and to help the selection of experimental case. A commercial CFD code, ANSYS CFX is used for the analyses

Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

Directory of Open Access Journals (Sweden)

Po Hu

2014-01-01

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

Study on core flow distribution of the reference core design Mark-III of experimental multi-purpose VHTR

International Nuclear Information System (INIS)

Satoh, Sadao; Arai, Taketoshi; Miyamoto, Yoshiaki; Hirano, Mitsumasa

1977-01-01

Concerning the coolant flow distribution between fuel channels and other flow paths in the core, designated as Reference Core Mark-III of the Multi-purpose Experimental Very High Temperature Reactor, thermal analysis has been made of the control rods and other steel structures around the core to find the coolant flow rates (bypass flow) necessary to cool them to their safe operating temperatures. Calculations showed that adequate cooling could be achieved in the Mark-III Core by the bypass flow of 8% of the total reactor coolant flow, 4% each for the control-rod channels and for other structures. The thermal and coolant flow design bases, including the assumption of a 10% bypass flow, were thus confirmed to first approximation. (auth.)

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

Flowing and freezing of molten core materials during unprotected loss of flow accidents in sodium cooled fast reactors

International Nuclear Information System (INIS)

Maschek, W.; Royl, P.

1988-09-01

Flowing and freezing of mobile core materials change the fissile material distribution and core-inventory under hypothetical accident conditions and determine the path to permanent shutdown of the neutronic events and the energetic potentials. The report classifies the bondary conditions for such flowing and freezing processes by going through the different situations under which these processes can occur in the scenario of the unprotected loss of flow (ULOF) accident. The classification is based on ULOF-accident simulations for a homogeneous reactor core concept of a 300 MWe LMFBR (e. g. SNR-300), but many boundary conditions are also characteristic for other core designs. A review of the relevant experiments is then made to correlate the available experimental information with these classified boundary conditions and to look at the resulting flowing and freezing processes. Boundary conditions that have been experimentally shown to be important are assigned high priorities. The data are specifically valued in relation to these boundary conditions of high priorities. The review includes the major experimental programs with published results. The discussion shows that the results from most clean condition tests for melt relocations are valuable for a better understanding of basic phenomena and analytical model development, but are not directly applicable to real accident conditions. The database for relevant boundary conditions from the ULOF scenario is limited and largely included in integral sequence tests from which quantitative information for modelling is difficult to obtain. Needs for additional investigations are identified. The suggestions are mainly restricted to investigations of the early phase of fuel removal. They are given with reference to candidate facilities and include relocations in the subassemblies and in the inter-subassembly gaps. Particular emphasis is put on the leading edge properties and possible driving forces to which more attention

Reactor vessel and core two-phase flow ultrasonic densitometer

International Nuclear Information System (INIS)

Arave, A.E.

1979-01-01

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

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.

Evolution of - and Core-Dominated Lava Flows Using Scaling Analysis

Science.gov (United States)

Castruccio, A.; Rust, A.; Sparks, R. S.

2010-12-01

We investigated the front evolution of simple lava flows on a slope using scaling arguments. For the retarding force acting against gravity, we analyzed three different cases: a flow controlled by a Newtonian viscosity, a flow controlled by the yield strength of a diffusively growing crust and a flow controlled by its core yield strength. These models were tested using previously published data of front evolution and volume discharge of 10 lava flow eruptions from 6 different volcanoes. Our analysis suggests that for basaltic eruptions with high effusion rate and low crystal content, (Hawaiian eruptions), the best fit of the data is with a Newtonian viscosity. For basaltic eruptions with lower effusion rates (Etna eruptions) or long duration andesitic eruptions (Lonquimay eruption, Chile) the flow is controlled by the yield strength of a growing crust. Finally, for very high crystalline lavas (Colima, Santiaguito) the flow is controlled by its core yield strength. The order of magnitude of the viscosities from our analysis is in the same range as previous studies using field measurements on the same lavas. The yield strength values for the growing crust and core of the flow are similar and with an order of magnitude of 10^5 Pa. This number is similar to yield strength values found in lava domes by different authors. The consistency of yield strength ~10^5 Pa is because larger stresses cause fracturing of very crystalline magma, which drastically reduces its effective strength. Furthermore, we used a 2-D analysis of a Bingham fluid flow on a slope to conclude that, for lower yield strength values, the flow is controlled mainly by its plastic viscosity and the lava can be effectively modelled as Newtonian. Our analysis provides a simple tool to evaluate the main controlling forces in the evolution of a lava flow, as well as the magnitude of its rheological properties, for eruptions of different compositions and conditions and may be useful to predict the evolution of

Core thermal response during Semiscale Mod-1 blowdown heat transfer tests

International Nuclear Information System (INIS)

Larson, T.K.

1976-06-01

Selected experimental data and results calculated from experimental data obtained from the Semiscale Mod-1 PWR blowdown heat transfer test series are analyzed. These tests were designed primarily to provide information on the core thermal response to a loss-of-coolant accident. The data are analyzed to determine the effect of core flow on the heater rod thermal response. The data are also analyzed to determine the effects of initial operating conditions on the rod cladding temperature behavior during the transient. The departure from nucleate boiling and rewetting characteristics of the rod surfaces are examined for radial and axial patterns in the response. Repeatability of core thermal response data is also investigated. The test data and the core thermal response calculated with the RELAP4 code are compared

A study on bypass flow gap distribution in a prismatic VHTR core

International Nuclear Information System (INIS)

Kim, M. H.; Jo, C. K.; Lim, H. S.

2010-01-01

Core bypass flow in VHTR is one of the key issues for core thermal margins and efficiency. The bypass flow in the prismatic core varies during core cycles due to the irradiation shrinkage and thermal expansion of the graphite blocks. A procedure to evaluate the local gap size variation between graphite blocks was developed and applied to a prismatic core VHTR. The influence of the core restraint mechanism on the bypass flow gap was evaluated. The predicted gap size is as much as 8 mm when the graphite block is exposed to its allowable limit of irradiation fluence. The analysis for the core bypass flow and hot spot was carried out based on the calculated gap distributions. The results indicate that the bypass flow and the location of core hot spots are closely related and a measure to reduce the bypass flow is necessary. (authors)

Alteration of helical vortex core without change in flow topology

DEFF Research Database (Denmark)

Velte, Clara Marika; Okulov, Valery; Hansen, Martin Otto Laver

2011-01-01

topology. The helical symmetry as such is preserved, although the characteristic parameters of helical symmetry of the vortex core transfer from a smooth linear variation to a different trend under the influence of a non-uniform pressure gradient, causing an increase in helical pitch without changing its......The abrupt expansion of the slender vortex core with changes in flow topology is commonly known as vortex breakdown. We present new experimental observations of an alteration of the helical vortex core in wall bounded turbulent flow with abrupt growth in core size, but without change in flow...

Influence of core sand properties on flow dynamics of core shooting process based on experiment and multiphase simulation

Directory of Open Access Journals (Sweden)

Chang-jiang Ni

2017-03-01

Full Text Available The influence of core sand properties on flow dynamics was investigated synchronously with various core sands, transparent core-box and high-speed camera. To confirm whether the core shooting process has significant turbulence, the flow pattern of sand particles in the shooting head and core box was reproduced with colored core sands. By incorporating the kinetic theory of granular flow (KTGF, kinetic-frictional constitutive correlation and turbulence model, a two-fluid model (TFM was established to study the flow dynamics of the core shooting process. Two-fluid model (TFM simulations were then performed and a areasonable agreement was achieved between the simulation and experimental results. Based on the experimental and simulation results, the effects of turbulence, sand density, sand diameter and binder ratio were analyzed in terms of filling process, sand volume fraction (αs and sand velocity (Vs.

BWR core response to fluctuations in coolant flow and pressure, with implications on noise diagnosis and stability monitoring

International Nuclear Information System (INIS)

Blomstrand, J.H.; Andersson, S.A.

1982-01-01

Reactor dynamic tests, utilizing sinuosidal oscillations in pressure and recirculation flow, have been conducted in operating BWRs in Sweden and Finland. Test data recorded, as well as recordings of process noise, have been analyzed in terms of dynamic core properties. The results obtained show good qualitative agreement with model predictions of BWR core dynamics. Model studies can often support interpretation of dynamic information obtained from operating plants. Comparisons between model studies, dynamic tests and process noise may also provide improved understanding of test results and noise patterns; in this way it can be demonstrated that some neutron flux noise is caused by noise in coolant flow and steam flow. From reactor test data nd noise recordings, core stability parameters have been evaluated by a number of methods. These have been found to provide essentially the same results. The cores investigated were found to be very stable under normal operating conditions. In special operating points, outside the normal operating range, higher decay ratios may occur. The experience indicates that for BWR cores, operated at decay ratios above quarter damping, the stability parameters may be identified from the oscillatory behavior of the autocorrelation in the time domain of the neutron flux noise

Evaluation of the influence of bypass flow gap distribution on the core hot spot in a prismatic VHTR core

International Nuclear Information System (INIS)

Kim, Min-Hwan; Lim, Hong-Sik

2011-01-01

Highlights: → A procedure to evaluate the local gap size variation between graphite blocks was developed and applied to a prismatic core VHTR. → The analysis for the core bypass flow and hot spot was carried out based on the calculated gap distributions. → The predicted gap size is large enough to affect the flow distribution in the core. → The bypass gap and flow distributions are closely related to the local hot spot temperature and its location. → The core restraint mechanism preventing outward movement of graphite block reduces the bypass gap size and hot spot temperature. - Abstract: Core bypass flow in VHTR is one of the key issues for core thermal margins and efficiency. The bypass flow in the prismatic core varies during core cycles due to the irradiation shrinkage/swelling and thermal expansion of the graphite blocks. A procedure to evaluate the local gap size variation between graphite blocks was developed and applied to a prismatic core VHTR. The influence of the core restraint mechanism on the bypass flow gap was evaluated. The predicted gap size is as much as 8 mm when the graphite block is exposed to its allowable limit of fast neutron fluence. The analysis for the core bypass flow and hot spot was carried out based on the calculated gap distributions. The results indicate that the bypass gap and flow distributions are closely related to the local hot spot and its location and the core restraint mechanism preventing outward movement of the graphite block by a fastening device reduces the bypass gap size, which results in the decrease of maximum fuel temperature not less than 100 deg. C, when compared to the case without it.

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.

Experimental investigation of the vibration response of a flexible tube due to simulated reactor core, cross and annular exit flows

International Nuclear Information System (INIS)

Haslinger, K.H.; Martin, M.L.; Higgins, W.H.; Rossano, F.V.

1989-01-01

Instrumentation tubes in pressurized nuclear reactors have experienced wear due to excessive flow-induced vibrations. Experiments to identify the predominant flow excitation mechanism at a particular plant, and to develop a sleeve design to remedy the wear problem are reported. An instrumented flow visualization model enabled simulation of a wide range of individual or combined reactor core flow, cross flow and thimble flow conditions. The instrumentation scheme adopted for these experiments used proximity displacement transducers and a force transducer to measure respectively tube motion and contact/impact forces at the wear region. Extensive testing of the original, in-plant configuration identified the normal core flow as the primary source of excitation. Shielding the In-Core-Instrumentation thimble tube from the normal core flow curtailed vibration amplitudes; however, thimble flow excitation then became more pronounced. Various outlet nozzle configurations were investigated. An internal cavity combined with radial outlet slots became the optimum solution for the problem. The paper presents typical test data in the form of orbital tube motion, spectrum analysis and time history collages. The effectiveness of shielding the instrumentation tube from the flow is demonstrated. (author)

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)

Bypass Flow and Hot Spot Analysis for PMR200 Block-Core Design with Core Restraint Mechanism

International Nuclear Information System (INIS)

Lim, Hong Sik; Kim, Min Hwan

2009-01-01

The accurate prediction of local hot spot during normal operation is important to ensure core thermal margin in a very high temperature gas-cooled reactor because of production of its high temperature output. The active cooling of the reactor core determining local hot spot is strongly affected by core bypass flows through the inter-column gaps between graphite blocks and the cross gaps between two stacked fuel blocks. The bypass gap sizes vary during core life cycle by the thermal expansion at the elevated temperature and the shrinkage/swelling by fast neutron irradiation. This study is to investigate the impacts of the variation of bypass gaps during core life cycle as well as core restraint mechanism on the amount of bypass flow and thus maximum fuel temperature. The core thermo fluid analysis is performed using the GAMMA+ code for the PMR200 block-core design. For the analysis not only are some modeling features, developed for solid conduction and bypass flow, are implemented into the GAMMA+ code but also non-uniform bypass gap distribution taken from a tool calculating the thermal expansion and the shrinkage/swell of graphite during core life cycle under the design options with and without core restraint mechanism is used

Fluid flow near the surface of earth's outer core

Science.gov (United States)

Bloxham, Jeremy; Jackson, Andrew

1991-01-01

This review examines the recent attempts at extracting information on the pattern of fluid flow near the surface of the outer core from the geomagnetic secular variation. Maps of the fluid flow at the core surface are important as they may provide some insight into the process of the geodynamo and may place useful constraints on geodynamo models. In contrast to the case of mantle convection, only very small lateral variations in core density are necessary to drive the flow; these density variations are, by several orders of magnitude, too small to be imaged seismically; therefore, the geomagnetic secular variation is utilized to infer the flow. As substantial differences exist between maps developed by different researchers, the possible underlying reasons for these differences are examined with particular attention given to the inherent problems of nonuniqueness.

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.)

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)

The Effect of the Holes Size Change of Lower-Support-Structure-Bottom Plate on the Reactor Core-Inlet Flow-Distribution

International Nuclear Information System (INIS)

Lee, Gong Hee; Bang, Young Seok; Cheong, Ae Ju

2015-01-01

Complex thermal-hydraulic phenomena exist inside PWR because reactor interiors include a fuel assembly, control rod assembly, ICI (In-Core Instrumentation), and other internal structures. Because changes to reactor design may influence interior, thermal-hydraulic characteristics, licensing applicants commonly conduct a flow-distribution test and use test results (e.g., core-inlet flow-rate distribution) as the input data for a core thermal-margin analysis program. Because the APR+ (Advanced Power Reactor Plus) had more fuel assemblies (241EA → 257EA) and the design of some internal structures was changed (from those of APR1400), the core-inlet flow-rate distribution for a 1/5 scaled-down reactor model was measured and high flow-rates were found especially near the outer region of the reactor core. In this study, to examine the effect of the holes size change (i.e. smaller diameter) in the outer region of the LSSBP, not a 50% blockage of the flow holes, on the reactor core-inlet flow-distribution, simulations were conducted with the commercial CFD (Computational Fluid Dynamics) software, ANSYS CFX R.14. The predicted results were compared with those of the original LSSBP. In this study, to examine the effect of the holes size change (smaller diameter) in the outer region of the LSSBP on the reactor core-inlet flow-distribution, simulations were conducted with the commercial CFD software, ANSYS CFX R.14. The predicted results were compared with those of the original LSSBP. Through these comparisons it was concluded that a more uniform distribution of the mass-flow rate at the core-inlet plane could be obtained by reducing the holes size in the outer region of the LSSBP

SCTF Core-I test results

International Nuclear Information System (INIS)

Adachi, Hiromichi; Sudo, Yukio; Iwamura, Takamichi; Osakabe, Masahiro; Ohnuki, Akira; Hirano, Kemmei

1982-07-01

The Slab Core Test Facility (SCTF) of Japan Atomic Energy Research Institute (JAERI) was constructed to investigate two-dimensional thermohydrodynamics in the core and the communication 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). In the present report, effects of system pressure on reflooding phenomena shall be discussed based on the data of Tests S1-SH2, S1-01 and S1-02 which are the parameteris tests for system pressure effects belonging to the SCTF Core-I forced flooding test series. Major items discussed in this report are (1) hydrodynamic behavior in the system, (2) core thermal behavior, (3) core heat transfer and (4) two-dimensional hydrodynamic behavior in the pressure vessel including the core. (author)

Evaluation report on CCTF core-I reflood tests Cl-17(Run 36) and Cl-20(Run 39)

International Nuclear Information System (INIS)

Murao, Yoshio; Iguchi, Tadashi

1983-02-01

In the safety analysis of the reflood phase of a PWR LOCA, the core thermo-hydrodynamic behavior is analyzed as a phenomena in a single channel core. In other words, the core thermo-hydrodynamic behavior is treated one-dimensionally. In order to confirm the validity of the one-dimensional treatment, tests named Asymmetric power test Cl-17(Run 36) and Asymmetric temperature test Cl-20(Run 39) were performed, whose test conditions were similar to the base case test Cl-5(Run 14) (the reference test for the parametric effect tests) except for the power distribution and the initial temperature distribution in core, respectively. First the results of the base case test were investigated. And the results of Asymmetric power test and Asymmetric temperature test were compared with the results of the base case test. The main conclusions are as follows: (1) The water accumulation was observed along the whole core even above quench front almost simultaneously just after reflood initiation and almost terminated within 20 to 30 seconds. The flow pattern was recognized as a slug flow. (2) In the lower two-thirds of the core, bottom quench was observed and the core water accumulation was one-dimensional even under the thermally asymmetric conditions. (3) In the upper portion of the core, the multi-dimensional effect was observed, i.e. the top quench occurred locally. (4) The water accumulation behavior or void fraction in the lower two-thirds of the core, i.e. our concerning region for the peak clad temperature analysis, and the core behavior for system analysis can be one-dimensionally analyzed with a representing single channel core. (author)

Analysis of resonance oscillation of the neutron flow in a BWR-core

International Nuclear Information System (INIS)

Storm, J.

1987-09-01

This is a thesis which has been made within the institution of automatic control in Lund. Two programs, 'Blackie' and 'Test' have been written in Fortran. These two programs are to be used for the evaluation of ASEA-ATOMs resonance test in different nuclear reactors. In these tests the condition of the reactor becomes more and more unstable because the coolant flow decreases at the same time as the power gradually increases. This leads to resonance in the neutron flow. This flow is measured by detectors placed in different parts of the reactor core. 'Blackie' receives and stores the values sampled by the detectors. The same program also carries out a Fourier analysis. Amplitudes and phase angles from the different oscillations are calculated. These results are then used as inputs for 'Test'. 'Test' is a plotting program. It draws the reactor and plots arrows where the detectors are situated. The size and direction of the arrows are measurements of the amplitudes and phase angles of the neutron flow oscillations. From these arrow diagrams you can come to conclusions about the oscillations in the neutron flow and how the affect the reactor. (author)

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)

Preliminary Estimation of Local Bypass Flow Gap Sizes for a Prismatic VHTR Core

International Nuclear Information System (INIS)

Kim, Min Hwan; Jo, Chang Keun; Lee, Won Jae

2009-01-01

The Very High Temperature Reactor (VHTR) has been selected for the Nuclear Hydrogen Development and Demonstration (NHDD) project. In the VHTR design, core bypass flow has been one of key issues for core thermal margins and target temperature of the core outlet. The core bypass flow in the prismatic VHTR varies with the core life due to the irradiation shrinkage/ swelling and thermal expansion of the graphite blocks, which could be a significant proportion of the total core flow. Thus, accurate prediction of the bypass flow is of major importance in assuring the core thermal margin. To predict the bypass flow, first of all, local gap sizes between graphite blocks in the core should be determined. The objectives of this work are to develop a methodology for determining the gap sizes and to perform a preliminary evaluation for a reference reactor

Test plan: Air intake shaft performance test -- Addendum for obtaining cores in the Culebra for radionuclide retardation studies

International Nuclear Information System (INIS)

Gelbard, F.

1991-10-01

Core samples are needed for obtaining data on radionuclide retardation. The cores will be used to first determine local basic properties of Culebra rock such as permeability, structural integrity, fracture spacing, and fracture size. These quantities will then be used to design a laboratory experimental program to determine radionuclide retardation in a column flow apparatus using the cores obtained in this project. This addendum covers only the coring activities necessary to retrieve Culebra cores. The laboratory work will be documented in a separate test plan. It is anticipated that Culebra rock samples will be highly fractured, with a fracture spacing on the order of 2 to 3 inches To obtain representative core samples that are intact, horizontal cores about 6 inches in diameter and several feet long will be needed. These cores will provide a good indication of Culebra rock fracturing and provide several samples needed to conduct column flow experiments. If the rock is so fractured that only rubble is obtained, then the rubble will be used in the column experiments. In addition, as a byproduct of the coring operation, natural groundwater collected from the holes will be used to develop a synthetic brine for the laboratory experiments

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.

Breakdown voltage at the electric terminals of GCFR-core flow test loop fuel rod simulators in helium and air

International Nuclear Information System (INIS)

Huntley, W.R.; Conley, T.B.

1979-12-01

Tests were performed to determine the ac and dc breakdown voltage at the terminal ends of a fuel rod simulator (FRS) in helium and air atmospheres. The tests were performed at low pressures (1 to 2 atm) and at temperatures from 20 to 350 0 C (68 to 660 0 F). The area of concern was the 0.64-mm (0.025-in.) gap between the coaxial conductor of the FRS and the sheaths of the four internal thermocouples as they exit the FRS. The tests were prformed to ensure a sufficient safety margin during Core Flow Test Loop (CFTL) operations that require potentials up to 350 V ac at the FRS terminals. The primary conclusion from the test results is that the CFTL cannot be operated safely if the terminal ends of the FRSs are surrounded by a helium atmosphere but can be operated safely in air

In-core fuel element temperature and flow measurment of HFETR

International Nuclear Information System (INIS)

Chen Daolong; Jiang Pei

1988-02-01

The HFETR in-core fuel element temperature-flow measurement facility and its measurement system are expounded. The applications of the instrumented fuel element to stationary and transient states measurements during the lift of power, the operation test of all lifetime at first load, and the deepening burn-up test at second load are described. The method of determination of the hot point temperature under the fin is discussed. The error analysis is made. The fuel element out-of-pile water deprivation test is described. The development of this measurement facility and succesful application have made important contribution to high power and deep burn-up safe operation at two load, in-core fuel element irradiation, and varied investigation of HFETR. After operation at two loads, the integrated power of this instrumented fuel element arrives at 90.88 MWd, its maximum point burn-up is about 64.9%, so that the economy of fuel use of HFETR is raised very much

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.)

Numerical and experimental investigation on labyrinth seal mechanism for bypass flow reduction in prismatic VHTR core

Energy Technology Data Exchange (ETDEWEB)

Yoon, Su-Jong, E-mail: paper80@snu.ac.r [Department of Nuclear Engineering, Seoul National University, San 56-1, Daehak-Dong, Kwanak-Gu, Seoul 151-742 (Korea, Republic of); Lee, Jeong-Hun [Department of Nuclear Engineering, Seoul National University, San 56-1, Daehak-Dong, Kwanak-Gu, Seoul 151-742 (Korea, Republic of); Lee, Sang-Moon [Department of Mechanical Engineering, Inha University, 253 Yonghyun-Dong, Nam-Gu, Incheon 402-751 (Korea, Republic of); Tak, Nam-il; Kim, Min-Hwan [Korea Atomic Energy Research Institute, 150-1 Deokjin-Dong, 1045 Daedeokdaero, Yuseong, Daejeon 305-353 (Korea, Republic of); Kim, Kwang-Yong [Department of Mechanical Engineering, Inha University, 253 Yonghyun-Dong, Nam-Gu, Incheon 402-751 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University, San 56-1, Daehak-Dong, Kwanak-Gu, Seoul 151-742 (Korea, Republic of)

2013-09-15

Highlights: • Bypass flow reduction method was developed by applying labyrinth seal mechanism. • Grooves on side walls of replaceable reflector block were made. • Design of the grooved wall of the reflector block was optimized by the RSA method. • The flow resistance of the bypass gap rose from 18.04 to 26.24 by the optimization. • The bypass ratios at the inlet and outlet were reduced by 36.19% and 14.66%, respectively. -- Abstract: Core bypass flow in block type very high temperature reactor (VHTR) occurs due to the inevitable gaps between the hexagonal core blocks for the block installation and refueling. Since the core bypass flow affects the reactor safety and efficiency, it should be minimized to enhance the core thermal margin. In this regard, the core bypass flow reduction method applying the labyrinth seal mechanism was developed and optimized by using the single-objective shape optimization method. Response surface approximation (RSA) method was adopted as the optimization method. Side wall of the replaceable reflector block was redesigned and response surface approximate model was adopted to optimize the shape of the reflector wall. Computational fluid dynamics (CFD) analyses were carried out not only to assess the limitation of existing method of bypass flow reduction, but also to optimize the design of a newly developed reduction method. The experiment with Seoul National University (SNU) multi-block experimental facility was performed to demonstrate the performance of the reduction method. It was found that the effect of the existing bypass flow reduction method by sealing the bypass gap exit was restricted nearby the lower region of the core. However, the flow resistance factor of the bypass gap increased from 18.04 to 26.24 by the optimized reduction method. The results of the performance test showed that the bypass flow distribution was reduced throughout the entire core regions. The bypass flow ratios at the inlet and the outlet were

Experimental study of core thermohydraulics in fast reactors during transition from forced to natural circulation. Influence of inter-wrapper flow

International Nuclear Information System (INIS)

Kamide, H.; Hayashi, K.; Momoi, K.

1997-01-01

The evaluation of core thermohydraulics under natural circulation conditions is important to utilize inherent safety features of Fast Reactors. When heat exchangers of a decay heat removal system are operated in an upper plenum of reactor vessel, cold sodium is provided by the heat exchangers. Core-plenum interactions will occur during a natural circulation condition due to this cold sodium in the upper plenum, e.g., it can penetrate into gap regions between fuel subassemblies (inter-wrapper flow, IWF) and the flow may reverse in low power core channels. These interactions will significantly modify the flow and temperature distributions in the core. Sodium experiments were carried out to study these phenomena. In a test section, seven subassemblies are housed and connected to an upper plenum. The influences of core-plenum interactions on the core thermohydraulics were investigated under steady state conditions and also in transitions from forced to natural circulation. Cooling effects of IWF on the fuel subassemblies were found in spite of natural circulation flow reduction in the primary loop due to temperature decreases in the upper non-heated section in the core. The inter-wrapper flow can effectively cool the core under extreme conditions of low flow rates through the core. (author)

Heat Transfer Analysis of the European Pressurized Water Reactor (EPR) Core Catcher Test Facility Volley

Energy Technology Data Exchange (ETDEWEB)

Pikkarainen, Mika; Laine, Jani; Purhonen, Heikki; Kyrki-Rajamaeki, Riitta [Lappeenranta University of Technology, P.O. 20 53851 Lappeenranta (Finland); Sairanen, Risto [Radiation and Nuclear Safety Authority, P.O. 14 00881 Helsinki (Finland)

2008-07-01

The EPR is designed to cope with severe accidents, involving core meltdown. A specific melt spreading area has been designed within the containment. This core catcher will be flooded by water, which transfers the decay heat to the containment heat removal system. To improve cooling, horizontal flow channels made of cast iron are located also below the core catcher. STUK, the radiation and nuclear safety authority in Finland, wanted an independent study of the functionality of the core catcher design. Effect of the presence of insulation material and boric acid in the cooling water was to be studied, as well as the general behavior of the system in different phases of the flooding of the core melt spreading area. To verify the function of the core catcher design, a scaled down test facility was built at Lappeenranta University of Technology. Since there are some physical restrictions of a test facility computational tools were applied especially for the tests where steady state conditions could not be reached without endangering the integrity of the test facility. This paper introduces the Volley test facility, computational simulations and compares them with the test results. Simulated temperatures of those Volley tests, which could be run until steady state conditions, are very close to the measured temperatures. It can be concluded also, that the temperatures are evidently below the cast iron melting point with heat fluxes used in the tests, if there is a small flow inside the cooling channels or even in case when only a few adjacent cooling channels are totally dry. (authors)

Heat Transfer Analysis of the European Pressurized Water Reactor (EPR) Core Catcher Test Facility Volley

International Nuclear Information System (INIS)

Pikkarainen, Mika; Laine, Jani; Purhonen, Heikki; Kyrki-Rajamaeki, Riitta; Sairanen, Risto

2008-01-01

The EPR is designed to cope with severe accidents, involving core meltdown. A specific melt spreading area has been designed within the containment. This core catcher will be flooded by water, which transfers the decay heat to the containment heat removal system. To improve cooling, horizontal flow channels made of cast iron are located also below the core catcher. STUK, the radiation and nuclear safety authority in Finland, wanted an independent study of the functionality of the core catcher design. Effect of the presence of insulation material and boric acid in the cooling water was to be studied, as well as the general behavior of the system in different phases of the flooding of the core melt spreading area. To verify the function of the core catcher design, a scaled down test facility was built at Lappeenranta University of Technology. Since there are some physical restrictions of a test facility computational tools were applied especially for the tests where steady state conditions could not be reached without endangering the integrity of the test facility. This paper introduces the Volley test facility, computational simulations and compares them with the test results. Simulated temperatures of those Volley tests, which could be run until steady state conditions, are very close to the measured temperatures. It can be concluded also, that the temperatures are evidently below the cast iron melting point with heat fluxes used in the tests, if there is a small flow inside the cooling channels or even in case when only a few adjacent cooling channels are totally dry. (authors)

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

Science.gov (United States)

Bragg-Sitton, Shannon M.

2005-01-01

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

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

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

UPTF-TRAM test A2. Formation of stratified flow in the hot leg

International Nuclear Information System (INIS)

Tenckhoff; Brand, B.; Weiss, P.

1992-10-01

The separate effect UPTF TRAM Test A2 consisting of six runs was designed to investigate flow regimes in the hot leg of a pressurized water reactor under two-phase natural circulation conditions. In particular, the following phenomena were investigated: - Formation of different flow regimes, e.g. stratified and slug flow in the hot leg under different boundary conditions; -Correlation between flow regime and boundary conditions of the system (mass flows, water level etc.); - Mechanism of the transport of water into the steam generator. The test runs are divided into two groups: a) Test Runs 01a, 01b and 02b with steam injection through the core simulator: In these test runs the steam injection through the core simulator was increased stepwise. In each step the steam injection was kept constant for about 100 s in order to observe steady water distribution in the hot leg and SG-simulator of broken loop. b) Test Runs 03c, 04c and 04d with steam and water injection through the core simulator: These test runs were performed at a constant steam injection rate and the water injection rate was increased stepwise. In order to verify the consistency of scaling with the pressure, the test runs were carried out at different pressures as: a) Runs 01a and 01b at 15 bar, and Run 02b at 3 bar b) Runs 03c, 04c and 04d at 15, 3 and 5 bar respectively. A preliminary evaluation of the test is presented in the Quick Look Report. (orig.) [de

Predictions of the Bypass Flows in the HTR-PM Reactor Core

International Nuclear Information System (INIS)

Sun Jun; Chen Zhipeng; Zheng Yanhua; Shi Lei; Li Fu

2014-01-01

In the HTR-PM reactor core, the basic structure materials are large amount of graphite reflectors and carbon bricks. Small gaps among those graphite and carbon bricks are widespread in the reactor core so that the cold helium flow may be bypassed and not completely heated. The bypass flows in relative lower temperature would change the flow and temperature distributions in the reactor core, therefore, the accurate prediction of bypass flows need to be carried out carefully to evaluate the influence to the reactor safety. Based on the characteristics of the bypass flow problem, hybrid method of the flow network and the CFD tools was employed to represent the connections and calculate flow distributions of all the main flow and bypass flow paths. In this paper, the hybrid method was described and applied to specific bypass flow problem in the HTR-PM. Various bypass flow paths in the HTR-PM were reviewed, figured out, and modeled by the flow network and the CFD methods, including the axial vertical gaps in the side reflectors, control rod channels, absorber sphere channels and radial gap flow through keys around the hot helium plenum. The bypass flow distributions and its flow rate ratio to the total flow rate in the primary loop were also calculated, discussed and evaluated. (author)

Reverse Flow Engine Core Having a Ducted Fan with Integrated Secondary Flow Blades

Science.gov (United States)

Kisska, Michael K. (Inventor); Princen, Norman H. (Inventor); Kuehn, Mark S. (Inventor); Cosentino, Gary B. (Inventor)

2014-01-01

Secondary air flow is provided for a ducted fan having a reverse flow turbine engine core driving a fan blisk. The fan blisk incorporates a set of thrust fan blades extending from an outer hub and a set of integral secondary flow blades extending intermediate an inner hub and the outer hub. A nacelle provides an outer flow duct for the thrust fan blades and a secondary flow duct carries flow from the integral secondary flow blades as cooling air for components of the reverse flow turbine engine.

Application of the extended Kalman filtering for the estimation of core coolant flow rate in pressurized water reactors

International Nuclear Information System (INIS)

Shieh, D.J.; Upadhyaya, B.R.

1986-01-01

In-core neutron detector and core-exit temperature signals in a pressurized water reactor (PWR) satisfy the condition of observability of the core dynamic system, and can be used to estimate nonmeasurable state variables and model parameters. The extension of the Kalman filtering technique is very useful for direct parameter estimation. This approach is applied to the determination of core coolant mass flow rate in PWRs and is evaluated using in-core measurements at the Loss-of-Fluid Test (LOFT) reactor. The influence of model uncertainties on the estimation accuracy was studied using the ambiguity function analysis. A sequential discretization method was developed to achieve faster convergence to the true value, avoiding model discretization at each sample point. The performance of the extended Kalman filter and the computational innovations were evaluated using a reduced order core dynamic model of the LOFT reactor and random data simulation. The technique was then applied to the determination of LOFT core coolant flow rate from operational data at 100% and 65% flow conditions

Neutronics of a mixed-flow gas-core reactor

International Nuclear Information System (INIS)

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

1977-11-01

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

Sodium flow distribution test of the air cooler tubes

International Nuclear Information System (INIS)

Uchida, Hiroyuki; Ohta, Hidehisa; Shimazu, Hisashi

1980-01-01

In the heat transfer tubes of the air cooler which is installed in the auxiliary core cooling system of the fast breeder prototype plant reactor ''Monju'', sodium freezing may be caused by undercooling the sodium induced by an extremely unbalanced sodium flow in the tubes. Thus, the sodium flow distribution test of the air cooler tubes was performed to examine the flow distribution of the tubes and to estimate the possibility of sodium freezing in the tubes. This test was performed by using a one fourth air cooler model installed in the water flow test facility. As the test results show, the flow distribution from the inlet header to each tube is almost equal at any operating condition, that is, the velocity deviation from normalized mean velocity is less than 6% and sodium freezing does not occur up to 250% air velocity deviation at stand-by condition. It was clear that the proposed air cooler design for the ''Monju'' will have a good sodium flow distribution at any operating condition. (author)

Evaluation of Core Bypass Flow in the Prismatic VHTR with a Multi-block Experiment

International Nuclear Information System (INIS)

Lee, Jeong Hun; Yoon, Su Jong; Park, Goon Cherl; Kim, Min Hwan

2010-01-01

The core of Prismatic Modular Reactor (PMR) consists of assemblies of hexagonal graphite fuel and reflector elements. The core bypass flow of Very High Temperature Reactor (VHTR) is defined as the core flow that does not pass through the coolant channels but passes through the bypass gap between fuel elements. The increase in bypass flow makes the decrease in effective coolant flow. Since the core bypass flow has a negative impact on safety and efficiency of VHTR, core bypass phenomena have to be investigated to improve the core thermal margin of VHTR. For this purpose, the international project, I-NERI project, has been carried out since 2008. I-NERI project is collaborative project that KAERI and SNU of Korea side and INL, ANL and TAMU of U.S side are involved. In order to evaluate the core bypass flow, the multicolumn and multi-layer experimental facility is designed by SNU. In this experiment, the effect of cross-flow and local variation of bypass gap on the bypass flow distribution is investigated. Furthermore, the experimental data will be used for validation of CFD code or thermal hydraulic analysis codes such as GAMMA or GAS-NET

A High-Resolution Continuous Flow Analysis System for Polar Ice Cores

DEFF Research Database (Denmark)

Dallmayr, Remi; Goto-Azuma, Kumiko; Kjær, Helle Astrid

2016-01-01

of Polar Research (NIPR) in Tokyo. The system allows the continuous analysis of stable water isotopes and electrical conductivity, as well as the collection of discrete samples from both inner and outer parts of the core. This CFA system was designed to have sufficiently high temporal resolution to detect...... signals of abrupt climate change in deep polar ice cores. To test its performance, we used the system to analyze different climate intervals in ice drilled at the NEEM (North Greenland Eemian Ice Drilling) site, Greenland. The quality of our continuous measurement of stable water isotopes has been......In recent decades, the development of continuous flow analysis (CFA) technology for ice core analysis has enabled greater sample throughput and greater depth resolution compared with the classic discrete sampling technique. We developed the first Japanese CFA system at the National Institute...

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

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)

Development of flow network analysis code for block type VHTR core by linear theory method

International Nuclear Information System (INIS)

Lee, J. H.; Yoon, S. J.; Park, J. W.; Park, G. C.

2012-01-01

VHTR (Very High Temperature Reactor) is high-efficiency nuclear reactor which is capable of generating hydrogen with high temperature of coolant. PMR (Prismatic Modular Reactor) type reactor consists of hexagonal prismatic fuel blocks and reflector blocks. The flow paths in the prismatic VHTR core consist of coolant holes, bypass gaps and cross gaps. Complicated flow paths are formed in the core since the coolant holes and bypass gap are connected by the cross gap. Distributed coolant was mixed in the core through the cross gap so that the flow characteristics could not be modeled as a simple parallel pipe system. It requires lot of effort and takes very long time to analyze the core flow with CFD analysis. Hence, it is important to develop the code for VHTR core flow which can predict the core flow distribution fast and accurate. In this study, steady state flow network analysis code is developed using flow network algorithm. Developed flow network analysis code was named as FLASH code and it was validated with the experimental data and CFD simulation results. (authors)

Measurement of two-phase flow at the core upper plenum interface under simulated reflood conditions

International Nuclear Information System (INIS)

Thomas, D.G.; Combs, S.K.; Bagwell, M.E.

1980-01-01

Objectives of the Instrument Development Loop program were to simulate flows at the core/upper plenum interface during the reflood phase of a LOCA and to develop instruments for measuring mass-flows at this interface. A tie plate drag body was developed and tested successfully, and the data obtained were shown to be equivalent to pressure drops. The tie-plate drag body gave useful measurements in pure downflow, and the drag/turbine combination correlates with mass flow for high upflow

Fort St. Vrain core performance

International Nuclear Information System (INIS)

McEachern, D.W.; Brown, J.R.; Heller, R.A.; Franek, W.J.

1977-07-01

The Fort St. Vrain High Temperature Gas Cooled Reactor core performance has been evaluated during the startup testing phase of the reactor operation. The reactor is graphite moderated, helium cooled, and uses coated particle fuel and on-line flow control to each of the 37 refueling regions. Principal objectives of startup testing were to determine: core and control system reactivity, radial power distribution, flow control capability, and initial fission product release. Information from the core demonstrates that Technical Specifications are being met, performance of the core and fuel is as expected, flow and reactivity control are predictable and simple for the operator to carry out

Flow with boiling in four-cusp channels simulating damaged core in PWR type reactors

International Nuclear Information System (INIS)

Esteves, M.M.

1985-01-01

The study of subcooled nucleate flow boiling in non-circular channels is of great importance to engineering applications in particular to Nuclear Engineering. In the present work, an experimental apparatus, consisting basically of a refrigeration system, running on refrigerant-12, has been developed. Preliminary tests were made with a circular tube. The main objective has been to analyse subcooled flow boiling in four-cusp channels simulating the flow conditions in a PWR core degraded by accident. Correlations were developed for the forced convection film coefficient for both single-phase and subcooled flow boiling. The incipience of boiling in such geometry has also been studied. (author) [pt

RAMONA-3B calculations of core-wide and regional power/flow oscillations - comparison with Oskarshamn 3 natural circulation test data

International Nuclear Information System (INIS)

Andersson, S.; Stepniewski, M.

1991-01-01

The 3-D BWR Dynamics Code RAMONA-3D is used for a study of power and flow fluctuations at conditions close to observed limit-cycle out-of-phase oscillations during the start-up stability test at the Oskarshamn 3 internal pump BWR. The purpose of the work, which is sponsored by the Swedish Nuclear Inspectorate, SKI, is to see if the observed first order azimuthal oscillations can be reproduced in a simulation using no a-priori information on the oscillation pattern when setting up the model. The experimental data contain dynamic information on the local neutron flux at 32 LPRM detector locations and the inlet flow to 8 fuel channels. The oscillation patterns, i.e. amplitudes and phases at the resonance frequency for local power and flow, are evaluated by direct Fourier transformation of small time slices of the data. A large RAMONA-3B model, using 350 and 268 channels in the neutronic and hydraulic models respectively, each divided into 25 axial nodes, was used for the analyses. The agreement between measured and calculated local flux and flow oscillation frequency and patterns at the limit-cycle threshold is excellent, taking into account the limitations of the half-core model used

Experimental Investigation on Cross Flow of Wedge-shaped Gap in the core of Prismatic VHTR

International Nuclear Information System (INIS)

Lee, Jeong Hun; Park, Goon Cherl; Cho, Hyoung Kyu; Yoon, Su Jong

2014-01-01

The core of the PMR type reactor consists of assemblies of hexagonal graphite blocks. The graphite blocks have lots of advantages for neutron economy and high temperature structural integrity. The height and flat-to-flat width of fuel bock are 793 mm and 360 mm, respectively. Each block has 108 coolant channels of which the diameter is 16 mm. And there are gaps between blocks not only vertically but also horizontally for reloading of the fuel elements. The vertical gap induces the bypass flow and through the horizontal gap the cross flow is formed. Since the complicated flow distribution occurs by the bypass flow and cross flow, flow characteristics in the core of the PMR reactor cannot be treated as a simple pipe flow. The fuel zone of the PMR core consists of multiple layers of fuel blocks. The shape change of the fuel blocks could be caused by the thermal expansion and fast-neutron induced shrinkage. It could make different axial shrinkage of fuel block and this leads to wedge-shaped gaps between two stacked fuel blocks. The cross flow is often considered as a leakage flow through the horizontal gap between stacked fuel blocks and it complicates the flow distribution in the reactor core by connecting the coolant channel and the bypass gap. Moreover, the cross flow could lead to uneven coolant distribution and consequently cause superheating of individual fuel element zones with increased fission product release. Since the core cross flow has a negative impact on safety and efficiency of VHTR, core cross flow phenomena have to be investigated to improve the core thermal margin of VHTR. To develop the cross flow loss coefficient model for determination of the flow distribution for PMR core analysis codes, study on cross flow for PMR200 core is essential. In particular, to predict the amount of flow through the cross flow gap, obtaining accurate flow loss coefficient is important. In this study, the full-scale cross flow experimental facility was constructed to

Calculation of pressure drop and flow redistribution in the LMFBR core

International Nuclear Information System (INIS)

Morgado, O.J.

1984-01-01

The flow redistribution through fuel assemblies of LMFBRs: for the correct calculation of mass flow rates and pressure drop, are studied. Using a quasi-static formulation of conservation equations of mass and energy, a computer program was developed to simulate any arbitrary number of flow channels, operating at different linear power levels. Therefore f flow channels, operating at different linear power levels. Therefore, it was possible to perform thermal transient calculations for the Clinch River reactor core. The results of the calculations agree with the data found in the literature and supply accurate information about flow redistribution, average temperature, and pressure drop in the core, when the reactor is operated at conditions from the designed flow conditions, as is always the case in a load changing operation, or during transients. (Autor) [pt

Simulant-material experimental investigation of flow dynamics in the CRBR Upper-Core Structure

International Nuclear Information System (INIS)

Wilhelm, D.; Starkovich, V.S.; Chapyak, E.J.

1982-09-01

The results of a simulant-material experimental investigation of flow dynamics in the Clinch River Breeder Reactor (CRBR) Upper Core Structure are described. The methodology used to design the experimental apparatus and select test conditions is detailed. Numerous comparisons between experimental data and SIMMER-II Code calculations are presented with both advantages and limitations of the SIMMER modeling features identified

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

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

Computational fluid dynamic analysis of core bypass flow phenomena in a prismatic VHTR

International Nuclear Information System (INIS)

Sato, Hiroyuki; Johnson, Richard; Schultz, Richard

2010-01-01

The core bypass flow in a prismatic very high temperature reactor (VHTR) is an important design consideration and can have considerable impact on the condition of reactor core internals including fuels. The interstitial gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The occurrence of hot spots in the core and lower plenum and hot streaking in the lower plenum (regions of very hot gas flow) are affected by bypass flow. In the present study, three-dimensional computational fluid dynamic (CFD) calculations of a typical prismatic VHTR are conducted to better understand bypass flow phenomena and establish an evaluation method for the reactor core using the commercial CFD code FLUENT. Parametric calculations changing several factors in a one-twelfth sector of a fuel column are performed. The simulations show the impact of each factor on bypass flow and the resulting flow and temperature distributions in the prismatic core. Factors include inter-column gap-width, turbulence model, axial heat generation profile and geometry change from irradiation-induced shrinkage in the graphite block region. It is shown that bypass flow provides a significant cooling effect on the prismatic block and that the maximum fuel and coolant channel outlet temperatures increase with an increase in gap-width, especially when a peak radial factor is applied to the total heat generation rate. Also, the presence of bypass flow causes a large lateral temperature gradient in the block and also dramatically increases the variation in coolant channel outlet temperatures for a given block that may have repercussions on the structural integrity of the graphite, the neutronics and the potential for hot streaking and hot spots occurring in the lower plenum.

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

Long-term pumping test in borehole KR24 flow measurements

Energy Technology Data Exchange (ETDEWEB)

Rouhiainen, P.; Poellaenen, J. [PRG-Tec Oy, Espoo (Finland)

2005-09-15

The Difference Flow method can be used for the relatively fast determination of transmissivity and hydraulic head in fractures or fractured zones in cored boreholes. In this study, the Difference Flow method was used for hydraulic crosshole interference tests. The tests were performed in boreholes KR24 (pumped borehole) KR4, KR7, KR8, KRlO, KR14, KR22, KR22B, KR26, KR27, KR27B, KR28 and KR28B at Olkiluoto during the first and second quarters of 2004. The distance between the boreholes varies from approximately tens of meters to hundreds of meters. All the measurements were carried out in open boreholes, i.e. no packers were used. For interpretation, a normal single hole test was first performed in each borehole. Flow rates and drawdown were first measured both without pumping and with pumping the borehole under test. For practical reasons, the data set is neither complete nor similar in all tested boreholes. Connected flow to borehole KR24 was detected in all these boreholes. These flow responses were concentrated on a few zones. (orig.)

Evaluation of advanced two-phase flow instrumentation in SCTF Core-1

International Nuclear Information System (INIS)

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

1984-03-01

In the Slab Core Test Facility (SCTF) Core-I, advanced two-phase flow instruments have been provided by the USNRC to measure the thermohydraulic behavior in the primary system including pressure vessel during the end of blowdown, refill and reflood phases of a postulated loss-of-coolant accident in a pressurized water reactor. The advanced instruments are turbine meters, drag disks, γ-densitometers, spool pieces, liquid level detectors (LLD), fluid distribution grids (FDG), impedance probes (flag, prong and string probes), film probes, and video optical probes. This report presents evaluated results of the data from these instruments. Some instruments are quantitatively evaluated by comparing with the data from the conventional instruments or the other advanced instruments. Main conclusions are as follows: (1) The spool pieces and the γ-densitometers work well and provide satisfactory results; (2) Some of the turbine meters, the impedance probes and the film probes give partially reasonable results, but still more improvements are required; (3) Most of the LLDs, the FDGs, the impedance probes, and the film probes do not work well due to a hard cable corrosion, and (4) The video optical probes give clear image of the flow pattern. (author)

Integral manifolding structure for fuel cell core having parallel gas flow

Science.gov (United States)

Herceg, Joseph E.

1984-01-01

Disclosed herein are manifolding means for directing the fuel and oxidant gases to parallel flow passageways in a fuel cell core. Each core passageway is defined by electrolyte and interconnect walls. Each electrolyte and interconnect wall consists respectively of anode and cathode materials layered on the opposite sides of electrolyte material, or on the opposite sides of interconnect material. A core wall projects beyond the open ends of the defined core passageways and is disposed approximately midway between and parallel to the adjacent overlaying and underlying interconnect walls to define manifold chambers therebetween on opposite sides of the wall. Each electrolyte wall defining the flow passageways is shaped to blend into and be connected to this wall in order to redirect the corresponding fuel and oxidant passageways to the respective manifold chambers either above or below this intermediate wall. Inlet and outlet connections are made to these separate manifold chambers respectively, for carrying the fuel and oxidant gases to the core, and for carrying their reaction products away from the core.

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)

ASCOT-1, Thermohydraulics of Axisymmetric PWR Core with Homogeneous Flow During LOCA

International Nuclear Information System (INIS)

1978-01-01

1 - Nature of the physical problem solved: ASCOT-1 is used to analyze the thermo-hydraulic behaviour in a PWR core during a loss-of-coolant accident. 2 - Method of solution: The core is assumed to be axisymmetric two-dimensional and the conservation laws are solved by the method of characteristics. For the temperature response of fuel in the annular regions into which the core is divided, the heat conduction equations are solved by an explicit method with averaged flow conditions. 3 - Restrictions on the complexity of the problem: Axisymmetric two-dimensional homogeneous flows

Forces on zonal flows in tokamak core turbulence

International Nuclear Information System (INIS)

Hallatschek, K.; Itoh, K.

2005-01-01

The saturation of stationary zonal flows (ZF) in the core of a tokamak has been analyzed in numerical fluid turbulence computer studies. The model was chosen to properly represent the kinetic global plasma flows, i.e., undamped stationary toroidal or poloidal flows and Landau damped geodesic acoustic modes. Reasonable agreement with kinetic simulations in terms of magnitude of transport and occurrence of the Dimits shift was verified. Contrary to common perception, in the final saturated state of turbulence and ZFs, the customary perpendicular Reynolds stress continues to drive the ZFs. The force balance is established by the essentially quasilinear parallel Reynolds stress acting on the parallel return flows required by incompressibility. (author)

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)

Vortex flow during early and late left ventricular filling in normal subjects: quantitative characterization using retrospectively-gated 4D flow cardiovascular magnetic resonance and three-dimensional vortex core analysis.

Science.gov (United States)

Elbaz, Mohammed S M; Calkoen, Emmeline E; Westenberg, Jos J M; Lelieveldt, Boudewijn P F; Roest, Arno A W; van der Geest, Rob J

2014-09-27

LV diastolic vortex formation has been suggested to critically contribute to efficient blood pumping function, while altered vortex formation has been associated with LV pathologies. Therefore, quantitative characterization of vortex flow might provide a novel objective tool for evaluating LV function. The objectives of this study were 1) assess feasibility of vortex flow analysis during both early and late diastolic filling in vivo in normal subjects using 4D Flow cardiovascular magnetic resonance (CMR) with retrospective cardiac gating and 3D vortex core analysis 2) establish normal quantitative parameters characterizing 3D LV vortex flow during both early and late ventricular filling in normal subjects. With full ethical approval, twenty-four healthy volunteers (mean age: 20±10 years) underwent whole-heart 4D Flow CMR. The Lambda2-method was used to extract 3D LV vortex ring cores from the blood flow velocity field during early (E) and late (A) diastolic filling. The 3D location of the center of vortex ring core was characterized using cylindrical cardiac coordinates (Circumferential, Longitudinal (L), Radial (R)). Comparison between E and A filling was done with a paired T-test. The orientation of the vortex ring core was measured and the ring shape was quantified by the circularity index (CI). Finally, the Spearman's correlation between the shapes of mitral inflow pattern and formed vortex ring cores was tested. Distinct E- and A-vortex ring cores were observed with centers of A-vortex rings significantly closer to the mitral valve annulus (E-vortex L=0.19±0.04 versus A-vortex L=0.15±0.05; p=0.0001), closer to the ventricle's long-axis (E-vortex: R=0.27±0.07, A-vortex: R=0.20±0.09, p=0.048) and more elliptical in shape (E-vortex: CI=0.79±0.09, A-vortex: CI=0.57±0.06; vortex. The circumferential location and orientation relative to LV long-axis for both E- and A-vortex ring cores were similar. Good to strong correlation was found between vortex shape and

Influence of core flows on the decade variations of the polar motion

Science.gov (United States)

Hulot, G.; Le Huy, M.; Le Mouël, J.-L.

We address the possibility for the core flows that generate the geomagnetic field to contribute significantly to the decade variations of the mean pole position (generally called the Markowitz wobble). This assumption is made plausible by the observation that the flow at the surface of the core-estimated from the geomagnetic secular variation models-experiences important changes on this time scale. We discard the viscous and electromagnetic core-mantle couplings and consider only the pressure torque pf resulting from the fluid flow overpressure acting on the non-spherical core-mantle boundary (CMB) at the bottom of the mantle, and the gravity torque gf due to the density heterogeneity driving the core flow. We show that forces within the core balance each other on the time scale considered and, using global integrals over the core, the mantle and the whole Earth, we write Euler's equation for the mantle in terms of two more useful torques Pgeo and . The "geostrophic torque", γ Pgeo incorporates γpf and part of γgf, while γ is another fraction of γgf. We recall how the geostrophic pressure pgeo, and thus γPgeo for a given topography, can be derived from the flow at the CMB and compute the motion of the mean pole from 1900 to 1990, assuming in a first approach that the unknown γ can be neglected. The amplitude of the computed pole motion is three to ten times less than the observed one and out of the phase with it. In order to estimate the possible contribution of γ we then use a second approach and consider the case in which the reference state for the Earth is assumed to be the classical axisymmetric ellipsoidal figure with an almost constant ellipticity within the core. We show that (γPgeo + γ) is then equal to a pseudo-electromagnetic torque γL3, the torque exerted on the core by the component of the Lorentz force along the axis of rotation (this torque exists even though the mantle is assumed insulating). This proves that, at least in this case and

CoreFlow: a computational platform for integration, analysis and modeling of complex biological data.

Science.gov (United States)

Pasculescu, Adrian; Schoof, Erwin M; Creixell, Pau; Zheng, Yong; Olhovsky, Marina; Tian, Ruijun; So, Jonathan; Vanderlaan, Rachel D; Pawson, Tony; Linding, Rune; Colwill, Karen

2014-04-04

A major challenge in mass spectrometry and other large-scale applications is how to handle, integrate, and model the data that is produced. Given the speed at which technology advances and the need to keep pace with biological experiments, we designed a computational platform, CoreFlow, which provides programmers with a framework to manage data in real-time. It allows users to upload data into a relational database (MySQL), and to create custom scripts in high-level languages such as R, Python, or Perl for processing, correcting and modeling this data. CoreFlow organizes these scripts into project-specific pipelines, tracks interdependencies between related tasks, and enables the generation of summary reports as well as publication-quality images. As a result, the gap between experimental and computational components of a typical large-scale biology project is reduced, decreasing the time between data generation, analysis and manuscript writing. CoreFlow is being released to the scientific community as an open-sourced software package complete with proteomics-specific examples, which include corrections for incomplete isotopic labeling of peptides (SILAC) or arginine-to-proline conversion, and modeling of multiple/selected reaction monitoring (MRM/SRM) results. CoreFlow was purposely designed as an environment for programmers to rapidly perform data analysis. These analyses are assembled into project-specific workflows that are readily shared with biologists to guide the next stages of experimentation. Its simple yet powerful interface provides a structure where scripts can be written and tested virtually simultaneously to shorten the life cycle of code development for a particular task. The scripts are exposed at every step so that a user can quickly see the relationships between the data, the assumptions that have been made, and the manipulations that have been performed. Since the scripts use commonly available programming languages, they can easily be

CFD Validation with a Multi-Block Experiment to Evaluate the Core Bypass Flow in VHTR

International Nuclear Information System (INIS)

Yoon, Su Jong; Lee, Jeong Hun; Park, Goon Cherl; Kim, Min Hwan

2010-01-01

Core bypass flow of Very High Temperature Reactor (VHTR) is defined as the ineffective coolant which passes through the bypass gaps between the block columns and the crossflow gaps between the stacked blocks. This flows lead to the variation of the flow distribution in the core and affect the core thermal margin and the safety of VHTR. Therefore, bypass flow should be investigated and quantified. However, it is not a simple question, because the flow path of VHTR core is very complex. In particular, since dimensions of the bypass gap and the crossflow gap are of the order of few millimeters, it is very difficult to measure and to analyze the flow field at those gaps. Seoul National University (SNU) multi-block experiment was carried out to evaluate the bypass flow distribution and the flow characteristics. The coolant flow rate through outlet of each block column was measured, but the local flow field was measured restrictively in the experiment. Instead, CFD analysis was carried out to investigate the local phenomena of the experiment. A commercial CFD code CFX-12 was validated by comparing the simulation results and the experimental data

Numerical solution of heat transfer process in a prismatic VHTR core accompanying bypass and cross flows

International Nuclear Information System (INIS)

Wang, Li; Liu, Qiusheng; Fukuda, Katsuya

2016-01-01

Highlights: • Three-dimensional CFD analysis is conducted for the thermal analysis in the reactor core. • Hot spot temperature, coolant channel outlet temperature distribution are affected by bypass flow. • Bypass gap size has significant influence on temperature and flow distribution in the core. • Cross flow has some effect on the temperature distribution of the coolant in the core due to flow mixing in the cross gaps. - Abstract: Bypass flow and cross flow gaps both exist in the core of a very high temperature gas-cooled reactor (VHTR), which is inevitable owing to tolerances in manufacturing, thermal expansion and irradiation shrinkage. The coolant mass flow rate distribution, temperature distribution, and hot spot temperature are significantly affected by bypass and cross flows. In the present study, three-dimensional CFD analysis is conducted for thermal analysis of the reactor core. A validation study for the turbulence model is performed by comparing the friction coefficient with published correlations. A sensitivity study of the near wall mesh is conducted to ensure mesh quality. Parametric studies are performed by changing the size of the bypass and cross gaps using a one-twelfth sector of a fuel block. Simulation results show the influence of the bypass gap size on temperature distribution and coolant mass flow rate distribution in the prismatic core. It is shown that the maximum fuel and coolant channel outlet temperatures increase with an increase in the gap size, which may lead to a structural risk to the fuel block. The cross flow is divided into two types: the cross flow from the bypass gap to the coolant channels and the cross flow from the high-pressure coolant channels to low-pressure coolant channels. These two types of flow have an opposing influence on the temperature gradient. It is found that the presence of the cross flow gaps may have a significant effect on the distribution of the coolant in the core due to flow mixing in the

Study of core flow distribution for small modular natural circulation lead or lead-alloy cooled fast reactors

International Nuclear Information System (INIS)

Chen, Zhao; Zhao, Pengcheng; Zhou, Guangming; Chen, Hongli

2014-01-01

Highlights: • A core flow distribution calculation code for natural circulation LFRs was developed. • The comparison study between the channel method and the CFD method was conducted. • The core flow distribution analysis and optimization design for a 10MW natural circulation LFR was conducted. - Abstract: Small modular natural circulation lead or lead-alloy cooled fast reactor (LFR) is a potential candidate for LFR development. It has many attractive advantages such as reduced capital costs and inherent safety. The core flow distribution calculation is an important issue for nuclear reactor design, which will provide important input parameters to thermal-hydraulic analysis and safety analysis. The core flow distribution calculation of a natural circulation LFR is different from that of a forced circulation reactor. In a forced circulation reactor, the core flow distribution can be controlled and adjusted by the pump power and the flow distributor, while in a natural circulation reactor, the core flow distribution is automatically adjusted according to the relationship between the local power and the local resistance feature. In this paper, a non-uniform heated parallel channel flow distribution calculation code was developed and the comparison study between the channel method and the CFD method was carried out to assess the exactness of the developed code. The core flow distribution analysis and optimization design for a 10MW natural circulation LFR was conducted using the developed code. A core flow distribution optimization design scheme for a 10MW natural circulation LFR was proposed according to the optimization analysis results

Evaluation of tests for coastdown of reactor coolant flow and measure of primary circuit flow of Angra-1 nuclear power plant

International Nuclear Information System (INIS)

Galetti, M.R.S.; Camargo, C.T.M.; Pontedeiro, A.C.

1987-05-01

The Angra 1 Nuclear Power Plant first reload license was issued after several technical discussions among CNEN, FURNAS and KWU. During the license process CNEN has established that the plant could return to anormal operation if the requirements described in the letter CNEN-DExL-C 06/86 were satisfied. The requirements according to the CNEN Transient and Thermohydraulic Group Analysis were to do again the following tests: 'Primary Flow Measurement' to check if the excess flow measured in the first cycle was held; and Pump Coastdown' to check if the Westinghouse and KWU fuel elements are thermo-hydraulicaly compatibles during transients. The mixed core must keep at least the same safety margin presented on Angra 1 FSAR for the original core. The tests and the analysis of results are described. (Author) [pt

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...

Development of a detailed core flow analysis code for prismatic fuel reactors

International Nuclear Information System (INIS)

Bennett, R.G.

1990-01-01

The detailed analysis of the core flow distribution in prismatic fuel reactors is of interest for modular high-temperature gas-cooled reactor (MHTGR) design and safety analyses. Such analyses involve the steady-state flow of helium through highly cross-connected flow paths in and around the prismatic fuel elements. Several computer codes have been developed for this purpose. However, since they are proprietary codes, they are not generally available for independent MHTGR design confirmation. The previously developed codes do not consider the exchange or diversion of flow between individual bypass gaps with much detail. Such a capability could be important in the analysis of potential fuel block motion, such as occurred in the Fort St. Vrain reactor, or for the analysis of the conditions around a flow blockage or misloaded fuel block. This work develops a computer code with fairly general-purpose capabilities for modeling the flow in regions of prismatic fuel cores. The code, called BYPASS solves a finite difference control volume formulation of the compressible, steady-state fluid flow in highly cross-connected flow paths typical of the MHTGR

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.

Technical description of the test section SUCOT to investigate a water/steam two-phase flow

International Nuclear Information System (INIS)

Daubner, M.; Janssens-Maenhout, G.; Knebel, J.U.

2002-02-01

Within the POOLTHY Project of the European Union (Euratom Fourth Framework Programme contract FJ4J-CT95-0003) an active/passive concept (SUCO-Programme) was investigated which controls the heat removal after a potential core melt-down accident in an evolutionary light water reactor by spreading and stabilising the core melt in the reactor sump and flooding the melt with sump water from above. The experiments were performed in the test facility SUCOT which was designed and erected at the Institute for Nuclear and Energy Technologies (IKET). The report gives an overview on the SUCO-Programme and the scaling analysis, which was applied to design the test facility SUCOT. A detailed technical description of the test facility SUCOT is given, in which the natural circulation driven two-phase flow within the reactor sump and relevant phenomena such as flow boiling, disperse bubbly flow with and without mass transfer, and geysering are investigated. The major components of the test facility, the three-loop system and the instrumentation are described. Finally, a perspective for future application of the gained knowledge is given. (orig.) [de

Decadal variability in core surface flows deduced from geomagnetic observatory monthly means

DEFF Research Database (Denmark)

Whaler, K. A.; Olsen, Nils; Finlay, Chris

2016-01-01

. On the annual to decadal timescale, the SV is generated primarily by advection in the fluid outer core. We demonstrate the utility of the revised monthly means by calculating models of the core surface advective flow between 1997 and 2013 directly from the SV data. One set of models assumes flow......Monthly means of the magnetic field measurements at ground observatories are a key data source for studying temporal changes of the core magnetic field. However, when they are calculated in the usual way, contributions of external (magnetospheric and ionospheric) origin may remain, which make them...... less favourable for studying the field generated by dynamo action in the core. We remove external field predictions, including a new way of characterising the magnetospheric ring current, from the data and then calculate revised monthly means using robust methods. The geomagnetic secular variation (SV...

Calculation of local flow conditions in the lower core of a PWR with code-Saturne

International Nuclear Information System (INIS)

Fournier, Y.

2003-01-01

In order to better understand the stresses to which fuel rods are subjected, we need to improve our knowledge of the fluid flow inside the core. A code specialized for calculations in tube bundles is used to calculate the flow inside the whole of the core, with a resolution at the assembly level. Still, it is necessary to obtain realistic entry conditions, and these depend on the flow in the downcomer and lower plenum. Also, the flow in the first stages of the core features 4 incoming jets per assembly, and requires a resolution much finer than that used for the whole core calculation. A series of calculations are thus run with our incompressible Navier-Stokes solver, Code-Saturne, using a classical Ranse turbulence model. The first calculations involve a detailed geometry, including part of the cold legs, downcomer, lower plenum, and lower core of a pressurized water reactor. The level of detail includes most obstacles below the core. The lower core plate, being pierced with close to 800 holes, cannot be realistically represented within a practical mesh size, so that a head loss model is used. The lower core itself requiring even more detail is also represented with head losses. We make full use of Code-Saturne's non conforming mesh possibilities to represent a complex geometry, being careful to retain a good mesh quality. Starting just under the lower core, the mesh is aligned with fuel rod assemblies, so that different types of assemblies can be represented through different head loss coefficients. These calculations yield steady-state or near steady-state results, which are compared to experimental data, and should be sufficient to yield realistic entry conditions for full core calculations at assembly width resolution, and beyond those mechanical strain calculations. We are also interested in more detailed flow conditions and fluctuations in the lower core area, so as to better quantify vibrational input. This requires a much higher resolution, which we limit

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)

Ultrasonic density detector for vessel and reactor core two-phase flow measurements

International Nuclear Information System (INIS)

Arave, A.E.

1979-01-01

A local ultrasonic density (LUD) detector has been developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory for the Loss-of-Fluid Test (LOFT) reactor vessel and core two-phase flow density measurements. The principle of operating the sensor is the change in propagation time of a torsional ultrasonic wave in a metal transmission line as a function of the density of the surrounding media. A theoretical physics model is presented which represents the total propagation time as a function of the sensor modulus of elasticity and polar moment of inertia

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)

Effects of lower plenum flow structure on core inlet flow of ABWR

International Nuclear Information System (INIS)

Watanabe, Shun; Abe, Yutaka; Kaneko, Akiko; Watanabe, Fumitoshi; Tezuka, Kenichi

2010-01-01

The evaluation of coolant flow structure at a lower plenum of an advanced boiling water reactor (ABWR) in which there are many structures is very important in order to improve generating power. Although the simulation results by CFD (Computational Fluid Dynamics) codes can predict such complicated flow in the lower plenum, it is required to establish the database of flow structure in lower plenum of ABWR experimentally for the benchmark of the CFD codes. In the model of the lower plenum, we measured velocity profiles with LDV and PIV. And differential pressure of constructed model is measured with differential pressure instrument. It was identified that the velocity and differential pressure profiles also showed the tendency to be flat in the core inlet. Moreover, vortexes were observed around side entry orifice by PIV measurement. (author)

Cores, Joins and the Fano-Flow Conjectures

Directory of Open Access Journals (Sweden)

Jin Ligang

2018-02-01

Full Text Available The Fan-Raspaud Conjecture states that every bridgeless cubic graph has three 1-factors with empty intersection. A weaker one than this conjecture is that every bridgeless cubic graph has two 1-factors and one join with empty intersection. Both of these two conjectures can be related to conjectures on Fano-flows. In this paper, we show that these two conjectures are equivalent to some statements on cores and weak cores of a bridgeless cubic graph. In particular, we prove that the Fan-Raspaud Conjecture is equivalent to a conjecture proposed in [E. Steffen, 1-factor and cycle covers of cubic graphs, J. Graph Theory 78 (2015 195–206]. Furthermore, we disprove a conjecture proposed in [G. Mazzuoccolo, New conjectures on perfect matchings in cubic graphs, Electron. Notes Discrete Math. 40 (2013 235–238] and we propose a new version of it under a stronger connectivity assumption. The weak oddness of a cubic graph G is the minimum number of odd components (i.e., with an odd number of vertices in the complement of a join of G. We obtain an upper bound of weak oddness in terms of weak cores, and thus an upper bound of oddness in terms of cores as a by-product.

Solid oxide fuel cell having monolithic cross flow core and manifolding

International Nuclear Information System (INIS)

Poeppel, R.B.; Dusek, J.T.

1984-01-01

This invention discloses a monolithic core construction having the flow passageways for the fuel and for the oxidant gases extended transverse to one another, whereby full face core manifolding can be achieved for these gases and their reaction products. The core construction provides that only anode material surround each fuel passageway and only cathode material surround each oxidant passageway, each anode and each cathode further sandwiching at spaced opposing sides electrolyte and interconnect materials to define electrolyte and interconnect walls. Webs of the cathode and anode material hold the electrolyte and interconnect walls spaced apart to define the flow passages. The composite anode and cathode wall structures are further alternately stacked on one another (with the separating electrolyte or interconnect material typically being a single common layer) whereby the fuel passageway and the oxidant passageways are disposed transverse to one another

Preferential flow through intact soil cores: Effects of matrix head

Energy Technology Data Exchange (ETDEWEB)

Langner, H.W.; Gaber, H.M.; Wraith, J.M.; Huwe, B.; Inskeep, W.P.

1999-12-01

Continuous soil pores may act as pathways for preferential flow depending on their size and water status (filled or drained), the latter being largely controlled by the soil matrix head (h). The literature contains a wide range of proposed minimal pore sizes that may contribute to preferential flow. The objective of this study was to examine the relationship between h (and corresponding pore sizes) and preferential solute transport in a naturally structured soil. Tracer ({sup 3}H{sub 2}O and pentafluorobenzoic acid, [PFBA]) miscible displacement experiments were performed at several h values in intact soil cores (15-cm diameter, 30-cm length) using an apparatus especially suited to maintain constant h while collecting large effluent volumes. To test for the occurrence of preferential flow, observed breakthrough curves (BTCs) were evaluated for physical nonequilibrium (PNE) using a comparison between fitted local equilibrium (PNE) and PNE models. Fitting results of the observed BTCs indicated absence of PNE in all solute transport experiments at h {le} {minus}10 cm. Experiments at h {ge} {minus}5 cm consistently exhibited PNE conditions, indicating the presence of preferential flow. These results suggest that soil pores with effective radii of 150 {micro}m and smaller (water-filled at h = {minus}10 cm) do not contribute to preferential flow. Observed pore water velocities were not indicative of the presence or absence of preferential flow conditions. Continuous measurements of soil water content ({theta}) using time domain reflectometry (TDR) revealed that at h = {minus}10 cm, <2% of the soil volume had drained.

Experimental study on two-phase flow natural circulation in a core catcher cooling channel for EU-APR1400 using air-water system

Energy Technology Data Exchange (ETDEWEB)

Song, Ki Won [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Korea Atomic Energy Research Institute, Daejeon 34057 (Korea, Republic of); Nguyen, Thanh Hung [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47906 (United States); Ha, Kwang Soon; Kim, Hwan Yeol; Song, Jinho [Korea Atomic Energy Research Institute, Daejeon 34057 (Korea, Republic of); Park, Hyun Sun [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Revankar, Shripad T., E-mail: shripad@postech.ac.kr [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); School of Nuclear Engineering, Purdue University, West Lafayette, IN 47906 (United States); Kim, Moo Hwan [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Korea Institute of Nuclear Safety, Daejeon 305-338 (Korea, Republic of)

2017-05-15

Highlights: • Two-phase flow regimes and transition behavior were observed in the coolant channel. • Test were conducted for natural circulation with air-water. • Data were obtained on flow regime, void fraction, flow rates and re-wetting time. • The data were related to a cooling capability of core catcher system. - Abstract: Ex-vessel core catcher cooling system driven by natural circulation is designed using a full scaled air-water system. A transparent half symmetric section of a core catcher coolant channel of a pressurized water reactor was designed with instrumentations for local void fraction measurement and flow visualization. Two designs of air-water top separator water tanks are studied including one with modified ‘super-step’ design which prevents gas entrainment into down-comer. In the experiment air flow rates are set corresponding to steam generation rate for given corium decay power. Measurements of natural circulation flow rate, spatial local void fraction distribution and re-wetting time near the top wall are carried out for various air flow rates which simulate boiling-induced vapor generation. Since heat transfer and critical heat flux are strongly dependent on the water mass flow rate and development of two-phase flow on the heated wall, knowledge of two-phase flow characteristics in the coolant channel is essential. Results on flow visualization showing two phase flow structure specifically near the high void accumulation regions, local void profiles, rewetting time, and natural circulation flow rate are presented for various air flow rates that simulate corium power levels. The data are useful in assessing the cooling capability of and safety of the core catcher system.

NASA Glenn Research Center, Propulsion Systems Laboratory: Plan to Measure Engine Core Flow Water Vapor Content

Science.gov (United States)

Oliver, Michael

2014-01-01

This presentation will be made at the 92nd AIAA Turbine Engine Testing Working Group (TETWoG), a semi-annual technical meeting of turbine engine testing professionals. The objective is to describe an effort by NASA to measure the water vapor content on the core airflow in a full scale turbine engine ice crystal icing test and to open a discussion with colleagues how to accurately conduct the measurement based on any previous collective experience with the procedure, instruments and nature of engine icing testing within the group. The presentation lays out the schematics of the location in the flow path from which the sample will be drawn, the plumbing to get it from the engine flow path to the sensor and several different water vapor measurement technologies that will be used: Tunable diode laser and infrared spectroscopy.

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)

Characterization of interfacial waves in horizontal core-annular flow

Science.gov (United States)

Tripathi, Sumit; Bhattacharya, Amitabh; Singh, Ramesh; Tabor, Rico F.

2016-11-01

In this work, we characterize interfacial waves in horizontal core annular flow (CAF) of fuel-oil and water. Experimental studies on CAF were performed in an acrylic pipe of 15.5mm internal diameter, and the time evolution of the oil-water interface shape was recorded with a high speed camera for a range of different flow-rates of oil (Qo) and water (Qw). The power spectrum of the interface shape shows a range of notable features. First, there is negligible energy in wavenumbers larger than 2 π / a , where a is the thickness of the annulus. Second, for high Qo /Qw , there is no single dominant wavelength, as the flow in the confined annulus does not allow formation of a preferred mode. Third, for lower Qo /Qw , a dominant mode arises at a wavenumber of 2 π / a . We also observe that the power spectrum of the interface shape depends weakly on Qw, and strongly on Qo, perhaps because the net shear rate in the annulus appears to depend weakly on Qw as well. We also attempt to build a general empirical model for CAF by relating the interfacial stress (calculated via the mean pressure gradient) to the flow rate in the annulus, the annular thickness and the core velocity. Authors are thankful to Orica Mining Services (Australia) for the financial support.

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.

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)

Applying CFD in the Analysis of Heavy-Oil Transportation in Curved Pipes Using Core-Flow Technique

Directory of Open Access Journals (Sweden)

S Conceição

2017-06-01

Full Text Available Multiphase flow of oil, gas and water occurs in the petroleum industry from the reservoir to the processing units. The occurrence of heavy oils in the world is increasing significantly and points to the need for greater investment in the reservoirs exploitation and, consequently, to the development of new technologies for the production and transport of this oil. Therefore, it is interesting improve techniques to ensure an increase in energy efficiency in the transport of this oil. The core-flow technique is one of the most advantageous methods of lifting and transporting of oil. The core-flow technique does not alter the oil viscosity, but change the flow pattern and thus, reducing friction during heavy oil transportation. This flow pattern is characterized by a fine water pellicle that is formed close to the inner wall of the pipe, aging as lubricant of the oil flowing in the core of the pipe. In this sense, the objective of this paper is to study the isothermal flow of heavy oil in curved pipelines, employing the core-flow technique. A three-dimensional, transient and isothermal mathematical model that considers the mixture and k-e  turbulence models to address the gas-water-heavy oil three-phase flow in the pipe was applied for analysis. Simulations with different flow patterns of the involved phases (oil-gas-water have been done, in order to optimize the transport of heavy oils. Results of pressure and volumetric fraction distribution of the involved phases are presented and analyzed. It was verified that the oil core lubricated by a fine water layer flowing in the pipe considerably decreases pressure drop.

Construction of the HTTR in-core components

International Nuclear Information System (INIS)

Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Jinza, K.; Miki, T.

1996-01-01

The reactor internals of HTTR consist of graphite and metallic core support structures and shielding blocks and are designed to support core elements and to shield neutron fluence. They also have functions to restrict by-pass flow for ensuring the core cooling performance and to maintain the temperature of metallic core support structures within their design limits. The detailed design of the HTTR core support structure was approved by the government through safety review, 1990-1991. Machining of all graphite components, which consist of about 150 large blocks, was finished in September 1994 successfully. Machining and fabricating of the metallic components were also finished in September. Prior to their installation in the reactor pressure vessel (RPV), the assembly test of actual reactor internals was performed at the works to confirm above mentioned functions. The assembly test was conducted by examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the RPV and reactor internals as well as under the core support plates with respect to structural integrity, and measuring by-pass flow rate through gaps between graphite components which may degrade core performance. The another purpose of the assembly test was to confirm the installation procedure of those components. All components were assembled at the works according to the planned procedure, and the tests were executed while assembling. As a result of the tests, measured level difference and gap width between reactor internals were negligible from core thermal and hydraulic performance point of view. Coolant flows uniformly in circumferential direction at any axial level in the RPV. By-pass flow rate was found to be suppressed sufficiently and far less than the design limit. (J.P.N.)

Performance Tests of Three Flow Distributors Using SMART-ITL with 1-Train CMT

Energy Technology Data Exchange (ETDEWEB)

Bae, Hwang; Ryu, Sung Uk; Shin, Yong-Cheol; Ko, Yung-Joo; Min, Kyoung-Ho; Ryu, Hyo Bong; Park, Jong-Kuk; Bang, Yun-Gon; Chae, Young-Jong; Yi, Sung-Jae; Park, Hyun-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

Passive safety systems (PSSs) are key tools to remove the heat from the core or containment. Safety improvements for SMART have been studied since the Standard Design Approval (SDA) for SMART was certificated in 2012. Active safety systems such as safety injection pumps are replaced by a passive system, which is a kind of the gravity injection system with core makeup tanks (CMT) and safety injection tanks (SIT). All tanks for the passive safety systems are located higher than a pressurized reactor vessel, whose injection nozzles are located around the reactor coolant pumps (RCP). An Integral Test Loop for the SMART design (SMART-ITL) has been constructed and its commissioning tests finished in 2012. SMART-ITL is scaled down by the volume scaling methodology. Its height is conserved and its volume scale ratio is 1/49. The SMART-ITL has all fluid systems of SMART together with a break system and instruments. Recently, a test program to validate the performance of SMART Passive Safety System (PSS) was launched. A scaled-down test facility for SMART PSS was additionally installed at the existing SMART-ITL facility and a set of validation tests were performed. In this paper, the performance tests of the flow distributors using SMART-ITL with 1-train CMT will be discussed. A 1-train passive safety system including a CMT and SIT, which is operated only by gravity force, was additionally installed in the SMART-ITL to replace the active safety system for the SMART design. Several performance tests for the flow distributors were carried out to estimate a designed flow rate. 1. The peak flow rate in a hot test does not reach the value in a cold test, and the approaching time to peak is also delayed during the early stage of gravity injection. 2.. It is verified that the flow rate from a gravity injection depends on the differential pressure in the injection pipe line including a friction and form drag, which can be adjusted by controlling the resistance coefficient.

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.

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)

Evaluation report on CCTF core-II reflood test C2 - 18 (Run 78)

International Nuclear Information System (INIS)

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

1987-03-01

This report presents the result of the upper plenum injection (UPI) test C2 - 18 (Run 78), which was conducted on November 13, 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 phase. The objectives of the test are to investigate the refill behavior with UPI condition and to investigate the reflood behavior with UPI Best-Estimate (BE) condition. The test was performed to simulate refill/reflood behavior with UPI and BE conditions (However, the LPCI flow rate was determined based on single failure of LPCI pumps.). The result of the test showed the followings. (1) Little special phenomena were recognized under UPI and BE conditions in comparison with those under UPI and Evaluation-Model (EM) conditions, although certain special phenoma (i.e. significant fluid oscillation) were recognized under Cold-Leg-Injection (CLI) and BE conditions in comparison with those under CLI and EM conditions. (2) Water inventory in lower plenum increased smoothly due to water injected into both upper plenum and cold leg during refill phase, similarly to that in refill-simulation test with CLI condition. Small difference in refill behavior with UPI condition is the existing of steam condensation in upper plenum, resulting in lower steam binding and higher core cooling during early reflood phase. This indicates the conservatism of UPI against CLI during early reflood phase. (3) The good core-cooling capability was confirmed under UPI and BE conditions. (author)

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…

R and D on thermal hydraulics of core and core-bottom structure

International Nuclear Information System (INIS)

Inagaki, Yoshiyuki; Hino, Ryutaro; Kunitomi, Kazuhiko; Takase, Kazuyuki; Ioka, Ikuo; Maruyama, So

2004-01-01

Thermal hydraulic tests on the core and core-bottom structure of the high-temperature engineering test reactor (HTTR) were carried out with the helium engineering demonstration loop (HENDEL) under simulated reactor operating conditions. The HENDEL was composed of helium gas circulation loops, mother sections (M 1 and M 2 ) and adaptor section (A), and two test sections, i.e. the fuel stack test section (T 1 ) and in-core structure test section (T 2 ). In the T 1 test section simulating a fuel stack of the core, thermal and hydraulic performances of helium gas flowing through a fuel block were investigated for thermal design of the HTTR core. In the T 2 test section simulating the core-bottom structure, demonstration tests were performed to verify the structural integrity of graphite and metal components, seal performance against helium gas leakage among the graphite permanent blocks and thermal mixing performance of helium gas. The above test results in the T 1 and T 2 test sections were applied to the detailed design and licensing works of the HTTR and the HENDEL-loop was dismantled in 1999

Development of a detailed core flow analysis code for prismatic fuel reactors

International Nuclear Information System (INIS)

Bennett, R.G.

1990-01-01

The development of a computer code for the analysis of the detailed flow of helium in prismatic fuel reactors is reported. The code, called BYPASS, solves, a finite difference control volume formulation of the compressible, steady state fluid flow in highly cross-connected flow paths typical of the Modular High-Temperature Gas Cooled Reactor (MHTGR). The discretization of the flow in a core region typically considers the main coolant flow paths, the bypass gap flow paths, and the crossflow connections between them. 16 refs., 5 figs

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

Structural organization of the quiescent core region in a turbulent channel flow

International Nuclear Information System (INIS)

Yang, Jongmin; Hwang, Jinyul; Sung, Hyung Jin

2016-01-01

Highlights: • The structural organization of the quiescent core region in a turbulent channel flow is explored. • The quiescent core region is the uniform momentum zone located at the center of the channel. • The boundary of the quiescent core region can be identified from the probability density function of the streamwise modal velocity. • The prograde and retrograde vortices form a counter-rotating vortex pair at the boundary of the core region. - Abstract: The structural organization of the quiescent core region in a turbulent channel flow was explored using direct numerical simulation data at Re_τ = 930. The quiescent core region is the uniform momentum zone located at the center of the channel, and contains the highest momentum with a low level of turbulence. The boundary of the quiescent core region can be identified from the probability density function of the streamwise modal velocity. The streamwise velocity changes abruptly near the boundary of the core region. The abrupt jump leads the increase of the velocity gradient, which is similar to the vorticity thickness of the laminar superlayer at the turbulent/non-turbulent interface. The strong shear induced from the abrupt change is originated from the vortical structure lying on the boundary of the core region. The spanwise population densities of the prograde and retrograde vortices have a local maximum near the boundary of the core region. The prograde vortex dominantly contributes to the total mean shear near the core boundary and the contribution to the total mean shear rapidly decreases within the core region. The prograde and retrograde vortices form a counter-rotating vortex pair at the boundary of the core region associated with the nibbling mechanism. The boundary of the core region contains large-scale concave and convex features. The concave (convex) core interface is organized by the negative-u (positive-u) regions which induce the ejections (sweeps) around the core boundary.

Peach Bottom Cycle 2 Low Flow Stability Tests analysis using RELAP5/PARCS

International Nuclear Information System (INIS)

Costa, A.L.; Salah, A.B.; D'Auria, F.

2004-01-01

Nowadays, the coupled codes technique, which consists in incorporating threedimensional (3D) neutron modeling of the reactor core into system codes, is extensively used for simulating transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal-hydraulics. So, in this work, the coupled codes technique using RELAP5/3.3-PARCS is applied to simulate stability transients in a BWR (Boiling Water Reactor). Validation has been performed against Peach Bottom-2 Low-Flow Stability Tests. In these transients dynamically complex neutron kinetics coupling with thermal-hydraulics events take place in response to a core pressure perturbation. The calculated coupled code results are herein compared against the available experimental data. (author)

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)

Analysis of fractures in volcanic cores from Pahute Mesa, Nevada Test Site

International Nuclear Information System (INIS)

Drellack, S.L. Jr.; Prothro, L.B.; Roberson, K.E.

1997-09-01

The Nevada Test Site (NTS), located in Nye County, southern Nevada, was the location of 828 announced underground nuclear tests, conducted between 1951 and 1992. Approximately one-third of these tests were detonated near or below the water table. An unavoidable consequence of these testing activities was introducing radionuclides into the subsurface environment, impacting groundwater. Groundwater flows beneath the NTS almost exclusively through interconnected natural fractures in carbonate and volcanic rocks. Information about these fractures is necessary to determine hydrologic parameters for future Corrective Action Unit (CAU)-specific flow and transport models which will be used to support risk assessment calculations for the U.S. Department of Energy, Nevada Operations Office (DOE/NV) Underground Test Area (UGTA) remedial investigation. Fracture data are critical in reducing the uncertainty of the predictive capabilities of CAU-specific models because of their usefulness in generating hydraulic conductivity values and dispersion characteristics used in transport modeling. Specifically, fracture aperture and density (spacing) are needed to calculate the permeability anisotropy of the formations. Fracture mineralogy information is used qualitatively to evaluate diffusion and radionuclide retardation potential in transport modeling. All these data can best be collected through examination of core samples

Analysis of fractures in volcanic cores from Pahute Mesa, Nevada Test Site

Energy Technology Data Exchange (ETDEWEB)

Drellack, S.L. Jr.; Prothro, L.B.; Roberson, K.E. [and others

1997-09-01

The Nevada Test Site (NTS), located in Nye County, southern Nevada, was the location of 828 announced underground nuclear tests, conducted between 1951 and 1992. Approximately one-third of these tests were detonated near or below the water table. An unavoidable consequence of these testing activities was introducing radionuclides into the subsurface environment, impacting groundwater. Groundwater flows beneath the NTS almost exclusively through interconnected natural fractures in carbonate and volcanic rocks. Information about these fractures is necessary to determine hydrologic parameters for future Corrective Action Unit (CAU)-specific flow and transport models which will be used to support risk assessment calculations for the U.S. Department of Energy, Nevada Operations Office (DOE/NV) Underground Test Area (UGTA) remedial investigation. Fracture data are critical in reducing the uncertainty of the predictive capabilities of CAU-specific models because of their usefulness in generating hydraulic conductivity values and dispersion characteristics used in transport modeling. Specifically, fracture aperture and density (spacing) are needed to calculate the permeability anisotropy of the formations. Fracture mineralogy information is used qualitatively to evaluate diffusion and radionuclide retardation potential in transport modeling. All these data can best be collected through examination of core samples.

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

Flow velocity calculation to avoid instability in a typical research reactor core

International Nuclear Information System (INIS)

Oliveira, Carlos Alberto de; Mattar Neto, Miguel

2011-01-01

Flow velocity through a research reactor core composed by MTR-type fuel elements is investigated. Core cooling capacity must be available at the same time that fuel-plate collapse must be avoided. Fuel plates do not rupture during plate collapse, but their lateral deflections can close flow channels and lead to plate over-heating. The critical flow velocity is a speed at which the plates collapse by static instability type failure. In this paper, critical velocity and coolant velocity are evaluated for a typical MTR-type flat plate fuel element. Miller's method is used for prediction of critical velocity. The coolant velocity is limited to 2/3 of the critical velocity, that is a currently used criterion. Fuel plate characteristics are based on the open pool Australian light water reactor. (author)

Device for measuring flow rate in a nuclear reactor core

International Nuclear Information System (INIS)

Hamano, Jiro.

1980-01-01

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

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…

CoreFlow: A computational platform for integration, analysis and modeling of complex biological data

DEFF Research Database (Denmark)

Pasculescu, Adrian; Schoof, Erwin; Creixell, Pau

2014-01-01

between data generation, analysis and manuscript writing. CoreFlow is being released to the scientific community as an open-sourced software package complete with proteomics-specific examples, which include corrections for incomplete isotopic labeling of peptides (SILAC) or arginine-to-proline conversion......A major challenge in mass spectrometry and other large-scale applications is how to handle, integrate, and model the data that is produced. Given the speed at which technology advances and the need to keep pace with biological experiments, we designed a computational platform, CoreFlow, which...... provides programmers with a framework to manage data in real-time. It allows users to upload data into a relational database (MySQL), and to create custom scripts in high-level languages such as R, Python, or Perl for processing, correcting and modeling this data. CoreFlow organizes these scripts...

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)

Correction for Delay and Dispersion Results in More Accurate Cerebral Blood Flow Ischemic Core Measurement in Acute Stroke.

Science.gov (United States)

Lin, Longting; Bivard, Andrew; Kleinig, Timothy; Spratt, Neil J; Levi, Christopher R; Yang, Qing; Parsons, Mark W

2018-04-01

This study aimed to assess how the ischemic core measured by perfusion computed tomography (CTP) was affected by the delay and dispersion effect. Ischemic stroke patients having CTP performed within 6 hours of onset were included. The CTP data were processed twice, generating standard cerebral blood flow (sCBF) and delay- and dispersion-corrected CBF (ddCBF), respectively. Ischemic core measured by the sCBF and ddCBF was then compared at the relative threshold core were used: acute diffusion-weighted imaging or 24-hour diffusion-weighted imaging in patients with complete recanalization. Difference of core volume between CTP and diffusion-weighted imaging was estimated by Mann-Whitney U test and limits of agreement. Patients were also classified into favorable and unfavorable CTP patterns. The imaging pattern classification by sCBF and ddCBF was compared by the χ 2 test; their respective ability to predict good clinical outcome (3-month modified Rankin Scale score) was tested in logistic regression. Fifty-five patients were included in this study. Median sCBF ischemic core volume was 38.5 mL (12.4-61.9 mL), much larger than the median core volume of 17.2 mL measured by ddCBF (interquartile range, 5.5-38.8; P core much closer to diffusion-weighted imaging core references, with the mean volume difference of -0.1 mL (95% limits of agreement, -25.4 to 25.2; P =0.97) and 16.7 mL (95% limits of agreement, -21.7 to 55.2; P core measurement on CTP. © 2018 American Heart Association, Inc.

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

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)

Effect of a Central Graphite Column on a Pebble Flow in a Pebble Bed Core

International Nuclear Information System (INIS)

In, W. K.; Lee, W. J.; Chang, J. H.

2006-01-01

A pebble bed reactor(PBR) uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. The pebble bed core is configured as cylindrical or annular depending on the reactor power. It is well known that an annular core can increase a cores' thermal power. The annular inner core zone is typically filled with movable graphite balls or a fixed graphite column. The first problem with this conventional annular core is that it is difficult to maintain a boundary between the central graphite ball zone and the outer fuel zone. The second problem is that it is expensive to replace the central fixed graphite column after several tens of years of reactor operation. In order to resolve these problems, a PBR with a central graphite column in a low core is invented. This paper presents the effect of the central graphite column on a pebble flow by using the computational fluid dynamics(CFD) code, CFX-10

Continuous-flow leaching studies of crushed and cored SYNROC

International Nuclear Information System (INIS)

Coles, D.G.; Bazan, F.

1980-01-01

Both crushed (150 to 300 μm) and cored 1.8 mm diameter) samples of SYNROC have been leached with the single-pass continuous-flow leaching equipment. Crushed samples of Cs-hollandite were also leached in a similar experiment. Temperatures used were 25 0 C and 75 0 C and leachates were 0.03 N NaHCO 3 and distilled water. Leaching rates from SYNROC C were ranked Cs > Sr greater than or equal to Ca > Ba > Zr. A comparison of leaching rates is made between crushed SYNROC, cored SYNROC, and PNL 76-68 glass beads. Problems encountered when comparing the leaching rates of different waste forms are discussed

Aespoe Hard Rock Laboratory. Backfill and Plug test. Hydraulic testing of core drilled boreholes in the ZEDEX drift

Energy Technology Data Exchange (ETDEWEB)

Ludvigson, Jan-Erik; Nordqvist, Rune; Ekman, Lennart; Hansson, Kent (GEOSIGMA AB, Uppsala (Sweden))

2009-07-01

The present report documents the performance and results of hydraulic testing in selected core boreholes in the Zedex drift. The holes will be used as rock instrumentation boreholes during the Backfill and Plug Test at Aespoe HRL. The testing involves both 1 m long boreholes with 56 mm diameter as well as longer boreholes c. 5 m, 8 m and 25 m long with 56 mm or 76 mm diameter. Only single-hole tests were performed. The tests were carried out as short-time constant head injection tests since all boreholes tested (except one) were non-flowing before tests. The injection phase was followed by a pressure recovery phase. Furthermore, the tests were carried out as single-packer tests. A specially designed test system was used for the tests. The main evaluation of the tests was performed on data from the recovery phase by a new approach based on a non-linear regression technique combined with a flow simulation model (SUTRA). The tests in the 1 m-holes (testing the interval c. 0.3-0.7 m in the rock perpendicular to the tunnel face) show that the hydraulic conductivity of the superficial rock around the Zedex drift in general is low. However, during testing in some boreholes, visible leakage in the rock occurred through superficial fractures into the tunnel. These fractures were mainly located in the floor of the Zedex drift and are probably blast-induced. These fractures have a high hydraulic conductivity. The tests in the longer boreholes show that the hydraulic conductivity further into the rock in general is below c. 1x10-10 m/s. Increased hydraulic conductivity (c.1.5x10-8 m/s) was only observed in the flowing borehole KXZSD8HL.

Scramjet test flow reconstruction for a large-scale expansion tube, Part 2: axisymmetric CFD analysis

Science.gov (United States)

Gildfind, D. E.; Jacobs, P. A.; Morgan, R. G.; Chan, W. Y. K.; Gollan, R. J.

2017-11-01

This paper presents the second part of a study aiming to accurately characterise a Mach 10 scramjet test flow generated using a large free-piston-driven expansion tube. Part 1 described the experimental set-up, the quasi-one-dimensional simulation of the full facility, and the hybrid analysis technique used to compute the nozzle exit test flow properties. The second stage of the hybrid analysis applies the computed 1-D shock tube flow history as an inflow to a high-fidelity two-dimensional-axisymmetric analysis of the acceleration tube. The acceleration tube exit flow history is then applied as an inflow to a further refined axisymmetric nozzle model, providing the final nozzle exit test flow properties and thereby completing the analysis. This paper presents the results of the axisymmetric analyses. These simulations are shown to closely reproduce experimentally measured shock speeds and acceleration tube static pressure histories, as well as nozzle centreline static and impact pressure histories. The hybrid scheme less successfully predicts the diameter of the core test flow; however, this property is readily measured through experimental pitot surveys. In combination, the full test flow history can be accurately determined.

Test Facility Construction for Flow Visualization on Mixing Flow inside Subchannels of PWR Rod Bundle

Energy Technology Data Exchange (ETDEWEB)

Kim, Seok; Jeon, Byong-Guk; Youn, Young-Jung; Choi, Hae-Seob; Euh, Dong-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Flow inside rod bundles has a similarity with flow in porous media. To ensure thermal performance of a nuclear reactor, detailed information of the heat transfer and turbulent mixing flow phenomena taking place within the subchannels is required. The subchannel analysis is one of the key thermal-hydraulic calculations in the safety analysis of the nuclear reactor core. At present, subchannel computer codes are employed to simulate fuel elements of nuclear reactor cores and predict the performance of cores under normal operating and hypothetical accident conditions. The ability of these subchannels codes to predict both the flow and enthalpy distribution in fuel assemblies is very important in the design of nuclear reactors. Recently, according to the modern tend of the safety analysis for the nuclear reactor, a new component scale analysis code, named CUPID, and has been developed in KAERI. The CUPID code is based on a two-fluid and three-field model, and both the open and porous media approaches are incorporated. The PRIUS experiment has addressed many key topics related to flow behaviour in a rod bundle. These issues are related to the flow conditions inside a nuclear fuel element during normal operation of the plant or in accident scenarios. From the second half of 2016, flow visualization will be performed by using a high speed camera and image analysis technique, from which detailed information for the two-dimensional movement of single phase flow is quantified.

IR1 flow tube and In-Pile Test Section Pressure drop test for the 3-Pin Fuel Test Loop

Energy Technology Data Exchange (ETDEWEB)

Lee, H. H.; Park, K. N.; Chi, D. Y.; Sim, B. S.; Park, S. K.; Lee, J. M.; Lee, C. Y.; Kim, H. N

2006-02-15

The in-pile Section (IPS) of 3-pin Fuel Test Loop(FTL) shall be installed in the vertical hole call IR1 of HANARO reactor core. In order to verify the pressure drop and flow rate both the inside region of IPS at the annular region between IPS and IR1 flow tube, a pressure drop was measured by varing the flow rate on both regions. The measured pressure drop in the annular region is 209kpa at 14.9kg/s which meets the limiting condition of operation of 200kpa. The measured pressure drop in side the IPS becomes 260.25kpa which is lower than the designed value of 306.65kpa. As the pressure drop is lower than the design value, it is quite conservative from the safety and operating point of view.

Analysis of Semiscale Mod-1 integral test with asymmetrical break (Test S-29-1)

International Nuclear Information System (INIS)

Langerman, M.A.

1977-03-01

Selected experimental data obtained from Semiscale Mod-1 cold leg break Test S-29-1 and results obtained from analytical codes are analyzed. This test was the first integral blowdown reflood test conducted with the Mod-1 system and was a special test designed specifically to evaluate the sensitivity of the early Mod-1 core thermal response (0 to 5 sec after rupture) to the magnitude and direction of the core flow. To achieve this specific objective in Test S-29-1, the vessel side break area was reduced to approximately one-half the scaled break area associated with a 200 percent cold leg break test. The reduction in break area significantly reduced the core flow reversal that took place immediately after rupture and resulted in periods of positive core flow in the early portion of the test. The results obtained from this test are compared with results obtained from a 200 percent cold leg break test and the effect of core flow on early core thermal response is evaluated. Since Test S-29-1 was the first integral blowdown reflood test conducted with the Mod-1 system, data are also presented through the reflood stage of the test and the results are analyzed. The test data and the core thermal response calculated with the RELAP4 code are also compared

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)

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

James Webb Space Telescope Core 2 Test - Cryogenic Thermal Balance Test of the Observatorys Core Area Thermal Control Hardware

Science.gov (United States)

Cleveland, Paul; Parrish, Keith; Thomson, Shaun; Marsh, James; Comber, Brian

2016-01-01

The James Webb Space Telescope (JWST), successor to the Hubble Space Telescope, will be the largest astronomical telescope ever sent into space. To observe the very first light of the early universe, JWST requires a large deployed 6.5-meter primary mirror cryogenically cooled to less than 50 Kelvin. Three scientific instruments are further cooled via a large radiator system to less than 40 Kelvin. A fourth scientific instrument is cooled to less than 7 Kelvin using a combination pulse-tube Joule-Thomson mechanical cooler. Passive cryogenic cooling enables the large scale of the telescope which must be highly folded for launch on an Ariane 5 launch vehicle and deployed once on orbit during its journey to the second Earth-Sun Lagrange point. Passive cooling of the observatory is enabled by the deployment of a large tennis court sized five layer Sunshield combined with the use of a network of high efficiency radiators. A high purity aluminum heat strap system connects the three instrument's detector systems to the radiator systems to dissipate less than a single watt of parasitic and instrument dissipated heat. JWST's large scale features, while enabling passive cooling, also prevent the typical flight configuration fully-deployed thermal balance test that is the keystone of most space missions' thermal verification plans. This paper describes the JWST Core 2 Test, which is a cryogenic thermal balance test of a full size, high fidelity engineering model of the Observatory's 'Core' area thermal control hardware. The 'Core' area is the key mechanical and cryogenic interface area between all Observatory elements. The 'Core' area thermal control hardware allows for temperature transition of 300K to approximately 50 K by attenuating heat from the room temperature IEC (instrument electronics) and the Spacecraft Bus. Since the flight hardware is not available for test, the Core 2 test uses high fidelity and flight-like reproductions.

CANFLEX fuel bundle cross-flow endurance test (test report)

International Nuclear Information System (INIS)

Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D.

1997-04-01

As part of the normal refuelling sequence of CANDU nuclear reactor, both new and irradiated bundles can be parked in the cross-flow region of the liner tubes. This situation occurs normally for a few minutes. The fuel bundle which is subjected to the cross-flow should be capable of withstanding the consequences of cross flow for normal periods, and maintain its mechanical integrity. The cross-flow endurance test was conducted for CANFLEX bundle, latest developed nuclear fuel, at CANDU-Hot Test Loop. The test was carried out during 4 hours at the inlet cross-flow region. After the test, the bundle successfully met all acceptance criteria after the 4 hours cross-flow test. (author). 2 refs., 3 tabs

CANFLEX fuel bundle cross-flow endurance test (test report)

Energy Technology Data Exchange (ETDEWEB)

Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D.

1997-04-01

As part of the normal refuelling sequence of CANDU nuclear reactor, both new and irradiated bundles can be parked in the cross-flow region of the liner tubes. This situation occurs normally for a few minutes. The fuel bundle which is subjected to the cross-flow should be capable of withstanding the consequences of cross flow for normal periods, and maintain its mechanical integrity. The cross-flow endurance test was conducted for CANFLEX bundle, latest developed nuclear fuel, at CANDU-Hot Test Loop. The test was carried out during 4 hours at the inlet cross-flow region. After the test, the bundle successfully met all acceptance criteria after the 4 hours cross-flow test. (author). 2 refs., 3 tabs.

Experimental investigation of flow dynamics in the SNR-upper-core structure

International Nuclear Information System (INIS)

Meyer, L.

1985-03-01

This report describes the results of a simulant-material experimental investigation of flow dynamics in the upper-core (UCS) during a HCDA of a LMFBR. The experiments were designed to verify some of the thermal-hydraulic models in SIMMER-II. Four different liquids were used to simulate the flashing U0 2 ; and numerous parameter variations were made regarding initial pressure, temperature, and configurations of the test apparatus. The experiments showed the large effect of the heat transfer in the UCS and the relatively small effect of friction. The reduction in final kinetic energy by the presence of the UCS is shown as a function of the initial pressure and the temperature difference between core and UCS. Calculations with SIMMER-II for the wide range of experiments produced results for the kinetic energy within a factor of 2 of the experimental results without changing the crucial input parameters. The minimum droplet size during the flashing process and the structure-side heat transfer coefficient were determined to be the crucial and most sensitive parameters. This reflects deficiencies in modeling of both the flashing process and the transient heat conduction in the structure. (orig./HP) [de

Data report of BWR post-CHF tests. Transient core thermal-hydraulic test program. Contract research

International Nuclear Information System (INIS)

Iguchi, Tadashi; Itoh, Hideo; Kiuchi, Toshio; Watanabe, Hironori; Kimura, Mamoru; Anoda, Yoshinari

2001-03-01

JAERI has been performing transient core thermal-hydraulic test program. In the program, authors performed BWR/ABWR DBE simulation tests with a test facility, which can simulate BWR/ABWR transients. The test facility has a 4 x 4 bundle core simulator with 15-rod heaters and one non-heated rod. Through the tests, authors quantified the thermal safety margin for core cooling. In order to quantify the thermal safety margin, authors collected experimental data on post-CHF. The data are essential for the evaluation of clad temperature transient when core heat-up occurs during DBEs. In comparison with previous post-CHF tests, present experiments were performed in much wider experimental condition, covering high clad temperature, low to high pressure and low to high mass flux. Further, data at wider elevation (lower to higher elevation of core) were obtained in the present experiments, which make possible to discuss the effect of axial position on thermal-hydraulics, while previous works usually discuss the thermal-hydraulics at the position where the first heat-up occurs. This data report describes test procedure, test condition and major experimental data of post-CHF tests. (author)

Direct numerical simulations of reacting flows with detailed chemistry using many-core/GPU acceleration

KAUST Repository

Herná ndez Pé rez, Francisco E.; Mukhadiyev, Nurzhan; Xu, Xiao; Sow, Aliou; Lee, Bok Jik; Sankaran, Ramanan; Im, Hong G.

2018-01-01

A new direct numerical simulation (DNS) code for multi-component gaseous reacting flows has been developed at KAUST, with the state-of-the-art programming model for next generation high performance computing platforms. The code, named KAUST Adaptive Reacting Flows Solver (KARFS), employs the MPI+X programming, and relies on Kokkos for “X” for performance portability to multi-core, many-core and GPUs, providing innovative software development while maintaining backward compatibility with established parallel models and legacy code. The capability and potential of KARFS to perform DNS of reacting flows with large, detailed reaction mechanisms is demonstrated with various model problems involving ignition and turbulent flame propagations with varying degrees of chemical complexities.

Direct numerical simulations of reacting flows with detailed chemistry using many-core/GPU acceleration

KAUST Repository

Hernández Pérez, Francisco E.

2018-03-29

A new direct numerical simulation (DNS) code for multi-component gaseous reacting flows has been developed at KAUST, with the state-of-the-art programming model for next generation high performance computing platforms. The code, named KAUST Adaptive Reacting Flows Solver (KARFS), employs the MPI+X programming, and relies on Kokkos for “X” for performance portability to multi-core, many-core and GPUs, providing innovative software development while maintaining backward compatibility with established parallel models and legacy code. The capability and potential of KARFS to perform DNS of reacting flows with large, detailed reaction mechanisms is demonstrated with various model problems involving ignition and turbulent flame propagations with varying degrees of chemical complexities.

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.

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

Accidents and transients analyses of a super fast reactor with single flow pass core

International Nuclear Information System (INIS)

Sutanto,; Oka, Yoshiaki

2014-01-01

Highlights: • Safety analysis of a Super FR with single flow pass core is conducted. • Loss of feed water flow leads to a direct effect on the loss of fuel channel flow. • The core pressure is sensitive to LOCA accidents due to the direct effect. • Small LOCA introduces a critical break. • The safety criteria for all selected events are satisfied. - Abstract: The supercritical water cooled fast reactor with single flow pass core has been designed to simplify refueling and the structures of upper and lower mixing plenums. To evaluate the safety performance, safety analysis has been conducted with regard to LOCA and non-LOCA accidents including transient events. Safety analysis results show that the safety criteria are satisfied for all selected events. The total loss of feed water flow is the most important accident which the maximum cladding surface temperature (MCST) is high due to a direct effect of the accident on the total loss of flow in all fuel assemblies. However, actuation of the ADS can mitigate the accident. Small LOCA also introduces a critical break at 7.8% break which results high MCST at BOC because the scram and ADS are not actuated. Early ADS actuation is effective to mitigate the accident. In large LOCA, 100% break LOCA results a high MCST of flooding phase at BOC due to high power peaking at the bottom part. Use of high injection flow rate by 2 LPCI units is effective to decrease the MCST

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

The role of natural circulation in the FFTF [Fast Flux Test Facility] passive safety tests

International Nuclear Information System (INIS)

Stover, R.L.; Padilla, A.; Burke, T.M.; Knecht, W.L.

1987-03-01

A series of tests were completed at the Fast Flux Test Facility to demonstrate the passive safety characteristics of liquid metal reactors with natural circulation flow. The first test consisted of transition from forced to natural circulation flow at an initial decay power of 0.3%. The second test represented an unprotected loss-of-flow transient to natural circulation from 50% power with the control rods prevented from scramming into the core. The third test was a steady-state, natural circulation condition with core fission powers up ato about 2.3%. Core sodium data and results of single and multi-channel computer models confirmed the reliability and effectiveness of natural circulation flow for liquid metal reactor safety

Prediction, analysis and solution of flow inversion phenomenon in a typical MTR reactor with upward core cooling

International Nuclear Information System (INIS)

El-Morshedy, Salah El-Din

2010-01-01

Research reactors of power greater than 20 MW are usually designed to be cooled with upward coolant flow direction inside the reactor core. This is mainly to prevent flow inversion problems following a pump coast down. However, in some designs and under certain operating conditions, flow inversion phenomenon is predicted. In the present work, the best-estimate Material Testing Reactors Thermal-Hydraulic Analysis program (MTRTHA) is used to simulate a typical MTR reactor behavior with upward cooling under a hypothetical case of loss of off-site power. The flow inversion phenomenon is predicted under certain decay heat and/or pool temperature values below the design values. The reactor simulation under loss of off-site power is performed for two cases namely; two-flap valves open and one flap-valve fails to open. The model results for the flow inversion phenomenon prediction is analyzed and a solution of the problem is suggested. (orig.)

Scaling of Core Material in Rubble Mound Breakwater Model Tests

DEFF Research Database (Denmark)

Burcharth, H. F.; Liu, Z.; Troch, P.

1999-01-01

The permeability of the core material influences armour stability, wave run-up and wave overtopping. The main problem related to the scaling of core materials in models is that the hydraulic gradient and the pore velocity are varying in space and time. This makes it impossible to arrive at a fully...... correct scaling. The paper presents an empirical formula for the estimation of the wave induced pressure gradient in the core, based on measurements in models and a prototype. The formula, together with the Forchheimer equation can be used for the estimation of pore velocities in cores. The paper proposes...... that the diameter of the core material in models is chosen in such a way that the Froude scale law holds for a characteristic pore velocity. The characteristic pore velocity is chosen as the average velocity of a most critical area in the core with respect to porous flow. Finally the method is demonstrated...

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.)

CFD simulation on reactor flow mixing phenomena

International Nuclear Information System (INIS)

Kwon, T.S.; Kim, K.H.

2016-01-01

A pre-test calculation for multi-dimensional flow mixing in a reactor core and downcomer has been studied using a CFD code. To study the effects of Reactor Coolant Pump (RCP) and core zone on the boron mixing behaviors in a lower downcomer and core inlet, a 1/5-scale CFD model of flow mixing test facility for the APR+ reference plant was simulated. The flow paths of the 1/5-scale model were scaled down by the linear scaling method. The aspect ratio (L/D) of all flow paths was preserved to 1. To preserve a dynamic similarity, the ratio of Euler number was also preserved to 1. A single phase water flow at low pressure and temperature conditions was considered in this calculation. The calculation shows that the asymmetric effect driven by RCPs shifted the high velocity field to the failed pump's flow zone. The borated water flow zone at the core inlet was also shifted to the failed RCP side. (author)

French experimental facilities for measurements of transverse flows and assessment of the corresponding risk of vibrations in heterogeneous cores

International Nuclear Information System (INIS)

Le Borgne, E.; Mattei, A.; Oceraies, Y.; Fardeau, P.

1994-01-01

Due to insertion of a limited number of new assemblies at each cycle, the cores in Pressurized Water Reactors are not homogeneous. Referring only to the impact on coolant flow, these differences can range from variable hydraulic resistances in the assembly, which depend on the geometric changes occurring during preceding cycles, to coexistence of assemblies with new design structures. Deviations in resistance between neighboring fuel assemblies causes the flow rates to be distributed differently between the assembly rods. This results in development of transverse flows from the main axial flow, and changes in the axial velocity gradients. These particularities of coolant flow have an effect on both vibration levels and cooling of the fuel rods, and also on the axial forces exerted on the assemblies in the core cavity. Since 1985, French Atomic Energy Commission (CEA) has gradually acquired experimental and measuring facilities that have allowed it to engage in research and development programs in these areas, in cooperation with industry partners in the nuclear field. Two complementary test loops have been constructed, called ARIANE and HERMES T. Use of these experimental facilities allows to obtain complete and detailed information on the hydraulic and vibratory phenomena specific to heterogeneous cores. In particular it is possible to establish a direct assessment of the actual compatibility between two different assemblies. By making a few specific changes, these facilities can also be used as a unique tool for assembly behaviour studies under seismic conditions with simulation of the flow effects. Also, a source of information in thus made available for qualification of computation codes for vibratory mechanics and multidimensional fluid mechanics under development at CEA and also used in the field of nuclear fuel. (authors). 6 figs., 1 ref

Behaviour of steel pipe exposed to fouling by heavy oil during core-annular flow; Comportamento de tubo de aco exposto a sujeira de oleo pesado durante escoamento nucleo-anular

Energy Technology Data Exchange (ETDEWEB)

Barbosa, Adriana; Bannwart, Antonio C. [Universidade Estadual de Campinas, SP (Brazil). Faculdade de Engenharia Mecanica. Dept. de Engenharia de Petroleo

2004-07-01

The use of water-assisted technologies such as core-annular flow to the pipelines of viscous oils has been proposed as an attractive alternative for production and transportation of heavy crudes in both onshore and offshore scenarios. Usually, core-annular flow can be created by injecting a relatively small water flow rate laterally in the pipe, so as to form a thin water annulus surrounding the viscous oil, which is pumped through the center. The reduction in friction losses obtained thanks to lubrication by water is significant, since the pressure drop in a steady state core flow becomes comparable to water flow only. For a complete assessment of core flow technology, however, unwanted effects associated with possible oil adhesion onto the pipe wall should be investigated, since these may cause severe fouling of the wall and pressure drop increase. It has been observed that oil adhesion on metallic surfaces may occur for certain types of crude and oilphilic pipe materials. In this work we present results of pressure drop monitoring during 35 hour-operation of a heavy oil-water core annular flow in a 26.08 mm. i.d. horizontal steel pipe. The oil used is described in terms of its main components and the results of static wet ability tests are also presented for comparison (author)

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

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.

Contribution to the physical validation of computer programs for reactor cores flows

International Nuclear Information System (INIS)

Bourgeois, Pierre

1998-01-01

A κ-ε turbulence model was implemented in the FLICA computer code which is devoted to thermal-hydraulic analysis of nuclear reactor cores flows. Foreseen applications concern single-phase flows in rod bundles. First-moment closure principles are reminded. Low Reynolds wall effects are accounted for by a two-layer approach. A computational method for the distance from the wall must have been developed to do so. Two two-layer κ-ε models are proposed and studied: the classical isotropic version, based on the Boussinesq's hypothesis, and an original anisotropic version which supposes a non-linear relation between Reynolds stresses and mean deformation rate. The second one permits the treatment of anisotropy, which is encountered in non-circular ducts in general, and in rod bundles in particular. Turbulent solver is linearized implicit, based on a finite volume method - VF9 scheme for the viscous part, upwind scheme for passive scalar for the convective part, centered scheme for the source terms. Several numerical simulations on 2D and 3D configurations were conducted (validation standard test, industrial application). (author) [fr

Continuous-flow leaching studies of crushed and cored SYNROC

International Nuclear Information System (INIS)

Coles, D.G.; Bazan, F.

1982-01-01

Both crushed (150- to 300- μm) and cored (1,8- mm-diam) samples of SYNROC have been leached with single-pass continuous-flow leaching equipment. Crushed samples of cesium-hollandite were also leached in a similar experiment. Temperatures used were 25 0 and 75 0 C and leachates were 0.03 N NaHCO 3 and distilled water. Leaching rates from SYNROC-C were ranked cesium > strontium greater than or equal to calcium > barium > zirconium. A comparison of leaching rates is made between crushed SYNROC, cored SYNROC, and Pacific Northwest Laboratory 76-68 glass beads. This comparison depends on how the surface areas are determined for each sample. Based on geometric surface areas for SYNROC cores and glass beads, cesium leach rates from SYNROC compare well with both sodium and neptunium leached from the glass. The other elements leached from SYNROC are lower than sodium and neptunium leached from glass. They also vary for each element, while glass shows nearly the same leach rate for both sodium and neptunium

Continuous-flow leaching studies of crushed and cored SYNROC

International Nuclear Information System (INIS)

Coles, D.G.; Bazan, F.

1981-01-01

Both crushed (150 to 300 μm) and cored (1.8 mm diameter) samples of SYNROC have been leached with the single-pass continuous-flow leaching equipment. Crushed samples of Cs-hollandite were also leached in a similar experiment. Temperatures used were 25 and 75 0 C and leachates were 0.03 N NaHCO 3 and distilled water. Leaching rates from SYNROC C were ranked Cs > Sr greater than or equal to Ca > Ba > Zr. A comparison of leaching rates is made between crushed SYNROC, cored SYNROC, and PNL 76-68 glass beads. This comparison depends on how the surface areas are determined for each sample. Based on geometric surface areas for SYNROC cores and glass beads Cs leach rates from SYNROC compare well with both Na and Np leached from the glass. The other elements leached from SYNROC are lower than Na and Np leached from glass. They also vary for each element while glass shows nearly the same leach rate for both Na and Np

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

Deconvolution-based resolution enhancement of chemical ice core records obtained by continuous flow analysis

DEFF Research Database (Denmark)

Rasmussen, Sune Olander; Andersen, Katrine K.; Johnsen, Sigfus Johann

2005-01-01

Continuous flow analysis (CFA) has become a popular measuring technique for obtaining high-resolution chemical ice core records due to an attractive combination of measuring speed and resolution. However, when analyzing the deeper sections of ice cores or cores from low-accumulation areas...... of the data for high-resolution studies such as annual layer counting. The presented method uses deconvolution techniques and is robust to the presence of noise in the measurements. If integrated into the data processing, it requires no additional data collection. The method is applied to selected ice core...

Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

International Nuclear Information System (INIS)

Liu, Y.Y.; Tsai, H.; Billone, M.C.; Kramer, J.M.

1992-01-01

The next step in the development of metal fuels for the integral fast reactor (IFR) is the conversion of the Experimental Breeder Reactor II (EBR-II) core to one containing the ternary U-20 Pu-10 Zr alloy clad with HT-9 cladding, i.e., the Mk-V core. This paper presents results of three hot-cell furnace simulation tests on irradiated Mk-V-type fuel elements (U-19 Pu-10 Zr/HT-9), which were performed to support the safety case for the Mk-V core. These tests were designed to envelop an umbrella (bounding) unlikely loss-of-flow (LOF) event in EBR-II during which the calculated peak cladding temperature would reach 776 degree C for < 2 min. The principal objectives of these tests were (a) demonstration of the safety margin of the fuel element, (b) investigation of cladding breaching behavior, and (c) provision of data for validation of the FPIN2 and LIFE-METAL codes

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.

Assessment of RELAP5-3D copyright using data from two-dimensional RPI flow tests

International Nuclear Information System (INIS)

Davis, C.B.

1998-01-01

The capability of the RELAP5-3D copyright computer code to perform multi-dimensional thermal-hydraulic analysis was assessed using data from steady-state flow tests conducted at Rensselaer Polytechnic Institute (RPI). The RPI data were taken in a two-dimensional test section in a low-pressure air/water loop. The test section consisted of a thin vertical channel that simulated a two-dimensional slice through the core of a pressurized water reactor. Single-phase and two-phase flows were supplied to the test section in an asymmetric manner to generate a two-dimensional flow field. A traversing gamma densitometer was used to measure void fraction at many locations in the test section. High speed photographs provided information on the flow patterns and flow regimes. The RPI test section was modeled using the multi-dimensional component in RELAP5-3D Version BF06. Calculations of three RPI experiments were performed. The flow regimes predicted by the base code were in poor agreement with those observed in the tests. The two-phase regions were observed to be in the bubbly and slug flow regimes in the test. However, nearly all of the junctions in the horizontal direction were calculated to be in the stratified flow regime because of the relatively low velocities in that direction. As a result, the void fraction predictions were also in poor agreement with the measured values. Significantly improved results were obtained in sensitivity calculations with a modified version of the code that prevented the horizontal junctions from entering the stratified flow regime. These results indicate that the code's logic in the determination of flow regimes in a multi-dimensional component must be improved. The results of the sensitivity calculations also indicate that RELAP5-3D will provide a significant multi-dimensional hydraulic analysis capability once the flow regime prediction is improved

CoreFlow: Enriching Bro security events using network traffic monitoring data

NARCIS (Netherlands)

Koning, R.; Buraglio, N.; de Laat, C.; Grosso, P.

Attacks against network infrastructures can be detected by Intrusion Detection Systems (IDS). Still reaction to these events are often limited by the lack of larger contextual information in which they occurred. In this paper we present CoreFlow, a framework for the correlation and enrichment of IDS

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

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

International Nuclear Information System (INIS)

Ishii, M.; De Jarlais, G.

1987-01-01

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 inlet section consists of specially designed coaxial nozzles for gas and liquid such that the ideal inverted annular flow can be generated. The roll wave formation, droplet entrainment from wave crests, agitated sections with large interfacial areas, classical sinuous jet instability, jet break-up into multiple liquid ligaments and drop formation from liquid ligaments have been observed in detail. (orig.)

Calculation of the flow distribution for the new core of the RA-6 reactor

International Nuclear Information System (INIS)

Garcia, J.C.; Delmastro, Dario F.

2007-01-01

In this work the pressure drop, the flow distribution, effective cooling flow rate and the velocity in the subchannels that cool fuel plates for the new core of RA-6 research reactor were calculated. These calculations were performed for a flow of 340 m 3 /hr and water temperatures of 12 C degrees, of 35 C degrees and 42 C degrees. The flow distribution was calculated without considering either safety factors or geometric changes. All the calculations were performed considering the flow as isothermal. (author) [es

Breccia-cored columnar rosettes in a rubbly pahoehoe lava flow, Elephanta Island, Deccan Traps, and a model for their origin

Directory of Open Access Journals (Sweden)

Hetu Sheth

2017-11-01

Full Text Available Rubbly pahoehoe lava flows are abundant in many continental flood basalts including the Deccan Traps. However, structures with radial joint columns surrounding cores of flow-top breccia (FTB, reported from some Deccan rubbly pahoehoe flows, are yet unknown from other basaltic provinces. A previous study of these Deccan “breccia-cored columnar rosettes” ruled out explanations such as volcanic vents and lava tubes, and showed that the radial joint columns had grown outwards from cold FTB inclusions incorporated into the hot molten interiors. How the highly vesicular (thus low-density FTB blocks might have sunk into the flow interiors has remained a puzzle. Here we describe a new example of a Deccan rubbly pahoehoe flow with FTB-cored rosettes, from Elephanta Island in the Mumbai harbor. Noting that (1 thick rubbly pahoehoe flows probably form by rapid inflation (involving many lava injections into a largely molten advancing flow, and (2 such flows are transitional to ‘a’ā flows (which continuously shed their top clinker in front of them as they advance, we propose a model for the FTB-cored rosettes. We suggest that the Deccan flows under study were shedding some of their FTB in front of them as they advanced and, with high-eruption rate lava injection and inflation, frontal breakouts would incorporate this FTB rubble, with thickening of the flow carrying the rubble into the flow interior. This implies that, far from sinking into the molten interior, the FTB blocks may have been rising, until lava supply and inflation stopped, the flow began solidifying, and joint columns developed outward from each cold FTB inclusion as already inferred, forming the FTB-cored rosettes. Those rubbly pahoehoe flows which began recycling most of their FTB became the ‘a’ā flows of the Deccan.

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

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

Scale-model characterization of flow-induced vibrational response of FFTF reactor internals

International Nuclear Information System (INIS)

Ryan, J.A.; Mahoney, J.J.

1980-10-01

Fast Test Reactor core internal and peripheral components were assessed for flow-induced vibrational characteristics under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup as an integral part of the Fast Test Reactor Vibration Program. The Hydraulic Core Mockup was an 0.285 geometric scale model of the Fast Test Reactor internals designed to simulate prototype vibrational and hydraulic characteristics. Using water to simulate sodium coolant, vibrational characteristics were measured and determined for selected model components over the scaled flow range of 36 to 110%. Additionally, in-situ shaker tests were conducted on selected Hydraulic Core Mockup outlet plenum components to establish modal characteristics. Most components exhibited resonant response at all test flow rates; however, the measured dynamic response was neither abnormal nor anomalously flow-rate dependent, and the predicted prototype components' response were deemed acceptable

Relationship of core exit-temperature noise to thermal-hydraulic conditions in PWRs

International Nuclear Information System (INIS)

Sweeney, F.J.; Upadhyaya, B.R.

1983-01-01

Core exit thermocouple temperature noise and neutron detector noise measurements were performed at the Loss of Fluid Test Facility (LOFT) reactor and a Westinghouse, 1148 MW(e) PWR to relate temperature noise to core thermal-hydraulic conditions. The noise analysis results show that the RMS of the temperature noise increases linearly with increasing core δT at LOFT and the commercial PWR. Out-of-core test loop temperature noise has shown similar behavior. The phase angle between core exit temperature noise and in-core or ex-core neutron noise is directly related to the core coolant flow velocity. However, if the thermocouple response time is slow, compared to the coolant transit time between the sensors, velocities inferred from the phase angle are lower than measured coolant flow velocities

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)

CFD analysis of the effect of rolling motion on the flow distribution at the core inlet

International Nuclear Information System (INIS)

Yan, B.H.; Zhang, G.; Gu, H.Y.

2012-01-01

Highlights: ► The flow distribution at the core inlet in rolling motion is investigated. ► In rolling motion, the variation of flow distribution factor is not regular. ► The minimum flow distribution factor could be decreased by rolling motion. ► The effect of rolling motion diminishes with Reynolds number increasing. ► Effect of rolling motion in single loop operation is more significant. - Abstract: The flow distribution at the core inlet in rolling motion is investigated with software CFX12.0. The calculation results were in agreement with experimental data in steady state. As the increasing of rolling amplitude and the decreasing of rolling period, the effect of rolling motion on the flow distribution factor and the flowing behavior increases. In rolling motion, the variation of flow distribution factor is not regular. The rolling motion could decrease the minimum flow distribution factor. The effect of rolling motion on the coolant field and flow distribution diminishes with the Reynolds number increasing. The effect of rolling motion on the flow distribution in the case of single loop operation is more significant than that in the case of double loops operation.

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.)

Short-term variations in core surface flow resolved from an improved method of calculating observatory monthly means

Science.gov (United States)

Olsen, Nils; Whaler, Kathryn A.; Finlay, Christopher C.

2014-05-01

Monthly means of the magnetic field measurements taken by ground observatories are a useful data source for studying temporal changes of the core magnetic field and the underlying core flow. However, the usual way of calculating monthly means as the arithmetic mean of all days (geomagnetic quiet as well as disturbed) and all local times (day and night) may result in contributions from external (magnetospheric and ionospheric) origin in the (ordinary, omm) monthly means. Such contamination makes monthly means less favourable for core studies. We calculated revised monthly means (rmm), and their uncertainties, from observatory hourly means using robust means and after removal of external field predictions, using an improved method for characterising the magnetospheric ring current. The utility of the new method for calculating observatory monthly means is demonstrated by inverting their first differences for core surface advective flows. The flow is assumed steady over three consecutive months to ensure uniqueness; the effects of more rapid changes should be attenuated by the weakly conducting mantle. Observatory data are inverted directly for a regularised core flow, rather than deriving it from a secular variation spherical harmonic model. The main field is specified by the CHAOS-4 model. Data from up to 128 observatories between 1997 and 2013 were used to calculate 185 flow models from the omm and rmm, for each possible set of three consecutive months. The full 3x3 (non-diagonal) data covariance matrix was used, and two-norm (least squares) minimisation performed. We are able to fit the data to the target (weighted) misfit of 1, for both omm and rmm inversions, provided we incorporate the full data covariance matrix, and produce consistent, plausible flows. Fits are better for rmm flows. The flows exhibit noticeable changes over timescales of a few months. However, they follow rapid excursions in the omm that we suspect result from external field contamination

Identification of Columbia River basalt flows from deep cores in the Pasco Basin based on trace element abundances

International Nuclear Information System (INIS)

Fruchter, J.S.; Rancitelli, L.A.

1976-01-01

Samples of basalt from three deep core holes drilled in the Pasco Basin, Washington (DDH-3, DH-4, DH-5) were analyzed by instrumental neutron activation (INAA) for up to fifteen trace and major elements. These analyses were used to assign each basalt flow to one of a series of previously defined chemical types of the Columbia River Basalt Group. All of the flows except the two flows at the bottom of well DDH-3 were clearly assignable to one of the defined chemical types. These two flows apparently represent new, as yet undefined chemical types. Average values and standard deviations for compositions of each of the chemical types found in the three wells are presented along with two-element variation diagrams for the geochemically important pair La-Cr, La-Fe, La-Th and La-Sc. The assignment of the flows to known chemical types accomplished in this study was very helpful in relating the basalts in the core holes to stratigraphically defined basalt flows in surface sections. A correlation diagram relating the flows in the core holes to one another on the basis of chemical type is presented

UPTF/TEST10B/RUN081, Steam/Water Flow Phenomena Reflood PWR Cold Leg Break LOCA

International Nuclear Information System (INIS)

1998-01-01

1 - Description of test facility: The Upper Plenum Test Facility (UPTF) is a geometrical full-scale simulation of the primary system of the four-loop 1300 MWe Siemens/KWU pressurized water reactor (PWR) at Grafenrheinfeld. The test vessel, upper plenum and its internals, downcomer, primary loops, pressurizer and surge line are replicas of the reference plant. The core, coolant pumps, steam generators and containment of a PWR are replaced by simulators which simulate the boundary and initial conditions during end-of-blowdown, refill and reflood phase following a loss-of-coolant accident (LOCA) with a hot or cold leg break. The break size and location can be simulated in the broken loop. The emergency core coolant (ECC) injection systems at the UPTF are designed to simulate the various ECC injection modes, such as hot leg, upper plenum, cold leg, downcomer or combined hot and cold leg injection of different ECC systems of German and US/Japan PWRs. Moreover, eight vent valves are mounted in the core barrel above the hot leg nozzle elevation for simulation of ABB and B and W PWRs. The UPTF primary system is divided into the investigation and simulation areas. The investigation areas, which are the exact replicas of a GPWR, consist of the upper plenum with internals, hot legs, cold legs and downcomer. The realistic thermal-hydraulic behavior in the investigation areas is assured by appropriate initial and boundary conditions of the area interface. The boundary conditions are realized by above mentioned simulators, the setup and the operation of which are based on small-scale data and mathematical models. The simulation areas include core simulator, steam generator simulators, pump simulators and containment simulator. The steam production and entrainment in a real core during a LOCA are simulated by steam and water injection through the core simulator. 2 - Description of test: Investigation of steam/water flow phenomena at the upper tie plate and in the upper plenum and

FFTF scale-model characterization of flow-induced vibrational response of reactor internals

International Nuclear Information System (INIS)

Ryan, J.A.; Julyk, L.J.

1977-01-01

As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36 percent to 111 percent of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable

FFTF scale-model characterization of flow induced vibrational response of reactor internals

Energy Technology Data Exchange (ETDEWEB)

Ryan, J A; Julyk, L J [Hanford Engineering Development Laboratory, Richland, WA (United States)

1977-12-01

As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36% to 111% of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable. (author)

FFTF scale-model characterization of flow induced vibrational response of reactor internals

International Nuclear Information System (INIS)

Ryan, J.A.; Julyk, L.J.

1977-01-01

As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36% to 111% of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable. (author)

Analysis of the TREAT loss-of-flow tests L6 and L7 using SAS3D

International Nuclear Information System (INIS)

Morris, E.E.; Simms, R.; Gruber, E.E.

1985-01-01

The TREAT loss-of-flow tests L6 and L7 have been analyzed using the SAS3D accident analysis code. The impetus for the analysis was the need for experimentally supported fuel motion modeling in whole core accident studies performed in support of licensing of the Clinch River Breeder Reactor Project. The input prescription chosen for the SAS3D/SLUMPY fuel motion model gave reasonable agreement with the test results. Tests L6 and L7, each conducted with a cluster of three fuel pins, were planned to simulate key events in the loss-of-flow accident scenario for the Clinch River homogeneous reactor

Flow visualization study of post critical heat flux region for inverted bubbly, slug and annular flow regimes

International Nuclear Information System (INIS)

Denten, J.G.; Ishii, M.

1988-11-01

A visual study of film boiling using still photographic and high- speed motion picture methods was carried out in order to analyze the post-CHF hydrodynamics for steady-state inlet pre-CHF two-phase flow regimes. Pre-CHF two-phase flow regimes were established by introducing Freon 113 liquid and nitrogen gas into a jet core injection nozzle. An idealized, post-CHF two-phase core initial flow geometry (cylindrical multiphase jet core surrounded by a coaxial annulus of gas) was established at the nozzle exit by introducing nitrogen gas into the annular gap between the jet nozzle two-phase effluent and the heated test section inlet. For the present study three basic post-CHF flow regimes have been observed: the rough wavy regime (inverted annular flow preliminary break down), the agitated regime (transition between inverted annular and dispersed droplet flow), and the dispersed ligament/droplet regime. For pre-CHF bubbly flow in the jet nozzle, the post-CHF flow (beginning from jet nozzle exit/heated test section inlet) consists of the rough wavy regime, followed by the agitated and then the dispersed ligament/droplet regime. In the same way, for pre-CHF slug flow in the jet core, the post-CHF flow is comprised of the agitated regime at the nozzle exit, followed by the dispersed regime. Pre-CHF annular jet core flow results in a small, depleted post-CHF agitated flow regime at the nozzle exit, immediately followed by the dispersed ligament/droplet regime. Observed post dryout hydrodynamic behavior is reported, with particular attention given to the transition flow pattern between inverted annular and dispersed droplet flow. 43 refs., 20 figs., 5 tabs

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...

Flow distribution experimental study on the emergency core cooling system of the IEA-R1m - IPEN-CNEN/SP - Brazil

International Nuclear Information System (INIS)

Torres, Walmir Maximo; Baptista Filho, Benedito Dias; Ting, Daniel Kao Sun

1999-01-01

This paper presents a brief description of Emergency Core Cooling System designed by the IEA-R1m Reactor and the experimental results of flow distribution over the core. Several parameters were evaluated, such as: relative position of spray header to the reactor core; type and quantity of spray nozzles; spray nozzles position on spray header; and total spray flow. The main conclusions are presented. (author)

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

Loss reduction in axial-flow compressors through low-speed model testing

Science.gov (United States)

Wisler, D. C.

1984-01-01

A systematic procedure for reducing losses in axial-flow compressors is presented. In this procedure, a large, low-speed, aerodynamic model of a high-speed core compressor is designed and fabricated based on aerodynamic similarity principles. This model is then tested at low speed where high-loss regions associated with three-dimensional endwall boundary layers flow separation, leakage, and secondary flows can be located, detailed measurements made, and loss mechanisms determined with much greater accuracy and much lower cost and risk than is possible in small, high-speed compressors. Design modifications are made by using custom-tailored airfoils and vector diagrams, airfoil endbends, and modified wall geometries in the high-loss regions. The design improvements resulting in reduced loss or increased stall margin are then scaled to high speed. This paper describes the procedure and presents experimental results to show that in some cases endwall loss has been reduced by as much as 10 percent, flow separation has been reduced or eliminated, and stall margin has been substantially improved by using these techniques.

Features and validation of discrete element method for simulating pebble flow in reactor core

International Nuclear Information System (INIS)

Xu Yong; Li Yanjie

2005-01-01

The core of a High-Temperature Gas-cooled Reactor (HTGR) is composed of big number of fuel pebbles, their kinetic behaviors are of great importance in estimating the path and residence time of individual pebble, the evolution of the mixing zone for the assessment of the efficiency of a reactor. Numerical method is highlighted in modern reactor design. In view of granular flow, the Discrete Element Model based on contact mechanics of spheres was briefly described. Two typical examples were presented to show the capability of the DEM method. The former is piling with glass/steel spheres, which provides validated evidences that the simulated angles of repose are in good coincidence with the experimental results. The later is particle discharge in a flat- bottomed silo, which shows the effects of material modulus and demonstrates several features. The two examples show the DEM method enables to predict the behaviors, such as the evolution of pebble profiles, streamlines etc., and provides sufficient information for pebble flow analysis and core design. In order to predict the cyclic pebble flow in a HTGR core precisely and efficiently, both model and code improvement are needed, together with rational specification of physical properties with proper measuring techniques. Strategic and methodological considerations were also discussed. (authors)

Ring waves as a mass transport mechanism in air-driven core-annular flows.

Science.gov (United States)

Camassa, Roberto; Forest, M Gregory; Lee, Long; Ogrosky, H Reed; Olander, Jeffrey

2012-12-01

Air-driven core-annular fluid flows occur in many situations, from lung airways to engineering applications. Here we study, experimentally and theoretically, flows where a viscous liquid film lining the inside of a tube is forced upwards against gravity by turbulent airflow up the center of the tube. We present results on the thickness and mean speed of the film and properties of the interfacial waves that develop from an instability of the air-liquid interface. We derive a long-wave asymptotic model and compare properties of its solutions with those of the experiments. Traveling wave solutions of this long-wave model exhibit evidence of different mass transport regimes: Past a certain threshold, sufficiently large-amplitude waves begin to trap cores of fluid which propagate upward at wave speeds. This theoretical result is then confirmed by a second set of experiments that show evidence of ring waves of annular fluid propagating over the underlying creeping flow. By tuning the parameters of the experiments, the strength of this phenomenon can be adjusted in a way that is predicted qualitatively by the model.

Two-phase flow experiments in emergency core cooling feed through the hot leg for developing numerical models

International Nuclear Information System (INIS)

Staebler, T.; Meyer, L.; Schulenberg, T.; Laurien, E.

2006-01-01

When a leakage, a 'loss-of-coolant accident', occurs in a light water reactor, the emergency cooling system is able to supply large amounts of coolant to ensure residual heat removal. This supply can be routed through a special emergency cooling pipe, the 'scoop', into the horizontal section of the main coolant pipe, the 'hot leg'. At the same time, hot steam from the superheated, partly voided core flows against the coolant. This gives rise to a two-phase flow in the opposite direction. A factor of primary interest in this situation is whether the coolant supplied by the emergency cooling system will reach the reactor core. The research project is being conducted in order to compute the rate of water supply by numerical methods. The WENKA test facility has been designed and built at the Karlsruhe Research Center to verify numerical calculations. It can be used to study the fluid dynamics phenomena expected to arise in emergency coolant feeding into the hot leg; the necessary local data can be determined experimentally. An extensive database for validating the numerical calculations is then available to complete the experimental work. (orig.)

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)

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)

Combustion of metal agglomerates in a solid rocket core flow

Science.gov (United States)

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

2013-12-01

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

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.

Flow Boiling on a Downward-Facing Inclined Plane Wall of Core Catcher

International Nuclear Information System (INIS)

Kim, Hyoung Tak; Bang, Kwang Hyun; Suh, Jung Soo

2013-01-01

In order to investigate boiling behavior on downward-facing inclined heated wall prior to the CHF condition, an experiment was carried out with 1.2 m long rectangular channel, inclined by 10 .deg. from the horizontal plane. High speed video images showed that the bubbles were sliding along the heated wall, continuing to grow and combining with the bubbles growing at their nucleation sites in the downstream. These large bubbles continued to slide along the heated wall and formed elongated slug bubbles. Under this slug bubble thin liquid film layer on the heated wall was observed and this liquid film prevents the wall from dryout. The length, velocity and frequency of slug bubbles sliding on the heated wall were measured as a function of wall heat flux and these parameters were used to develop wall boiling model for inclined, downward-facing heated wall. One approach to achieve coolable state of molten core in a PWR-like reactor cavity during a severe accident is to retain the core melt on a so-called core catcher residing on the reactor cavity floor after its relocation from the reactor pressure vessel. The core melt retained in the core catcher is cooled by water coolant flowing in an inclined cooling channel underneath as well as the water pool overlaid on the melt layer. Two-phase flow boiling with downward-facing heated wall such as this core catcher cooling channel has drawn a special attention because this orientation of heated wall may reach boiling crisis at lower heat flux than that of a vertical or upward-facing heated wall. Nishikawa and Fujita, Howard and Mudawar, Qiu and Dhir have conducted experiments to study the effect of heater orientation on boiling heat transfer and CHF. SULTAN experiment was conducted to study inclined large-scale structure coolability by water in boiling natural convection. In this paper, high-speed visualization of boiling behavior on downward-facing heated wall inclined by 10 .deg. is presented and wall boiling model for the

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

Science.gov (United States)

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

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

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...

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)

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)

Numerical study on coolant flow distribution at the core inlet for an integral pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Sun, Lin; Peng, Min Jun; Xia, Genglei; Lv, Xing; Li, Ren [Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin (China)

2017-02-15

When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

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

Calculation of pressure drop and flow redistribution in the core of LMFBR type reactors

International Nuclear Information System (INIS)

Botelho, D.A.; Morgado, O.J.

1985-01-01

It is studied the flow redistribution through of fuel elements to the pressure drop calculation in the core of sodium cooled reactors. Using the quasi-static formulation of equations of the conservation of mass, energy and momentum, it was developed a computer program to flow redistribution calculations and pressure drop for different power levels and total flow simulating an arbitrary number of channels for sodium flowing . An optimization of the number of sufficient channels for calculations of this nature is done. The method is applied in studies of transients in the same reactor. (M.C.K.) [pt

High-resolution continuous-flow analysis setup for water isotopic measurement from ice cores using laser spectroscopy

Science.gov (United States)

Emanuelsson, B. D.; Baisden, W. T.; Bertler, N. A. N.; Keller, E. D.; Gkinis, V.

2015-07-01

Here we present an experimental setup for water stable isotope (δ18O and δD) continuous-flow measurements and provide metrics defining the performance of the setup during a major ice core measurement campaign (Roosevelt Island Climate Evolution; RICE). We also use the metrics to compare alternate systems. Our setup is the first continuous-flow laser spectroscopy system that is using off-axis integrated cavity output spectroscopy (OA-ICOS; analyzer manufactured by Los Gatos Research, LGR) in combination with an evaporation unit to continuously analyze water samples from an ice core. A Water Vapor Isotope Standard Source (WVISS) calibration unit, manufactured by LGR, was modified to (1) enable measurements on several water standards, (2) increase the temporal resolution by reducing the response time and (3) reduce the influence from memory effects. While this setup was designed for the continuous-flow analysis (CFA) of ice cores, it can also continuously analyze other liquid or vapor sources. The custom setups provide a shorter response time (~ 54 and 18 s for 2013 and 2014 setup, respectively) compared to the original WVISS unit (~ 62 s), which is an improvement in measurement resolution. Another improvement compared to the original WVISS is that the custom setups have a reduced memory effect. Stability tests comparing the custom and WVISS setups were performed and Allan deviations (σAllan) were calculated to determine precision at different averaging times. For the custom 2013 setup the precision after integration times of 103 s is 0.060 and 0.070 ‰ for δ18O and δD, respectively. The corresponding σAllan values for the custom 2014 setup are 0.030, 0.060 and 0.043 ‰ for δ18O, δD and δ17O, respectively. For the WVISS setup the precision is 0.035, 0.070 and 0.042 ‰ after 103 s for δ18O, δD and δ17O, respectively. Both the custom setups and WVISS setup are influenced by instrumental drift with δ18O being more drift sensitive than δD. The �

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 ...

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

International Nuclear Information System (INIS)

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

1984-01-01

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

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

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)

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

Inverted annular flow experimental study

International Nuclear Information System (INIS)

De Jarlais, G.; Ishii, M.

1985-04-01

Steady-state inverted annular flow of Freon 113 in up flow was established in a transparent test section. Using a special inlet configuration consisting of long aspect-ratio liquid nozzles coaxially centered within a heated quartz tube, idealized inverted annular flow initial geometry (cylindrical liquid core surrounded by coaxial annulus of gas) could be established. Inlet liquid and gas flowrates, liquid subcooling, and gas density (using various gas species) were measured and varied systematically. The hydrodynamic behavior of the liquid core, and the subsequent downstream break-up of this core into slugs, ligaments and/or droplets of various sizes, was observed. In general, for low inlet liquid velocities it was observed that after the initial formation of roll waves on the liquid core surface, an agitated region of high surface area, with attendant high momentum and energy transfers, occurs. This agitated region appears to propagate downsteam in a quasi-periodic pattern. Increased inlet liquid flow rates, and high gas annulus flow rates tend to diminish the significance of this agitated region. Observed inverted annular flow (and subsequent downstream flow pattern) hydrodynamic behavior is reported, and comparisons are drawn to data generated by previous experimenters studying post-CHF flow

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…

Spatial connectivity in a highly heterogeneous aquifer: From cores to preferential flow paths

Science.gov (United States)

Bianchi, M.; Zheng, C.; Wilson, C.; Tick, G.R.; Liu, Gaisheng; Gorelick, S.M.

2011-01-01

This study investigates connectivity in a small portion of the extremely heterogeneous aquifer at the Macrodispersion Experiment (MADE) site in Columbus, Mississippi. A total of 19 fully penetrating soil cores were collected from a rectangular grid of 4 m by 4 m. Detailed grain size analysis was performed on 5 cm segments of each core, yielding 1740 hydraulic conductivity (K) estimates. Three different geostatistical simulation methods were used to generate 3-D conditional realizations of the K field for the sampled block. Particle tracking calculations showed that the fastest particles, as represented by the first 5% to arrive, converge along preferential flow paths and exit the model domain within preferred areas. These 5% fastest flow paths accounted for about 40% of the flow. The distribution of preferential flow paths and particle exit locations is clearly influenced by the occurrence of clusters formed by interconnected cells with K equal to or greater than the 0.9 decile of the data distribution (10% of the volume). The fraction of particle paths within the high-K clusters ranges from 43% to 69%. In variogram-based K fields, some of the fastest paths are through media with lower K values, suggesting that transport connectivity may not require fully connected zones of relatively homogenous K. The high degree of flow and transport connectivity was confirmed by the values of two groups of connectivity indicators. In particular, the ratio between effective and geometric mean K (on average, about 2) and the ratio between the average arrival time and the arrival time of the fastest particles (on average, about 9) are consistent with flow and advective transport behavior characterized by channeling along preferential flow paths. ?? 2011 by the American Geophysical Union.

Investigations on flow reversal in stratified horizontal flow

International Nuclear Information System (INIS)

Staebler, T.; Meyer, L.; Schulenberg, T.; Laurien, E.

2005-01-01

The phenomena of flow reversal in stratified flows are investigated in a horizontal channel with application to the Emergency Core Cooling System (ECCS) in Pressurized Water Reactors (PWR). In case of a Loss-of-Coolant-Accident (LOCA), coolant can be injected through a secondary pipe within the feeding line of the primary circuit, the so called hot leg, counter-currently to the steam flow. It is essential that the coolant reaches the reactor core to prevent overheating. Due to high temperatures in such accident scenarios, steam is generated in the core, which escapes from the reactor vessel through the hot leg. In case of sufficiently high steam flow rates, only a reduced amount of coolant or even no coolant will be delivered to the reactor core. The WENKA test facility at the Institute for Nuclear and Energy Technologies (IKET) at Forschungszentrum Karlsruhe is capable to investigate the fluid dynamics of two-phase flows in such scenarios. Water and air flow counter-currently in a horizontal channel made of clear acrylic glass to allow full optical access. Flow rates of water and air can be varied independently within a wide range. Once flow reversal sets in, a strong hysteresis effect must be taken into account. This was quantified during the present investigations. Local experimental data are needed to expand appropriate models on flow reversal in horizontal two-phase flow and to include them into numerical codes. Investigations are carried out by means of Particle Image Velocimetry (PIV) to obtain local flow velocities without disturbing the flow. Due to the wavy character of the flow, strong reflections at the interfacial area must be taken into account. Using fluorescent particles and an optical filter allows eliminating the reflections and recording only the signals of the particles. The challenges in conducting local investigations in stratified wavy flows by applying optical measurement techniques are discussed. Results are presented and discussed allowing

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

Development of the loss coefficient correlation for cross flow between graphite fuel blocks in the core of prismatic very high temperature reactor-PMR200

International Nuclear Information System (INIS)

Lee, Jeong-Hun; Cho, Hyoung-Kyu; Park, Goon-Cherl

2016-01-01

Highlights: • Cross flow experimental data are produced with wedge-shaped and parallel gaps. • The results of a CFD analysis and experimental data are in good agreement. • Pressure loss coefficient for the cross gap between fuel blocks in PMR200 is found. • A new correlation of the cross flow loss coefficient for PMR200 is proposed. - Abstract: The core of the very high temperature reactor (VHTR) PMR200 (a prismatic modular reactor rated at 200 MW of thermal power) consists of hexagonal prismatic fuel blocks and reflector blocks made of graphite. If the core bypass flow ratio increases, the coolant channel flow is decreased and can then lower the heat removal efficiency, resulting in a locally increased fuel block temperature. The coolant channels in the fuel blocks are connected to bypass gaps by the cross gap, complicating flow distribution in the VHTR core. Therefore, reliable estimation of the bypass flow is highly important for the design and safety analysis of the VHTR core. Because of the complexity of the core geometry and gap configuration, it is challenging to predict the flow distribution in the VHTR core. To analyze this flow distribution accurately, it is necessary to determine the cross flow phenomena, and the loss coefficient across the cross gap has to be evaluated to determine the flow distribution in the VHTR core when a lumped parameter code or a flow network analysis code that uses the correlation of the loss coefficient is employed. The purpose of this paper is to develop a loss coefficient correlation applicable to the cross gap in the PMR200 core. The cross flow was evaluated experimentally using the difference between the measured inlet and outlet mass flow rates. Next, the applicability of a commercial computational fluid dynamics (CFD) code, CFX 15, was confirmed by comparing the experimental data and CFD analysis results. To understand the cross flow phenomena, the loss coefficient was evaluated; in the high Reynolds number region

Development of the loss coefficient correlation for cross flow between graphite fuel blocks in the core of prismatic very high temperature reactor-PMR200

Energy Technology Data Exchange (ETDEWEB)

Lee, Jeong-Hun, E-mail: huny12@snu.ac.kr; Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr; Park, Goon-Cherl, E-mail: parkgc@snu.ac.kr

2016-10-15

Highlights: • Cross flow experimental data are produced with wedge-shaped and parallel gaps. • The results of a CFD analysis and experimental data are in good agreement. • Pressure loss coefficient for the cross gap between fuel blocks in PMR200 is found. • A new correlation of the cross flow loss coefficient for PMR200 is proposed. - Abstract: The core of the very high temperature reactor (VHTR) PMR200 (a prismatic modular reactor rated at 200 MW of thermal power) consists of hexagonal prismatic fuel blocks and reflector blocks made of graphite. If the core bypass flow ratio increases, the coolant channel flow is decreased and can then lower the heat removal efficiency, resulting in a locally increased fuel block temperature. The coolant channels in the fuel blocks are connected to bypass gaps by the cross gap, complicating flow distribution in the VHTR core. Therefore, reliable estimation of the bypass flow is highly important for the design and safety analysis of the VHTR core. Because of the complexity of the core geometry and gap configuration, it is challenging to predict the flow distribution in the VHTR core. To analyze this flow distribution accurately, it is necessary to determine the cross flow phenomena, and the loss coefficient across the cross gap has to be evaluated to determine the flow distribution in the VHTR core when a lumped parameter code or a flow network analysis code that uses the correlation of the loss coefficient is employed. The purpose of this paper is to develop a loss coefficient correlation applicable to the cross gap in the PMR200 core. The cross flow was evaluated experimentally using the difference between the measured inlet and outlet mass flow rates. Next, the applicability of a commercial computational fluid dynamics (CFD) code, CFX 15, was confirmed by comparing the experimental data and CFD analysis results. To understand the cross flow phenomena, the loss coefficient was evaluated; in the high Reynolds number region

Fundamental study on flow characteristics of disrupted core pool at a low energy level (Joint research)

International Nuclear Information System (INIS)

Morita, Koji; Liu, Ping; Matsumoto, Tatsuya; Fukuda, Kenji; Tobita, Yoshiharu; Yamano, Hidemasa; Sato, Ikken

2009-09-01

Dynamic behaviors of solid-particle dominant multiphase flows were investigated to model the mobility of core materials in a low-energy disrupted core of a liquid metal fast reactor. Two series of experiments were performed, those were dam-break experiments and bubble visualization experiments. Verification of fluid-dynamics models used in the fast reactor safety analysis code SIMMER-III was also conducted based on the numerical simulations of these experiments. The experimental analyses show that SIMMER-III can represent effects of solid particle interaction on multiphase flow behaviors by adjusting model parameters of the particle jamming model if the particles are immersed in liquid phase. Further improvement of SIMMER-III with more generalized models is necessary to appropriately simulate interactions between solid particles in a wider range of flow conditions. (author)

Hot Hydrogen Test Facility

International Nuclear Information System (INIS)

W. David Swank

2007-01-01

The core in a nuclear thermal rocket will operate at high temperatures and in hydrogen. One of the important parameters in evaluating the performance of a nuclear thermal rocket is specific impulse, ISp. This quantity is proportional to the square root of the propellant's absolute temperature and inversely proportional to square root of its molecular weight. Therefore, high temperature hydrogen is a favored propellant of nuclear thermal rocket designers. Previous work has shown that one of the life-limiting phenomena for thermal rocket nuclear cores is mass loss of fuel to flowing hydrogen at high temperatures. The hot hydrogen test facility located at the Idaho National Lab (INL) is designed to test suitability of different core materials in 2500 C hydrogen flowing at 1500 liters per minute. The facility is intended to test non-uranium containing materials and therefore is particularly suited for testing potential cladding and coating materials. In this first installment the facility is described. Automated Data acquisition, flow and temperature control, vessel compatibility with various core geometries and overall capabilities are discussed

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

International Nuclear Information System (INIS)

Sobajima, Makoto

1985-08-01

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

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.

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.

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.)

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)

CANFLEX fuel bundle cross-flow endurance test 2 (test procedure)

Energy Technology Data Exchange (ETDEWEB)

Hong, Sung Deok; Chung, C. H.; Chang, S. K. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-04-01

This report describes test procedure of cross-flow 2 test for CANFLEX fuel. In October 1996. a cross-flow test was successfully performed in the KAERI Hot Test Loop for four hours at a water flow rate of 31kg/s, temperature of 266 deg C and inlet pressure of 11MPa, but it is requested more extended time periods to determine a realistic operational margin for the CANFLEX bundle during abnormal refuelling operations. The test shall be conducted for twenty two hours under the reactor conditions. After an initial period of ten hours, the test shall be stopped at the intervals of four hours for bundle inspection and inspect the test bundle end-plate to end-cap welds for failure or crack propagation using liquid penetrant examination. 2 refs., 1 fig. (Author)

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...

ROSA-II test data report, 11

International Nuclear Information System (INIS)

1978-02-01

Results of the ROSA-II tests simulating a loss-of-coolant accident (LOCA) and effects of an emergency core cooling system (ECCS) in a pressurized water reactor (PWR) are presented including the test conditions and interpretations of the data in test runs 327,328,329 and 330. Each test was performed with large double-ended hot leg break and effect of the break area distribution (break diameter are 25.0 mm at one end and 37.5 mm at the other end of break) and of pump circulation upon coolant flow in the core were studied. The following are the results: In the case of a smaller break on the steam generator side, core cooling was achieved due to upward coolant flow in the core and early reflooding by ACC water injected into the cold leg. In the case of a smaller break area on the vessel side, on the other hand, coolant flow in the core was stagnant and the heater rods were mostly exposed to steam, so that core cooling was not as good. Effect of the coolant circulation by acting pump on the core cooling during a blowdown was not significant except that in a steam generator side small break the core cooling was improved. (auth.)

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.

Downscale cascades in tracer transport test cases: an intercomparison of the dynamical cores in the Community Atmosphere Model CAM5

Directory of Open Access Journals (Sweden)

J. Kent

2012-12-01

Full Text Available The accurate modeling of cascades to unresolved scales is an important part of the tracer transport component of dynamical cores of weather and climate models. This paper aims to investigate the ability of the advection schemes in the National Center for Atmospheric Research's Community Atmosphere Model version 5 (CAM5 to model this cascade. In order to quantify the effects of the different advection schemes in CAM5, four two-dimensional tracer transport test cases are presented. Three of the tests stretch the tracer below the scale of coarse resolution grids to ensure the downscale cascade of tracer variance. These results are compared with a high resolution reference solution, which is simulated on a resolution fine enough to resolve the tracer during the test. The fourth test has two separate flow cells, and is designed so that any tracer in the western hemisphere should not pass into the eastern hemisphere. This is to test whether the diffusion in transport schemes, often in the form of explicit hyper-diffusion terms or implicit through monotonic limiters, contains unphysical mixing.

An intercomparison of three of the dynamical cores of the National Center for Atmospheric Research's Community Atmosphere Model version 5 is performed. The results show that the finite-volume (CAM-FV and spectral element (CAM-SE dynamical cores model the downscale cascade of tracer variance better than the semi-Lagrangian transport scheme of the Eulerian spectral transform core (CAM-EUL. Each scheme tested produces unphysical mass in the eastern hemisphere of the separate cells test.

Two-phase flow visualization under reservoir conditions for highly heterogeneous conglomerate rock: A core-scale study for geologic carbon storage.

Science.gov (United States)

Kim, Kue-Young; Oh, Junho; Han, Weon Shik; Park, Kwon Gyu; Shinn, Young Jae; Park, Eungyu

2018-03-20

Geologic storage of carbon dioxide (CO 2 ) is considered a viable strategy for significantly reducing anthropogenic CO 2 emissions into the atmosphere; however, understanding the flow mechanisms in various geological formations is essential for safe storage using this technique. This study presents, for the first time, a two-phase (CO 2 and brine) flow visualization under reservoir conditions (10 MPa, 50 °C) for a highly heterogeneous conglomerate core obtained from a real CO 2 storage site. Rock heterogeneity and the porosity variation characteristics were evaluated using X-ray computed tomography (CT). Multiphase flow tests with an in-situ imaging technology revealed three distinct CO 2 saturation distributions (from homogeneous to non-uniform) dependent on compositional complexity. Dense discontinuity networks within clasts provided well-connected pathways for CO 2 flow, potentially helping to reduce overpressure. Two flow tests, one under capillary-dominated conditions and the other in a transition regime between the capillary and viscous limits, indicated that greater injection rates (potential causes of reservoir overpressure) could be significantly reduced without substantially altering the total stored CO 2 mass. Finally, the capillary storage capacity of the reservoir was calculated. Capacity ranged between 0.5 and 4.5%, depending on the initial CO 2 saturation.

Steady-state thermal hydraulic analysis and flow channel blockage accident analysis of JRR-3 silicide core

International Nuclear Information System (INIS)

Kaminaga, Masanori

1997-03-01

JRR-3 is a light water moderated and cooled, beryllium and heavy water reflected pool type research reactor using low enriched uranium (LEU) plate-type fuels. Its thermal power is 20 MW. The core conversion program from uranium-aluminum (UAl x -Al) dispersion type fuel (aluminide fuel) to uranium-silicon-aluminum (U 3 Si 2 -Al) dispersion type fuel (silicide fuel) is currently conducted at the JRR-3. This report describes about the steady-state thermal hydraulic analysis results and the flow channel blockage accident analysis result. In JRR-3, there are two operation mode. One is high power operation mode up to 20 MW, under forced convection cooling using the primary and the secondary cooling systems. The other is low power operation mode up to 200 kW, under natural circulation cooling between the reactor core and the reactor pool without the primary and the secondary cooling systems. For the analysis of the flow channel blockage accident, COOLOD code was used. On the other hand, steady-state thermal hydraulic analysis for both of the high power operation mode under forced convection cooling and low power operation under natural convection cooling, COOLOD-N2 code was used. From steady-state thermal hydraulic analysis results of both forced and natural convection cooling, fuel temperature, minimum DNBR etc. meet the design criteria and JRR-3 LEU silicide core has enough safety margin under normal operation conditions. Furthermore, flow channel blockage accident analysis results show that one channel flow blockage accident meet the safety criteria for accident conditions which have been established for JRR-3 LEU silicide core. (author)

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

Directory of Open Access Journals (Sweden)

NAM-IL TAK

2013-11-01

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

Structure, Mechanics and Flow Properties of Fractured Shale: Core-Scale Experimentation and In-situ Imaging

Science.gov (United States)

Abdelmalek, B. F.; Karpyn, Z.; Liu, S.

2014-12-01

Over the last several years, hydrocarbon exploitation and development in North America has been heavily centered on shale gas plays. However, the physical attributes of shales and their manifestation on transport properties and storage capacity remain poorly understood. Therefore, more experimentally based data are needed to fill the gaps in understanding both transport and storage of fluids in shale. The proposed work includes installation and testing of an experimental system which is capable of monitoring the dynamic evolution of shale core permeability under variable loading conditions and in coordination with X-ray microCT imaging. The goal of this study is to better understand and quantify fluid flow patterns and associated transport dynamics of fractured shale samples. The independent variables considered in this study are: mechanical loading and pore pressure. The mechanical response of shale core is captured for different loading paths. To best replicate the in-situ production scenario, the pore pressure is progressively depleted to mimic pressure decline. During the course of experimentation, permeability is estimated using the pulse-decay method under tri-axial stress boundary conditions. Simultaneously, X-ray microCT imaging is used with a tracer gas that is allowed to flow through the sample as an illuminating agent. In the presence of an illuminating agent, either Xenon or Krypton, the X-ray CT scanner can image fractures, global pathways and diffusional fronts in the matrix, as well as sorption sites that reflect heterogeneities in the sample and localized deformation. Anticipated results from these experiments will help quantify permeability evolution as a function of different loading conditions and pore pressure depletion. Also, the X-ray images will help visualize the change of flow patterns and the intensity of sorption as a function of mechanical loading and pore pressure.

A new model for turbidity current behavior based on integration of flow monitoring and precision coring in a submarine canyon

Science.gov (United States)

Symons, William O.; Sumner, Esther J.; Paull, Charles K.; Cartigny, Matthieu J.B.; Xu, Jingping; Maier, Katherine L.; Lorenson, Thomas; Talling, Peter J.

2017-01-01

Submarine turbidity currents create some of the largest sediment accumulations on Earth, yet there are few direct measurements of these flows. Instead, most of our understanding of turbidity currents results from analyzing their deposits in the sedimentary record. However, the lack of direct flow measurements means that there is considerable debate regarding how to interpret flow properties from ancient deposits. This novel study combines detailed flow monitoring with unusually precisely located cores at different heights, and multiple locations, within the Monterey submarine canyon, offshore California, USA. Dating demonstrates that the cores include the time interval that flows were monitored in the canyon, albeit individual layers cannot be tied to specific flows. There is good correlation between grain sizes collected by traps within the flow and grain sizes measured in cores from similar heights on the canyon walls. Synthesis of flow and deposit data suggests that turbidity currents sourced from the upper reaches of Monterey Canyon comprise three flow phases. Initially, a thin (38–50 m) powerful flow in the upper canyon can transport, tilt, and break the most proximal moorings and deposit chaotic sands and gravel on the canyon floor. The initially thin flow front then thickens and deposits interbedded sands and silty muds on the canyon walls as much as 62 m above the canyon floor. Finally, the flow thickens along its length, thus lofting silty mud and depositing it at greater altitudes than the previous deposits and in excess of 70 m altitude.

Hypervapotron flow testing with rapid prototype models

International Nuclear Information System (INIS)

Driemeyer, D.; Hellwig, T.; Kubik, D.; Langenderfer, E.; Mantz, H.; McSmith, M.; Jones, B.; Butler, J.

1995-01-01

A flow test model of the inlet section of a three channel hypervapotron plate that has been proposed as a heat sink in the ITER divertor was prepared using a rapid prototyping stereolithography process that is widely used for component development in US industry. An existing water flow loop at the University of Illinois is being used for isothermal flow tests to collect pressure drop data for comparison with proposed vapotron friction factor correlations. Differential pressure measurements are taken, across the test section inlet manifold, the vapotron channel (about a seven inch length), the outlet manifold and the inlet-to-outlet. The differential pressures are currently measured with manometers. Tests were conducted at flow velocities from 1--10 m/s to cover the full range of ITER interest. A tap was also added for a small hypodermic needle to inject dye into the flow channel at several positions to examine the nature of the developing flow field at the entrance to the vapotron section. Follow-on flow tests are planned using a model with adjustable flow channel dimensions to permit more extensive pressure drop data to be collected. This information will be used to update vapotron design correlations for ITER

The impact of interpreted flow regimes during constant head injection tests on the estimated transmissivity from injection tests and difference flow logging

Energy Technology Data Exchange (ETDEWEB)

Hjerne, Calle; Ludvigsson, Jan-Erik; Harrstroem, Johan [Geosigma AB, Uppsala (Sweden)

2013-04-15

A large number of constant head injection tests were carried out in the site investigation at Forsmark using the Pipe String System, PSS3. During the original evaluation of the tests the dominating transient flow regimes during both the injection and recovery period were interpreted together with estimation of hydraulic parameters. The flow regimes represent different flow and boundary conditions during the tests. Different boreholes or borehole intervals may display different distributions of flow regimes. In some boreholes good agreement was obtained between the results of the injection tests and difference flow logging with Posiva flow log (PFL) but in other boreholes significant discrepancies were found. The main objective of this project is to study the correlation between transient flow regimes from the injection tests and other borehole features such as transmissivity, depth, geology, fracturing etc. Another subject studied is whether observed discrepancies between estimated transmissivity from difference flow logging and injection tests can be correlated to interpreted flow regimes. Finally, a detailed comparison between transient and stationary evaluation of transmissivity from the injection tests in relation to estimated transmissivity from PFL tests in corresponding sections is made. Results from previous injection tests in 5 m sections in boreholes KFM04, KFM08A and KFM10A were used. Only injection tests above the (test-specific) measurement limit regarding flow rate are included in the analyses. For all of these tests transient flow regimes were interpreted. In addition, results from difference flow logging in the corresponding 5 m test sections were used. Finally, geological data of fractures together with rock and fracture zone properties have been used in the correlations. Flow regimes interpreted from the injection period of the tests are generally used in the correlations but deviations between the interpreted flow regimes from the injection and

High-resolution continuous flow analysis setup for water isotopic measurement from ice cores using laser spectroscopy

Science.gov (United States)

Emanuelsson, B. D.; Baisden, W. T.; Bertler, N. A. N.; Keller, E. D.; Gkinis, V.

2014-12-01

Here we present an experimental setup for water stable isotopes (δ18O and δD) continuous flow measurements. It is the first continuous flow laser spectroscopy system that is using Off-Axis Integrated Cavity Output Spectroscopy (OA-ICOS; analyzer manufactured by Los Gatos Research - LGR) in combination with an evaporation unit to continuously analyze sample from an ice core. A Water Vapor Isotopic Standard Source (WVISS) calibration unit, manufactured by LGR, was modified to: (1) increase the temporal resolution by reducing the response time (2) enable measurements on several water standards, and (3) to reduce the influence from memory effects. While this setup was designed for the Continuous Flow Analysis (CFA) of ice cores, it can also continuously analyze other liquid or vapor sources. The modified setup provides a shorter response time (~54 and 18 s for 2013 and 2014 setup, respectively) compared to the original WVISS unit (~62 s), which is an improvement in measurement resolution. Another improvement compared to the original WVISS is that the modified setup has a reduced memory effect. Stability tests comparing the modified WVISS and WVISS setups were performed and Allan deviations (σAllan) were calculated to determine precision at different averaging times. For the 2013 modified setup the precision after integration times of 103 s are 0.060 and 0.070‰ for δ18O and δD, respectively. For the WVISS setup the corresponding σAllan values are 0.030, 0.060 and 0.043‰ for δ18O, δD and δ17O, respectively. For the WVISS setup the precision is 0.035, 0.070 and 0.042‰ after 103 s for δ18O, δD and δ17O, respectively. Both the modified setups and WVISS setup are influenced by instrumental drift with δ18O being more drift sensitive than δD. The σAllan values for δ18O of 0.30 and 0.18‰ for the modified (2013) and WVISS setup, respectively after averaging times of 104 s (2.78 h). The Isotopic Water Analyzer (IWA)-modified WVISS setup used during the

RELAPS choked flow model and application to a large scale flow test

International Nuclear Information System (INIS)

Ransom, V.H.; Trapp, J.A.

1980-01-01

The RELAP5 code was used to simulate a large scale choked flow test. The fluid system used in the test was modeled in RELAP5 using a uniform, but coarse, nodalization. The choked mass discharge rate was calculated using the RELAP5 choked flow model. The calulations were in good agreement with the test data, and the flow was calculated to be near thermal equilibrium

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)

Fuel assembly outlet temperature profile influence on core by-pass flow and power distribution determination in WWER -440 reactors

International Nuclear Information System (INIS)

Petenyi, V.; Klucarova, K.; Remis, J.

2003-01-01

The in core instrumentation of the WWER-440 reactors consists of the thermocouple system and the system of self powered detectors (SPD). The thermocouple systems are positioned about 50 cm above the fuel bundle upper flow-mixing grid. The usual assumption is that, the coolant is well mixed in the Tc location, i.e. the temperature is constant through the flow cross-section area. The present evaluations by using the FLUENT 5.5.14 code reveal that, this assumption is not fulfilled. There exists a temperature profile that depends on fuel assembly geometry and on inner power profile of the fuel assembly. The paper presents the estimation of this effect and its influence on the core power distribution and the core by-pass flow determination. Comparison with measurements in Mochovce NPP will also be a part of this presentation (Authors)

Analysis of two-phase flow and boiling heat transfer in inclined channel of core-catcher

International Nuclear Information System (INIS)

Tahara, M.; Suzuki, Y.; Abe, N.; Kurita, T.; Hamazaki, R.; Kojima, Y.

2008-01-01

Passive Corium Cooling System (CCS) provides a function of ex-vessel debris cooling and molten core stabilization during a severe accident. CCS features inclined cooling channels arranged axi-symmetrically below the core-catcher basin. In order to estimate the coolability of the inclined cooling channel, it is indispensable to identify the flow pattern of the two-phase flow in the cooling channel. Several former studies for the two-phase flow pattern in the inclined channel are referred. Taitel and Dukler (1976) developed a prediction method of the flow pattern transition in horizontal and near horizontal tubes. Barnea et al. (1980) showed the flow pattern map of upward flow with 10 degrees inclination. Sakaguti et al. (1996) observed the two-phase flow patterns in the horizontal pipe connected with slightly upward pipe, in which the flow pattern in the pipe with a bending part was expressed by the combination of a basic flow pattern and some auxiliary flow patterns. Then we investigated these studies In order to identify the flow patterns observed in the inclined cooling channel of CCS. Furthermore we experimentally observed the flow patterns in the inclined cooling channel with various inlet conditions. As a result of the investigation and observation, typical flow patterns in the inclined cooling channel were identified. Two typical flow patterns were observed depending on the steam flow rate, one of which is 'elongated bubble 'flow, and the other is 'churn with collapsing backward and upward slug 'flow The flow and heat transfer in the inclined channel of CCS is analyzed by using a two-phase analysis code employing two-fluid model in which the constitutive equations for the two-phase flow in inclined channels are incorporated. That is, drift flux parameter for each of the elongated bubble flow, and the churn with collapsing backward and upward slug flow are incorporated to the two-phase analysis code, which are based on the rising velocity of the long bubble in

Assessment of the MARS Code Using the Two-Phase Natural Circulation Experiments at a Core Catcher Test Facility

Directory of Open Access Journals (Sweden)

Dong Hun Lee

2017-01-01

Full Text Available A core catcher has been developed to maintain the integrity of nuclear reactor containment from molten corium during a severe accident. It uses a two-phase natural circulation for cooling molten corium. Flow in a typical core catcher is unique because (i it has an inclined cooling channel with downwards-facing heating surface, of which flow processes are not fully exploited, (ii it is usually exposed to a low-pressure condition, where phase change causes dramatic changes in the flow, and (iii the effects of a multidimensional flow are very large in the upper part of the core catcher. These features make computational analysis more difficult. In this study, the MARS code is assessed using the two-phase natural circulation experiments that had been conducted at the CE-PECS facility to verify the cooling performance of a core catcher. The code is a system-scale thermal-hydraulic (TH code and has a multidimensional TH component. The facility was modeled by using both one- and three-dimensional components. Six experiments at the facility were selected to investigate the parametric effects of heat flux, pressure, and form loss. The results show that MARS can predict the two-phase flow at the facility reasonably well. However, some limitations are obviously revealed.

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

Ultrasonic flow-meter test in sodium

International Nuclear Information System (INIS)

Ishii, Y.; Uno, O.; Kamei, M.

1978-01-01

As a part of the R and D programme for the prototype fast breeder reactor MONJU, an ultrasonic flow-meter (USFM) test is being carried out in sodium in the O-Arai Engineering Center of PNC. Prior to the present test, an in-water test was done at the manufacturer's as a preliminary investigation. The results reported here are the results up to the present. Calibration tests using the actual fluid were conducted on a 12-inch ultrasonic flow-meter with guide rods fabricated for sodium flow measurement. The test conditions in sodium were a temperature of 200 approximately 400 0 C and flow-rates of 0 approximately 6m/s. The main results are: (1) The linearity of output signal was good and accuracy was within 1%; (2) The alternating type of the USFM was much better than the fixed type in temperature change; (3) 2MHz of transducer frequency was better than 3MHz in sodium; (4) The S/N ratio of the ultrasonic signal and the length/diameter effect in a wide range in sodium surpassed the in-water test. (author)

Analytical study on coolant temperature of several leak flows in the experimental VHTr core

International Nuclear Information System (INIS)

Fumizawa, Motoh; Arai, Taketoshi; Miyamoto, Yoshiaki

1982-08-01

This report describes heat transfer analysis of several leak flows which bypass main coolant flow path in the experimental VHTR core. The analysis contains the leak flow at permanent reflectors, replaceable reflectors and gaps between fuel columns. The summary of the results are as follows: (1) the temperature of the leak flow gas increases up to the surface temperature of permanent reflectors, (2) the gas temperature at replaceable reflectors increases at least 40 0 C in case of the worst analytical condition, (3) the gas temperature increases remarkably with decreasing equivalent diameter which is changed by the angle of bevel edge of the reflector, (4) while the gas temperature is low at the upper part of the fuel element, the temperature increases rapidly when it flow down along the gap of the fuel columns. (author)

Detailed evaluation of two phase natural circulation flow in the cooling channel of the ex-vessel core catcher for EU-APR1400

Energy Technology Data Exchange (ETDEWEB)

Park, Rae-Joon, E-mail: rjpark@kaeri.re.kr; Ha, Kwang-Soon; Rhee, Bo-Wook; Kim, Hwan Yeol

2016-03-15

Highlights: • Ex-vessel core catcher of PECS is installed in EU-APR1400. • CE-PECS has been conducted to test a cooling capability of the PECS. • Two phase flow in CE-PECS and PECS was analyzed using RELAP5/MOD3. • RELAP5 results are very similar to the CE-PECS data. • The super-step design is suitable for steam injection into the downcomer in PECS. - Abstract: The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1 kg/m{sup 2} s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.

Detailed evaluation of two phase natural circulation flow in the cooling channel of the ex-vessel core catcher for EU-APR1400

International Nuclear Information System (INIS)

Park, Rae-Joon; Ha, Kwang-Soon; Rhee, Bo-Wook; Kim, Hwan Yeol

2016-01-01

Highlights: • Ex-vessel core catcher of PECS is installed in EU-APR1400. • CE-PECS has been conducted to test a cooling capability of the PECS. • Two phase flow in CE-PECS and PECS was analyzed using RELAP5/MOD3. • RELAP5 results are very similar to the CE-PECS data. • The super-step design is suitable for steam injection into the downcomer in PECS. - Abstract: The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1 kg/m"2 s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.

Test facilities for HTR, (2)

International Nuclear Information System (INIS)

Ishizuka, Hiroshi; Hayakawa, Hitoshi; Miki, Toshiya.

1981-01-01

The core of the multi-purpose high temperature gas-cooled experimental reactor is a circular column as a whole, in which the columns of stacked graphite blocks of hexagonal prism are arranged. The blocks in a column are doweled so as not to move horizontally, but adjacent columns vibrate while colliding mutually at the time of an earthquake because there is a gap between them. For the purpose to know the vibrating characteristics of a column surrounded by gap, Fuji Electric Co., Ltd., carried out the experiment. The tested column, the testing setup and the test result are reported. The distribution of flow rate in the core must be clarified, and the design data must be established early for confirming the feasibility of core design. The core structure tester was installed in Japan Atomic Energy Research Institute. The 1/2.75 scale model of the reactor bed was used, and the sealing performance of the block assemblies was tested. The sealing tester is related also to the distribution of flow rate in the core, and the basic performance of seal elements and the cross flow in fuel blocks were tested. The one-column tester and the seal element/two-column tester, the piping unit and the blower filter unit compose this tester. (Kako, I.)

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.

Study on corrosion test techniques in lead bismuth eutectic flow. Joint research report in JFY2002

International Nuclear Information System (INIS)

Takahashi, Minoru; Sekimoto, Hiroshi

2003-03-01

The evaluation of corrosion behaviors of core and structural materials in lead bismuth eutectic is one of the key issues for the utilization of lead bismuth eutectic as a coolant of the primary loops of lead bismuth cooled fast breeder reactors (FBRs) and the intermediate heat transport media of new-type steam generators of the sodium cooled FBRs. The purpose of the present study is to establish corrosion test techniques in lead bismuth eutectic flow. The techniques of steel corrosion test and oxygen control in flowing lead bismuth eutectic, and the technologies of a lead bismuth flow test at high temperature and high velocity were developed through corrosion test using a lead bismuth flow test loop of the Tokyo Institute of Technology in JFY2002. The major results are summarized as follows: (1) Techniques of fabrication, mount and rinse of corrosion specimens, measurement method of weight loss, and SEM/EDX analysis method have been established through lead bismuth corrosion test. (2) Weight losses were measured, corrosion and lead bismuth-adhered layers and eroded parts were observed in two 1000 hr-corrosion tests, and the results were compared with each other for twelve existing steels including ODS, F82H and SUH-3. (3) An oxygen sensor made of zirconia electrolyte structurally resistant to thermal stress and thermal shock was developed and tested in the lead bismuth flow loop. Good performance has been obtained. (4) An oxygen control method by injecting argon and hydrogen mixture gas containing steam into lead bismuth was applied to the lead bismuth flow loop, and technical issues for the development of the oxygen control method were extracted. (5) Technical measures for freezing and leakage of lead bismuth in the flow loop were accumulated. (6) Technical measures for flow rate decrease/blockage due to precipitation of oxide and corrosion products in a low temperature section of the lead bismuth flow loop were accumulated. (7) Electromagnetic flow meters with MI

Characterization of interfacial waves and pressure drop in horizontal oil-water core-annular flows

Science.gov (United States)

Tripathi, Sumit; Tabor, Rico F.; Singh, Ramesh; Bhattacharya, Amitabh

2017-08-01

We study the transportation of highly viscous furnace-oil in a horizontal pipe as core-annular flow (CAF) using experiments. Pressure drop and high-speed images of the fully developed CAF are recorded for a wide range of flow rate combinations. The height profiles (with respect to the centerline of the pipe) of the upper and lower interfaces of the core are obtained using a high-speed camera and image analysis. Time series of the interface height are used to calculate the average holdup of the oil phase, speed of the interface, and the power spectra of the interface profile. We find that the ratio of the effective velocity of the annular fluid to the core velocity, α , shows a large scatter. Using the average value of this ratio (α =0.74 ) yields a good estimate of the measured holdup for the whole range of flow rate ratios, mainly due to the low sensitivity of the holdup ratio to the velocity ratio. Dimensional analysis implies that, if the thickness of the annular fluid is much smaller than the pipe radius, then, for the given range of parameters in our experiments, the non-dimensional interface shape, as well as the non-dimensional wall shear stress, can depend only on the shear Reynolds number and the velocity ratio. Our experimental data show that, for both lower and upper interfaces, the normalized power spectrum of the interface height has a strong dependence on the shear Reynolds number. Specifically, for low shear Reynolds numbers, interfacial modes with large wavelengths dominate, while, for large shear Reynolds numbers, interfacial modes with small wavelengths dominate. Normalized variance of the interface height is higher at lower shear Reynolds numbers and tends to a constant with increasing shear Reynolds number. Surprisingly, our experimental data also show that the effective wall shear stress is, to a large extent, proportional to the square of the core velocity. Using the implied scalings for the holdup ratio and wall shear stress, we can derive

Experimental study of core bypass flow in a prismatic VHTR based on a two-layer block model

Energy Technology Data Exchange (ETDEWEB)

Wang, Huhu, E-mail: huhuwang@tamu.edu; Hassan, Yassin A., E-mail: y-hassan@tamu.edu; Dominguez-Ontiveros, Elvis, E-mail: elvisdom@tamu.edu

2016-09-15

Bypass flow in a prismatic very high temperature gas-cooled nuclear reactor (VHTR) plays an important role in determining the coolant distribution in the core region. Efficient removal of heat from the core relies on the majority of coolant passing through the coolant channels instead of the bypass gaps. Consequently, the bypass flow fraction and its flow characteristic are important in the design process of the prismatic VHTR. The objective of this study is to experimentally investigate the flow behavior including the turbulence characteristics inside the bypass gaps using laser Doppler velocimetry (LDV), bypass fraction and pressure drops in the system. The experiment facility constructed at Texas A&M University is a scaled model consisting of two layers of fuel blocks. The distributions of the mean streamwise velocity, turbulence intensity and turbulence kinetic energy within the bypass gap at two different elevations under different Reynolds number were investigated. Uncertainties in the bypass flow fraction estimation were evaluated. The velocity and turbulence study in this work is considered to be unique, and may serve as a benchmark for the related numerical calculations.

Coupling of the 3D neutron kinetic core model DYN3D with the CFD software ANSYS-CFX

International Nuclear Information System (INIS)

Grahn, Alexander; Kliem, Sören; Rohde, Ulrich

2015-01-01

Highlights: • Improved thermal hydraulic description of nuclear reactor cores. • Possibility of three-dimensional flow phenomena in the core, such as cross flow, flow reversal, flow around obstacles. • Simulation at higher spatial resolution as compared to system codes. - Abstract: This article presents the implementation of a coupling between the 3D neutron kinetic core model DYN3D and the commercial, general purpose computational fluid dynamics (CFD) software ANSYS-CFX. In the coupling approach, parts of the thermal hydraulic calculation are transferred to CFX for its better ability to simulate the three-dimensional coolant redistribution in the reactor core region. The calculation of the heat transfer from the fuel into the coolant remains with DYN3D, which incorporates well tested and validated heat transfer models for rod-type fuel elements. On the CFX side, the core region is modeled based on the porous body approach. The implementation of the code coupling is verified by comparing test case results with reference solutions of the DYN3D standalone version. Test cases cover mini and full core geometries, control rod movement and partial overcooling transients

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

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)

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

Optimization of High-Resolution Continuous Flow Analysis for Transient Climate Signals in Ice Cores

DEFF Research Database (Denmark)

Bigler, Matthias; Svensson, Anders; Kettner, Ernesto

2011-01-01

Over the past two decades, continuous flow analysis (CFA) systems have been refined and widely used to measure aerosol constituents in polar and alpine ice cores in very high-depth resolution. Here we present a newly designed system consisting of sodium, ammonium, dust particles, and electrolytic...... meltwater conductivity detection modules. The system is optimized for high- resolution determination of transient signals in thin layers of deep polar ice cores. Based on standard measurements and by comparing sections of early Holocene and glacial ice from Greenland, we find that the new system features...

Calculations of flow oscillations during reflood using RELAP4/MOD6

International Nuclear Information System (INIS)

Chen, Y.S.; Fischer, S.R.; Sullivan, L.H.

1979-01-01

RELAP4/MOD6 is an analytical computer code which can be used for best-estimate analysis of LWR reactor system blowdown and reflood response to a postulated LOCA. In this study, flow oscillations in the PKL reflood test K5A were investigated using RELAP4/MOD6. Both calculated and measured oscillations exhibited transient characteristics of density-wave and pressure-drop oscillations. The calculated average core mixture level rising rate agrees closely with the test data. Several mechanisms which appear to be responsible for initiation and continuation of calculated or experimental reflood flow oscillations are (a) the coupling between the vapor generation in the core channel and the U-tube geometrical arrangement of a downcomer and a heated core; (b) the inherent low core inlet resistance and the high system outlet resistance; (c) the dependence of heat transfer rate on mass flow rate especially in the dispersed flow ially in the dispersed flow regime; (d) the amount of the liquid entrainment fraction of the heated core channel

Experience with the instrumentation tests in large sodium test facilities

International Nuclear Information System (INIS)

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

1976-01-01

A facility is described for fast breeder core components (AKB) to test specially instrumented fuel dummies and blanket elements, and also absorber elements under simulated normal and extreme reactor conditions. In addition to endurance testing of a special sodium and high temperature sub-assembly, instrumentation is provided to investigate thermohydraulic and vibrational behaviour of core elements. During tests of > 3000 h at temperatures above 820 K the main sub-assembly characteristics, e.g. pressure drop, leakage flow, vibration and noise spectra can be reproduced. The use of eddy current flow meters, strain gauges, magnetostrictive noise sensors, pressure transducers, thermocouples, and acoustic surveillance devices, are described. (U.K.)

Role of wall-attached structures in the interface of the quiescent core region in turbulent pipe flow

Science.gov (United States)

Yang, Jongmin; Hwang, Jinyul; Sung, Hyung Jin

2017-11-01

The effects of low- and high-speed structures on the interface of the quiescent core region are explored using direct numerical simulation data of turbulent pipe flow. The quiescent core region is a uniform momentum zone located at the center of the pipe flow, which contains the highest streamwise momentum with a low level of turbulence. The interface of the quiescent core region can be identified from the probability density function of the streamwise modal velocity. In the vicinity of the interface of the quiescent core region, the streamwise velocity changes abruptly. The abrupt jump in velocity causes an increase of the velocity gradient. The interface of the quiescent core region is similar to the laminar superlayer in turbulent/non-turbulent interface. The interface of the quiescent core region contains the low- and high-speed structures. They can be classified into wall-attached and detached structures depending on the distance between the structures and the wall. The influence of the detached structures accounted for most of the number of detected structures is negligible due to its small volume. Conversely, the wall-attached structures adjacent to the interface have a huge influence on the statistical amount of the interface, such as entrainment characteristics. This work was supported by the Creative Research Initiatives (No. 2017-013369) program of the National Research Foundation of Korea (MSIP).

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...

Testing to expand the rotary-mode core sampling system operating envelope

International Nuclear Information System (INIS)

Witwer, K.S.

1998-01-01

Rotary sampling using the Rotary Mode Core Sampling System (RMCSS) is constrained by what is referred to as the ''Operating Envelope''. The Operating Envelop defines the maximum downward force, maximum rotational speed and minimum purge gas flow allowed during operation of the RMCSS. The original values of 1170 lb. down force, 55 RPM rotational speed, and 30 SCFM nitrogen purge gas were determined during original envelope testing. This envelope was determined by observing the temperature rise on the bitface while drilling into waste simulants. The maximum temperature in single-shell tanks (SSTS) is considered to be approximately 9O C and the critical drill bit temperature, which is the temperature at which an exothermic reaction could be initiated in the tank waste, was previously determined to be 150 C. Thus, the drill bit temperature increase was limited to 60 C. Thermal properties of these simulants approximated typical properties of waste tank saltcake. Later, more detailed envelope testing which used a pumice block simulant, showed a notably higher temperature rise while drilling. This pumice material, which simulated a ''worst case'' foreign object embedded in the waste, has lower thermal conductivity and lower thermal diffusivity than earlier simulants. These properties caused a slower heat transfer in the pumice than in the previous simulants and consequently a higher temperature rise. The maximum downward force was subsequently reduced to 750 lb (at a maximum 55 RPM and minimum 30 SCFM purge gas flow) which was the maximum value at which the drill bit could be operated and still remain below the 60 C temperature rise

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)

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)

Two-Dimensional Cutting (TDC Vitrectome: In Vitro Flow Assessment and Prospective Clinical Study Evaluating Core Vitrectomy Efficiency versus Standard Vitrectome

Directory of Open Access Journals (Sweden)

Mitrofanis Pavlidis

2016-01-01

Full Text Available Purpose. To evaluate comparative aspiration flow performance and also vitrectomy operating time efficiency using a double-cutting open port vitreous cutting system incorporated in a two-dimensional cutting (TDC, DORC International vitrectome design versus standard vitreous cutter. Methods. In vitro investigations compared aspiration flow rates in artificial vitreous humor at varying cutter speeds and vacuum levels using a TDC vitrectome and a standard vitrectome across different aspiration pump systems. A prospective single-centre clinical study evaluated duration of core vitrectomy in 80 patients with macular pucker undergoing 25-gauge or 27-gauge vitrectomy using either a TDC vitrectome at 16,000 cuts per minute (cpm or standard single-cut vitrectome, combined with a Valve Timing intelligence (VTi pump system (EVA, DORC International. Results. Aspiration flow rates remained constant independent of TDC vitrectome cut rate, while flow rates decreased linearly at higher cutter speeds using a classic single-blade vitrectome. Mean duration of core vitrectomy surgeries using a TDC vitreous cutter system was significantly (p<0.001 shorter than the mean duration of core vitrectomy procedures using a single-cut vitrectome of the same diameter (reduction range, 34%–50%. Conclusion. Vitrectomy surgery performed using a TDC vitrectome was faster than core vitrectomy utilizing a standard single-action vitrectome at similar cut speeds.

Experiment data report for semiscale Mod-1 test S-02-3 (blowdown heat transfer test)

International Nuclear Information System (INIS)

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

1975-09-01

Recorded test data are presented for Test S-02-3 of the Semiscale Mod-1 blowdown heat transfer test series. Test S-02-3 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,263 psig. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient with a moderately heated core (75 percent design power level). An electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core. System flow was also set at the 75 percent design level to achieve full core temperature differential. The flow resistance of the intact loop was based on core area scaling. During system depressurization, core power was reduced from the initial level of 1.2 MW in such a manner as to simulate the surface heat flux response of the LOFT nuclear fuel rods until such time that departure from nucleate boiling (DNB) occurs. Blowdown to the pressure suppression system was accomplished without simulated emergency core coolant injection or pressure suppression system coolant spray

Review of coaxial flow gas core nuclear rocket fluid mechanics

International Nuclear Information System (INIS)

Weinstein, H.

1976-01-01

In a prematurely aborted attempt to demonstrate the feasibility of using a gas core nuclear reactor as a rocket engine, NASA initiated a number of studies on the relevant fluid mechanics problems. These studies were carried out at NASA laboratories, universities and industrial research laboratories. Because of the relatively sudden termination of most of this work, a unified overview was never presented which demonstrated the accomplishments of the program and pointed out the areas where additional work was required for a full understanding of the cavity flow. This review attempts to fulfill a part of this need in two important areas

Gas Test Loop Booster Fuel Hydraulic Testing

International Nuclear Information System (INIS)

Gas Test Loop Hydraulic Testing Staff

2006-01-01

The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3

Gas Test Loop Booster Fuel Hydraulic Testing

Energy Technology Data Exchange (ETDEWEB)

Gas Test Loop Hydraulic Testing Staff

2006-09-01

The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3.

Reactor core flow measurements during plant start-up using non-intrusive flow meter CROSSFLOW

Energy Technology Data Exchange (ETDEWEB)

Kanda, V.; Sharp, B.; Gurevich, A., E-mail: vkanda@amag-inc.com, E-mail: bsharp@amag-inc.com, E-mail: agurevich@amag-inc.com [Advanced Measurement & Analysis Group Inc., Ontario (Canada); Gurevich, Y., E-mail: yuri.gurevich@daystartech.ca [Daystar Technologies Inc., Ontario (Canada); Selvaratnarajah, S.; Lopez, A., E-mail: sselvaratnarajah@amag-inc.com, E-mail: alopez@amag-inc.com [Advanced Measurement & Analysis Group Inc., Ontario (Canada)

2013-07-01

For the first time, direct measurements of the total reactor coolant flow and the flow distribution between the inner reactor zone and the outer zone were conducted using the non-intrusive clamp on ultrasonic cross-correlation flow meter, CROSSFLOW, developed and manufactured by Advanced Measurement & Analysis Group Inc. (AMAG). The measurements were performed at Bruce Power A Unit 1 on the Pump Discharge piping of the Primary Heat Transport (PHT) system during start-up. This paper describes installation processes, hydraulic testing, uncertainty analysis and traceability of the measurements to certified standards. (author)

Effect of wettability on scale-up of multiphase flow from core-scale to reservoir fine-grid-scale

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.C.; Mani, V.; Mohanty, K.K. [Univ. of Houston, TX (United States)

1997-08-01

Typical field simulation grid-blocks are internally heterogeneous. The objective of this work is to study how the wettability of the rock affects its scale-up of multiphase flow properties from core-scale to fine-grid reservoir simulation scale ({approximately} 10{prime} x 10{prime} x 5{prime}). Reservoir models need another level of upscaling to coarse-grid simulation scale, which is not addressed here. Heterogeneity is modeled here as a correlated random field parameterized in terms of its variance and two-point variogram. Variogram models of both finite (spherical) and infinite (fractal) correlation length are included as special cases. Local core-scale porosity, permeability, capillary pressure function, relative permeability functions, and initial water saturation are assumed to be correlated. Water injection is simulated and effective flow properties and flow equations are calculated. For strongly water-wet media, capillarity has a stabilizing/homogenizing effect on multiphase flow. For small variance in permeability, and for small correlation length, effective relative permeability can be described by capillary equilibrium models. At higher variance and moderate correlation length, the average flow can be described by a dynamic relative permeability. As the oil wettability increases, the capillary stabilizing effect decreases and the deviation from this average flow increases. For fractal fields with large variance in permeability, effective relative permeability is not adequate in describing the flow.

Study on natural convection in core barrel. Experimental and numerical results for band type spacer pads

International Nuclear Information System (INIS)

Hayashi, Kenji; Kawamata, Nobuhiro; Kamide, Hideki

2003-03-01

In a fast reactor an Inter-Wrapper Flow (IWF) is one of significant phenomena for decay heat removal under natural circulation condition, when a direct reactor auxiliary cooling system (DRACS) is adopted for decay heat removal system. Cold coolant provided by dipped heat exchangers (DHX) of DRACS can penetrate into the core barrel (region between the subassemblies) and it makes natural convection int he core barrel. Such IWF will depend on a spacer pad geometry of subassemblies. Water experiment, TRIF (Test Rig for Inter-wrapper Flow), was carried out for IWF in a reactor core. The test section modeled a 1/12th sector of the core and upper plenum of reactor vessel. Experimental parameters were the spacer pad geometry and flow path geometries connecting the upper plenum and core barrel. Numerical simulation using AQUA code was also performed to confirm applicability of a simulation method. An experimental series using a button type spacer pad had been carried out. Here a band type spacer pad was examined. Temperatures at subassembly wall were measured with parameter of the flow path geometries; one was a connection pipe between the upper plenum and core barrel and the other was flow hole in core former plates between the outermost subassemblies and the core barrel. It was found that these flow paths were effective to remove heat in the core in case of the band type spacer pad. A general purpose three dimensional analysis code, AQUA, was applied to the experimental analysis. Each subassembly and inter wrapper gap region were modeled by slab mesh geometry. Pressure loss coefficient at the pacer pad was set based on the geometry. The numerical simulation results were in good agreement with measured temperature profiles in the core. (author)

Drilling history of core hole DB-15

International Nuclear Information System (INIS)

Diediker, L.D.; Ledgerwood, R.K.

1980-09-01

This core hole was drilled to obtain hydrologic and chemical data on the permeable zones of the Saddle Mountains and Wanapum Formations. These data were obtained by testing the zones that were penetrated during drilling. This testing-as-drilled method reduced the potential problems of interflow and water contamination. This report summarizes the drilling and coring operations; geologic logging, hydrologic testing, and geophysical logging activities; and cementing operations of DB-15 during drilling. The successful completion of DB-15 demonstrated that hydrologic testing can be conducted during core drilling operations. More reliable head measurements and uncontaminated representative water samples from isolated permeable zones, which have not been exposed to potential open borehole cross-flow and head equilibration problems, were benefits derived from the testing-as-drilled method. Disadvantages of the technique were a longer time to complete the borehole caused by time required for testing and increased drilling costs due to rig standby time during testing. Extension of the testing-as-drilled method to the drilling of future core holes is recommended. An evaluation should be made of the required hydrologic data and expected borehole stratigraphy before and during drilling to allow uninterrupted drilling in zones of low permeability that can be tested after drilling is complete

Operability test report for core sample truck number one flammable gas modifications

International Nuclear Information System (INIS)

Akers, J.C.

1997-01-01

This report primarily consists of the original test procedure used for the Operability Testing of the flammable gas modifications to Core Sample Truck No. One. Included are exceptions, resolutions, comments, and test results. This report consists of the original, completed, test procedure used for the Operability Testing of the flammable gas modifications to the Push Mode Core Sample Truck No. 1. Prior to the Acceptance/Operability test the truck No. 1 operations procedure (TO-080-503) was revised to be more consistent with the other core sample truck procedures and to include operational steps/instructions for the SR weather cover pressurization system. A draft copy of the operations procedure was used to perform the Operability Test Procedure (OTP). A Document Acceptance Review Form is included with this report (last page) indicating the draft status of the operations procedure during the OTP. During the OTP 11 test exceptions were encountered. Of these exceptions four were determined to affect Acceptance Criteria as listed in the OTP, Section 4.7 ACCEPTANCE CRITERIA

Core status computing system

International Nuclear Information System (INIS)

Yoshida, Hiroyuki.

1982-01-01

Purpose: To calculate power distribution, flow rate and the like in the reactor core with high accuracy in a BWR type reactor. Constitution: Total flow rate signals, traverse incore probe (TIP) signals as the neutron detector signals, thermal power signals and pressure signals are inputted into a process computer, where the power distribution and the flow rate distribution in the reactor core are calculated. A function generator connected to the process computer calculates the absolute flow rate passing through optional fuel assemblies using, as variables, flow rate signals from the introduction part for fuel assembly flow rate signals, data signals from the introduction part for the geometrical configuration data at the flow rate measuring site of fuel assemblies, total flow rate signals for the reactor core and the signals from the process computer. Numerical values thus obtained are given to the process computer as correction signals to perform correction for the experimental data. (Moriyama, K.)

Test Methodologies for Hydrogen Sensor Performance Assessment: Chamber vs. Flow Through Test Apparatus: Preprint

Energy Technology Data Exchange (ETDEWEB)

Buttner, William J [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Hartmann, Kevin S [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Schmidt, Kara [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Cebolla, Rafeal O [Joint Research Centre, Petten, the Netherlands; Weidner, Eveline [Joint Research Centre, Petten, the Netherlands; Bonato, Christian [Joint Research Centre, Petten, the Netherlands

2017-11-06

Certification of hydrogen sensors to standards often prescribes using large-volume test chambers [1, 2]. However, feedback from stakeholders such as sensor manufacturers and end-users indicate that chamber test methods are often viewed as too slow and expensive for routine assessment. Flow through test methods potentially are an efficient, cost-effective alternative for sensor performance assessment. A large number of sensors can be simultaneously tested, in series or in parallel, with an appropriate flow through test fixture. The recent development of sensors with response times of less than 1s mandates improvements in equipment and methodology to properly capture the performance of this new generation of fast sensors; flow methods are a viable approach for accurate response and recovery time determinations, but there are potential drawbacks. According to ISO 26142 [1], flow through test methods may not properly simulate ambient applications. In chamber test methods, gas transport to the sensor can be dominated by diffusion which is viewed by some users as mimicking deployment in rooms and other confined spaces. Alternatively, in flow through methods, forced flow transports the gas to the sensing element. The advective flow dynamics may induce changes in the sensor behaviour relative to the quasi-quiescent condition that may prevail in chamber test methods. One goal of the current activity in the JRC and NREL sensor laboratories [3, 4] is to develop a validated flow through apparatus and methods for hydrogen sensor performance testing. In addition to minimizing the impact on sensor behaviour induced by differences in flow dynamics, challenges associated with flow through methods include the ability to control environmental parameters (humidity, pressure and temperature) during the test and changes in the test gas composition induced by chemical reactions with upstream sensors. Guidelines on flow through test apparatus design and protocols for the evaluation of

Experimental study of fuel bundle vibrations with rods subjected to mixed axial flow and cross-flow provided by a narrow gap (baffle jetting interaction)

International Nuclear Information System (INIS)

Boulanger, P.; Jacques, Y.; Fardeau, P.; Barbier, D.; Rigaudeau, J.

1997-01-01

The Hydraulic Core Laboratory (LHC) performs experimental studies of PWR fuel assembly mechanical behaviour submitted to representative flows in PWR core. Cross-flows prove particularly troublesome by generating on rods, in special cases, vibratory levels high enough to induce early grid to rod fretting. The fluid-structure interaction under mixed axial and cross-flow is also a major topic for analysis. The authors present a test loop devoted to the mixed axial-cross-flow fluid-structure interaction on representative half-scale mockup which is able to simulate, under ambient conditions, any complex flow (direction and flow rates) representative of PWR core flows. Despite its reduced size, the mockup retains the overall structure of a PWR fuel assembly. Rods displacement/velocity and velocity flow field are measured by laser techniques

Measurement of two-phase flow variables in a BWR by analysis of in-core neutron detector noise signals

International Nuclear Information System (INIS)

Stekelenburg, A.J.C.; Hagen, T.H.J.J. van der

1996-01-01

In this paper, the state of the art of the measurement of two-phase flow variables in a boiling water reactor (BWR) by analysis of in-core neutron detector noise signals is given. It is concluded that the neutronic processes involved in neutron noise are quite well understood, but that little is known about the density fluctuations in two-phase flow which are the main cause of the neutron noise. For this reason, the neutron noise measurements, like the well known two-detector velocity measurements, are still difficult to interpret. By analyzing neutron noise measurements in a natural circulation cooled BWR, it is illustrated that, once a theory on the density fluctuations is developed, two-phase flow can be monitored with a single in-core detector. (author). 70 refs, 4 figs

A fast alternative to core plug tests for optimising injection water salinity for EOR

DEFF Research Database (Denmark)

Hassenkam, Tue; Andersson, Martin Peter; Hilner, Emelie Kristin Margareta

2014-01-01

of the clays which would lead to permanent reservoir damage but evidence of effectiveness at moderate salinity would offer the opportunity to dispose of produced water. The goal is to define boundary conditions so injection water salinity is high enough to prevent reservoir damage and low enough to induce...... the low salinity effect while keeping costs and operational requirements at a minimum. Traditional core plug testing for optimising conditions has some limitations. Each test requires a fresh sample, core testing requires sophisticated and expensive equipment, and reliable core test data requires several...... experiments can be done relatively quickly on very little material, it gives the possibility of testing salinity response on samples from throughout a reservoir and for gathering statistics. Our approach provides a range of data that can be used to screen core plug testing conditions and to provide extra data...

Interfacial friction in low flowrate vertical annular flow

International Nuclear Information System (INIS)

Kelly, J.M.; Freitas, R.L.

1993-01-01

During boil-off and reflood transients in nuclear reactors, the core liquid inventory and inlet flowrate are largely determined by the interfacial friction in the reactor core. For these transients, annular flow occurs at relatively modest liquid flowrates and at the low heat fluxes typical of decay heat conditions. The resulting low vapor Reynolds numbers, are out of the data range used to develop the generally accepted interfacial friction relations for annular flow. In addition, most existing annular flow data comes from air/liquid adiabatic experiments with fully developed flows. By contrast, in a reactor core, the flow is continuously developing along the heated length as the vapor flowrate increases and the flow regimes evolve from bubbly to annular flow. Indeed, the entire annular flow regime may exist only over tens of L/D's. Despite these limitations, many of the advanced reactor safety analysis codes employ the Wallis model for interfacial friction in annular flow. Our analyses of the conditions existing at the end-of-reflood in the PERICLES tests have indicated that the Wallis model seriously underestimates the interfacial shear for low vapor velocity cocurrent upflow. To extend the annular flow data base to diabatic low flowrate conditions, the DADINE tests were re-analyzed. In these tests, both pressure drop and local cross-section averaged void fractions were measured. Thus, both the wall and interfacial shear can be deduced. Based on the results of this analysis, a new correlation is proposed for interfacial friction in annular flow. (authors). 5 figs., 12 refs

Evaluation of aseismic integrity in HTTR core-bottom structure. Pt. 1. Aseismic test for core-bottom structure

International Nuclear Information System (INIS)

Iyoku, T.; Futakawa, M.; Ishihara, M.

1994-01-01

The aseismic tests were carried out using (1)/(5)-scale and (1)/(3)-scale models of the core-bottom structure of the HTTR to quantitatively evaluate the response of acceleration, strain, impact load etc. The following conclusions are obtained. (i) The frequency response of the keyway strain is correlative with that of the impact acceleration on the hot plenum block. (ii) It was confirmed through (1)/(5)-scale and (1)/(3)-scale model tests that the applied similarity law is valid to evaluate the seismic response characteristics of the core-bottom structure. (ii) The stress of graphite components estimated from the scale model test using S 2 -earthquake excitation was sufficiently lower than the allowable stress used as the design criterion. ((orig.))

FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

Energy Technology Data Exchange (ETDEWEB)

Hawkes, Grant L.; Jones, Warren F.; Marcum, Wade; Weiss, Aaron; Howard, Trevor

2017-06-01

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.

Flow visualization on a natural circulation inter-wrapper flow. Experimental and numerical results under a geometric condition of button type spacer pads

Energy Technology Data Exchange (ETDEWEB)

Yasuda, A.; Miyakoshi, H.; Hayashi, K.; Nishimura, M.; Kamide, H.; Hishida, K. [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

1999-04-01

Investigations on the inter-wrapper flow (IWF) in a liquid metal cooled fast breeder reactor core have been carried out. The IWF is a natural circulation flow between wrapper tubes in the core barrel where cold fluid is coming from a direct heat exchanger (DHX) in the upper plenum. It was shown by the sodium experiment using 7-subassembly core model that the IWF can cool the subassemblies. To clarify thermal-hydraulic characteristics of the IWF in the core, the water experiment was performed using the flow visualization technique. The test rig for IWF (TRIF) has the core simulating the fuel subassemblies and radial reflectors. The subassemblies are constructed featuring transparent heater to enable both Joule heating and flow visualization. The transparent heater was made of glass with thin conductor film coating of tin oxide, and the glass heater was embedded on the wall of modeled wrapper tube made of acrylic plexiglass. In the present experiment, influences of peripheral geometric parameters such as flow holes of core formers on the thermal-hydraulic field were investigated with the button type spacer pads of the wrapper tube. Through the water tests, flow patterns of the IWF were revealed and velocity fields were quantitatively measured with a particle image velocimetry (PIV). Also, no substantial influence of peripheral geometry was found on the temperature field of the IWF, as far as the button type spacer pad was applied. Numerical simulation was applied to the experimental analysis of IWF by using multidimensional code with porous body model. The numerical results reproduced the flow patterns within TRIF and agreed well to experimental temperature distributions, showing capability of predicting IWF with porous body model. (author)