Assessment of RELAP/MOD3 using BETHSY 6.2TC 6-inch cold leg side break comparative test

International Nuclear Information System (INIS)

Chung, Young-Jong; Jeong, Jae-Jun; Chang, Won-Pyo; Kim, Dong-Su

1996-10-01

This report presents the results of the RELAP5/MOD3 Version 7j assessment on BETHSY 6.2TC. BETHSY 6.2TC test corresponding to a six inch cold leg break LOCA of the Pressurizer Water Reactor(PWR). The primary objective of the test was to provide reference data of two facilities of different scales (BETHSY and LSTF facility). On the other hand, the present calculation aims at analysis of RELAP5/N4OD3 capability on the small break LOCA simulation, The results of calculation have shown that the RELAP5/MOD3 reasonably predicts occurrences as well as trends of the major phenomena such as primary pressure, timing of loop seal clearing, liquid hold up, etc. However, some disagreements also have been found in the predictions of loop seal clearing, collapsed core water level after loop seal clearing, and accumulator injection behaviors. For better understanding of discrepancies in same predictions, several sensitivity calculations have been performed as well. These include the changes of two-phase discharge coefficient at the break junction and some corrections of the interphase drag term. As result, change of a single parameter has not improved the overall predictions and it has been found that the interphase drag model has still large uncertainties

ATLAS Cold Leg Top Slot Break Analysis using RELAP5

Energy Technology Data Exchange (ETDEWEB)

Sung, Haejung; Lee, Sang Ik; Park, Ju-Hyun; Choi, Tong-Soo [KEPCO NF, Daejeon (Korea, Republic of)

2016-10-15

U.S. Nuclear Regulatory Commission (US-NRC) has been reviewing the design certification application for APR1400 submitted by Korea Electric Power Corporation (KEPCO). The main concern about cold leg top slot break is that cladding temperature might be increased by core uncover due to four loop seal reformation following flooding of safety injection water. An integral effect test for cold leg top slot break was performed by KAERI (Korea Atomic Energy Research Institute) using ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation), which is a scaled down experimental facility for APR1400. In this study, RELAP5/MOD3.3/Patch04 is assessed by experimental result of ATLAS cold leg top slot break. Also, thermal hydraulic phenomena by four loop seals reformation is observed by RELAP5 result. The RELAP5/MOD3.3/Patch04 is assessed by the experimental result of ATLAS cold leg top slot break. The top slot break is described by offtake model, and the mass flow rate is fairly well estimated. The RELAP5 well predicts the correlation between general trend and four loop seal reformation. The pressure of the core region and the cladding temperature tends to increase during four loop seal reformation due to steam path blockage on four loop seals. It is presumed that the code cannot estimate two phase phenomena by loop seal clearing as same as experiments. In terms of cladding temperature, loop seal reformation due to loop seal elevation of APR1400 does not need to be the issue, since the void fraction at the active top core is maintained over 0.4.

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

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

2010-02-01

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

A study on effective system depressurization during a PWR vessel bottom break LOCA with HPI failure and gas inflow prevention. ROSA-V/LSTF test SB-PV-05

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

2006-11-01

A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-05, simulated a PWR vessel bottom SBLOCA with a rupture of nine instrument tubes, which is equivalent to 0.18% cold leg break. It is clarified that AM actions with steam generator (SG) depressurization to achieve a primary loop cooling rate at -55 K/h and auxiliary feedwater supply for 30 minutes are effective to avoid core uncovery by actuating the low pressure injection (LPI) system. It is also shown through the comparison with the previous experiment of SB-PV-03 that prevention of non-condensable gas inflow from the accumulator injection system (AIS) is very important to actuate the LPI to achieve adequate core cooling. This report presents experiment results of SB-PV-05 in detail and shows the effects of gas inflow prevention on core cooling through the estimation of primary coolant mass and energy balance in the primary system. (author)

Summary of ROSA-4 LSTF first phase test program and station blackout (TMLB) test results

International Nuclear Information System (INIS)

Tasaka, K.; Kukita, Y.; Anoda, Y.

1990-01-01

This paper summarizes major test results obtained at the ROSA-4 Large Scale Test Facility (LSTF) during the first phase of the test program. The results from a station blackout (TMLB) test conducted at the end of the first-phase program are described in some detail. The LSTF is an integral test facility being operated by the Japan Atomic Energy Research Institute for simulation of pressurized water reactor (PWR) thermal-hydraulic responses during small-break loss-of-coolant accidents (SBLOCAs) and operational/abnormal transients. It is a 1/48 volumetrically scaled, full-height, full-pressure simulator of a Westinghouse-type 4-loop PWR. The facility includes two symmetric primary loops each one containing an active inverted-U tube steam generator and an active reactor coolant pump. The loop horizontal legs are sized to conserve the scaled (1/24) volumes as well as the length to the square root of the diameter ratio in order to simulate the two-phase flow regime transitions. The primary objective of the LSTF first-phase program was to define the fundamental PWR thermal-hydraulic responses during SBLOCAs and transients. Most of the tests were conducted with simulated component/operator failures, including unavailability of the high pressure injection system and auxiliary feedwater system, as well as operator failure to take corrective actions. The forty-two first phase tests included twenty-nine SBLOCA tests conducted mainly for cold leg breaks, three abnormal transient tests and ten natural circulation tests. Attempts were made in several of the SBLOCA tests to simulate the plant recovery procedures as well as candidate accident management measures for prevention of high-pressure core melt situation. The natural circulation tests simulated the single-phase and two-phase natural circulation as well as reflux condensation behavior in the primary loops in steady or quasi-steady states

ROSA/LSTF experiment report for RUN SB-CL-24 repeated core heatup phenomena during 0.5% cold leg break LOCA

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Mitsuhiro; Anoda, Yoshinari [Department of Reactor Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

2000-03-01

A small break loss-of-coolant accident (SBLOCA) in a Westinghouse-type four-loop PWR was simulated in an experiment (SB-CL-24) conducted at the Large-Scale Test Facility (LSTF) with an intention to study repeated core heatup during a long-term cooldown process. The experiment was conducted on February 28, 1990 with specified test conditions including failure assumptions both on the high pressure injection (HPI) and the auxiliary feedwater systems, and the intentional secondary system depressurization as an operator action. The secondary depressurization contributed to promote the primary depressurization and the actuation of accumulator injection system (AIS). A temporary core heatup was observed in each of three loopseal clearing (LSC) processes. A significant core heatup occurred in the following boil-off process after loss of the secondary coolant mass and the AIS termination due to increase of the primary pressure. By additional opening of the pressurizer relief valves and safety valves, the primary pressure rapidly decreased to result in the low pressure injection (LPI) which cooled the heated core. This report summarizes results of the experiment (SB-CL-24) in addition to typical responses of some accident indication systems including the core exit thermocouples (CETs) and the water level meters in the primary system. (author)

The Numerical Sensitivity Study of Cold Leg Top Slot Break for ATLAS

Energy Technology Data Exchange (ETDEWEB)

Sung, Hae Jung; Lee, Sang Ik; Park, Ju hyun; Choi, Tong Soo [KEPCO, Daejeon (Korea, Republic of)

2016-05-15

In this study, cold leg top slot break calculations are performed by RELAP5/MOD3.3 Patch04 for the ATLAS test which is the scaled down experimental facility for the APR1400. The test condition for base case is selected as 0.1016 m (4 in.) based on the break size of the APR1400. In addition, sensitivity studies about break size, break distance from vessel, and pressurizer location are performed until the initiation of simultaneous injection of 2.5 hours. The loop seal reformation occurs early, and the duration of final loop seal reformation is longer as the break is close to vessel. Nonetheless, PCT increased by loop seal reformation is not identified since core uncovery does not occur. In this study, it is confirmed through RELAP5 simulation of the ATLAS test that cold leg top slot break for the APR1400 is not a safety issue in perspective of the loop seal reformation.

Data report for ROSA-IV LSTF gravity-driven safety injection experiment run SB-CL-27

International Nuclear Information System (INIS)

Yonomoto, Taisuke; Saitou, Seishi; Kuroda, Takeshi

1994-03-01

Experimental data are presented for the passive injection test, Run SB-CL-27, conducted at the ROSA-IV Large Scale Test Facility (LSTF) on September 17, 1992. This experiment simulated thermal-hydraulic behavior of a gravity-driven, passive safety injection system during a small-break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). The injection system consisted of a gravity-driven injection tank, located above the reactor vessel, with connecting lines. The tank was initially filled with water of room temperature at the same pressure as the pressurizer. The connecting lines to the cold leg and to the vessel downcomer were opened at the test initiation. Then, a natural circulation flow developed in the loop which was formed by these lines and the injection tank. The hot water in the cold leg circulated into the upper part of tank and accumulated there causing a significant thermal stratification. This thermal stratification prevented direct-contact condensation of steam from occurring during the subsequent tank drain-down phase. Therefore, no condensation-induced depressurization of the tank, affecting adversely the injection performance, occurred. (author)

Analysis of the ATLAS Cold Leg Top-Slot Break Experiment Using the MARS Code

Energy Technology Data Exchange (ETDEWEB)

Ha, T. W.; Jeong, J. J. [Pusan National University, Busan (Korea, Republic of)

2016-10-15

During a small-break loss of coolant accident (SBLOCA) or intermediate-break loss of coolant accident (IBLOCA) in a PWR, such as the APR1400, the steam volume in the reactor vessel upper plenum may continue to expand until the liquid phase in the horizontal intermediate legs is released, called loop seal clearing (LSC), due to the increase of the pressure in the upper plenum. A domestic standard problem (DSP) exercise using the ATLAS facility was promoted in order to transfer the database to domestic nuclear industries. For 4th DSP (DSP-04), the ATLAS cold leg top-slot break experiment was postulated. For the DSP-04, main concerns are to predict the LSC and LSR having a significantly effect on the behavior of the system under long term cooling. In this study, we simulated the ATLAS cold leg top-slot break experiment using the MARS code and the predicted LSC and LSR are compared to experimental results. The LTC-CL-04R was simulated using the MARS code. Most of the predicted results agree well with the experimental data. However, the timing of LSC and LSR is slightly different from each other and, thus, the behavior of the primary system is slightly different. The core heat up was not observed in the experiment and the calculation.

Counter-current flow limitation at hot leg pipe during reflux condensation cooling after small-break LOCA

International Nuclear Information System (INIS)

Jeong, Hae Yong; Ha, Sang Jun; Jo, Yung Jo; Jun, Hwang Yong

1999-01-01

The possibility of hot leg flooding is evaluated in case of a small-break loss-of-coolant accident in Korean Next Generation Reactor (KNGR) operating at the core power of 3983 MW normally. The vapor and liquid velocities in hot leg and steam generator tubes are calculated during reflux condensation cooling with the accident scenarios of three typical break sizes, 0.13 %, 1.02 % and 10.19 % cold leg break. The calculated results are compared with the existing flooding correlations. It is predicted that the hot leg flooding is excluded when two steam generators are available. It is also shown that the possibility of hot leg flooding under the operation with one steam generator is very low. Therefore, it can be said that the occurrence of hot leg flooding is unexpected when the reflux condensation cooling is maintained in steam generator tubes

Prediction of Counter-Current Flow Limitation at Hot Leg Pipe During a Small-Break Loca

Energy Technology Data Exchange (ETDEWEB)

Jeong, H.Y. [Korea Electric Power Research Institute, Taejeon (Korea)

2001-07-01

The possibility of hot leg flooding during reflux condensation cooling after a small-break loss-of-coolant accident in a nuclear power plant is evaluated. The vapor and liquid velocities in hot leg and steam generator tubes are calculated during reflux condensation cooling with the accident scenarios of three typical break sizes, 0.13 %, 1.02 % and 10.19 % cold leg break. The effect of initial water level to counter-current flow limitation is taken into account. It is predicted that the hot leg flooding is precluded when all steam generators are available for heat removal. It is also shown the both hot leg flooding and SG flooding are possible under the operation of one steam generators. Therefore, it can be said that the occurrence of hot leg flooding under reflux condensation cooling is possible when the number of steam generators available for heat removal is limited. (author). 15 refs., 15 figs., 3 tabs.

Comparision of calculations for the ROSA-IV LSTF with RELAP5/MOD0 and RELAP5/MOD1 (cycle 1)

International Nuclear Information System (INIS)

Fineman, C.P.; Tanaka, Mitsugu; Tasaka, Kanji

1982-03-01

10% and 2.5% cold leg break analyses have been completed for the ROSA-IV Large Scale Test Facility (LSTF) with the RELAP5/MOD0 and RELAP5/MOD1, cycle 1, computer codes. Comparisons between the calculations were made to determine any differences in the results obtained from the two versions of RELAP5. Differences in the two calculations were found which can be attributed to changes in the flow regime maps and critical flow model. (author)

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

Analysis of ATLAS 6-inch cold leg break simulation with MARS-KS code

Energy Technology Data Exchange (ETDEWEB)

Kim, Se Yun; Jun, Hwang Yong; Ha, Sang Jun [Korea Electric Power Company, Daejeon (Korea, Republic of)

2011-05-15

A Domestic Standard Problem (DSP) exercise using ATLAS facility has been organized by KAERI. As the second DSP exercise, the 6-inch cold leg bottom break was determined. This experiment is the counterpart test to the DVI line break to verify the safety performance of the DVI method over the traditional CLI method. Compared with the large break LOCA, the phases of the small break LOCA prior to core recovery occur over a long period. The blowdown, natural circulation, loop seal clearance, boil-off, and core recovery phase should be investigated minutely with relevant models of safety analysis codes in order to predict these thermal hydraulic phenomena correctly. To investigate the ECC bypass phenomena, a finer study on the thermalhydraulic behavior in upper annulus downcomer was carried out

Loss of Coolant Accident Simulation for the Top-Slot break at Cold Leg Focusing on the Loop Seal Reformation under Long Term Cooling with the ATLAS

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Rok; Park, Yu Sun; Bae, Byoung Uhn; Choi, Nam Hyun; Kang, Kyoung Ho; Choi, Ki Yong [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In the present paper, loss of coolant accident for the top-slot break at cold leg was simulated with the ATLAS, which is a thermal-hydraulic integral effect test facility for evolutionary pressurized water reactors (PWRs) of an advanced power reactor of 1400 MWe (APR1400). The simulation was focused on the loop seal reformation under long term cooling condition. During a certain class of Loss of Coolant Accident (LOCA) in a PWR like an advanced power reactor of 1400 MWe (APR1400), the steam volume in the reactor vessel upper plenum and/or upper head may continue expanding until steam blows liquid out of the intermediate leg (U-shaped pump suction cold leg), called loop seal clearing (LSC), opening a path for the steam to be relieved from the break. Prediction of the LSC phenomena is difficult because they are varies for many parameters, which are break location, type, size, etc. This LSC is the major factor that affects the coolant inventory in the small break LOCA (SBLOCA) or intermediate break LOCA (IBLOCA). There is an issue about the loop seal reformation that liquid refills intermediate leg and blocks the steam path after LSC. During the SBLOCA or IBLOCA, the Emergency Core Cooling System (ECCS) is operated. For long term of the top slot small or intermediate break at cold leg, the primary steam condensation by SG heat transfer or SIP, SIT water flooding (reverse flow to loop seal) make loop seal reformation possibly. The primary pressure increase at the top core region due to the steam release blockage by loop seal reformation. And then core level decreases and partial core uncover may occur. The loss of coolant accident for the top-slot break at cold leg was simulated with the ATLAS. The loop seal clearing and loop seal reformation were occurred repeatedly.

Assessment of the MARS-KS Code Using Atlas 6-inch cold leg Break Test

Energy Technology Data Exchange (ETDEWEB)

Kang, D. G.; Kim, J. S.; Ahn, S. H.; Seul, K. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2012-03-15

An integral effect test on the SBLOCA (Small-Break Loss of Coolant Accident) aiming at 6-inch cold leg bottom break, SB-CL-09, was conducted with the Atlas on November, 13, 2009, by KAERI. In this study, the calculations using MARS-KS Vt1.2 code were conducted for 6-inch cold leg break test of Atlas (SB-CL-09) which is the second domestic standard problem (Dsp-02) to assess MARS-KS code capability to simulate the transient thermal-hydraulic behavior for SBLOCA. The steady state was determined by conducting a null transient calculation and the errors between the calculated and measured values are acceptable for almost primary/secondary system parameters. The predicted pressurizer pressure agrees relatively well with the experimental data and the predicted break flow and mass are in good agreement with experiment. In MARS-KS calculation, the decrease of core collapsed water level is predicted well in blowdown phase, but just before LSC, water level is higher than experiment. However, the sudden decrease and increase of water level is higher than experiment. However, the sudden decrease and increase of water level at the LSC are predicted qualitatively. After LSC, there is another water level dip at Sit injection time which is not in experiment. It is considered that this phenomenon is caused by rapid depressurization of downcomer due to significant condensation rate of vapor in downcomer when Sit water flows in it. For the downcomer water level is predicted well, however, it is significantly over-predicted at SIT injection time, water level is predicted well, however, it is significantly over-predicted at SIT injection time after SIT water flows in downcomer. Predicted cladding temperature generally agrees well with the experiment, while there is peak at SIT injection time in calculation which is not in experiment. The loop seals of 1A, 2B intermediate leg are cleared around 400 seconds in experiment, while only that of 1A is cleared in MARS-KS calculation at the

An experimental study on effective depressurization actions for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V test SB-PV-04)

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

2006-03-01

A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westing-house-type PWR (3423 MWt). The experiment, SB-PV-04, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. It is clarified that AM actions with steam generator (SG) rapid depressurization by fully opening relief valves and auxiliary feedwater supply are effective to avoid core uncovery by actuating the low pressure injection (LPI) system though the primary depressurization is degraded by non-condensable gas inflow to the primary loops from the accumulator injection system. The effective core cooling was established by the rapid depressurization which contributed to preserve larger primary coolant mass than in the previous experiment (SB-PV-03) which was conducted with smaller primary cooling rate of -55 K/h as AM actions. (author)

Code Assessment of SPACE 2.19 using LSTF 10% Main Steam-Line-Break Test

Energy Technology Data Exchange (ETDEWEB)

Kim, Minhee; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) has been developed in recent years by the Korea Hydro and Nuclear Power Co. through collaborative works with other Korean nuclear industries and research institutes. As a result of the development, the 2.19 version of the code was released through the successive various verification and validation works. In this study, results produced by the SPACE 2.19 code were compared with the experimental data from JAERI's LSTF Test Run SBSL- 01 for a 10% main steam line break transient in a pressurized water reactor. The LSTF 10% main steam line break test were simulated using the SPACE 2.19 for code V and V work. The overall comparisons between the SPACE 2.19 code prediction and the LSTF Test Run SB-SL-01 experimental data are reasonably satisfactory. The comparisons were conducted in terms of the variations of mass flow rate, void fraction, pressure, collapsed liquid level, temperature, and system flow rate for the transient. In addition, the input model was modified for simulation accuracy of PZR pressure based on the calculated results. The correction of PORV setpoint affects to simulate the PORV open and close phenomena similarly with experiments. From the modification, the computed results show a reasonable agreement with experimental data in overall transient time.

Leak-before-break due to fatigue cracks in the cold leg piping system

International Nuclear Information System (INIS)

Mayfield, M.E.; Collier, R.P.

1984-01-01

This review paper presents the results of a deterministic assessment of the margin of safety against a large break in the cold leg piping system of pressurized water reactors. The paper focuses on the computation of leak rates resulting from fatigue cracks that penetrate the full wall thickness. Results are presented that illustrate the sensitivity of the leak rate to stress level, crack shape and crack orientation. Further, the leak rates for specific conditions are contrasted to detection levels, shutdown criteria, make-up capacity and the leak rate associated with final failure of the piping system. The results of these computations indicate that, in general, leaks far in excess of the present detection sensitivities would result at crack sizes well below the critical crack sizes for the upset loadings on the cold leg piping system

Post-test analysis of the ROSA/LSTF and PKL counterpart test

Energy Technology Data Exchange (ETDEWEB)

Carlos, S., E-mail: scarlos@iqn.upv.es [Departament d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera, 14, València Spain (Spain); Querol, A., E-mail: anquevi@iqn.upv.es [Departament d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera, 14, València Spain (Spain); Instituto de Seguridad Industrial, Radiofísica y Medioambiental, Universitat Politècnica de València, Camí de Vera, 14, València (Spain); Gallardo, S., E-mail: sergalbe@iqn.upv.es [Departament d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera, 14, València Spain (Spain); Instituto de Seguridad Industrial, Radiofísica y Medioambiental, Universitat Politècnica de València, Camí de Vera, 14, València (Spain); Sanchez-Saez, F., E-mail: frasansa@etsii.upv.es [Departament d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera, 14, València Spain (Spain); and others

2016-02-15

Highlights: • TRACE modelization for PKL and ROSA/LSTF installations. • Secondary-side depressurization as accident management action. • CET vs PCT relation. • Analysis of differences in the vessel models. - Abstract: Experimental facilities are scaled models of commercial nuclear power plants, and are of great importance to improve nuclear power plants safety. Thus, the results obtained in the experiments undertaken in such facilities are essential to develop and improve the models implemented in the thermal-hydraulic codes, which are used in safety analysis. The experiments and inter-comparisons of the simulated results are usually performed in the frame of international programmes in which different groups of several countries simulate the behaviour of the plant under the accidental conditions established, using different codes and models. The results obtained are compared and studied to improve the knowledge on codes performance and nuclear safety. Thus, the Nuclear Energy Agency (NEA), in the nuclear safety work area, auspices several programmes which involve experiments in different experimental facilities. Among the experiments proposed in NEA programmes, one on them consisted of performing a counterpart test between ROSA/LSTF and PKL facilities, with the main objective of determining the effectiveness of late accident management actions in a small break loss of coolant accident (SBLOCA). This study was proposed as a result of the conclusion obtained by the NEA Working Group on the Analysis and Management of Accidents, which analyzed different installations and observed differences in the measurements of core exit temperature (CET) and maximum peak cladding temperature (PCT). In particular, the transient consists of a small break loss of coolant accident (SBLOCA) in a hot leg with additional failure of safety systems but with accident management measures (AM), consisting of a fast secondary-side depressurization, activated by the CET. The paper

Post-test analysis of the ROSA/LSTF and PKL counterpart test

International Nuclear Information System (INIS)

Carlos, S.; Querol, A.; Gallardo, S.; Sanchez-Saez, F.

2016-01-01

Highlights: • TRACE modelization for PKL and ROSA/LSTF installations. • Secondary-side depressurization as accident management action. • CET vs PCT relation. • Analysis of differences in the vessel models. - Abstract: Experimental facilities are scaled models of commercial nuclear power plants, and are of great importance to improve nuclear power plants safety. Thus, the results obtained in the experiments undertaken in such facilities are essential to develop and improve the models implemented in the thermal-hydraulic codes, which are used in safety analysis. The experiments and inter-comparisons of the simulated results are usually performed in the frame of international programmes in which different groups of several countries simulate the behaviour of the plant under the accidental conditions established, using different codes and models. The results obtained are compared and studied to improve the knowledge on codes performance and nuclear safety. Thus, the Nuclear Energy Agency (NEA), in the nuclear safety work area, auspices several programmes which involve experiments in different experimental facilities. Among the experiments proposed in NEA programmes, one on them consisted of performing a counterpart test between ROSA/LSTF and PKL facilities, with the main objective of determining the effectiveness of late accident management actions in a small break loss of coolant accident (SBLOCA). This study was proposed as a result of the conclusion obtained by the NEA Working Group on the Analysis and Management of Accidents, which analyzed different installations and observed differences in the measurements of core exit temperature (CET) and maximum peak cladding temperature (PCT). In particular, the transient consists of a small break loss of coolant accident (SBLOCA) in a hot leg with additional failure of safety systems but with accident management measures (AM), consisting of a fast secondary-side depressurization, activated by the CET. The paper

NPP Krsko small break LOCA analysis

International Nuclear Information System (INIS)

Mavko, B.; Petelin, S.; Peterlin, G.

1987-01-01

Parametric analysis of small break loss of coolant accident for the Krsko NPP was calculated by using RELAP5/MOD1 computer code. The model that was used in our calculations has been improved over several years and was previously tested in simulation (s) of start-up tests and known NPP Krsko transients. In our calculations we modelled automatic actions initiated by control, safety and protection systems. We also modelled the required operator actions as specified in emergency operating instructions. In small-break LOCA calculations, we varied break sizes in the cold leg. The influence of steam generator tube plugging on small break LOCA accidents was also analysed. (author)

Best estimate small break LOCA analysis for KNGR SIS optimization

International Nuclear Information System (INIS)

Song, JIn Ho; Lim, Hong Sik; Bae, Kyoo Hwan; Lee, Joon

1996-01-01

The KNGR has an advanced ECCS design feature which employs four mechanically-separated SI trains where each train consisting of one HPSI pump and one SIT injects ECC water directly into the reactor vessel downcomer annulus. To demonstrate that the KNGR ECCS design features meet the EPRI ALWR requirements of no core uncovery for a break of up to 6 inch diameter, small break LOCA cases with various break sizes were analyzed using a best-estimate analytical procedure. Two kinds of break locations are considered: cold leg and DVI line breaks. It was observed that the KNGR ECCS design can tolerate a cold leg break of up to 10 inches with no core uncovery. However, since DVI line break with 6 inch diameter undergoes slight core uncovery, further investigation is required for KNGR SIS optimization

An analysis methodology for hot leg break mass and energy release

International Nuclear Information System (INIS)

Song, Jin Ho; Kwon, Young Min; Kim, Taek Mo; Chung, Hae Yong; Lee, Sang Jong

1996-07-01

An analysis methodology for the hot leg break mass and energy release is developed. For the blowdown period a modified CEFLASH-4A analysis is suggested. For the post-blowdown period a new computer model named COMET is developed. Differently from previous post-blowdown analysis model FLOOD3, COMET is capable of analyzing both cold leg and hot leg break cases. The cold leg break model is essentially same as that of FLOOD3 with some improvements. The analysis results by the newly proposed hot leg break model in the COMET is in the same trend as those observed in scaled-down integral experiment. And the analyses results for the UCN 3 and 4 by COMET are qualitatively and quantitatively in good agreement with those predicted by best-estimate analysis by using RELAP5/MOD3. Therefore, the COMET code is validated and can be used for the licensing analysis. 6 tabs., 82 figs., 9 refs. (Author)

The 7.4 per cent cold leg break without accumulator operation

International Nuclear Information System (INIS)

Perneczky, L.; Toth, I.; Szabados, L.; Ezsoel, Gy.

1986-12-01

A simulation technique for the loss-of-coolant failure analysis of light-water-cooled nuclear reactor is described. It has been used to analyze transient processes during a hypothetical accident and to estimate the effectiveness of built-in safety systems. The model PMK-NHV was established for these types of simulation in the Paks Nuclear Power Plant, Hungary. The first test on this simulation facility is described: a 7.4 per cent cold leg break from full power covering the blowdown phase of the accident. The pre-test analysis using the RELAP4/mod6 computer code, the evaluation of the measured data, the interpretation of the test results and the post-test calculations are presented. The work was performed within the IAEA Standard Problem Exersice (SPE). (R.P.)

Comparison of a TRAC calculation to the data from LSTF run SB-CL-05

International Nuclear Information System (INIS)

Motley, F.; Schultz, R.

1986-01-01

Run SB-CL-05 is a 5% break in the side of the cold leg. The test results show that the core was uncovered briefly and that the rods overheated at certain core locations. Liquid holdup on the upflow side of the steam generator tubes was observed. When the loop seal cleared, the core refilled and the rods cooled. The TRAC results are in reasonable agreement with the test data, meaning that TRAC correctly predicted the major trends and phenomena. TRAC predicted the core uncovery, the resulting rod heatup, and the liquid holdup on the upflow side of the steam generator tubes correctly. The clearing of the loop seal allowed core recovery and cooled the overheated rods just as it had in the data, but TRAC predicted its occurrence 20 s late. The experimental and TRAC analysis results of run SB-CL-05 are similar to those for Semiscale Run S-UT-8. In both runs there was core uncovery, rod overheating, and steam generator liquid holdup. These results confirm scaling of these phenomena from Semiscale (1/1650) to LSTF (1/48)

Experiment and analyses on intentional secondary-side depressurization during PWR small break LOCA. Effects of depressurization rate and break area on core liquid level behavior

International Nuclear Information System (INIS)

Asaka, Hideaki; Ohtsu, Iwao; Anoda, Yoshinari; Kukita, Yutaka

1997-01-01

The effects of the secondary-side depressurization rate and break area on the core liquid level behavior during a PWR small-break LOCA were studied using experimental data from the Large Scale Test Facility (LSTF) and by using analysis results obtained with a JAERI modified version of RELAP5/MOD3 code. The LSTF is a 1/ 48 volumetrically scaled full-height integral model of a Westinghouse-type PWR. The code reproduced the thermal-hydraulic responses, observed in the experiment, for important parameters such as the primary and secondary side pressures and core liquid level behavior. The sensitivity of the core minimum liquid level to the depressurization rate and break area was studied by using the code assessed above. It was found that the core liquid level took a local minimum value for a given break area as a function of secondary side depressurization rate. Further efforts are, however, needed to quantitatively define the maximum core temperature as a function of break area and depressurization rate. (author)

Supplemental description of ROSA-IV/LSTF with No.1 simulated fuel-rod assembly

International Nuclear Information System (INIS)

1989-09-01

Forty-two integral simulation tests of PWR small break LOCA (loss-of-coolant accident) and transient were conducted at the ROSA-IV Large-Scale Test Facility (LSTF) with the No.1 simulated fuel-rod assembly between March 1985 and August 1988. Described in the report are supplemental information on modifications of the system hardware and measuring systems, results of system characteristics tests including the initial fluid mass inventory and heat loss distribution for the primary system, and thermal properties for the heater rod materials. These are necessary to establish the correct boundary conditions of each LSTF experiment with the No.1 core assembly in addition to the system data given in the system description report (JAERI-M 84-237). (author)

Simulation of a postulated 2% cold leg break in Angra 2 nuclear power plant

International Nuclear Information System (INIS)

Palmieiri, Elcio Tadeu; Azevedo, Carlos Vicente Goulart de; Aronne, Ivan Dionysio

2007-01-01

This paper presents the simulation of a 2% break in the cold leg pipe of Angra 2 nuclear power plant, with the computer code RELAP5/Mod3.3. The main boundary conditions specified for this simulation were: no injection from high pressure injection system; enhanced depressurization of the primary system by opening the pressure operated relief valve (PORV) and the safety relief valve (SRV) when core temperature reaches circa 100 K above saturation; and accumulator injection starting at 2.7 MPa. The specific objectives to be addressed with this simulation are: the core boil-off and dryout at relatively high pressure in the primary system; the phenomena during enhanced primary depressurization; the effectiveness of hot leg accumulator injection into the partially uncovered rod bundle; and the core rewetting. The results obtained were compared with the Lobi A1-93 test, which was performed under the same boundary conditions. This activity was executed in the scope of IAEA research project Evaluation of Uncertainties in the Simulation of Accidents in Angra 2 using RELAP5/MOD3 Code Applying CIAU Methodology (author)

Investigation of small break loss-of-coolant phenomena in a small scale nonnuclear test facility

International Nuclear Information System (INIS)

Cozzuol, J.M.; Fauble, T.J.; Harvego, E.A.

1980-01-01

A small-scale nonnuclear integral test facility designed to simulate a pressurized water reactor (PWR) system was used to evaluate the effects of a small break loss-of-coolant accident (LOCA) on the system thermal-hydraulic response. The experiment approximated a 2.5% (11-cm diameter) communicative break in the cold leg of a PWR, and included initial conditions which were similar to conditions in a PWR operating at full power. The 2.5% break size ensured that the nominal break flow rate was greater than the high pressure injection system (HPIS) flow rate, thus providing the potential for a continuous system depressurization. The sequence of events was similar to that used in evaluation model analysis of small break loss-of-coolant accidents, and included simulated reactor scram and loss of offsite power. Comparisions of experimental data with computer code calculations are used to demonstrate the capabilities and limitations of integral system calculations used to predict phenomena which can be important in the assessment of a small break LOCA in a PWR

Simulation of the SPE-4 small-break loss-of-coolant accident

International Nuclear Information System (INIS)

Cebull, P.; Hassan, Y.A.

1993-01-01

A small-break loss of coolant accident (SBLOCA) conducted at the PMK-2 integral test facility was analyzed using RELAP5/MOD3. 1. The experiment simulated a 7.4% break in the cold leg of a VVER-440/213-type nuclear power plant as part of the International Atomic Energy Agency's Fourth Standard Problem Exercise (SPE-4). The VVER design differs from pressurized water reactors (PWRS) of western origin, primarily in its use of horizontal steam generators, hot- and cold-leg loop seals, and safety injection tanks. Because of these differences, it will exhibit somewhat different transient behavior than most PWRS. The PMK-2 test facility, located at the KFKI Atomic Energy Research Institute (AEKI), is a scale model of the Paks nuclear power plant in Hungary with scaling factors of 1:2070 in power and volume and 1:1 in elevation. Primarily used to study SBLOCAs and natural circulation behavior of VVER reactors, it has been used in three previous SPEs

Evaluation of a postulated loss of coolant accident (LOCA) due to a 160 cm2 break in a cold leg of Angra 2 nuclear power plant

International Nuclear Information System (INIS)

Azevedo, Carlos Vicente Goulart de; Palmieri, Elcio Tadeu; Aronne, Ivan Dionysio

2002-01-01

The development of a qualified full nodalization of Angra2 NPP for RELAP5/Mod 3.2.2 gamma, aiming at the evaluation of a comprehensive number of accidents and transients, thus providing suitable safety analysis support for licensing purposes, is being carried out within the framework of CNEN internal technical cooperation, involving some of its institutes (CDTN, IPEN and IEN) and the Reactors Coordination (CODRE). This work presents a simulation of a postulated Angra2 small cold leg break loss of coolant accident (SBLOCA). A 160 cm 2 break is supposed to occur at one cold leg between the main coolant pump and the reactor vessel and is described in the Angra2 Final Safety Analysis Report, section 15.6.4.1.3.4. The simulation of several types of transients and accidents is necessary to verify the adequate performance of the modeled logic and systems. In general, the analysis of such and accident allows to demonstrate the safety Injection System performance and the reliable transition between the high pressure safety injection, the accumulator injection and the residual heat removal phases. Furthermore, it is assumed that some components are out of service due to fail or repair in order to make a conservative analysis. The results showed a compatible behavior of the molded systems and that the simulated Emergency Core Cooling System was able to provide sufficient cooling to avoid any damage to the core. (author)

Prediction of loop seal formation and clearing during small break loss of coolant accident

International Nuclear Information System (INIS)

Lee, Suk Ho; Kim, Hho Jung

1992-01-01

Behavior of loop seal formation and clearing during small break loss of coolant accident is investigated using the RELAP5/MOD2 and /MOD3 codes with the test of SB-CL-18 of the LSTF(Large Scale Test Facility). The present study examines the thermal-hydraulic mechanisms responsible for early core uncovery includeing the manometric effect due to an asymmetric coolant holdup in the steam generator upflow and downflow side. The analysis with the RELAP5/ MOD2 demonstrates the main phenomena occuring in the depressurization transient including the loop seal formation and clearing with sufficient accuracy. Nevertheless, several differences regarding the evolution of phenomena and their timing have been pointed out in the base calculations. The RELAP5/MOD3 predicts overall phenomena, particularly the steam generator liquid holdup better than the RELAP5/MOD2. The nodalization study in the components of the steam generator U-tubes and the cross-over legs with the RELAP5/MOD3 results in good prediction of the loop seal clearing phenomena and their timing. (Author)

Effects of the reactor coolant pumps following a small break in a Westinghouse PWR

International Nuclear Information System (INIS)

Koenig, J.E.

1983-10-01

Numerical simulations of the thermal-hydraulic events following a small cold-leg break in a Westinghouse pressurized water reactor were performed to address the pumps-on/off issue. The mode of pump operation was varied in each calculation to ascertain the optimum mode. It was found that pump operation was not critical for this break size and location because the fuel rods remained cool in all accidents analyzed. In terms of system mass, however, it was preferable to leave the pumps in operation

Water-hammer in the cold leg during an SBLOCA due to cold ECCS injection

International Nuclear Information System (INIS)

Ortiz, M.G.; Ghan, L.S.

1991-01-01

Water-hammer might occur in the cold leg of pressurized water reactors (PWR) during small break loss-of-coolant accidents (SBLOCA's), when cold emergency core cooling system (ECCS) water is injected into a pipe that may be partially filled with saturated steam. The water may mix with the steam and cause it to condense abruptly. Depending on the flow regime present, slugs of liquid may then be accelerated towards each other or against the piping structure. The possibility of this phenomenon is of concern to us because it may become a dominant phenomenon and change the character of the transient. In performing the code scaling, applicability, and uncertainty study (CSAU) on a SBLOCA scenario, we had to examine the possibility that the transient being analyzed could experience water-hammer and thus depart from the scope of the study. Two criteria for water-hammer initiation were investigated and tested using a RELAP5/MOD3 simulation of the transient. Our results indicated a very low likelihood of occurrence of the phenomenon. 8 refs., 6 figs

The analysis of 14.8 percent cold leg break without the application of hydroaccumulators in the PMK-NHV test facility

International Nuclear Information System (INIS)

Szabados, L.; Ezsoel, Gy.; Perneczky, L.

1990-12-01

A series of reactor safety tests have been performed in the experimental reactor simulation facility PMK-NHV of the Paks Nuclear Power Plant, Hungary, with and without the use of hydroaccumulator (SIT) operation. 14.8 percent cold leg break simulation experiments are reported without SITs in action, and the measurement results were analyzed using the RELAP5/mod2 computer code. The description of the experiment is followed by the parameter variations and their analysis, together with an interpretation of the measurement results. The lessons from the LOCA simulation tests are evaluated. (R.P.) 10 refs.; 48 figs.; 3 tabs

ROSA-V large scale test facility (LSTF) system description for the third and fourth simulated fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Mitsuhiro; Nakamura, Hideo; Ohtsu, Iwao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

2003-03-01

The Large Scale Test Facility (LSTF) is a full-height and 1/48 volumetrically scaled test facility of the Japan Atomic Energy Research Institute (JAERI) for system integral experiments simulating the thermal-hydraulic responses at full-pressure conditions of a 1100 MWe-class pressurized water reactor (PWR) during small break loss-of-coolant accidents (SBLOCAs) and other transients. The LSTF can also simulate well a next-generation type PWR such as the AP600 reactor. In the fifth phase of the Rig-of-Safety Assessment (ROSA-V) Program, eighty nine experiments have been conducted at the LSTF with the third simulated fuel assembly until June 2001, and five experiments have been conducted with the newly-installed fourth simulated fuel assembly until December 2002. In the ROSA-V program, various system integral experiments have been conducted to certify effectiveness of both accident management (AM) measures in beyond design basis accidents (BDBAs) and improved safety systems in the next-generation reactors. In addition, various separate-effect tests have been conducted to verify and develop computer codes and analytical models to predict non-homogeneous and multi-dimensional phenomena such as heat transfer across the steam generator U-tubes under the presence of non-condensable gases in both current and next-generation reactors. This report presents detailed information of the LSTF system with the third and fourth simulated fuel assemblies for the aid of experiment planning and analyses of experiment results. (author)

Performance of core exit thermocouple for PWR accident management action in vessel top break LOCA simulation experiment at OECD/NEA ROSA project

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

2009-01-01

Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary side in case of core temperature excursion. Test 6-1 is first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reason of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection. (author)

Sampling based uncertainty analysis of 10% hot leg break LOCA in large scale test facility

International Nuclear Information System (INIS)

Sengupta, Samiran; Kraina, V.; Dubey, S. K.; Rao, R. S.; Gupta, S. K.

2010-01-01

Sampling based uncertainty analysis was carried out to quantify uncertainty in predictions of best estimate code RELAP5/MOD3.2 for a thermal hydraulic test (10% hot leg break LOCA) performed in the Large Scale Test Facility (LSTF) as a part of an IAEA coordinated research project. The nodalisation of the test facility was qualified for both steady state and transient level by systematically applying the procedures led by uncertainty methodology based on accuracy extrapolation (UMAE); uncertainty analysis was carried out using the Latin hypercube sampling (LHS) method to evaluate uncertainty for ten input parameters. Sixteen output parameters were selected for uncertainty evaluation and uncertainty band between 5 th and 95 th percentile of the output parameters were evaluated. It was observed that the uncertainty band for the primary pressure during two phase blowdown is larger than that of the remaining period. Similarly, a larger uncertainty band is observed relating to accumulator injection flow during reflood phase. Importance analysis was also carried out and standard rank regression coefficients were computed to quantify the effect of each individual input parameter on output parameters. It was observed that the break discharge coefficient is the most important uncertain parameter relating to the prediction of all the primary side parameters and that the steam generator (SG) relief pressure setting is the most important parameter in predicting the SG secondary pressure

Cold leg condensation tests. Task C. Steam--water interaction tests

International Nuclear Information System (INIS)

Brodrick, J.R.; Loiselle, V.

1974-03-01

A report is presented of tests to determine the condensation efficiency of ECC water injected into a quality fluid mixture flowing through the cold leg. In particular, a specific objective was to determine if the mixture of ECC water and quality fluid reached thermodynamic equilibrium before exiting the cold leg. Further, the stability of the ECC water/quality fluid interaction would be assessed by interpretation of thermocouple records and utilization of a section of cold leg piping with view ports to film the interaction whenever possible. The cold leg condensation tests showed complete condensation of the 5 lbm/sec steam quality mixtures in the cold leg by the ECC water flows of the test matrix. The cold leg exit fluid temperature remained below the saturation temperature and had good agreement with the predicted cold leg outlet temperature, calculated assuming total condensation. (U.S.)

Report on the uncertainty methods study

International Nuclear Information System (INIS)

1998-06-01

The Uncertainty Methods Study (UMS) Group, following a mandate from CSNI, has compared five methods for calculating the uncertainty in the predictions of advanced 'best estimate' thermal-hydraulic codes: the Pisa method (based on extrapolation from integral experiments) and four methods identifying and combining input uncertainties. Three of these, the GRS, IPSN and ENUSA methods, use subjective probability distributions, and one, the AEAT method, performs a bounding analysis. Each method has been used to calculate the uncertainty in specified parameters for the LSTF SB-CL-18 5% cold leg small break LOCA experiment in the ROSA-IV Large Scale Test Facility (LSTF). The uncertainty analysis was conducted essentially blind and the participants did not use experimental measurements from the test as input apart from initial and boundary conditions. Participants calculated uncertainty ranges for experimental parameters including pressurizer pressure, primary circuit inventory and clad temperature (at a specified position) as functions of time

Identification of flow regimes and heat transfer modes in Angra-2 core during the simulation of the small break loss of coolant accident of 250 cm2 in the cold leg of primary loop using RELAP5 code

International Nuclear Information System (INIS)

Borges, Eduardo M.; Sabundjian, Gaiane

2017-01-01

The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2. gamma code in Angra-2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 250cm 2 of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of Angra-2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of Angra-2 during the postulated accident. The results obtained for Angra-2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core. (author)

LOFT/LP-SB-2, Loss of Fluid Test, Small Hot Leg Break LOCA, Delayed Pump

International Nuclear Information System (INIS)

1989-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The third OECD LOFT experiment was conducted on 14 July 1983. It simulated a 3-in (7.62 cm) equivalent break diameter located in the hot leg of the operating loop. The major objective of this experiment was to determine system transient characteristics for small hot leg break loss-of-coolant accidents with delayed pump trip. The experiment was conducted from initial temperature and pressure conditions representative of typical commercial PWRs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

LOFT/LP-SB-1, Loss of Fluid Test, Small Hot Leg Break LOCA, Early Pump

International Nuclear Information System (INIS)

1989-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The second OECD LOFT experiment was conducted on 23 June 1983. It simulated a 3-in (7.62 cm) equivalent break diameter located in the hot leg of the operating loop. The major objective of this experiment was to determine system transient characteristics for small hot leg break loss-of-coolant accidents with early pump trip. The experiment was conducted from initial temperature and pressure conditions representative of typical commercial PWRs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

Potential for boron dilution during small-break LOCAs in PWRs

International Nuclear Information System (INIS)

Nourbakhsh, H.P.; Cheng, Z.

1995-01-01

This paper documents the results of a scoping study of boron dilution and mixing phenomena during small break loss of coolant accidents (LOCAs) in pressurized water reactors (PWRs). Boron free condensate can accumulate in the cold leg loop seals when the reactor is operating in a reflux/boiler condenser mode. A problem may occur when the subsequent change in flow conditions such as loop seal clearing or re-establishment of natural circulation flow drive the diluted water in the loop seals into the reactor core without sufficient mixing with the highly borated water in the reactor downcomer and lower plenum. The resulting low boron concentration coolant entering the core may cause a power excursion leading to fuel failure. The mixing processes associated with a slow moving stream of diluted water through the loop seal to the core were examined in this report. A bounding evaluation of the range of boron concentration entering the core during a small break LOCA in a typical Westinghouse-designed, four-loop plant is also presented in this report

Identification of flow regimes and heat transfer modes in Angra-2 core during the simulation of the small break loss of coolant accident of 250 cm{sup 2} in the cold leg of primary loop using RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Borges, Eduardo M.; Sabundjian, Gaiane, E-mail: borges.em@hotmail.com, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

2017-07-01

The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2. gamma code in Angra-2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 250cm{sup 2} of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of Angra-2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of Angra-2 during the postulated accident. The results obtained for Angra-2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core. (author)

Mixing phenomena of interest to boron dilution during small break LOCAs in PWRs

International Nuclear Information System (INIS)

Nourbakhsh, H.P.; Cheng, Z.

1995-01-01

This paper presents the results of a study of mixing phenomena related to boron dilution during small break loss of coolant accidents (LOCAs)in pressurized water reactors (PWRs). Boron free condensate can accumulate in the cold leg loop seals when the reactor is operating in a reflux/boiler-condenser mode. A problem may occur when subsequent change in flow conditions such as loop seal clearing or re-establishment of natural circulation flow drive the diluted water in the loop seals into the reactor core without sufficient mixing with the highly borated water in the reactor downcomer and lower plenum. The resulting low boron concentration coolant entering the core may cause a power excursion leading to fuel failure. The mixing processes associated with a slow moving stream of diluted water through the loop seal to the core are examined in this paper. Bounding calculations for boron concentration of coolant entering the core during a small break LOCA in a typical Westinghouse-designed four-loop plant are also presented

CATHARE 2 analysis of the small break LOCA experiment SP-SB-03, conducted in SPES facility

International Nuclear Information System (INIS)

Meloni, P.

1995-01-01

SPES integral test facility is a scale model of a commercial three-loop PWR plant, making the simulation of a wide range of accident scenarios possible. A Small Break Loss of Coolant test was carried out in this facility in 1991 to serve as a counterpart of tests conducted on BETHSY (France), LSTF (Japan) and LOBI (EC) facilities. A post-test analysis of this test, performed with CATHARE 2 code was realized by ENEA in the framework of the co-operation ENEA-CEA on advanced reactors. This paper presents a survey of the results of the post-test calculation. (author). 5 refs, 11 figs, 3 tabs

International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA

Directory of Open Access Journals (Sweden)

N. Aksan

2008-01-01

Full Text Available Best-estimate thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. OECD/NEA Committee on the Safety of Nuclear Installations (CSNI has promoted, over the last twenty-nine years, some forty-eight international standard problems (ISPs. These ISPs were performed in different fields as in-vessel thermal-hydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermal-hydraulic behaviour. 80% of these ISPs were related to the working domain of principal working group no.2 on coolant system behaviour (PWG2 and were one of the major PWG2 activities for many years. A global review and synthesis on the contribution that ISPs have made to address nuclear reactor safety issues was initiated by CSNI-PWG2 and an overview on the subject of small break LOCA ISPs is given in this paper based on a report prepared by a writing group. In addition, the relevance of small break LOCA in a PWR with relation to nuclear reactor safety and the reorientation of the reactor safety program after TMI-2 accident are shortly summarized. The experiments in four integral test facilities, LOBI, SPES, BETHSY, ROSA IV/LSTF and the recorded data during a steam generator tube rupture transient in the DOEL-2 PWR (Belgium were the basis of the five small break LOCA related ISP exercises, which deal with the phenomenon typical of small break LOCAs in Western design PWRs. Some lessons learned from these small break LOCA ISPs are identified in relation to code deficiencies and capabilities, progress in the code capabilities, possibility of scaling, and various additional aspects

TRAC analysis of an 80% pump-side, cold-leg, large-break loss-of-coolant accident for the Westinghouse AP600 advanced reactor design

International Nuclear Information System (INIS)

Lime, J.F.; Boyack, B.E.

1996-01-01

An updated TRAC 80% pump-side, cold-leg, large-break (LB) loss-of-coolant accident (LOCA) has been calculated for the Westinghouse AP600 advanced reactor design. The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The break size and location were calculated by Westinghouse to be the most severe LBLOCA for the AP600 design. The LBLOCA transient was calculated to 280 s, which is the time of in-containment refueling water-storage-tank injection. All fuel rods were quenched completely by 240 s. Peak cladding temperatures (PCTs) were well below the licensing limit of 1,478 K (2,200 F) but were very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCTs for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown and refill periods. The reasons for these differences are still being investigated

ROSA-IV Large Scale Test Facility (LSTF) system description for second simulated fuel assembly

International Nuclear Information System (INIS)

1990-10-01

The ROSA-IV Program's Large Scale Test Facility (LSTF) is a test facility for integral simulation of thermal-hydraulic response of a pressurized water reactor (PWR) during small break loss-of-coolant accidents (LOCAs) and transients. In this facility, the PWR core nuclear fuel rods are simulated using electric heater rods. The simulated fuel assembly which was installed during the facility construction was replaced with a new one in 1988. The first test with this second simulated fuel assembly was conducted in December 1988. This report describes the facility configuration and characteristics as of this date (December 1988) including the new simulated fuel assembly design and the facility changes which were made during the testing with the first assembly as well as during the renewal of the simulated fuel assembly. (author)

Analysis of small break loss of coolant accident for Chinese CPR1000

Energy Technology Data Exchange (ETDEWEB)

Kim, Ju Youl [FNC Technology Co., Yongin (Korea, Republic of); Cilier, Anthonie [North-West University, Mahikeng (South Africa); Poc, Li-chi Cliff [Micro-Simulation Technology, Montville (United States)

2016-05-15

This research analyses the small break loss of coolant accident (LOCA) on a Chinese CPR1000 type reactor. LOCA accident is used as benchmark for the PCTRAN/CPR1000 code by comparing the effects and results to the Manshaan FSAR accident analysis. LOCA is a design basis accident in which a guillotine break is postulated to occur in one of the cold legs of a pressurized water reactor (PWR). Consequently, the primary system pressure would drop and almost all the reactor coolant would be discharged into the reactor containment. The drop in pressure would activate the reactor protection system and the reactor would trip. The simulation of a 3-inch small break loss of coolant accident using the PCTRAN/CPR1000 has revealed this code's effectiveness as well as weaknesses in specific simulation applications. The code has the ability to run at 16 times real time and produce very accurate results. The results are consistently producing the same trends as licensed codes used in Safety Assessment Reports. It is however able to produce these results in a fraction of the time and also provides a whole plant simulation coupling the various thermal, hydraulic, chemical and neutronic systems together with a plant specific control system.

Data report of ROSA/LSTF experiment TR-LF-07. Loss-of-feedwater transient with primary feed-and-bleed operation

International Nuclear Information System (INIS)

Takeda, Takeshi

2016-07-01

An experiment TR-LF-07 was conducted on June 23, 1992 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-IV (ROSA-IV) Program. The ROSA/LSTF experiment TR-LF-07 simulated a loss-of-feedwater transient in a pressurized water reactor (PWR) under assumptions of primary feed-and-bleed operation and total failure of auxiliary feedwater system. A safety injection (SI) signal was generated when steam generator (SG) secondary-side collapsed liquid level decreased to 3 m. Primary depressurization was initiated by fully opening a power-operated relief valve (PORV) of pressurizer (PZR) 30 min after the SI signal. High pressure injection (HPI) system was started in loop with PZR 12 s after the SI signal, while it was initiated in loop without PZR when the primary pressure decreased to 10.7 MPa. The primary and SG secondary pressures were kept almost constant because of cycle opening of the PZR PORV and SG relief valves. The PZR liquid level began to drop steeply following the PORV full opening, which caused liquid level formation at the hot leg. The PZR and hot leg liquid levels recovered due to the HPI actuation in both loops. The primary pressure became lower than the SG secondary pressure, which resulted in the actuation of accumulator (ACC) system in both loops. The PZR and hot legs became full of liquid again after the ACC actuation. The primary feed-and-bleed operation by use of the PORV, HPI and ACC systems was effective to core cooling because of no core uncovery. The experiment was terminated when the continuous core cooling was confirmed due to the successive coolant injection from the HPI system even after the ACC termination. The obtained data would be useful to study operator actions and procedures in the PWR multiple fault events which behaviors in the PZR affect. This report summarizes the test procedures, conditions and major observation in the ROSA/LSTF experiment TR-LF-07. (author)

Lessons learned from OECD/CSNI ISP on small break LOCA: final report

International Nuclear Information System (INIS)

1996-07-01

This document presents an overview of the results obtained from five recent OECD/CSNI International Standard Problems (ISPs) dealing with phenomenologies typical of Small Break LOCA in PWR nuclear power plants of western design. The experiment in four Integral test Facilities, Lobi, Spes, Bethsy and Lstf and the recorded data from a steam generator tube rupture transient in the Belgian PWR of Doel, were taken as reference for the calculations. Relevant hardware characteristics of the facilities and of the plant are firstly given, including the correlation between key thermalhydraulic phenomena and the reference experimental scenarios. A statistical evaluation of the general data connected with each ISP is then presented. The lessons learned from the ISPs are then considered. Four areas have been identified: code deficiencies and capabilities, scaling of the data, progress in code capabilities and various additional aspects

Comparisons of TRAC-PF-1 calculations with semiscale Mod-3 small-break tests S-SB-P1 and S-SB-P7

International Nuclear Information System (INIS)

Sahota, M.S.

1982-01-01

Semiscale Tests S-SB-P1 and S-SB-P7 conducted in the Semiscale Mod-3 facility at the Idaho National Engineering Laboratory are analyzed using the latest released version of the Transient Reactor Analysis Code (TRAC-PF1). The results are used to assess TRAC-PF1 predictions of thermal-hydraulic phenomena and the effects of break size and pump operation on system response during slow transients. Tests S-SB-P1 and S-SB-P7 simulated an equivalent pressurized-water-reactor (PWR) 2.5% communicative cold-leg break for early and late pump trips, respectively, with only high-pressure injection (HPI) into the cold legs. The parameters examined include break flow, primary-system pressure response, primary-system mass distribution, and core characteristics

LOFT/LP-SB-3, Loss of Fluid Test, Cold Leg Break LOCA, No High Pressure injection System (HPIS)

International Nuclear Information System (INIS)

1989-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The sixth OECD LOFT experiment was conducted on 5 March 1984. It simulated a 1.8-in cold leg break LOCA with no HPIS available. This experiment was designed mainly for investigation of plant recovery effectiveness using secondary bleed and feed during core uncover and addressed accumulator injection at low pressure differentials. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

International Standard Problems and Small Break Loss-Of-Coolant Accident (SBLOCA)

International Nuclear Information System (INIS)

Aksan, N.

2008-01-01

, ROSA IV/LSTF and the recorded data during a steam generator tube rupture transient in the DOEL-2 PWR (Belgium) were the basis of ISP calculations. The statistical evaluation of the general data obtained from these ISPs is summarized. Some lessons learned from these small break LOCA ISPs are identified in relation to code deficiencies and capabilities, progress in the code capabilities, possibility of scaling, and various additional aspects. ISPs are providing unique material and benefits for some safety related issues. Some of the technical findings and benefits provided by small break LOCA ISPs are provided as conclusions and recommendations.

Pressurized thermal shock. CNA-I behavior when a hot leg breaks of 50 cm2 is produced

International Nuclear Information System (INIS)

Rosso, Ricardo D.; Ventura, Mirta A.

2002-01-01

Pressurized thermal shock (PTS) phenomena in the CNA-I pressurize heavy water reactor is analyzed in this paper. The initiating event is a hypothetical 50 cm 2 break of the line connecting the pressurizer and the primary system. The calculation procedure for obtaining the local thermal-hydraulic parameters in the reactor pressure vessel downcomer is described firstly. Results obtained lead to conclusions in different subjects. The first conclusion is that a simple tool of easy application is available to analyze PTS phenomena in cases of breaks in the primary system in cold and hot legs. This methodology is fully independent of the methodology utilized by the Utility. Another important conclusion comes from the analysis of the temperature evolution of the fluid below the cold leg level in the RPV downcomer, as a function of the T HPI temperature of the TJ system injected water from. It is also concluded that the results obtained with the methodology adopted agree with the ones obtained with the methodologies validated against experiments in the UPTF facility. It is possible to observe that when T HPI increase, the conditions suitable for PTS occurrence in a LOCA accident tend to diminish. The maximum value to the T HPI may be fixed from the maximum temperature allowed to preserve the structural integrity of the fuel cladding. (author)

Comparisons of TRAC-PD2 calculations with Semiscale Mod-3 small-break tests

International Nuclear Information System (INIS)

Gilbert, J.S.; Sahota, M.S.; Boyack, B.E.; Booker, C.P.; Meier, J.K.

1981-01-01

Five experiments conducted in the Semiscale Mod-3 facility at the Idaho National Engineering Laboratory (INEL) were calculated using the latest released version of the Transient Reactor Analysis Code (TRAC-PD2). The results were used to assess TRAC-PD2 predictions of thermal-hydraulic phenomena and the effects of pump operation on system response during slow transients. Tests S-SB-P1, S-SB-P2, and S-SB-P7 simulated equivalent 2.5% communicative cold-leg breaks for early pump-trip (pumps-off), intermediate pump-trip (pumps-on), and late pump-trip (pumps-on) operation, respectively. Tests S-SB-P3 and S-SB-P4 simulated equivalent 2.5% communicative hot-leg breaks for pumps-off and pumps-on operation, respectively. Parameters examined in the study included primary system mass distribution, mass inventory, and void fraction distribution

Reconstruction of core inlet temperature distribution by cold leg temperature measurements

International Nuclear Information System (INIS)

Saarinen, S.; Antila, M.

2010-01-01

The reduced core of Loviisa NPP contains 33 thermocouple measurements measuring the core inlet temperature. Currently, these thermocouple measurements are not used in determining the inlet temperature distribution. The average of cold leg temperature measurements is used as inlet temperature for each fuel assembly. In practice, the inlet temperature distribution is not constant. Thus, using a constant inlet temperature distribution induces asymmetries in the measured core power distribution. Using a more realistic inlet temperature distribution would help us to reduce virtual asymmetries of the core power distribution and increase the thermal margins of the core. The thermocouples at the inlet cannot be used directly to measure the inlet temperature accurately because the calibration of the thermocouples that is done at hot zero power conditions is no longer valid at full power, when there is temperature change across the core region. This is due to the effect of neutron irradiation on the Seebeck coefficient of the thermocouple wires. Therefore, we investigate in this paper a method to determine the inlet temperature distribution based on the cold leg temperature measurements. With this method we rely on the assumption that although the core inlet thermocouple measurements do not measure the absolute temperature accurately they do measure temperature changes with sufficient accuracy particularly in big disturbances. During the yearly testing of steam generator safety valves we observe a large temperature increase up to 12 degrees in the cold leg temperature. The change in the temperature of one of the cold legs causes a local disturbance in the core inlet temperature distribution. Using the temperature changes observed in the inlet thermocouple measurements we are able to fit six core inlet temperature response functions, one for each cold leg. The value of a function at an assembly inlet is determined only by the corresponding cold leg temperature disturbance

LOFT/L3-, Loss of Fluid Test, 7. NRC L3 Small Break LOCA Experiment

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the seventh in the NRC L3 Series of small-break LOCA experiments. A 2.5-cm (10-in.) cold-leg non-communicative-break LOCA was simulated. The experiment was conducted on 20 June 1980

Assessment of RELAP5/MOD3.1 with the LSTF SB-SG-06 experiment simulating a steam generator tube rupture transient

International Nuclear Information System (INIS)

Seul, K.W.; Bang, Y.S.; Lee, S.; Kim, H.J.

1996-09-01

The objective of the present work is to identify the predictability of RELAP5/MOD3.1 regarding thermal-hydraulic behavior during a steam generator tube rupture (SGTR). To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR that occurred at the Mihama Unit 2 in 1991 are used. Also, some sensitivity studies of the code change in RELAP5, the break simulation model, and the break valve discharge coefficient are performed. The calculation results indicate that the RELAP5/MOD3.1 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system (RCS) cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator (SG) relief valve, and so on. However, there are some differences from the experimental data in the number of the relief valve cycling in the affected SG, and the flow regime of the hot leg with the pressurizer, and the break flow rates. Finally, the calculation also indicates that the coolant in the core could remain in a subcooled state as a result of the heat transfer caused by the natural circulation flow even if the reactor coolant pumps (RCPs) turned off and that the affected SG could be properly isolated to minimize the radiological release after the SGTR

Assessment of RELAP5/MOD3.1 with the LSTF SB-SG-06 experiment simulating a steam generator tube rupture transient

Energy Technology Data Exchange (ETDEWEB)

Seul, K.W.; Bang, Y.S.; Lee, S.; Kim, H.J. [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

1996-09-01

The objective of the present work is to identify the predictability of RELAP5/MOD3.1 regarding thermal-hydraulic behavior during a steam generator tube rupture (SGTR). To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR that occurred at the Mihama Unit 2 in 1991 are used. Also, some sensitivity studies of the code change in RELAP5, the break simulation model, and the break valve discharge coefficient are performed. The calculation results indicate that the RELAP5/MOD3.1 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system (RCS) cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator (SG) relief valve, and so on. However, there are some differences from the experimental data in the number of the relief valve cycling in the affected SG, and the flow regime of the hot leg with the pressurizer, and the break flow rates. Finally, the calculation also indicates that the coolant in the core could remain in a subcooled state as a result of the heat transfer caused by the natural circulation flow even if the reactor coolant pumps (RCPs) turned off and that the affected SG could be properly isolated to minimize the radiological release after the SGTR.

Influence of liquid holdup in steam generator U-tubes on small break LOCA severity

International Nuclear Information System (INIS)

Leonard, M.T.; Perryman, J.L.; Johnson, G.W.

1983-01-01

The severity of small cold leg break loss-of-coolant accidents has been shown to be influenced by liquid holdup in steam generator U-tubes during pump suction loop seal formation in two experiments performed in the Semiscale Mod-2A facility. The core coolant level can be depressed lower than previously thought possible due to a positive hydrostatic head across the steam generators caused by delayed drainage of liquid from the upflow side of the U-tubes. The significance of a lower core coolant level depression is the potential for a more severe temperature excursion occurring during the coolant boiloff phase subsequent to loop seal clearing and prior to accumulator injection. Presented in this paper are the experimental data analysis and supporting computer code calculations that led to these conclusions

LOFT/L3-6, Loss of Fluid Test, 6. NRC L3 Small Break LOCA Experiment

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the sixth in the NRC L3 Series of small-break LOCA experiments. A 10-cm (2.5-in.) cold-leg non-communicative-break LOCA was simulated. Pumps were running. The experiment was conducted on 10 December 1980

LOFT/L3-5, Loss of Fluid Test, 5. NRC L3 Small Break LOCA Experiment

International Nuclear Information System (INIS)

1991-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the fifth in the NRC L3 Series of small-break LOCA experiments. A 10-cm (2.5-in.) cold-leg non-communicative-break LOCA was simulated. Pumps were shut off. The experiment was conducted on 29 September 1980

The phenomenology of a small break LOCA in a complex thermal hydraulic loop

International Nuclear Information System (INIS)

Di Marzo, M.; Almenas, K.K.; Hsu, Y.Y.; Wang, Z.

1988-01-01

A phenomenological description of the thermal hydraulics events that take place during a simulated Small Break Loss of Coolant Accident (SB-LOCA) is presented. The SB-LOCA transient is described in detail and the various mass and energy transport modes are identified. Similar behavior is observed in other facilities designed for the simulation of this type of accidents. Previous investigations suggest a simple modelling of the phenomena based on fluid mechanic considerations. An extensive experimental program conducted at the experimental facility of the University of Maryland reveals that condensation is a dominant driving force for this type of transients. This finding has significant implications in the modelling of enthalpy transport for some of the flow modes which occur during the transient. In particular it affects the Interruption and Resumption Mode (IRM) during which enthalpy is transported by periodic flow of a two phase mixture. The efforts to predict the flow interruption based on fluid mechanic criteria of phase separation in the hot leg are shown to be misdirected since thermodynamic phenomena taking place in the horizontal portion of the cold legs and in the reactor vessel downcomer are mostly responsible for that transition. For flow resumption to occur the liquid-vapor mixture swelling in the vertical portion of the hot leg determines the occurrence of the liquid spill over the top of the candy cane. (orig.)

ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA

Directory of Open Access Journals (Sweden)

Andi Sofrany Ekariansyah

2015-06-01

, mixture level, temperatur kelongsong, small break LOCA, RELAP5.   ABSTRACT ANALYSIS ON THE CORE CONDITION OF AP1000 ADVANCED POWER REACTOR DURING SMALL BREAK LOCA. Accident due to the loss of coolant from the reactor boundary is an anticipated design basis event in the design of power reactor adopting the Generation II up to IV technology. Small break LOCA leads to more significant impact on safety compared to the large break LOCA as shown in the Three-Mile Island (TMI. The focus of this paper is the small break LOCA analysis on the Generation III+ advanced power reactor of AP1000 by simulating three different initiating events, which are inadvertent opening of Automatic Depressurization System (ADS, double-ended break on one of Direct Vessel Injection (DVI pipe, and 10 inch diameter split break on one of cold leg pipe. Methodology used is by simulating the events on the AP1000 model developed using RELAP5/SCDAP/Mod3.4. The impact analyzed is the core condition during the small break LOCA consisting of the mixture level occurrence and the fuel cladding temperature transient. The results show that the mixture level for all small break LOCA events are above the active core height, which indicates no core uncovery event. The development of the mixture level affect the fuel cladding temperature transient, which shows a decreasingly trend after the break, and the effectifeness of core cooling. Those results are identical with the results of other code of NOTRUMP. The resulted core cooling is also due to the function of coolant injection from passive safety feature, which is unique in the AP1000 design. In overall, the result of analysis shows that the AP1000 model developed by the RELAP5 can be used for analysis of design basis accident considered in the AP1000 advanced power reactor. Keywords: analysis, mixture level, fuel cladding temperature, small break LOCA, RELAP5.

ISP-27 OECD/NEA/CSNI International Standard Problem n.27. Bethsy experiment 9.1 B. 2. cold leg break without HPSI and with delayed ultimate procedure. Comparison report. Volume 1 + 2

International Nuclear Information System (INIS)

1992-11-01

This report is the final comparison report for ISP-27, a blind problem which is based on the BETHSY test 9.1b performed in december 1989 at the Nuclear Research Center in Grenoble (France). The BETHSY integral test facility is a scaled down model of a 3 loop 900 e MW FRAMATOME PWR; the overall scaling factor applied to every volume, mass flowrate and power level is close to 1/100, while elevations are 1/1 in order to preserve the gravitational heads. The cold leg break is combined with the High Pressure Injection System (HPIS) failure. In that case, the state oriented approach requires operators to start an Ultimate Procedure, which consists in fully opening the Steam Generator (SG) atmospheric dumps as soon as they are informed of the unavailability of the HPIS. The presently studied scenario assumes a delayed application of this procedure, which is started only when the core outlet temperature rises significantly higher than the saturation temperature. The BETHSY Test 9.1b addresses, besides typical problems relevant to Small Break Loss Of Coolant Accidents (SBLOCA) such as critical 2-phase flow, loop seal clearing, heat-transfer during boil-off or accumulator injection, specific aspects related to the fast depressurization (primary to secondary and structural heat transfer), uncovered core behavior when intense condensation takes place in the SG, and primary side refilling by the Low Pressure Injection System (LPIS)

Small break critical discharge: The roles of vapor and liquid entrainment in a stratified two-phase region upstream of the break

International Nuclear Information System (INIS)

Schrock, V.E.; Revankar, S.T.; Mannheimer, R.; Wang, C.H.

1986-12-01

The main objective of this research program was to perform an experimental investigation on the phenomena of two-phase critical flow through small break from a horizontal pipe which contained a stratified two phase flow. Stagnation conditions investigated were saturated steam-water, and air-cold water at pressures ranging from 0.37 MPa to 1.07 MPa. The small breaks employed were cylindrical tubes of diameters 3.96 mm, 6.32 mm, and 10.1 mm with sharp-edged entrance. For breaks resulting from a small hole in a primary coolant pipe or in a small pipe, a sharp-edged orifice or a sharp-edged tube can be the approximation

Study on severe accident induced by large break loss of coolant accident for pressureized water reactor

International Nuclear Information System (INIS)

Zhang Longfei; Zhang Dafa; Wang Shaoming

2007-01-01

Using the best estimate computer code SCDAP/RELAP5/MOD3.2 and taking US Westinghouse corporation Surry nuclear power plant as the reference object, a typical three-loop pressurized water reactor severe accident calculation model was established and 25 cm large break loss of coolant accident (LBLOCA) in cold and hot leg of primary loop induced core melt accident was analyzed. The calculated results show that core melt progression is fast and most of the core material melt and relocated to the lower plenum. The lower head of reactor pressure vessel failed at an early time and the cold leg break is more severe than the hot leg break in primary loop during LBLOCA. (authors)

Development of analysis methodology for hot leg break mass and energy release

International Nuclear Information System (INIS)

Song, Jin Ho; Kim, Cheol Woo; Kwon, Young Min; Kim, Sook Kwan

1995-04-01

A study for the development of an analysis methodology for hot leg break mass and energy release is performed. For the blowdown period a modified CEFLASH-4A methodology is suggested. For the post blowdown period a modified CONTRAST boil-off model is suggested. By using these computer code improved mass and energy release data are generated. Also, a RELAP5/MOD3 analysis for finally the FLOOD-3 computer code has been modified for use in the analysis of hot leg break. The results of analysis using modified FLOOD-3 are reasonable as we expected and their trends are good. 66 figs., 8 tabs. (Author) .new

Assessment of SPACE Code Using the LSTF 10% MSLB Test

International Nuclear Information System (INIS)

Kim, Yo Han; Yang, Chang Keun; Ha, Sang Jun

2012-01-01

The Korea Nuclear Hydro and Nuclear Power Co. (KHNP) has developed a multipurpose nuclear safety analysis code called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE). The SPACE is a best-estimated two-phase three-field thermal-hydraulic analysis code used to analyze the safety and performance of pressurized water reactors (PWRs). As in the second phase of the project, the beta version of the code has been developed through the validation and verification (V and V) using integral loop test data or plant operating data and the complement of code to solve the SPACE code user problem and resolution reports. In this study, the Large Scale Test Facility (LSTF) 10% main steam line break (MSLB) test, SB-SL-01, was simulated as a V and V work. The results were compared with the experimental data and those of the RELAP5/MOD3.1 code simulation

Code Assessment of SPACE 2.19 using LSTF Steam Generator Tube Rupture Test

Energy Technology Data Exchange (ETDEWEB)

Kim, Minhee; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The SPACE is a best estimated two-phase three-field thermal-hydraulic analysis code used to analyze the safety and performance of pressurized water reactors. As a result of the development, the 2.19 version of the code was released through the successive various verification and validation works. The present work is on the line of expanding the work by Kim et al. In this study, results produced by the SPACE 2.19 code were compared with the experimental data from JAERI's LSTF Test Run LSTF SB-SG-06 experiment simulating a Steam Generator Tube Rupture (SGTR) transient. In order to identify the predictability of SPACE 2.19, the LSTF steam generator tube rupture test was simulated. To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR and the RELAP5/ MOD3.1 are used. The calculation results indicate that the SPACE 2.19 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator relief valve.

Analysis, by Relap5 code, of boron dilution phenomena in a Small Break Loca Transient, performed in PKL III E 2.2 test

International Nuclear Information System (INIS)

Rizzo, G.; Vella, G.

2007-01-01

The present work is finalized to investigate the E2.2 thermal-hydraulics transient of the PKL III facility, which is a scaled reproduction of a typical German PWR, operated by FRAMATOME-ANP in Erlangen, Germany, within the framework of an international cooperation (OECD/SETH project). The main purpose of the project is to study boron dilution events in Pressurized Water Reactors and to contribute to the assessment of thermal-hydraulic system codes like Relap5. The experimental test PKL III E2.2 investigates the behavior of a typical PWR after a Small Break Loss Of Coolant Accident (SB-LOCA) in a cold leg and an immediate injection of borated water in two cold legs. The main purpose of this work is to simulate the PKL III test facility and particularly its experimental transient by Relap5 system code. The adopted nodalization, already available at Department of Nuclear Engineering (DIN), has been reviewed and applied with an accurate analysis of the experimental test parameters. The main result relies in a good agreement of calculated data with experimental measures for a number of main important variables. (author)

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

Small break LOCA [loss of coolant accident] mitigation for Bellefonte

International Nuclear Information System (INIS)

Bayless, P.D.; Dobbe, C.A.

1986-01-01

Several 5-cm (2-in.) diameter cold leg break loss coolant accidents for the Bellefonte nuclear plant were analyzed as part of the Severe Accident Sequence Analysis Program. The transients assumed various system failures, and included the S 2 D sequence. Operator actions to mitigate the S 2 D transient were also investigated. The transients were analyzed until either core damage began or long-term decay heat removal was established. The S 2 D sequence was analyzed into the core damage phase of the transient. The analyses showed that the flow from one high pressure injection pump was necessary and sufficient to prevent core damage in the absence of operator actions. Operator actions were also able to prevent core damage for the S 2 D sequence

Analysis of breaks in pipe connections to the hot leg and to the loop seal in the primary system of Ringhals 2 PWR

International Nuclear Information System (INIS)

Nilsson, L.; Sjoeberg, A.

1987-01-01

Analysis has been made of seven different cases of breaks in pipes connected to the hot leg and to the loop seal in Ringhals 2 PWR. The pipes, in which guillotine breaks are postulated, have nominal diameters ranging from 1 to 14 inches. The purpose of the analysis is to supplement the basis for a review of the inspection procedures for the safety of pressure vessels prescribed by SKI. A similar analysis already exists concerning breaks in the cold leg connections. The analysis has been made using the thermal hydraulic computer code RELAPS/MOD2 and with best estimate assumptions. This means that normal operating conditions have been adopted for the input to the calculations. However, the capacity of the safety injection system was assumed to be reduced by having one pump not operating each of the low pressure and high pressure safety injection system. The results of the analysis are presented in tables and as computer plots. The analysis shows that the consequences with respect to increased fuel rod and cladding temperatures are quite harmless. Only the two cases with the largest break sizes lead to critical heat flux (CHF) and increased temperatures for the hottest rods in the core. The peak cladding temperature was 636 degrees C for the 12 inch break. In both cases rewetting occurred within 25 s of accident initiation. In the cases with breaks in connections of 6 inch diameter and smaller the RELAP5 calculations indicated a substantial margin to CHF throughout the transient. (authors)

Analysis of unmitigated large break loss of coolant accidents using MELCOR code

Science.gov (United States)

Pescarini, M.; Mascari, F.; Mostacci, D.; De Rosa, F.; Lombardo, C.; Giannetti, F.

2017-11-01

In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a generic Pressurized Water Reactor of 900 MWe has been developed. The aim of this paper is to present the analysis of MELCOR code calculations concerning two independent unmitigated large break loss of coolant accident transients, occurring in the cited type of reactor. In particular, the analysis and comparison between the transients initiated by an unmitigated double-ended cold leg rupture and an unmitigated double-ended hot leg rupture in the loop 1 of the primary cooling system is presented herein. This activity has been performed focusing specifically on the in-vessel phenomenology that characterizes this kind of accidents. The analysis of the thermal-hydraulic transient phenomena and the core degradation phenomena is therefore here presented. The analysis of the calculated data shows the capability of the code to reproduce the phenomena typical of these transients and permits their phenomenological study. A first sequence of main events is here presented and shows that the cold leg break transient results faster than the hot leg break transient because of the position of the break. Further analyses are in progress to quantitatively assess the results of the code nodalization for accident management strategy definition and fission product source term evaluation.

Integrity assessment of the cold leg piping system in a PWR

International Nuclear Information System (INIS)

Mayfield, M.E.; Leis, B.N.

1981-01-01

The purpose of this paper is to examine the integrity of a nuclear piping system, designed in accordance with Section III, in the context of a damage tolerance analysis procedure. Such a procedure directly addresses the defects and cyclic loadings that are responsible for the above noted exceptions. The analysis and results reported here are for a fatigue life analysis of the Cold Leg piping in a PWR. This piping system is particularly important from a safety standpoint since a large break is a possible initiator of a core meltdown accident. The analysis employs LEFM concepts to determine the time between the initial start-up and (1) formation of a leak, (2) detection of the leak, and (3) the final fracture of the piping. Both longitudinal and circumferential defects are considered. The defects are assumed to propagate from the pipe I.D. in a self-similar manner. Inputs to the analysis were derived from information supplied by plant operators and vendors, published data, and 'expert opinions'. The life was computed using a linear damage accumulation. (orig./GL)

An investigation of core liquid level depression in small break loss-of-coolant accidents

International Nuclear Information System (INIS)

Schultz, R.R.; Watkins, J.C.; Motley, F.E.; Stumpf, H.; Chen, Y.S.

1991-08-01

Core liquid level depression can result in partial core dryout and heatup early in a small break loss-of-coolant accident (SBLOCA) transient. Such behavior occurs when steam, trapped in the upper regions of the reactor primary system (between the loop seal and the core inventory), moves coolant out of the core region and uncovers the rod upper elevations. The net result is core liquid level depression. Core liquid level depression and subsequent core heatups are investigated using subscale data from the ROSA-IV Program's 1/48-scale Large Scale Test Facility (LSTF) and the 1/1705-scale Semiscale facility. Both facilities are Westinghouse-type, four-loop, pressurized water reactor simulators. The depression phenomena and factors which influence the minimum core level are described and illustrated using examples from the data. Analyses of the subject experiments, conducted using the TRAC-PF1/MOD1 (Version 12.7) thermal-hydraulic code, are also described and summarized. Finally, the response of a typical Westinghouse four-loop plant (RESAR-3S) was calculated to qualitatively study coal liquid level depression in a full-scale system. 31 refs., 37 figs., 6 tabs

Break location influence in pressure vessel SBLOCA scenarios

Energy Technology Data Exchange (ETDEWEB)

Querol, Andrea; Gallardo, Sergio; Verdú, Gumersindo, E-mail: anquevi@upv.es, E-mail: sergalbe@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Instituto Universitario de Seguridad Industrial, Radiofísica y Medioambiental (ISIRYM), Universitat Politècnica de València (Spain)

2017-07-01

The inspections performed in Davis Besse and in the South Texas Project Unit-I reactors pointed out safety issues regarding the structural integrity of the Pressure Vessel (PV). In these inspections, two anomalies were found: a wall thinning and degradation in the PV upper head of the Davis Besse reactor and a small amount of residue around of two instrument-tube penetration nozzles located in the PV lower plenum of the South Texas Project Unit-I reactor. The evolution of these defects could have resulted in Small Break Loss-Of-Coolant Accidents (SBLOCA) if they had not been detected in time. In this frame, the OECD/NEA considered the necessity to simulate these accidental sequences in the OECD/NEA ROSA Project using the Large Scale Test Facility (LSTF). This work is focused in simulating different hypothetical accidental scenarios in the PV using the thermalhydraulic code TRACE5. These simulations allow studying the break localization influence in the transient and the effectiveness of the accident management (AM) actions considered mitigating the consequences of these hypothetical accidental scenarios. (author)

RELAP5/MOD 3.2 Analysis of the Loss of RHR System Experiment Scaled to NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Bajs, T.; Prah, M.

1998-01-01

In the paper the RELAP5/MOD 3.2 analysis of the loss of Residual Heat Removal (RHR) system during midloop operation experiment performed at the Rig of Safety Assessment (ROSA)-IV/Large Scale Test Facility (LSTF) together with the analysis of the same test scenario scaled to NPP Krsko are presented. The experiment consisted in a loss of the RHR system at cold shutdown conditions along with a 5% cold leg break in the loop without pressurizer. The Safety Injection (SI) system was disable in the calculation. The aims of the work were to study the physical phenomena encountered under low power and low system pressure conditions while the upper part of the Reactor Coolant System (RCS) is filled with noncondensable. The impact of the bypass flow between upper plenum and downcomer inlet on transient responses was investigated. The transient was simulated for 6000 s. (author)

Analysis of a hot-leg small break loss-of-coolant accident in a three-loop westinghouse pressurized water reactor plant

International Nuclear Information System (INIS)

Peterson, C.E.; Chexal, V.K.; Clements, T.B.

1985-01-01

The RETRAN-02 computer code was used to perform a best-estimate analysis of a 7.52-cm-diam hotleg break in a three-loop Westinghouse pressurized water reactor. This break size produced a net primary coolant mass depletion through the early portion of the transient. The primary system started to refill only after the accumulator valves opened. As the primary system refilled, there were extreme temperature differentials around the system with cold, denser fluid collecting at the lower elevations and two-phase fluid at higher elevations

Modeling operator actions during a small break loss-of-coolant accident in a Babcock and Wilcox nuclear power plant

International Nuclear Information System (INIS)

Ghan, L.S.; Ortiz, M.G.

1991-01-01

A small break loss-of-accident (SBLOCA) in a typical Babcock and Wilcox (B ampersand W) nuclear power plant was modeled using RELAP5/MOD3. This work was performed as part of the United States Regulatory Commission's (USNRC) Code, Scaling, Applicability and Uncertainty (CSAU) study. The break was initiated by severing one high pressure injection (HPI) line at the cold leg. Thus, the small break was further aggravated by reduced HPI flow. Comparisons between scoping runs with minimal operator action, and full operator action, clearly showed that the operator plays a key role in recovering the plant. Operator actions were modeled based on the emergency operating procedures (EOPs) and the Technical Bases Document for the EOPs. The sequence of operator actions modeled here is only one of several possibilities. Different sequences of operator actions are possible for a given accident because of the subjective decisions the operator must make when determining the status of the plant, hence, which branch of the EOP to follow. To assess the credibility of the modeled operator actions, these actions and results of the simulated accident scenario were presented to operator examiners who are familiar with B ampersand W nuclear power plants. They agreed that, in general, the modeled operator actions conform to the requirements set forth in the EOPs and are therefore plausible. This paper presents the method for modeling the operator actions and discusses the simulated accident scenario from the viewpoint of operator actions

Assessment of SPACE code for multiple failure accident: 1% Cold Leg Break LOCA with HPSI failure at ATLAS Test Facility

Energy Technology Data Exchange (ETDEWEB)

Lee, Jong Hyuk; Lee, Seung Wook; Kim, Kyung-Doo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Design extension conditions (DECs) is a popular key issue after the Fukushima accident. In a viewpoint of the reinforcement of the defense in depth concept, a high-risk multiple failure accident should be reconsidered. The target scenario of ATLAS A5.1 test was LSTF (Large Scale Test Facility) SB-CL-32 test, a 1% SBLOCA with total failure of high pressure safety injection (HPSI) system of emergency core cooling system (ECCS) and secondary side depressurization as the accident management (AM) action, as a counterpart test. As the needs to prepare the DEC accident because of a multiple failure of the present NPPs are emphasized, the capability of SPACE code, just like other system analysis code, is required to expand the DEC area. The objectives of this study is to validate the capability of SPACE code for a DEC scenario, which represents multiple failure accident like as a SBLOCA with HPSI fail. Therefore, the ATLAS A5.1 test scenario was chosen. As the needs to prepare the DEC accident because of a multiple failure of operating NPPs are emphasized, the capability of SPACE code is needed to expand the DEC area. So the capability of SPACE code was validated for one of a DEC scenario. The target scenario was selected as the ATLAS A5.1 test, which is a 1% SBLOCA with total failure of HPSI system of ECCS and secondary side depressurization. Through the sensitivity study on discharge coefficient of break flow, the best fit of integrated mass was found. Using the coefficient, the ATLAS A5.1 test was analyzed using the SPACE code. The major thermal hydraulic parameters such as the system pressure, temperatures were compared with the test and have a good agreement. Through the simulation, it was concluded that the SPACE code can effectively simulate one of multiple failure accidents like as SBLOCA with HPSI failure accident.

Simulation of the IAEA's fourth Standard Problem Exercise small-break loss-of-coolant accident using RELAP5/MOD.3.1

International Nuclear Information System (INIS)

Cebull, P.P.; Hassan, Y.A.

1995-01-01

A small-break loss-of-coolant accident experiment conducted at the PMK-2 integral test facility in Hungary is analyzed using the RELAP5/MOD3.1 thermal-hydraulic code. The experiment simulated a 7.4% break in the cold leg of a VVER-440/213-type nuclear power plant as part of the International Atomic Energy Agency's Fourth Standard Problem Exercise (SPE-4). Blind calculations of the exercise are presented, and the timing of various events throughout the transient is discussed. A posttest analysis is performed in which the sensitivity of the calculated results is investigated. The code RELAP5 predicts most of the transient events well, although a few problems are noted, particularly the failure of RELAP5 to predict dryout in the core even through the collapsed liquid level fell below the top of the heated portion. A discrepancy between the predicted primary mass inventory distribution and the experimental data is identified. Finally, the primary and secondary pressures calculated by RELAP5 fell too rapidly during the latter part of the transient, resulting in rather large errors in the predicted timing of some pressure-actuated events

Small break loss of coolant accidents: Bottom and side break

International Nuclear Information System (INIS)

Hardy, P.G.; Richter, H.J.

1987-01-01

A LOCA can be caused, e.g. by a small break in the primary cooling system. The rate of fluid escaping through such a break will define the time until the core will be uncovered. Therefore the prediction of fluid loss and pressure transient is of major importance to plan for timely action in response to such an event. Stratification of the two phases might be present upstream of the break, thus, the location of the break relative to the vapor-liquid interface and the overall upstream fluid conditions are relevant for the calculation of fluid loss. Experimental results and analyses are presented here for small breaks at the bottom or at the side of a small pressure vessel. It was found that in such a case the onset of the so-called ''vapor pull through'' is important but swelling at sufficient depressurization rates of the liquid due to flashing is also of significance. It was also discovered that in the bottom break the flow rate is strongly dependent on the break entrance quality of the vapour-liquid mixture. The side break can be treated similarly to the bottom break if the interface level is above the break. The analyses developed on the basis of experimental observations showed reasonable agreement of predicted and measured pressure transients. It was possible to calculate the changing interface level and mixture void fraction history in a way compatible with the behavior observed during the experiments. Even though the experiments were performed at low pressures, this work should help to get a better understanding of physical phenomena occurring in a full scale small break LOCA. (orig./HP)

Human thermal responses during leg-only exercise in cold water.

Science.gov (United States)

Golden, F S; Tipton, M J

1987-10-01

1. Exercise during immersion in cold water has been reported by several authors to accelerate the rate of fall of core temperature when compared with rates seen during static immersion. The nature of the exercise performed, however, has always been whole-body in nature. 2. In the present investigation fifteen subjects performed leg exercise throughout a 40 min head-out immersion in water at 15 degrees C. The responses obtained were compared with those seen when the subjects performed an identical static immersion. 3. Aural and rectal temperatures were found to fall by greater amounts during static immersion. 4. It is concluded that 'the type of exercise performed' should be included in the list of factors which affect core temperature during cold water immersion.

Thermal-hydraulic behavior on break simulation of steam generator U-tube

International Nuclear Information System (INIS)

Seul, Kwang Won; Bang, Young Seok; Lee, Sukho; Kim, Hho Jung

1995-01-01

The thermal-hydraulic behavior depending on the break simulation in a steam generator U-tube was investigated and identified the code predictability on plant responses during SGTR accident. The calculated results were compared and assessed with LSTF SB-SG-06 test data. The RELAP5/MOD3.1 code well predicted the sequence of events and the significant phenomena, such as the asymmetric loop behavior, the RCS cooldown and heat transfer by the natural circulation, and system depressurization, even though there were some differences from the experimental data. The break flowrate was found to be sensitive to the break model and affected the system behavior

Analysis of a large break LOCA in the cold leg of the WWER-440/W-213 plant Griefswald, Unit 5

International Nuclear Information System (INIS)

Horche, W.

1993-01-01

The Gessellschaft fur Anlagen und Reaktorsicherheit (GRS) has performed a safety evaluation of the nuclear power plant (NPP) Greifswald, unit 5, of the Soviet type WWER-440/W-213, in cooperation with the French Institute de Protection of de Surete Nucleaire (IPSN) and other partners. Within this project an independent accident analysis is performed by GRS in order to assess the results of existing analysis and to supplement them. In this paper the analysis of the double-ended guillotine break (DEGB) of one cold leg of the main circulation pipe is described. The major objective of the calculation was the investigation of the accident sequence with reduced availability of the emergency core cooling system (single failure criterion). In addition, the simultaneous loss of onsite and offsite power and the failure of scram were assumed. The thermal-hydraulic system code ATHLET/FLUT, developed at GRS and already applied for the safety analysis of several WWER plants, was chosen again. The pressure in the confinement, the back pressure for the discharge model, was calculated as a function of time for this accident separately with GRS-Code RALOC. Furthermore, it was necessary to model the local concentration of direct accumulator injection into the reactor vessel with the help of a special two-channel model of the core and upper plenum. For this model, results were considered obtained from the 1:1 scaled test facility UPTF. It was assumed that only 25% of the upper plenum and core volume is directly penetrated by the injected water. The DEGB was defined in that loop, which is connected with one of three low-pressure injection subsystems. This means that this injected water flows towards the leak without passing the core. As single failure the failure of one of three diesel generators was assumed. The full paper will contain nodalization schemes, which are generated by the ATHLET-Input-Grafic

Estimation of break location and size for loss of coolant accidents using neural networks

International Nuclear Information System (INIS)

Na, Man Gyun; Shin, Sun Ho; Jung, Dong Won; Kim, Soong Pyung; Jeong, Ji Hwan; Lee, Byung Chul

2004-01-01

In this work, a probabilistic neural network (PNN) that has been applied well to the classification problems is used in order to identify the break locations of loss of coolant accidents (LOCA) such as hot-leg, cold-leg and steam generator tubes. Also, a fuzzy neural network (FNN) is designed to estimate the break size. The inputs to PNN and FNN are time-integrated values obtained by integrating measurement signals during a short time interval after reactor scram. An automatic structure constructor for the fuzzy neural network automatically selects the input variables from the time-integrated values of many measured signals, and optimizes the number of rules and its related parameters. It is verified that the proposed algorithm identifies very well the break locations of LOCAs and also, estimate their break size accurately

Case study: Investigating the causes of temperature breaks in South African summer fruit export cold chains

CSIR Research Space (South Africa)

Goedhals-Gerber, LL

2016-08-01

Full Text Available This study investigated the causes and extent of temperature breaks in the South African summer fruit export cold chain from the pack house to the vessel. Numerous causes of temperature breaks throughout the cold chain were found, resulting in many...

Analyzing the loss of coolant accident in PWR nuclear reactors with elevation change in cold leg by RELAP5/MOD3.2 system code

International Nuclear Information System (INIS)

Kheshtpaz, H.; Alison, C.

2006-01-01

As, the Russian designed VVER-1000 reactor of the Bushehr Nuclear Power Plant by taking into account the change from German technology to that of Russian technology, and with the design of elevation change in the cold legs has been developed; therefore safety assessment of these systems for loss of coolant accident in elevation change in the cold legs and comparison results for non change elevation in the cold legs for a typical reactor (normal design of nuclear reactors) is the main important factor to be considered for the safe operation. In this article, the main objective is the simulation of the loss of coolant accident scenario by the RELAP5/MOD3.2 code in two different cases; first, the elevation change in the cold legs, and the second, non change in it. After comparing and analyzing these two code calculations the results have been generalized for a new design feature of Bushehr reactor. The design and simulation of the elevation change in the cold legs process with RELAP5/MOD3.2 code for PWR reactor is performed for the first time in the country, where it is introducing several important results in this respect

Improvement of flavor and viscosity in hot and cold break tomato juice and sauce by peel removal.

Science.gov (United States)

Mirondo, Rita; Barringer, Sheryl

2015-01-01

Tomatoes are typically not peeled before being made into juice but the peels contain enzymes that affect the odor, flavor, and viscosity of the juice. The peels are removed in the finisher, but their presence during the break process may affect quality. Juice was processed from peeled and unpeeled tomatoes using hot or cold break. The juices were pasteurized by high temperature short time (HTST), low temperature long time (LTLT), or with a retort. The control samples were treated with 10% calcium chloride to stop enzymatic activity in the juice. Sauce was made from juice and the tomato products were analyzed for volatiles, color, viscosity, and by sensory. Cold break juice made with peel contained higher levels of some lipoxygenase-, carotenoid-, and amino acid-derived volatiles, than the juice made without peel. Because of the lack of enzyme activity, hot break juices had lower levels of these volatiles and there was no significant difference between hot break juices made with and without peel. CaCl2 -treated and HTST juice had higher levels of most of the volatiles than LTLT, including the lipoxygenase-derived volatiles. The presence of peel produced a significant decrease in the viscosity of the cold break juice and sauce. There was no significant difference in the hue angle, total soluble solids, pH, titratable acidity, and vitamin C for most of the treatments. The texture, flavor, and overall liking of cold break juice made without peel were preferred over cold break juice made with peel whereas the color was less preferred. Between the sauces no significant differences in preference were obtained. © 2014 Institute of Food Technologists®

Natural Cold Baryogenesis from Strongly Interacting Electroweak Symmetry Breaking

CERN Document Server

Konstandin, Thomas

2011-01-01

The mechanism of "cold electroweak baryogenesis" has been so far unpopular because its proposal has relied on the ad-hoc assumption of a period of hybrid inflation at the electroweak scale with the Higgs acting as the waterfall field. We argue here that cold baryogenesis can be naturally realized without the need to introduce any slow-roll potential. Our point is that composite Higgs models where electroweak symmetry breaking arises via a strongly first-order phase transition provide a well-motivated framework for cold baryogenesis. In this case, reheating proceeds by bubble collisions and we argue that this can induce changes in Chern-Simons number, which in the presence of new sources of CP violation commonly lead to baryogenesis. We illustrate this mechanism using as a source of CP violation an effective dimension-six operator which is free from EDM constraints, another advantage of cold baryogenesis compared to the standard theory of electroweak baryogenesis. Our results are general as they do not rely on...

Flow regimes and heat transfer modes identification in ANGRA 2 core, during small break in the primary loop with area of 100 cm2, simulated with RELAP5 code

International Nuclear Information System (INIS)

Borges, Eduardo M.; Sabundjian, Gaiane

2015-01-01

Identifying the flow regimes and the heat transfer modes is important for the analysis of accidents such as the Loss-of-Coolant Accident (LOCA). The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 100cm 2 -rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR - A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. (author)

RELAP5/SCDAPSIM model development for AP1000 and verification for large break LOCA

Energy Technology Data Exchange (ETDEWEB)

Trivedi, A.K. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Allison, C. [Innovative Systems Software, Idaho Falls, ID 83406 (United States); Khanna, A., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Munshi, P. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India)

2016-08-15

Highlights: • RELAP5/SCDAPSIM model of AP1000 has been developed. • Analysis involves a LBLOCA (double ended guillotine break) study in cold leg. • Results are compared with those of WCOBRA–TRAC and TRACE. • Concluded that PCT does not violate the safety criteria of 1477 K. - Abstract: The AP1000 is a Westinghouse 2-loop pressurized water reactor (PWR) with all emergency core cooling systems based on natural circulation. Its core design is very similar to a 3-loop PWR with 157 fuel assemblies. Westinghouse has reported their results of the safety analysis in its design control document (DCD) for a large break loss of coolant accident (LOCA) using WCOBRA/TRAC and for a small break LOCA using NOTRUMP. The current study involves the development of a representative RELAP5/SCDASIM model for AP1000 based on publically available data and its verification for a double ended cold leg (DECL) break in one of the cold legs in the loop containing core makeup tanks (CMT). The calculated RELAP5/SCDAPSIM results have been compared to publically available WCOBRA–TRAC and TRACE results of DECL break in AP1000. The objective of this study is to benchmark thermal hydraulic model for later severe accident analyses using the 2D SCDAP fuel rod component in place of the RELAP5 heat structures which currently represent the fuel rods. Results from this comparison provides sufficient confidence in the model which will be used for further studies such as a station blackout. The primary circuit pumps, pressurizer and steam generators (including the necessary secondary side) are modeled using RELAP5 components following all the necessary recommendations for nodalization. The core has been divided into 6 radial rings and 10 axial nodes. For the RELAP5 thermal hydraulic calculation, the six groups of fuel assemblies have been modeled as pipe components with equivalent flow areas. The fuel including the gap and cladding is modeled as a 1d heat structure. The final input deck achieved

Updated TRAC analysis of an 80% double-ended cold-leg break for the AP600 design

International Nuclear Information System (INIS)

Lime, J.F.; Boyack, B.E.

1995-01-01

An updated TRAC 80% large-break loss-of-coolant accident (LBLOCA) has been calculated for the Westinghouse AP600 advanced reactor design, The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The 80% break size was calculated by Westinghouse to be the most severe large-break size for the AP600 design. The LBLOCA transient was calculated to 144 s. Peak cladding temperatures (PCTS) were well below the Appendix K limit of 1,478 K (2,200 F), but very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCT for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown phase. The reasons for these differences are still being investigated. Additional break sizes and break locations need to be analyzed to confirm the most severe break postulated by Westinghouse

Flow regimes and heat transfer modes identification in ANGRA 2 core, during small break in the primary loop with area of 100 cm{sup 2}, simulated with RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Borges, Eduardo M.; Sabundjian, Gaiane, E-mail: gdgian@ipen.br, E-mail: borges.em@hotmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

Identifying the flow regimes and the heat transfer modes is important for the analysis of accidents such as the Loss-of-Coolant Accident (LOCA). The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 100cm{sup 2}-rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR - A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. (author)

Intermediate Leg SBLOCA - Long Lasting Pressure Transient

International Nuclear Information System (INIS)

Konjarek, D.; Bajs, T.; Vukovic, J.

2010-01-01

The basic phenomenology of Small Break Loss of Coolant Accident (SBLOCA) for PWR plant is described with focus on analysis of scenario in which reactor coolant pressure decreases below secondary system pressure. Best estimate light water reactor transient analysis code RELAP5/mod3.3 was used in calculation. Rather detailed model of the plant was used. The break occurs in intermediate leg on lowest elevation near pump suction. The size of the break is chosen to be small enough to cause cycling of safety valves (SVs) on steam generators (SGs) for some time, but, afterwards, it is large enough to remove decay heat through the break, causing cooling the secondary side. In this case of SBLOCA, when primary pressure decreases below secondary pressure, long lasting pressure transients with significant amplitude occur. Reasons for such behavior are explained.(author).

A study on the effect of fluidic device installed in a safety injection tank on thermal-hydraulic phenomena of large break loss of coolant accident

International Nuclear Information System (INIS)

Chung, Young Jong; Bae, Kyoo Hwan; Song, Jin Ho; Sim, Suk Ku; Park, Jong Kyun

1999-03-01

The performance of the Safety Injection Tank (SIT) with fluidic device (advanced SIT) is analyzed for the large break loss of coolant accident (LBLOCA) using RELAP5/MOD3.1-KREM. First the case is analyzed using the conventional SIT. Among various cases the case with 4-split downcomer, discharge coefficient Cd=0.6, MCP trip with reactor trip and break location of cold leg discharge side with the pressurizer is found to be the most limiting case. For the same condition, the advanced SIT results the similar PCT, however it can maintain adequately the liquid level in the downcomer. By changing the ECCS location from the current injection to the cold leg elevations, PCT is improved by 75 K. (Author). 6 refs., 4 tabs., 54 figs

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

Directory of Open Access Journals (Sweden)

Li Yuquan

2017-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-02-15

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

Modelling for great breaks accident analysis in the primary system of Angra 1 reactor

International Nuclear Information System (INIS)

Serrano, M.A.B.; Vanni, E.A.

1981-01-01

An analysis is made for a break in the cold leg, of the guillotine type with discharge coefficient C sub(D)=1.0, for the Angra 1 reactor. The computer codes, geometrical models and options used are described. A comparison between the method used and the requirements in the Appendix K of 10 CRF 50 is done. (Author) [pt

Analysis of cold leg LOCA with failed HPSI by means of integrated safety assessment methodology

International Nuclear Information System (INIS)

Gonzalez-Cadelo, J.; Queral, C.; Montero-Mayorga, J.

2014-01-01

Highlights: • Results of ISA for considered sequences endorse EOPs guidance in an original way. • ISA allows to obtain accurate available times for accident management actions. • RCP-trip adequacy and available time for beginning depressurization are evaluated. • ISA minimizes the necessity of expert judgment to perform safety assessment. - Abstract: The integrated safety assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal–hydraulic analysis of cold leg LOCA sequences with unavailable High Pressure Injection System in a Westinghouse 3-loop PWR. This analysis has been performed with TRACE 5.0 patch 1 code. ISA methodology allows obtaining the Damage Domain (the region of space of parameters where a safety limit is exceeded) as a function of uncertain parameters (break area) and operator actuation times, and provides to the analyst useful information about the impact of these uncertain parameters in safety concerns. In this work two main issues have been analyzed: the effect of reactor coolant pump trip and the available time for beginning of secondary-side depressurization. The main conclusions are that present Emergency Operating Procedures (EOPs) are adequate for managing this kind of sequences and the ISA methodology is able to take into account time delays and parameter uncertainties

Analysis of ATLAS Cold Leg SBLOCA Using SPACE Code

International Nuclear Information System (INIS)

Kang, Doo Hyuk; Suh, Jae Seung; Kim, Se Yun

2012-01-01

SPACE Code has been developed to predict the thermal-hydraulic responses of nuclear steam supply system to the anticipated transients and postulated accidents and adopted advanced physical modeling of two-phase flows, mainly two-fluid, three-field models that comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or non-structured meshes. In this paper, a cold-leg SBLOCA which is the experiment, SB-CL-09, of the ATLAS integral effect test facility during the second domestic stand problem (DSP-02) was analyzed. The results were compared with those of MARS-KS code simulations. The SPACE code with a 1.0 version was released by KHNP in 2012. The analysis has been performed in a desktop PC with Windows 7 environment

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

A synopsis of experimental activities on small-break LOCA

International Nuclear Information System (INIS)

Hein, D.

1984-01-01

Through reactor safety studies like WASH 1400 or the ''Deutsche Risiko-Studie'' the attention has turned from large break loss of coolant accidents to small breaks because of the high contribution of this type of accidents to core meltdown. But only after the TMI-2 accident were also the main activities in the experimental fields shifted world-wide to the small break LOCAs. Since TMI numerous research programs have either been finished or are underway. This review paper presents: a classification of the various types of transients according to break size; a discussion of major physical phenomena associated with a small break LOCA, and a description of a few selected research programs and the most important results achieved. (author)

Natural cold baryogenesis from strongly interacting electroweak symmetry breaking

International Nuclear Information System (INIS)

Konstandin, Thomas; Servant, Géraldine

2011-01-01

The mechanism of ''cold electroweak baryogenesis'' has been so far unpopular because its proposal has relied on the ad-hoc assumption of a period of hybrid inflation at the electroweak scale with the Higgs acting as the waterfall field. We argue here that cold baryogenesis can be naturally realized without the need to introduce any slow-roll potential. Our point is that composite Higgs models where electroweak symmetry breaking arises via a strongly first-order phase transition provide a well-motivated framework for cold baryogenesis. In this case, reheating proceeds by bubble collisions and we argue that this can induce changes in Chern-Simons number, which in the presence of new sources of CP violation commonly lead to baryogenesis. We illustrate this mechanism using as a source of CP violation an effective dimension-six operator which is free from EDM constraints, another advantage of cold baryogenesis compared to the standard theory of electroweak baryogenesis. Our results are general as they do not rely on any particular UV completion but only on a stage of supercooling ended by a first-order phase transition in the evolution of the universe, which can be natural if there is nearly conformal dynamics at the TeV scale. Besides, baryon-number violation originates from the Standard Model only

Small break LOCA analysis for RCP trip strategy for YGN 3 and 4 emergency procedure guidelines

International Nuclear Information System (INIS)

Suh, Jong Tae; Bae, Kyoo Hwan

1995-01-01

A continued operation of RCPs during a certain small break LOCA may increase unnecessary inventory loss from the RCS causing a severe core uncovery which might lead to a fuel failure. After TMI-2 accident, the CEOG developed RCP trip strategy called 'Trip-Two/Leave-Two' (T2/L2) in response to NRC requests and incorporated it in the generic EPG for CE plants. The T2/L2 RCP trip strategy consists of tripping the first two RCPs on low RCS pressure and then tripping the remaining two RCPs if a LOCA has occurred. This analysis determines the RCP trip setpoint and demonstrates the safe operational aspects of RCP trip strategy during a small break LOCA for YGN 3 and 4. The trip setpoint of the first two RCPs for YGN 3 and 4 is calculated to be 1775 psia in pressurizer pressure based on the limiting small break LOCA with 0.15 ft 2 break size in the hot leg. The analysis results show that YGN 3 and 4 can maintain the core coolability even if the operator fails to trip the second two RCPs or trips at worst time. Also, the YGN 3 and 4 RCP trip strategy demonstrates that both the 10 CFR 50.46 requirements on PCT and the ANSI standards 58.8 requirements on operator action time can be satisfied with enough margin. Therefore, it is concluded that the T2/L2 RCP trip strategy with a trip setpoint of 1775 psia for YGN 3 and 4 can provide improved operator guidance for the RCP operation during accidents. 11 figs., 4 tabs., 9 refs. (Author)

Condensation during gravity driven ECC: Experiments with PACTEL

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

This paper provides the results of the second series of gravity driven emergency core cooling (ECC) experiments with PACTEL (Parallel Channel Test Loop). The simulated accident was a small break loss-of-coolant accident (SBLOCA) with a break in a cold leg. The ECC flow was provided from a core makeup tank (CMT) located at a higher elevation than the main part of the primary system. The CMT was pressurized with pipings from the pressurizer and a cold leg. The tests indicated that steam condensation in the CMT can prevent ECC and lead to core uncovery.

Natural cold baryogenesis from strongly interacting electroweak symmetry breaking

Energy Technology Data Exchange (ETDEWEB)

Konstandin, Thomas; Servant, Géraldine, E-mail: tkonstan@cern.ch, E-mail: geraldine.servant@cern.ch [CERN Physics Department, Theory Division, CH-1211 Geneva 23 (Switzerland)

2011-07-01

The mechanism of ''cold electroweak baryogenesis'' has been so far unpopular because its proposal has relied on the ad-hoc assumption of a period of hybrid inflation at the electroweak scale with the Higgs acting as the waterfall field. We argue here that cold baryogenesis can be naturally realized without the need to introduce any slow-roll potential. Our point is that composite Higgs models where electroweak symmetry breaking arises via a strongly first-order phase transition provide a well-motivated framework for cold baryogenesis. In this case, reheating proceeds by bubble collisions and we argue that this can induce changes in Chern-Simons number, which in the presence of new sources of CP violation commonly lead to baryogenesis. We illustrate this mechanism using as a source of CP violation an effective dimension-six operator which is free from EDM constraints, another advantage of cold baryogenesis compared to the standard theory of electroweak baryogenesis. Our results are general as they do not rely on any particular UV completion but only on a stage of supercooling ended by a first-order phase transition in the evolution of the universe, which can be natural if there is nearly conformal dynamics at the TeV scale. Besides, baryon-number violation originates from the Standard Model only.

Effect on temperature of output of the core of the size of the break in the Upper Head of the vessel using TRACE5; Efecto sobre la Temperatura de Salida del Nucleo del Tamano de la Rotura en el Upper Head de la Vasija utilizando TRACE5

Energy Technology Data Exchange (ETDEWEB)

Querol, A.; Gallardo, S.; Verdu, G.

2013-07-01

Most (PWR) pressurized water reactors have thermocouples to detect overheating of the core since they are used to measure the temperature of exit of the nucleus (CET). However, it was found that in a small break (SBLOCA) located in the upper head of the vessel there is a delay between the measure of thermocouples and overheating of the core. This work is based on the simulation, using the code Thermo-hydraulic TRACE5, of the Test 6 - 1 the OECD/NEA rose project carried out in the experimental facility LSTF (Large Scale Test Facility). There have been different analyses in which geometric variables that can influence the model such as the size and location of the break, possible flow towards the break and the nodalization of the upper head of the vessel have been studied.

Investigating temperature breaks in the summer fruit export cold chain: A case study

Directory of Open Access Journals (Sweden)

Heinri W. Freiboth

2013-11-01

Full Text Available There is concern in the South African fruit industry that a large amount of fruit and money is lost every season due to breaks in the fruit export cold chain. The possibility of a large percentage of losses in a significant sector of the economy warranted further investigation. This article attempted to highlight some of the possible problem areas in the cold chain, from the cold store to the port, by analysing historic temperature data from different fruit export supply chains of apples, pears and grapes. In addition, a trial shipment of apples was used to investigate temperature variation between different pallets in the same container. This research has added value to the South African fruit industry by identifying the need to improve operational procedures in the cold chain.

Horizontal stratified flow model for the 1-D module of WCOBRA/TRAC-TF2: modeling and validation

Energy Technology Data Exchange (ETDEWEB)

Liao, J.; Frepoli, C.; Ohkawa, K., E-mail: liaoj@westinghouse.com [Westinghouse Electric Company LLC, LOCA Integrated Services I, Cranberry Twp, Pennsylvania (United States)

2011-07-01

For a two-phase flow in a horizontal pipe, the individual phases may separate by gravity. This horizontal stratification significantly impacts the interfacial drag, interfacial heat transfer and wall drag of the two phase flow. For a PWR small break LOCA, the horizontal stratification in cold legs is a highly important phenomenon during loop seal clearance, boiloff and recovery periods. The low interfacial drag in the stratified flow directly controls the time period for the loop clearance and the level of residual water in the loop seal. Horizontal stratification in hot legs also impacts the natural circulation stage of a small break LOCA. In addition, the offtake phenomenon and cold leg condensation phenomenon are also affected by the occurrence of horizontal stratification in the cold legs. In the 1-D module of the WCOBRA/TRAC-TF2 computer code, a horizontal stratification criterion was developed by combining the Taitel-Dukler model and the Wallis-Dobson model, which approximates the viscous Kelvin-Helmholtz neutral stability boundary. The objective of this paper is to present the horizontal stratification model implemented in the code and its assessment against relevant data. The adequacy of the horizontal stratification transition criterion is confirmed by examining the code-predicted flow regime in a horizontal pipe with the measured data in the flow regime map. The void fractions (or liquid level) for the horizontal stratified flow in cold leg or hot leg are predicted with a reasonable accuracy. (author)

Feasibility study of applying the passive safety system concept to fusion–fission hybrid reactor

International Nuclear Information System (INIS)

Yu, Zhang-cheng; Xie, Heng

2014-01-01

The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs

Investigating temperature breaks in the summer fruit export cold chain - a case study

CSIR Research Space (South Africa)

Freiboth, HW

2013-11-01

Full Text Available There is concern in the South African fruit industry that a large amount of fruit and money is lost every season due to breaks in the fruit export cold chain. The possibility of a large percentage of losses in a significant sector of the economy...

Stratification studies in components of nuclear power plants

International Nuclear Information System (INIS)

Randorf, J.A.

1997-01-01

The applicability of two stratification criteria during loss-of-coolant (LOCA) conditions was studied. The first criteria was developed for addressing cold water injection-induced stratification. The second criteria applied to downcomer/cold leg junction stratification. Both criteria provided predictions consistent with measured conditions during small break loss-of-coolant tests

Thermal hydraulic research on next generation PWRs using ROSA/LSTF

International Nuclear Information System (INIS)

Yonomoto, T.; Anoda, Y.

2000-01-01

A thermal-hydraulic research on next generation PWRs has been conducted at JAERI using the ROSA-V/Large Scale Test Facility (LSTF), focusing on phenomena related to passive safety systems. This paper describes two test results conducted for this research: a small break loss-of-coolant accident (SBLOCA) test and a low pressure steady-state natural circulation (NC) test. The former test investigated a combined use of a SG secondary-side automatic depressurization system (SADS) and a gravity-driven injection system (GDIS) to mitigate a SBLOCA. The results have shown that the primary loop can be depressurized to the GDIS actuation pressure of 0.2 MPa by the SADS alone, and then the stable long-term core cooling can be established by NC. Results of both tests showed a complicated nonuniform flow behavior among SG U-tubes during NC, which was characterized by the coexistence of concurrent condensing two-phase flow in some tubes and stagnant two-phase stratification in the others. The mechanism for the stratification was understood from the measured secondary side temperature distribution showing the lowest temperature at the top and bottom regions and the highest around the midplane. This was caused by the saturation temperature difference corresponding to the static pressure difference, and the recirculation in the secondary. This secondary side temperature distribution enabled the condensation occurring around the tube top to be balanced with evaporation occurring around the midplane in the U-tube with the stratification. Since the heat transfer occurs primarily through tubes with the concurrent flow, the nonuniform behavior directly affects the effective heat transfer area at SG. When the SG primary side was modeled with one lumped flow channel, the RELAP5 significantly over predicted the primary depressurization rate, and could not predict the stable long-term core cooling behavior at low pressure. In order to understand the mechanism of the nonuniform behavior, the

Thermal-Hydraulic Integral Effect Test with ATLAS for an Intermediate Break Loss of Coolant Accident at a Pressurizer Surge Line

Energy Technology Data Exchange (ETDEWEB)

Kang, Kyoung Ho; Seok Cho; Park, Hyun Sik; Choi, Nam Hyun; Park, Yu Sun; Kim, Jong Rok; Bae, Byoung Uhn; Kim, Yeon Sik; Kim, Kyung Doo; Choi, Ki Yong; Song, Chul Hwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

The main objectives of this test were not only to provide physical insight into the system response of the APR1400 during the pressurizer surge line break accident but also to produce an integral effect test data to validate the SPACE code. In order to simulate a double-ended guillotine break of a pressurizer surge line in the APR1400, the IB-SUR-01R test was performed with ATLAS. The major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Despite the core was uncovered, no excursion in the cladding temperature was observed. The pressurizer surge line break can be classified as a hot leg break from a break location point of view. Compared with a cold leg break, coolability in the core may be better in case of a hot leg break due to the enhanced flow in the core region. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for an IBLOCA simulation, especially for DVI-adapted plants. Redefinition of break size for design basis accident (DBA) based on risk information is being extensively investigated due to the potential for safety benefits and unnecessary burden reduction from current LBLOCA (large break loss of coolant accident)-based ECC (Emergency Core Cooling) Acceptance Criteria. As a transition break size (TBS), the rupture of medium-size pipe is considered to be more important than ever in risk-informed regulation (RIR)-relevant safety analysis. As plants age, are up-rated, and continue to seek improved operating efficiencies, the small break and intermediate break LOCA (IBLOCA) can become a concern. In particular, IBLOCA with DVI (Direct Vessel Injection) features will be addressed to support redefinition of a design-basis LOCA. With an aim of expanding code validation to address small

RELAP5 simulation of a large break Loss of Coolant Accident (LOCA) in the hot leg of the primary system in Angra 2 nuclear power plant

International Nuclear Information System (INIS)

Andrade, Delvonei Alves de; Sabundjian, Gaiane

2004-01-01

The objective of this work is to present the simulation of a large break loss of coolant accident - LBLOCA in the hot leg of the primary loop in Angra 2, with RELAP5/MOD3.2.2g code. This accident is described in the Final Safety Report Analysis of Angra 2 - FSAR and consists basically of the hot leg total break, in loop 20 of the plant. The area considered for the rupture is 4480 cm 2 , which corresponds to 100% of the pipe flow area. Besides, this work also has the objective of verifying the efficiency of the emergency core coolant system - ECCS in case of accidents and transients. The thermal-hydraulic processes inherent to the accident phenomenology, such as hot leg vaporization and consequently core vaporization causing an inappropriate flow distribution in the reactor core, can lead to a reduction in the liquid level, until the ECCS is capable to reflood it

Simulation and Analysis of Small Break LOCA for AP1000 Using RELAP5-MV and Its Comparison with NOTRUMP Code

Directory of Open Access Journals (Sweden)

Eltayeb Yousif

2017-01-01

Full Text Available Many reactor safety simulation codes for nuclear power plants (NPPs have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions. In reactor analysis, small break loss of coolant accident (SBLOCA is an important safety issue. RELAP5-MV Visualized Modularization software is recognized as one of the best estimate transient simulation programs of light water reactors (LWR. RELAP5-MV has new options for improved modeling methods and interactive graphics display. Though the same models incorporated in RELAP5/MOD 4.0 are in RELAP5-MV, the significant difference of the latter is the interface for preparing the input deck. In this paper, RELAP5-MV is applied for the transient analysis of the primary system variation of thermal hydraulics parameters in primary loop under SBLOCA in AP1000 NPP. The upper limit of SBLOCA (10 inches is simulated in the cold leg of the reactor and the calculations performed up to a transient time of 450,000.0 s. The results obtained from RELAP5-MV are in good agreement with those of NOTRUMP code obtained by Westinghouse when compared under the same conditions. It can be easily inferred that RELAP5-MV, in a similar manner to RELAP5/MOD4.0, is suitable for simulating a SBLOCA scenario.

OECD-LOFT large break LOCA experiments: phenomenology and computer code analyses

International Nuclear Information System (INIS)

Brittain, I.; Aksan, S.N.

1990-08-01

Large break LOCA data from LOFT are a very important part of the world database. This paper describes the two double-ended cold leg break tests LP-02-6 and LP-LB-1 carried out within the OECD-LOFT Programme. Tests in LOFT were the first to show the importance of both bottom-up and top-down quenching during blowdown in removing stored energy from the fuel. These phenomena are discussed in detail, together with the related topics of the thermal performance of nuclear fuel and its simulation by electric fuel rod simulators, and the accuracy of cladding external thermocouples. The LOFT data are particularly important in the validation of integral thermal-hydraulics codes such as TRAC and RELAP5. Several OECD partner countries contributed analyses of the large break tests. Results of these analyses are summarised and some conclusions drawn. 32 figs., 3 tabs., 45 refs

Evaluation of the effects of break nozzle configuration in the Semiscale Mod-1 system

International Nuclear Information System (INIS)

Hanson, R.G.

1977-08-01

The Semiscale Mod-1 Program has utilized two different break nozzle configurations in the test system. An evaluation has been made to determine the effect these break nozzle configurations have on system thermal-hydraulic response during a 200 percent double-ended cold leg break loss-of-coolant accident simulation. The first nozzle was a convergent-divergent nozzle (Henry nozzle) and the second, an elongated constant area throat nozzle. Analysis is confined primarily to system response phenomena observed to be affected by the nozzle configuration and concentrates on the fluid response at the break and the resulting core behavior during subcooled and saturated blowdown. The evaluation shows that considerable difference in system response occurs as a result of the difference in break nozzle configuration. The elongated throat nozzle was scaled from the Loss-of-Fluid Test (LOFT) nozzle geometry and since the LOFT counterpart tests were designed to provide results for the LOFT Program, the elongated throat nozzle was used in the subsequent LOFT counterpart tests

Packhouse to port: Investigating temperature breaks in the South African summer fruit export cold chain

CSIR Research Space (South Africa)

Freiboth, H

2014-10-01

Full Text Available A large amount of fruit and money is lost every season due to breaks in the South African fruit export cold chain. With food security becoming an ever growing concern, especially in developing countries like South Africa, a high percentage of losses...

Evaluation of a loss of residual heat removal event during mid-loop operation

International Nuclear Information System (INIS)

Seul, Kwang Won; Bang, Young Seok; Lee, Sukho; Kim, Hho Jung

1996-01-01

The potential for the RELAP5/MOD3.2 was assessed for the loss-of -RHR event during the mid-loop operation and the predictability of major thermal-hydraulic phenomena was also evaluated for the long term transient. The analysis results of the typical two cases(cold leg opening case and pressurizer opening case) were compared with experimental data which was conducted at ROSA-IV/LSTF in Japan. As a result, it was shown that the code was capable of simulating the thermal-hydraulic transport process with appropriate time step during the reduced inventory operation with the loss-of-RHR system

Assessment of RELAP/MOD2 using large break loss-of-coolant experimental data

International Nuclear Information System (INIS)

Kao, L.; Liao, L.Y.; Liang, K.S.; Wang, S.F.; Chen, Y.B.

1989-01-01

In this paper assessment of RELAP5/MOD2 using LOFT L2-5 and Semiscale S-06-3 tests are performed to provide information of the code capability and its limitation in analyzing large break LOCA of a nuclear power plant. Experiments L2-5 and S-06-3 are conducted to simulate a hypothetical LOCA which results from a 200% double-ended offset shear break in the cold-leg of a typical pressurized water reactor by utilizing scaling facilities of the LOFT and Semiscale Mod-1 systems, respectively. The RELAP5/MOD2 calculations for both tests begin with break initiation and subsequent blowdown, continue through lower plenum refill, core reflood, and terminate with corewide quench. Major phenomena of both large break loss-of-coolant tests are well predicted by RELAP5/MOD2. The results indicate that the break flow and system pressure are reasonably calculated. The cladding temperature response during blowdown period, which is the major importance to a large break LOCA, calculated by RELAP5/MOD2 shows good agreement with the test data

Experimental study on fundamental phenomena in HTGR small break air-ingress accident

International Nuclear Information System (INIS)

Kim, Jae Soon; Hwang, Jin-Seok; Kim, Eung Soo; Kim, Byung Jun; Oh, Chang Ho

2016-01-01

Highlights: • Air-ingress phenomena on the small break in a HTGR are experimentally investigated. • Experiment is investigated for various break sizes, angles, and density ratios. • Maximum air-ingress rate is observed at 120° in break angle. • This study reveals that air-ingress in the small break is governed by; buoyancy and flow inertia. • A non-dimensional parameter is newly proposed to determine the air-ingress flow regimes. • Newly proposed parameter is based on buoyancy versus inertia force. - Abstract: This study experimentally investigates fundamental phenomena in the HTGR small break air-ingress accident. Several important parameters including density ratio, break angle, break size, and main flow velocity are considered in the measurement and the analysis. The test-section is made of a circular pipe with small holes drilled around the surface and it is installed in the helium/air flow circulation loop. Oxygen concentrations and flow rates are recorded during the tests with fixed break angles, break sizes, and flow velocities for measurement of the air-ingress rates. According to the experimental results, the higher density difference leads to the higher rates of air-ingress with large sensitivity of the break angles. It is also found that the break angle significantly affects the air-ingress rates, which is gradually increased from 0° to 120° and suddenly decreased to 180°. The minimum air ingress rate is found at 0° and the maximum, at 110°. The air-ingress rate increases with the break size due to the increased flow-exchange area. However, it is not directly proportional to the break area due to the complexity of the phenomena. The increased flow velocity in the channel inside enhances the air-ingress process. However, among all the parameters, the main flow velocity exhibits the lowest effect on this process. In this study, the Froude Number relevant to the small break air-ingress conditions are newly defined considering both heavy

Post accidental small breaks analysis

International Nuclear Information System (INIS)

Depond, G.; Gandrille, J.

1980-04-01

EDF ordered to FRAMATOME by 1977 to complete post accidental long term studies on 'First Contrat-Programme' reactors, in order to demonstrate the safety criteria long term compliance, to get information on NSSS behaviour and to improve the post accidental procedures. Convenient analytical models were needed and EDF and FRAMATOME respectively developped the AXEL and FRARELAP codes. The main results of these studies is that for the smallest breaks, it is possible to manually undertake cooling and pressure reducing actions by dumping the steam generators secondary side in order to meet the RHR operating specifications and perform long term cooling through this system. A specific small breaks procedure was written on this basis. The EDF and FRAMATOME codes are continuously improved; the results of a French set of separate effects experiments will be incorporated as well as integral system verification

Updated pipe break analysis for Advanced Neutron Source Reactor conceptual design

International Nuclear Information System (INIS)

Wendel, M.W.; Chen, N.C.J.; Yoder, G.L.

1994-01-01

The Advanced Neutron Source Reactor (ANSR) is a research reactor to be built at the Oak Ridge National Laboratory that will supply the highest continuous neutron flux levels of any reactor in the world. It uses plate-type fuel with high-mass-flux and highly subcooled heavy water as the primary coolant. The Conceptual Safety Analysis for the ANSR was completed in June 1992. The thermal-hydraulic pipe-break safety analysis (performed with a specialized version of RELAP5/MOD3) focused primarily on double-ended guillotine breaks of the primary piping and some core-damage mitigation options for such an event. Smaller, instantaneous pipe breaks in the cold- and hot-leg piping were also analyzed to a limited extent. Since the initial analysis for the conceptual design was completed, several important changes to the RELAP5 input model have been made reflecting improvements in the fuel grading and changes in the elevation of the primary coolant pumps. Also, a new philosophy for pipe-break safety analysis (similar to that adopted for the New Production Reactor) accentuates instantaneous, limited flow area pipe-break accidents in addition to finite-opening-time, double-ended guillotine breaks of the major coolant piping. This paper discloses the results of the most recent instantaneous pipe-break calculations

Arm to leg coordination in elite butterfly swimmers.

Science.gov (United States)

Chollet, D; Seifert, L; Boulesteix, L; Carter, M

2006-04-01

This study proposed the use of four time gaps to assess arm-to-leg coordination in the butterfly stroke at increasing race paces. Fourteen elite male swimmers swam at four velocities corresponding to the appropriate paces for, respectively, the 400-m, 200-m, 100-m, and 50-m events. The different stroke phases of the arm and leg were identified by video analysis and then used to calculate four time gaps (T1: time gap between entry of the hands in the water and the high break-even point of the first undulation; T2: time gap between the beginning of the hands' backward movement and the low break-even point of the first undulation; T3: time gap between the hands' arrival in a vertical plane to the shoulders and the high break-even point of the second undulation; T4: time gap between the hands' release from the water and the low break-even point of the second undulation), the values of which described the changing relationship of arm to leg movements over an entire stroke cycle. With increases in pace, elite swimmers increased the stroke rate, the relative duration of the arm pull, the recovery and the first downward movement of the legs, and decreased the stroke length, the relative duration of the arm catch phase and the body glide with arms forward (measured by T2), until continuity in the propulsive actions was achieved. Whatever the paces, the T1, T3, and T4 values were close to zero and revealed a high degree of synchronisation at key motor points of the arm and leg actions. This new method to assess butterfly coordination could facilitate learning and coaching by situating the place of the leg undulation in relation with the arm stroke.

Analysis of the SBLOCAs in HANARO pool for the 3-pin fuel test loop

International Nuclear Information System (INIS)

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

2004-09-01

Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Small Break Loss Of Coolant Accidents (SBLOCAs) in HANARO pool for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the SBLOCAs. The location of the pipe break is assumed at the hill taps connecting the cold and hot legs in HANARO pool to the inlet and outlet nozzles of the In-Pile test Section (IPS). The break size is also assumed less than 20% of the cross section area of the pipe. The test fuels are heated up when the cold leg break occur. However, they are not heated up when the hot leg break occur. The maximum Peak Cladding Temperatures (PCT) are predicted to be about 906.9 .deg. C for the cold leg break accident in PWR fuel test mode and 971.9 .deg. C in CANDU fuel test mode respectively. The critical break size is about the 6% of the cross section area of the pipe for PWR fuel test mode and the 8% for CANDU fuel test mode. The PCTs meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C

Don't break a leg: running birds from quail to ostrich prioritise leg safety and economy on uneven terrain.

Science.gov (United States)

Birn-Jeffery, Aleksandra V; Hubicki, Christian M; Blum, Yvonne; Renjewski, Daniel; Hurst, Jonathan W; Daley, Monica A

2014-11-01

Cursorial ground birds are paragons of bipedal running that span a 500-fold mass range from quail to ostrich. Here we investigate the task-level control priorities of cursorial birds by analysing how they negotiate single-step obstacles that create a conflict between body stability (attenuating deviations in body motion) and consistent leg force-length dynamics (for economy and leg safety). We also test the hypothesis that control priorities shift between body stability and leg safety with increasing body size, reflecting use of active control to overcome size-related challenges. Weight-support demands lead to a shift towards straighter legs and stiffer steady gait with increasing body size, but it remains unknown whether non-steady locomotor priorities diverge with size. We found that all measured species used a consistent obstacle negotiation strategy, involving unsteady body dynamics to minimise fluctuations in leg posture and loading across multiple steps, not directly prioritising body stability. Peak leg forces remained remarkably consistent across obstacle terrain, within 0.35 body weights of level running for obstacle heights from 0.1 to 0.5 times leg length. All species used similar stance leg actuation patterns, involving asymmetric force-length trajectories and posture-dependent actuation to add or remove energy depending on landing conditions. We present a simple stance leg model that explains key features of avian bipedal locomotion, and suggests economy as a key priority on both level and uneven terrain. We suggest that running ground birds target the closely coupled priorities of economy and leg safety as the direct imperatives of control, with adequate stability achieved through appropriately tuned intrinsic dynamics. © 2014. Published by The Company of Biologists Ltd.

Simulation of a main steam line break with steam generator tube rupture using trace

Energy Technology Data Exchange (ETDEWEB)

Gallardo, S.; Querol, A.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Camino de Vera s/n, 46022, Valencia (Spain)

2012-07-01

A simulation of the OECD/NEA ROSA-2 Project Test 5 was made with the thermal-hydraulic code TRACE5. Test 5 performed in the Large Scale Test Facility (LSTF) reproduced a Main Steam Line Break (MSLB) with a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR). The result of these simultaneous breaks is a depressurization in the secondary and primary system in loop B because both systems are connected through the SGTR. Good approximation was obtained between TRACE5 results and experimental data. TRACE5 reproduces qualitatively the phenomena that occur in this transient: primary pressure falls after the break, stagnation of the pressure after the opening of the relief valve of the intact steam generator, the pressure falls after the two openings of the PORV and the recovery of the liquid level in the pressurizer after each closure of the PORV. Furthermore, a sensitivity analysis has been performed to know the effect of varying the High Pressure Injection (HPI) flow rate in both loops on the system pressures evolution. (authors)

Review on research of small break loss of coolant accident

International Nuclear Information System (INIS)

Bo Jinhai; Wang Fei

1998-01-01

The Small Break Loss of Coolant Accident (SBLOCA) and its research art-of -work are reviewed. A typical SBLOCA process in Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) and the influence of break size, break location and reactor coolant pump on the process are described. The existing papers are classified in two categories: experimental and numerical modeling, with the primary experimental apparatuses in the world listed and the research works on SBLOCA summarized

Evaluation of ATLAS 100% DVI Line Break Using TRACE Code

International Nuclear Information System (INIS)

Huh, Byung Gil; Bang, Young Seok; Cheong, Ae Ju; Woo, Sweng Woong

2011-01-01

ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation) is an integral effect test facility in KAERI. It had installed completely to simulate the accident for the OPR1000 and the APR1400 in 2005. After then, several tests for LBLOCA, DVI line break have been performed successfully to resolve the safety issues of the APR1400. Especially, a DVI line break is considered as another spectrum among the SBLOCAs in APR1400 because the DVI line is directly connected to the reactor vessel and the thermal hydraulic behaviors are expected to be different from those for the cold leg injection. However, there are not enough experimental data for the DVI line break. Therefore, integral effect data for the DVI line break of ATLAS is very useful and available for an improvement and validation of safety codes. For the DVI line break in ATLAS, several analyses using MARS and RELAP codes were performed in the ATLAS DSP (Domestic Standard Problem) meetings. However, TRACE code has still not used to simulate a DVI line break. TRACE code has developed as the unified code for the reactor thermal hydraulic analyses in USNRC. In this study, the 100% DVI line break in ATLAS was evaluated by TRACE code. The objectives of this study are to identify the prediction capability of TRACE code for the major thermal hydraulic phenomena of a DVI line break in ATLAS

A simple blowdown code for SUPER-SARA loop conditions

International Nuclear Information System (INIS)

Fritz, G.

1981-01-01

The Super Sara test programme (SSTP) is aimed to study in pile the fuel and cluster behaviour under two types of accident conditions: - the ''Large break loss of coolant'' condition (LB-Loca), - the ''Severe fuel damage'' (SFD) in a boildown caused by a small break. BIVOL was developed for the LB-Loca situation. This code is made for a loop where essentially two volumes define the thermohydraulics during the blowdown. In the SUPERSARA loop these two volumes are represented by the hot leg and cold leg pipings together with the respective upper and lower plenum of the test section

Small break LOCA analysis for YGN 5 and 6 RCP trip strategy in power mode operation

International Nuclear Information System (INIS)

Kim, Tech Mo; Choi, Han Rim

2001-01-01

A continued operation of Reactor Coolant Pumps(RCPs) during a Small Break Loss of Coolant Accident(SBLOCA) in all operation mode may increase unnecessary inventory loss from the Reactor Coolant System(RCS) causing a severe core uncovery which might lead to fuel failure. After Three Mile Island Unit 2(TMI-2) accident, the Combustion Engineering Owner Group(CEOG) developed RCP trip strategy called 'Trip-Two/Leave-Two' (T2/L2). The T2/L2 RCP trip strategy consists of tripping the first two RCPs on low RCS pressure and then tripping the remaining two RCPs if a LOCA has occurred. This analysis demonstrates the inherent safety of RCP trip strategy during an SBLOCA for Youggwang Nuclear Power Plant Unit 5 and 6(YGN 5 and 6). The trip setpoint of the first two RCPs for YGN 5 and 6 is calculated to be 1721 psia in pressurizer pressure based on the limiting SBLOCA with 0.15 ft 2 break size in the hot leg. The analysis results show that YGN 5 and 6 can maintain the core coolability even if the operator fails to trip the second two RCPs or trips at the worst time of minimum liquid inventory

Scaling effects concerning the analysis of small break experiments

International Nuclear Information System (INIS)

Austregesilo Filho, H.

1985-01-01

Some scaling effects related to the experimental facilities as well as to the analytical models used for the design and safety analysis of nuclear power plants are discussed or the basis of phenomena expected to occur during small-break loss - of - coolant accidents. The results of isolated small-break experiments should not be directly extrapolated to the safety analysis of commercial reactors, due to the scaling distortions inherent to the test facilities. With respect to the analytical models used to simulate thermohydraulic processes in experimental facilities, their eventual dependence relative to the system dimension should be examined in order to assess their applicability to the safety analysis of commercial power plants. (Author) [pt

Operator reliability analysis during NPP small break LOCA

International Nuclear Information System (INIS)

Zhang Jiong; Chen Shenglin

1990-01-01

To assess the human factor characteristic of a NPP main control room (MCR) design, the MCR operator reliability during a small break LOCA is analyzed, and some approaches for improving the MCR operator reliability are proposed based on the analyzing results

Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS line using MELCOR code

Energy Technology Data Exchange (ETDEWEB)

Lee, Seung Min; Sabundjian, Gaianê, E-mail: smlee@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2017-11-01

The aim of this work was to simulate a severe accident at a typical PWR caused by break in Emergency Core Cooling System (ECCS) line of a hot leg using the MELCOR code. The nodalization of this typical PWR was elaborated by the Global Research for Safety (GRS) and provided to the CNEN for analysis of the severe accidents at the Angra 2, which is similar to that PWR. Although both of them are not identical the results obtained for that typical PWR may be valuable because of the lack of officially published calculation for Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes after the break in the hot leg were calculated as well as degree of core degradation and hydrogen concentration in containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management (SAMG) by adding associated conditions for each measure in its input. (author)

Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS line using MELCOR code

International Nuclear Information System (INIS)

Lee, Seung Min; Sabundjian, Gaianê

2017-01-01

The aim of this work was to simulate a severe accident at a typical PWR caused by break in Emergency Core Cooling System (ECCS) line of a hot leg using the MELCOR code. The nodalization of this typical PWR was elaborated by the Global Research for Safety (GRS) and provided to the CNEN for analysis of the severe accidents at the Angra 2, which is similar to that PWR. Although both of them are not identical the results obtained for that typical PWR may be valuable because of the lack of officially published calculation for Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes after the break in the hot leg were calculated as well as degree of core degradation and hydrogen concentration in containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management (SAMG) by adding associated conditions for each measure in its input. (author)

FIST small break accident analysis with BWR TRACBO2-pretest predictions

International Nuclear Information System (INIS)

Alamgir, M.; Sutherland, W.A.

1984-01-01

The BWR Full Integral Simulation Test (FIST) program includes experimental simulation and analytical evaluation of BWR thermal-hydraulic phenomena during transient events. One such event is a small size break in the suction line of one of the recirculation pumps. The results from a test simulating this transient in the FIST facility are compared with a system analysis using the Transient Reactor Analysis Code (TRACB02). This comparison demonstrates BWR-TRAC capability for small break analyses and provides detailed understanding of the phenomena

Numerical modelling of the processes in the WWER-1000 containment building during cold leg LOCA using the CONTEMPT-LT/026 code

International Nuclear Information System (INIS)

Kolev, N.I.; Sybotinov, L.S.

1984-01-01

The CONTEMPT-LT/026 code has been used to produce numerical results for the processes in a WWER-1000 containment building during cold leg LOCA with break at the reactor vessel. The objective of the analysis is to estimate the maximal loads on the containment in case of LOCA. Available design data for the geometry and for the operational characteristics of the low-pressure ECC system and the sprinkler system have been used. Boundary conditions such as mass flow and enthalpies at the breach are given by a RELAP4/MOD6 computation. Hydrogen explosions in the containment are not considered. It is found that in case of normal functioning of the low-pressure ECC system the maximal pressure is 3,26±0,44 bar. In the case of malfunctioning of the low-pressure ECC system, the predicted maximal pressure is 4±0,44 bar, when: a) only 50% of the heat transfer surface of the heat exchanger is effectively used due to pollution; b) the main pipeline of the sprinkler is broken; c) the pipeline to the heat exchanger is partially broken so that the mass flow through the exchanger is only 50% of the nominal; and d) ECC low-pressure ECC system attains its maximal efficiency within 3 min, the predicted maximal pressure is 4±0,44 bar

Predictions of stratification in cold leg components using virtual noding schemes

International Nuclear Information System (INIS)

Piper, R.B.; Hassan, Y.A.; Banerjee, S.S.; Barsamian, H.R.; Cebull, P.P.

1996-01-01

In this investigation, a virtual noding scheme is used with RELAP5/MOD3.2 to capture thermal stratification effects in a small-break loss-of-coolant accident (LOCA) simulation. A three-dimensional code (CFD-ACE) has also been used to observe the stratification effects in a similar situation. Stratification temperature differences of the simulations compare well with that of the experiment. The Froude number was also evaluated

Leg blood flow is impaired during small muscle mass exercise in patients with COPD

DEFF Research Database (Denmark)

Iepsen, Ulrik Winning; Munch, Gregers Druedal Wibe; Rugbjerg, Mette

2017-01-01

to both endothelium-independent (SNP) and endothelium-dependent (ACh) stimulation. The results suggests that leg muscle blood flow is impaired during small muscle mass exercise in patients with COPD possibly due to impaired formation of prostacyclin and increased levels of endothelin-1.......Skeletal muscle blood flow is regulated to match the oxygen demand and dysregulation could contribute to exercise intolerance in patients with COPD. We measured leg hemodynamics and metabolites from vasoactive compounds in muscle interstitial fluid and plasma at rest, during one-legged knee...... the formation of interstitial prostacyclin (vasodilator) was only increased in the controls. There was no difference between groups in the nitrite/nitrate levels (vasodilator) in plasma or interstitial fluid during exercise. Moreover, patients and controls showed similar vasodilatory capacity in response...

ROSA-II test data report, 10

International Nuclear Information System (INIS)

1977-12-01

Results of the ROSA-II test simulating a loss-of coolant accident (LOCA) in a light water reactor (LWR) are presented, including the test conditions and interpretation of the phenomena for test runs 415, 417, 421 and 422. Even in small break at the cold leg, the core is exposed to void and the temperature rises. In small break of the hot leg, however, core cooling keeps without temperature rise, because there still remains much residual water and upward core flow exists. Direct effect of the HPCI on the depressurization rate is small, but it increases the accumulator injection rate, leading to early core reflooding and early core cooling from upward. Effects of the secondary system depressurization are increase of depressurization and discharge rates of the primary loop, which results in early initiation of the accumulator injection and core reflooding. (auth.)

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)

Simulation of small break loss of coolant accident using relap 5/ MOD 2 computer code

International Nuclear Information System (INIS)

Megahed, M.M.

1992-01-01

An assessment of relap 5 / MOD 2/Cycle 36.05 best estimate computer code capabilities in predicting the thermohydraulic response of a PWR following a small break loss of coolant accident is presented. The experimental data base for the evaluation is the results of Test S-N H-3 performed in the semi scale MOD-2 c Test facility which modeled a 0.5% small break loss of coolant accident with an accompanying failure of the high pressure injection emergency core cooling system. A conclusion was reached that the code is capable of making small break loss of coolant accident calculations efficiently. However, some of the small break loss of coolant accident related phenomena were not properly predicted by the code, suggesting a need for code improvement.9 fig., 3 tab

BEACON/MOD2A analysis of the Arkansas-1 reactor cavity during a hypothetical hot leg break

International Nuclear Information System (INIS)

Ramsthaler, J.A.

1979-01-01

As part of the evaluation of the new MOD2A version of the BEACON code, the Arkansas-1 reactor cavity was modeled during a hypothetical loss-of-coolant accident. Results of the BEACON analysis were compared with results obtained previously with the COMPARE containment code. Studies were also made investigating some of the BEACON interphasic, timestep control, and wall heat transfer options to assure that these models were working properly and to observe their effects on the results. Descriptions of the Arkansas-1 reactor cavity, initial assumptions during the hypothetical LOCA, and methods of modeling with BEACON are presented. Some of the problems encountered in accurately modeling the penetrations surrounding the hot and cold leg pipes are also discussed

Small accelerator-based pulsed cold neutron sources

International Nuclear Information System (INIS)

Lanza, Richard C.

1997-09-01

Small neutron sources could be used by individual researchers with the convenience of an adequate local facility. Although these sources would produce lower fluxes than the national facilities, for selected applications, the convenience and availability may overcome the limitations on source strength. Such sources might also be useful for preliminary testing of ideas before going to a larger facility. Recent developments in small, high-current pulsed accelerators makes possible such a local source for pulsed cold neutrons.

Pressure loss characteristics of LSTF steam generator heat-transfer tubes. Pressure loss increase due to tube internal instruments

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro

1994-11-01

The steam generator of the Large-Scale Test Facility (LSTF) includes 141 heat-transfer U-tubes with different lengths. Six U-tubes among them are furnished with 15 or 17 probe-type instruments (conduction probe with a thermocouple; CPT) protuberant into the primary side of the U-tubes. Other 135 U-tubes are not instrumented. This results in different hydraulic conditions between the instrumented and non-instrumented U-tubes with the same length. A series of pressure loss characteristics tests was conducted at a test apparatus simulating both types of U-tube. The following pressure loss coefficient (K CPT ) was reduced as a function of Reynolds number (Re) from these tests under single-phase water flow conditions. K CPT =0.16 5600≤Re≤52820, K CPT =60.66xRe -0.688 2420≤Re≤5600, K CPT =2.664x10 6 Re -2.06 1371≤Re≤2420. The maximum uncertainty is 22%. By using these results, the total pressure loss coefficients of full length U-tubes were estimated. It is clarified that the total pressure loss of the shortest instrumented U-tube is equivalent to that of the middle-length non-instrumented U-tube and also that a middle-length instrumented U-tube is equivalent to the longest non-instrumented U-tube. Concludingly. it is important to take account of the CPT pressure loss mentioned above in estimation of fluid behavior at the non-instrumented U-tubes either by using the LSTF experiment data from the CPT-installed U-tubes or by using any analytical codes. (author)

Symmetry breaking in small rotating clouds of trapped ultracold Bose atoms

International Nuclear Information System (INIS)

Dagnino, D.; Barberan, N.; Riera, A.; Osterloh, K.; Lewenstein, M.

2007-01-01

We study the signatures of rotational and phase symmetry breaking in small rotating clouds of trapped ultracold Bose atoms by looking at rigorously defined condensate wave function. Rotational symmetry breaking occurs in narrow frequency windows, where energy degeneracy between the lowest energy states of different total angular momentum takes place. This leads to a complex condensate wave function that exhibits vortices clearly seen as holes in the density, as well as characteristic local phase patterns, reflecting the appearance of vorticities. Phase symmetry (or gauge symmetry) breaking, on the other hand, is clearly manifested in the interference of two independent rotating clouds

Technical manual for COMET

International Nuclear Information System (INIS)

Song, Jin Ho; Kwon, Young Min; Kim, Taek Mo; Lee, Sang Jong; Jeong, Hae Yong

1996-07-01

The purpose of this report is to provide a description for a COMET computer code which is to be used in the analysis of mass and energy releases during post-blowdown phase of LOCA. The mass and energy data re to be used as input data for the containment functional design. This report contains a brief description of analytical models and guidelines for the usage of the computer code. This computer code is to be used for both cold leg and hot leg break analyses. A verification analyses are performed for Ulchin 3 and 4 cold and hot leg break. 11 figs (Author)

Relationship of the Cold-Heat Sensation of the Limbs and Abdomen with Physiological Biomarkers.

Science.gov (United States)

Pham, Duong Duc; Lee, JeongHoon; Kim, GaYul; Song, JiYeon; Kim, JiEun; Leem, Chae Hun

2016-01-01

The present study explored the relationship between the regional Cold-Heat sensation, the key indicator of the Cold-Heat patterns in traditional East Asian medicine (TEAM), and various biomarkers in Korean population. 734 apparently healthy volunteers aged 20 years and older were enrolled. Three scale self-report questions on the general thermal feel in hands, legs, and abdomen were examined. We found that 65% of women tended to perceive their body, particularly their hands and legs, to be cold, versus 25% of men. Energy expenditure and temperature load at resting state were lower in women, independently of body mass index (BMI). Those with warm hands and warm legs had a 0.74 and 0.52 kg/m 2 higher BMI than those with cold hands and cold legs, respectively, regardless of age, gender, and body weight. Norepinephrine was higher, whereas the dynamic changes in glucose and insulin during an oral glucose tolerance test were lower in those with cold extremities, particularly hands. No consistent differences in biomarkers were found for the abdominal dimension. These results suggest that gender, BMI, the sympathetic nervous system, and glucose metabolism are potential determinants of the Cold-Heat sensation in the hands and legs, but not the abdomen.

Relationship of the Cold-Heat Sensation of the Limbs and Abdomen with Physiological Biomarkers

Directory of Open Access Journals (Sweden)

Duong Duc Pham

2016-01-01

Full Text Available The present study explored the relationship between the regional Cold-Heat sensation, the key indicator of the Cold-Heat patterns in traditional East Asian medicine (TEAM, and various biomarkers in Korean population. 734 apparently healthy volunteers aged 20 years and older were enrolled. Three scale self-report questions on the general thermal feel in hands, legs, and abdomen were examined. We found that 65% of women tended to perceive their body, particularly their hands and legs, to be cold, versus 25% of men. Energy expenditure and temperature load at resting state were lower in women, independently of body mass index (BMI. Those with warm hands and warm legs had a 0.74 and 0.52 kg/m2 higher BMI than those with cold hands and cold legs, respectively, regardless of age, gender, and body weight. Norepinephrine was higher, whereas the dynamic changes in glucose and insulin during an oral glucose tolerance test were lower in those with cold extremities, particularly hands. No consistent differences in biomarkers were found for the abdominal dimension. These results suggest that gender, BMI, the sympathetic nervous system, and glucose metabolism are potential determinants of the Cold-Heat sensation in the hands and legs, but not the abdomen.

A loss-of -RHR event under the various plant configurations in low power or shutdown conditions

International Nuclear Information System (INIS)

Seul, Kwang Won; Bang, Young Seok; Lee, Suk Ho; Kim, Hho Jung

1997-01-01

A present study addresses a loss-of-RHR event as an initiating event under specific low power of shutdown conditions. Two typical plant configurations, cold leg opening case with water-filled steam generators and pressurizer opening case with emptied steam generators, were evaluated using the RELAP5/MOD3.2 code. The calculation was compared with the experiment conducted at ROSA-IV/LSTF in Japan. As a result, the code was capable of simulating the system transient behavior following the event. Especially, thermal hydraulic transport processes including non-condensable gas behavior were reasonably predicted with an appropriate time step and CPU time. However, there were some code deficiencies such as too large system mass errors and severe flow oscillations in core region

Desensitization of menthol-activated cold receptors in lower extremities during local cooling in young women with a cold constitution.

Science.gov (United States)

Yamazaki, Fumio; Sone, Ryoko

2017-03-01

To test the hypothesis that topical menthol-induced reactivity of cold sensation and cutaneous vasoconstriction to local cooling is augmented in individuals with a cold constitution, we examined thermal sensation and cutaneous vasoconstrictor responses at menthol-treated and untreated sites in the legs during local skin cooling in young women complaining of chilliness (C group) and young women with no complaint as a normal control group (N group). During local skin cooling, the sensitivity to cold sensation was greater in the C group than in the N group. The application of menthol enhanced the cold sensation at a low temperature in the N group, but not in the C group. Cutaneous vasoconstrictor responses to local skin cooling were not altered by menthol treatment in either of the two groups. These findings suggest the desensitization of menthol-activated cold receptors in the legs of C group subjects, and a minor role of cold receptor activity in cutaneous vasoconstrictor response to local cooling.

Reliability of CRBR primary piping: critique of stress-strength overlap method for cold-leg inlet downcomer

International Nuclear Information System (INIS)

Bari, R.A.; Buslik, A.J.; Papazoglou, I.A.

1976-04-01

A critique is presented of the strength-stress overlap method for the reliability of the CRBR primary heat transport system piping. The report addresses, in particular, the reliability assessment of WARD-D-0127 (Piping Integrity Status Report), which is part of the CRBR PSAR docket. It was found that the reliability assessment is extremely sensitive to the assumed shape for the probability density function for the strength (regarded as a random variable) of the cold-leg inlet downcomer section of the primary piping. Based on the rigorous Chebyschev inequality, it is shown that the piping failure probability is less than 10 -2 . On the other hand, it is shown that the failure probability can be much larger than approximately 10 -13 , the typical value put forth in WARD-D-0127

Experiment data report for LOFT large-break loss-of-coolant experiment L2-5

International Nuclear Information System (INIS)

Bayless, P.D.; Divine, J.M.

1982-08-01

Selected pertinent and uninterpreted data from the third nuclear large break loss-of-coolant experiment (Experiment L2-5) conducted in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large [approx. 1000 MW(e)] commercial PWR operations. Experiment L2-5 simulated a double-ended offset shear of a cold leg in the primary coolant system. The primary coolant pumps were tripped within 1 s after the break initiation, simulating a loss of site power. Consistent with the loss of power, the starting of the high- and low-pressure injection systems was delayed. The peak fuel rod cladding temperature achieved was 1078 +- 13 K. The emergency core cooling system re-covered the core and quenched the cladding. No evidence of core damage was detected

Break spectrum analyses for small break loss of coolant accidents in a RESAR-3S Plant

International Nuclear Information System (INIS)

Fletcher, C.D.; Kullberg, C.M.

1986-03-01

A series of thermal-hydraulic analyses were performed to investigate phenomena occurring during small break loss-of-coolant-accident (LOCA) sequences in a RESAR-3S pressurized water reactor. The analysis included simulations of plant behavior using the TRAC-PF1 and RELAP5/MOD2 computer codes. Series of calculations were performed using both codes for different break sizes. The analyses presented here also served an audit function in that the results shown here were used by the US Nuclear Regulatory Commission (NRC) as an independent confirmation of similar analyses performed by Westinghouse Electric Company using another computer code. 10 refs., 62 figs., 14 tabs

Measurement of L-[1-14C]leucine kinetics in splanchnic and leg tissues in humans. Effect of amino acid infusion

International Nuclear Information System (INIS)

Gelfand, R.A.; Glickman, M.G.; Castellino, P.; Louard, R.J.; DeFronzo, R.A.

1988-01-01

Although whole-body leucine flux is widely measured to study body protein turnover in humans, the contribution of specific tissues to the total-body measurement remains unknown. By combining the organ-balance technique with the systemic infusion of L-[1-14C]leucine, we quantitated leucine production and disposal by splanchnic and leg tissues and by the whole body, simultaneously, in six normal men before and during amino acid infusion. At steady state, disposal of arterial leucine by splanchnic and leg tissues was calculated from the percent extraction (E) of L-[1-14C]leucine counts: uptake = E x [Leu]a x flow. Tissue release of cold leucine (from protein turnover) into vein was calculated as the difference between leucine uptake and the net tissue leucine balance. In the postabsorptive state, despite substantial (P less than .01) extraction of L-[1-14C]leucine by splanchnic (23 +/- 1%) and leg (18 +/- 2%) tissues, net leucine balance across both tissue beds was small, indicating active simultaneous disposal and production of leucine at nearly equivalent rates. Splanchnic tissues accounted for approximately 50% of the measured total-body leucine flux. During amino acid infusion, the net leucine balance across splanchnic and leg tissues became positive, reflecting not only an increase in leucine uptake but also a marked suppression (by approximately 50%, P less than .02) of cold leucine release. This reduction in splanchnic and leg leucine release was indicated by a sharp decline in whole-body endogenous leucine flux

Special small-break applications with TRAC

International Nuclear Information System (INIS)

Dobranich, D.; DeMuth, N.S.; Henninger, R.J.; Burns, R.D. III.

1981-01-01

Input models for the Transient Reactor Analysis Code (TRAC) are described and applications of these models to reactor transients involving small breaks in the primary coolant pressure boundary are demonstrated. The operation of the primary overpressure protection system (relief and safety valves) and the thermal-hydraulic response of the reactor to these transients are obtained from numerical simulations. Also, the effects of steam generator recirculation, steam generator tube rupture, Emergency Core Cooling (ECC) injection and reactivity feedback on the course and consequences of these transients are investigated. These models allow reliable predictions of accident signatures that can help determine the adequacy of equipment and procedures at nuclear power plants to prevent and to control severe accidents

Cold leg injection reflood test results in the SCTF Core-I under constant system pressure

International Nuclear Information System (INIS)

Adachi, Hiromichi; Iwamura, Takamichi; Sobajima, Makoto; Osakabe, Masahiro; Ohnuki, Akira; Abe, Yutaka; Murao, Yoshio.

1990-08-01

The Slab Core Test Facility (SCTF) was constructed to investigate two-dimensional thermal-hydrodynamics in the core and the interaction in fluid behavior between the core and the upper plenum during the last part of blowdown, refill and reflood phases of a postulated loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). The present report describes the analytical results on the system behavior observed in the SCTF Core-I cold leg injection tests, S1-14 (Run 520), S1-15 (521), S1-16 (522), S1-17 (523), S1-20 (530), S1-21 (531), S1-23 (536) and S1-24 (537), performed under constant system pressure condition during transient. Major discussion items are: (1) steam binding, (2) U-tube oscillations, (3) bypass of ECC water (4) core cooling behavior, (5) effect of vent valve and (6) other parameter effects. These results give us very useful information and suggestion on reflood behavior. (author)

LOFT/L2-3, Loss of Fluid Test, 2. NRC L2 Large Break LOCA Experiment

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This experiment was the second of the NRC L2 Series of nuclear large Break LOCA experiments, and was conducted on 12 May 1979. It simulated a 100% cold leg break with a maximum heat generation of 39 kW/m

A study of thermal stratification in the cold legs during the subcooled blowdown phase of a loss of coolant accident in the OSU APEX thermal hydraulic testing facility

International Nuclear Information System (INIS)

Wachs, D. M.

1998-01-01

Thermal stratification, which has been linked to the occurrence of pressurized thermal shock (PTS), is observed to occur during the early stages of simulated loss of coolant accidents (LOCAS) in the Oregon State University Advanced Plant Experiment (OSU APEX) Thermal Hydraulic Test Facility. The OSU APEX Test Facility is a scaled model of the Westinghouse AP600 nuclear power plant. Analysis of the OSU APEX facility data has allowed the determination of an onset criteria for thermal stratification and has provided support for the postulated mechanisms leading to thermal stratification. CFX 4.1, a computational fluid dynamics code, was used to generate a model of the cold legs and the downcomer that described the phenomena occurring within them. Some mixing phenomena were predicted that lead to non-uniformity between the two cold legs attached to the steam generator on the side of the facility containing the Passive Residual Heat Removal (PRHR) injection system. The stratification was found to be two phase and unlikely to be a factor in PTS

Results of Semiscale Mod-2C small-break (5%) loss-of-coolant accident. Experiments S-LH-1 and S-LH-2

International Nuclear Information System (INIS)

Loomis, G.G.; Streit, J.E.

1985-11-01

Two experiments simulating small break (5%) loss-of-coolant accidents (5% SBLOCAs) were performed in the Semiscale Mod-2C facility. These experiments were identical except for downcomer-to-upper-head bypass flow (0.9% in Experiment S-LH-1 and 3.0% in Experiment S-LH-2) and were performed at high pressure and temperature [15.6 MPa (2262 psia) system pressure; 37 K (67 0 F) core differential temperature; 595 K(610 0 F) hot leg fluid temperature]. From the experimental results, the signature response and transient mass distribution are determined for a 5% SBLOCA. The core thermal-hydraulic response is characterized, including core void distribution maps, and the effect of core bypass flow on transient severity is assessed. Comparisons are made between postexperiment RELAP5 calculations and the experimental results, and the capability of RELAP5 to calculate the phenomena is assessed. 115 figs

Analysis of th SBLOCAs in the room 1 for the 3-pin fuel test loop

International Nuclear Information System (INIS)

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

2004-10-01

Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Small Break Loss of Coolant Accidents (SBLOCAs) in the Room 1 for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the SBLOCAs. The location of the pipe break is assumed at the downstream of the main cooling water pump and the upstream of the main cooler in the room 1. The break size is also assumed less than 20% of the cross section area of the pipe. The test fuels are heated up when the cold leg break occur. However, they are not heated up when the hot leg break occur. The maximum Peak Cladding Temperature (PCT) is predicted to be about 931.4 .deg. C for the cold leg break accident in PWR fuel test mode and 931.6 .deg. C in CANDU fuel test mode respectively. The critical break size is about the 8% of the cross section area of the pipe for PWR fuel test mode and the 10% for CANDU fuel test mode. The PCTs meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C

Best estimate prediction for LOFT nuclear experiment L3-2

International Nuclear Information System (INIS)

Kee, E.J.; Shinko, M.S.; Grush, W.H.; Condie, K.G.

1980-02-01

Comprehensive analyses using both the RELAP4 and the RELAP5 computer codes were performed to predict the LOFT transient thermal-hydraulic response for nuclear Loss-of-Coolant Experiment L3-2 to be performed in the Loss-of-Fluid Test (LOFT) facility. The LOFT experiment will simulate a small break in one of the cold legs of a large four-loop pressurized water reactor and will be conducted with the LOFT reactor operating at 50 MW. The break in LOCE L3-2 is sized to cause the break flow to be approximately equal to the high-pressure injection system flow at an intermediate pressure of approximately 7.6 MPa

The comparison of cold-water immersion and cold air therapy on maximal cycling performance and recovery markers following strength exercises

Directory of Open Access Journals (Sweden)

Kane J. Hayter

2016-03-01

Full Text Available This study examined the effects of cold-water immersion (CWI and cold air therapy (CAT on maximal cycling performance (i.e. anaerobic power and markers of muscle damage following a strength training session. Twenty endurance-trained but strength-untrained male (n = 10 and female (n = 10 participants were randomised into either: CWI (15 min in 14 °C water to iliac crest or CAT (15 min in 14 °C air immediately following strength training (i.e. 3 sets of leg press, leg extensions and leg curls at 6 repetition maximum, respectively. Creatine kinase, muscle soreness and fatigue, isometric knee extensor and flexor torque and cycling anaerobic power were measured prior to, immediately after and at 24 (T24, 48 (T48 and 72 (T72 h post-strength exercises. No significant differences were found between treatments for any of the measured variables (p > 0.05. However, trends suggested recovery was greater in CWI than CAT for cycling anaerobic power at T24 (10% ± 2%, ES = 0.90, T48 (8% ± 2%, ES = 0.64 and T72 (8% ± 7%, ES = 0.76. The findings suggest the combination of hydrostatic pressure and cold temperature may be favourable for recovery from strength training rather than cold temperature alone.

Generic evaluation of small break loss-of-coolant accident behavior in Babcock and Wilcox designed 177-FA operating plants

International Nuclear Information System (INIS)

1980-01-01

Slow system depressurization resulting from small break loss-of-coolant accidents (LOCAs) in the reactor coolant system have not, until recently, received detailed analytical study comparable to that devoted to large breaks. Following the TMI-2 accident, the staff had a series of meetings with Babcock and Wilcox (B and W) and the B and W licensees. The staff requested that B and W and the licensees: (1) systematically evaluate plant response for small break loss-of-coolant accidents; (2) address each of the concerns documented in the Michelson report; (3) validate the computer codes used against the TMI-2 accident; (4) extend the break spectrum analysis to very small breaks, giving special consideration to failure of pressurizer valves to close; (5) analyze degraded conditions where AFW is not available; (6) prepare design changes aimed at reducing the probability of loss-of-coolant accidents produced by the failure of a PORV to close; and (7) develop revised emergency procedures for small breaks. This report describes the review of the generic analyses performed by B and W based on the requests stated above

2D/3D program. Upper plenum test facility - UPTF. Test No. 1

International Nuclear Information System (INIS)

1987-01-01

Test No.1 was a quasi-steady state, separate effect test involving the UPTF-System with blocked break valves and blocked pump simulators. Initially the test vessel, the cold and hot leg nozzels as well as the pump seals were completely filled witht hot water in this test. This test was designed to investigate the fluid-fluid mixing phenomena and the development of the fluid and wall temperature fields in the cold leg and downcomer region of a PWR. The experiment was performed by injecting a cold water stream into one cold leg of UPTF while the system was initially filled with stagnant hot water. (orig.)

New theoretical model for two-phase flow discharged from stratified two-phase region through small break

International Nuclear Information System (INIS)

Yonomoto, Taisuke; Tasaka, Kanji

1988-01-01

A theoretical and experimental study was conducted to understand two-phase flow discharged from a stratified two-phase region through a small break. This problem is important for an analysis of a small break loss-of-coolant accident (LOCA) in a light water reactor (LWR). The present theoretical results show that a break quality is a function of h/h b , where h is the elevation difference between a bulk water level in the upstream region and break and b the suffix for entrainment initiation. This result is consistent with existing eperimental results in literature. An air-water experiment was also conducted changing a break orientation as an experimental parameter to develop and assess the model. Comparisons between the model and the experimental results show that the present model can satisfactorily predict the flow rate and the quality at the break without using any adjusting constant when liquid entrainment occurs in a stratified two-phase region. When gas entrainment occurs, the experimental data are correlated well by using a single empirical constant. (author)

SPES-2, AP600 intergral system test S01007 2 inch CL to core make-up tank pressure balance line break

Energy Technology Data Exchange (ETDEWEB)

Bacchiani, M.; Medich, C.; Rigamonti, M. [SIET S.p.A. Piacenza (Italy)] [and others

1995-09-01

The SPES-2 is a full height, full pressure experimental test facility reproducing the Westinghouse AP600 reactor with a scaling factor of 1/395. The experimental plant, designed and operated by SIET in Piacenza, consists of a full simulation of the AP600 primary core cooling system including all the passive and active safety systems. In 1992, Westinghouse, in cooperation with ENEL (Ente Nazionale per l` Energia Elettrica), ENEA (Enter per le numove Technlogie, l` Energia e l` Ambient), Siet (Societa Informazioni Esperienze Termoidraulich) and ANSALDO developed an experimental program to test the integrated behaviour of the AP600 passive safety systems. The SPES-2 test matrix, concluded in November 1994, has examined the AP600 passive safety system response for a range of small break LOCAs at different locations on the primary system and on the passive system lines; single steam generator tube ruptures with passive and active safety systems and a main steam line break transient to demonstrate the boration capability of passive safety systems for rapid cooldown. Each of the tests has provided detailed experimental results for verification of the capability of the analysis methods to predict the integrated passive safety system behaviour. Cold and hot shakedown tests have been performed on the facility to check the characteristics of the plant before starting the experimental campaign. The paper first presents a description of the SPES-2 test facility then the main results of S01007 test {open_quotes}2{close_quotes} Cold Leg (CL) to Core Make-up Tank (CMT) pressure balance line break{close_quotes} are reported and compared with predictions performed using RELAP5/mod3/80 obtained by ANSALDO through agreement with U.S.N.R.C. (U.S. Nuclear Regulatory Commission). The SPES-2 nodalization and all the calculations here presented were performed by ANSALDO and sponsored by ENEL as a part of pre-test predictions for SPES-2.

Comprehensive safety analysis code system for nuclear fusion reactors III: Ex-vessel LOCA analyses considering passive safety

International Nuclear Information System (INIS)

Honda, T.; Okazaki, T.; Maki, K.; Uda, T.; Seki, Y.; Aoki, I.; Kunugi, T.

1996-01-01

Ex-vessel loss-of-coolant accidents (LOCAs) in a fusion reactor have been analyzed to investigate the possibility of passive plasma shutdown. For this purpose, a hybrid code of the plasma dynamics and thermal characteristics of the reactor structures, which has been modified to include the impurity emission from plasma-facing components (PFCs), has been developed. Ex-vessel LOCAs of the cooling system during the ignition operation in the International Thermonuclear Experimental Reactor (ITER), in which graphite PFCs were employed in conceptual design activity, were assumed. When double-ended break occurs at the cold leg of the divertor cooling system, the copper cooling tube begins to melt within 3 s after the LOCA, even though the plasma is passively shut down at nearly 4 s. An active plasma shutdown system will be needed for such rapid transient accidents. On the other hand, when a small (1%) break LOCA occurs there, the plasma is passively shut down at nearly 36 s, which happens before the copper cooling tube begins to melt. When the double-ended break LOCA occurs at the cold leg of the first-wall cooling system, there is enough time (nearly 100 s) to shut down the plasma with a controllable method before the reactor structures are damaged. 21 refs., 8 figs

Subcritical to supercritical flow transition in a horizontal stratified flow

International Nuclear Information System (INIS)

Asaka, H.; Kukita, Y.

1995-01-01

The conditions for a transition from hydraulically subcritical to supercritical flow in the hot legs of a pressurized water reactor (PWR) were studied using data obtained from a two-phase natural circulation experiment conducted at the ROSA-IV Large Scale Test Facility (LSTF). The LSTF is a 1/48 volumetrically-scaled simulator of a Westinghouse-type PWR. The conditions for the transition were compared with the theory of Gardner. While the model explains the trend in the experimental data, the quantitative agreement was not satisfactory. It was found that the conditions for the transition from the subcritical to supercritical flow were predicted well by introducing energy loss term into the theory. (author)

LOFT/L2-5, Loss of Fluid Test, 3. NRC L2 Large Break LOCA Experiment

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This experiment was the third of the NRC L2 Series of nuclear large Break LOCA experiments, conducted on 16 June 1981. It simulated a 100% cold leg break with a maximum heat generation of 40 kW/m and rapid pump coast down

Monitoring the vaccine cold chain.

OpenAIRE

Cheriyan, E

1993-01-01

Maintaining the vaccine cold chain is an essential part of a successful immunisation programme. A continuous electronic temperature monitor helped to identify breaks in the cold chain in the community and the study led to the issue of proper guidelines and replacement of faulty equipment.

Orthostatic leg blood volume changes assessed by near-infrared spectroscopy

DEFF Research Database (Denmark)

Truijen, J; Kim, Y S; Krediet, C T P

2012-01-01

posture, volume accumulation in small blood vessels contributes significantly to the total fluid volume accumulated in the legs. Considering that near-infrared spectroscopy (NIRS) tracks postural blood volume changes within the small blood vessels of the lower leg, we evaluated the NIRS-determined changes......-linear accumulation of blood volume in the small vessels of the leg, with an initial fast phase followed by a more gradual increase at least partly contributing to the relocation of fluid during orthostatic stress....

Post-test analysis with RELAP5/MOD2 of ROSA-IV/LSTF natural circulation test ST-NC-02

International Nuclear Information System (INIS)

Chauliac, C.; Kukita, Yutaka; Kawaji, Masahiro; Nakamura, Hideo; Tasaka, Kanji.

1988-10-01

Results of post-test analysis for the ROSA-IV/LSTF natural circulation experiment ST-NC-02 are presented. The experiment consisted of many steady-state stages registered for different primary inventories. The calculation was done with RELAP5/MOD2 CYCLE 36.00. Discrepancies between the calculation and the experiment are observed: the core flow rate is overestimated at inventories between 80 % and 95 %; the inventory at which dryout occurs in the core is also much overestimated. The causes of these discrepancies are studies through sensitivity calculations and the following key parameters are pointed out: the interfacial friction and the form loss coefficients in the vessel riser, the SG U-tube multidimensional behaviour, the interfacial friction in the SG inlet plenum and in the pipe located underneath. (author)

INTRA-RATER RELIABILITY OF THE MULTIPLE SINGLE-LEG HOP-STABILIZATION TEST AND RELATIONSHIPS WITH AGE, LEG DOMINANCE AND TRAINING.

Science.gov (United States)

Sawle, Leanne; Freeman, Jennifer; Marsden, Jonathan

2017-04-01

Balance is a complex construct, affected by multiple components such as strength and co-ordination. However, whilst assessing an athlete's dynamic balance is an important part of clinical examination, there is no gold standard measure. The multiple single-leg hop-stabilization test is a functional test which may offer a method of evaluating the dynamic attributes of balance, but it needs to show adequate intra-tester reliability. The purpose of this study was to assess the intra-rater reliability of a dynamic balance test, the multiple single-leg hop-stabilization test on the dominant and non-dominant legs. Intra-rater reliability study. Fifteen active participants were tested twice with a 10-minute break between tests. The outcome measure was the multiple single-leg hop-stabilization test score, based on a clinically assessed numerical scoring system. Results were analysed using an Intraclass Correlations Coefficient (ICC 2,1 ) and Bland-Altman plots. Regression analyses explored relationships between test scores, leg dominance, age and training (an alpha level of p = 0.05 was selected). ICCs for intra-rater reliability were 0.85 for the dominant and non-dominant legs (confidence intervals = 0.62-0.95 and 0.61-0.95 respectively). Bland-Altman plots showed scores within two standard deviations. A significant correlation was observed between the dominant and non-dominant leg on balance scores (R 2 =0.49, ptest demonstrated strong intra-tester reliability with active participants. Younger participants who trained more, have better balance scores. This test may be a useful measure for evaluating the dynamic attributes of balance. 3.

A study of entrainment at a break and in the core during small break loss-of-coolant accidents in PWRs

International Nuclear Information System (INIS)

Yonomoto, Taisuke

1996-05-01

Objectives of the present study are to obtain a better understanding of entrainment at a break and in the core during small break loss-of-coolant-accidents (SBLOCAs) in PWRs, and to develop a means for the best evaluation of the phenomena. For the study of entrainment at a break, a theoretical model was developed, which was assessed by comparisons with several experimental data bases. By modifying a LOCA analysis code using the present model, experimental results obtained from SBLOCA experiments at a PWR large-scale simulator were reproduced very well. For the study of entrainment in the core, reflooding experiments were conducted at high pressure, from which the onset conditions were obtained. It was confirmed that the cooling behavior for a dry-out core is very simple under typical high pressure reflooding conditions for PWRs, because liquid entrainment does not occur in the core. (author)

Investigation of Loop Seal Clearing Phenomena for the ATLAS SBLOCA Long Term Cooling Test using TRACE and MARS-KS

Energy Technology Data Exchange (ETDEWEB)

Hwang, Min Jeong; Park, M. H.; Marigomen Ralph; Sim, S. K. [Environment and Energy Technology, Daejeon (Korea, Republic of)

2016-10-15

During Design Certificate(DC) review of the APR1400, USNRC raised a long term cooling safety issue on the effect of loop seal clearing during cold leg Small Break Loss Of Coolant Accident(SBLOCA) due to relatively deep cross-over loop compared to the US PWRs. The objective of this study is thus to investigate the loop seal clearing phenomena during cold leg slot break SBLOCA long term cooling and resolve the safety issue on the SBLOCA long term cooling related to the APR1400 DC. TRACE and MARS-KS were used to predict the test results and to perform sensitivity studies for the SBLOCA loop seal clearing phenomena. The calculation shows that the TRACE code well predict the sequence of Test LTC-CL-04R. However, compared to the experiment, the TRACE over predicts the primary pressure due to smaller break flow prediction. MARS-KS well predicts major thermal hydraulic parameters during the transient with reasonable agreement. MARS-KS better predicts ATLAS LTC-CL-04R test data with a good agreement than the TRACE due to better prediction of the break flow. Overall, compared to the experiment, the TRACE and MARS-KS Codes show a discrepancy in predicting the loop seal clearing and reformation time. Both TRACE and MARS-KS correctly predicts core water level and fuel cladding temperatures. From this study, it can be said that even though APR1400 cross-over leg design has slightly deeper loop seals, the effect on the safety of the SBLOCA long term cooling is minimal compared to the SBLOCA cladding failure criteria. Further study on the SBLOCA loop seal clearing phenomena is needed.

Multidimensional analysis of fluid flow in the loft cold leg blowdown pipe during a loss-of-coolant experiment

International Nuclear Information System (INIS)

Demmie, P.N.; Hofmann, K.R.

1979-03-01

A computer analysis of fluid flow in the Loss-of-Fluid Test (LOFT) cold leg blowdown pipe during a loss-of-coolant experiment (LOCE) was performed using the computer program K-FIX/MOD1. The purpose of this analysis was to evaluate the capability of K-FIX/MOD1 to calculate theoretical fluid quantity distributions in the blowdown pipe during a LOCE for possible application to the analysis of LOFT experimental data, the determination of mass flow, or the development of data reduction models. A rectangular section of a portion of the LOFT blowdown pipe containing measurement Station BL-1 was modeled using time-dependent boundary conditions. Fluid quantities were calculated during a simulation of the first 26 s of LOFT LOCE L1-4. Sensitivity studies were made to determine changes in void fractions and velocities resulting from specific changes in the inflow boundary conditions used for this simulation

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.)

Verification of a TRACE EPRTM model on the basis of a scaling calculation of an SBLOCA ROSA test

International Nuclear Information System (INIS)

Freixa, J.; Manera, A.

2011-01-01

Research highlights: → Verification of a TRACE input deck for the EPR TM generation III PWR. → Scaling simulation of an SBLOCA experiment of the integral test facility ROSA/LSTF. → The EPR TM model was compared with the TRACE results of the ROSA/LSTF model. - Abstract: In cooperation with the Finnish Radiation and Nuclear Safety Authority (STUK), a project has been launched at the Paul Scherrer Institute (PSI) aimed at performing safety evaluations of the Olkiluoto-3 nuclear power plant (NPP), the first EPR TM , a generation III pressurizer water reactor (PWR); with particular emphasis on small-and large-break loss-of-coolant-accidents (SB/LB-LOCAs) and main steam-line breaks. As a first step of this work, the best estimate system code TRACE has been used to develop a model of Olkiluoto-3. In order to test the nodalization, a scaling calculation from the rig of safety assessment (ROSA) test facility has been performed. The ROSA large scale test facility (LSTF) was built to simulate Westinghouse design pressurized water reactors (PWR) with a four-loop configuration. Even though there are differences between the EPR TM and the Westinghouse designs, the number of similarities is large enough to carry out scaling calculations on SBLOCA and LOCA cases from the ROSA facility; as a matter of fact, the main differences are located in the secondary side. Test 6-1 of the ROSA 1 programme, an SBLOCA with the break situated in the upper head of the reactor pressure vessel (RPV), was of special interest since a very good agreement with the experiment was obtained with a TRACE input deck. In order to perform such scaling calculation, the set-points of the secondary relief and safety valves in the EPR TM nodalization had to be changed to those used in the ROSA facility, the break size and the core power had to be scaled by a factor of 60 (according to the core power and core volume) and the pumps coast down had to be adapted to the ones of the test. The calculation showed

A role for small RNAs in DNA double-strand break repair

DEFF Research Database (Denmark)

Wei, W.; Ba, Z.; Wu, Y.

2012-01-01

Eukaryotes have evolved complex mechanisms to repair DNA double-strand breaks (DSBs) through coordinated actions of protein sensors, transducers, and effectors. Here we show that ∼21-nucleotide small RNAs are produced from the sequences in the vicinity of DSB sites in Arabidopsis and in human cells....... We refer to these as diRNAs for DSB-induced small RNAs. In Arabidopsis, the biogenesis of diRNAs requires the PI3 kinase ATR, RNA polymerase IV (Pol IV), and Dicer-like proteins. Mutations in these proteins as well as in Pol V cause significant reduction in DSB repair efficiency. In Arabidopsis, di...

Phase controlled metal–insulator transition in multi-leg quasiperiodic optical lattices

International Nuclear Information System (INIS)

Maiti, Santanu K.; Sil, Shreekantha; Chakrabarti, Arunava

2017-01-01

A tight-binding model of a multi-leg ladder network with a continuous quasiperiodic modulation in both the site potential and the inter-arm hopping integral is considered. The model mimics optical lattices where ultra-cold fermionic or bosonic atoms are trapped in double well potentials. It is observed that, the relative phase difference between the on-site potential and the inter-arm hopping integral, which can be controlled by the tuning of the interfering laser beams trapping the cold atoms, can result in a mixed spectrum of one or more absolutely continuous subband(s) and point like spectral measures. This opens up the possibility of a re-entrant metal–insulator transition. The subtle role played by the relative phase difference mentioned above is revealed, and we corroborate it numerically by working out the multi-channel electronic transmission for finite two-, and three-leg ladder networks. The extension of the calculation beyond the two-leg case is trivial, and is discussed in the work. - Graphical abstract: ▪ - Highlights: • Phase controlled metal–insulator transition is discussed. • An analytical prescription is given to understand MI transition. • Our work provides a way of designing experiments involving laser beams.

Transient computational fluid dynamics analysis of emergency core cooling injection at natural circulation conditions

International Nuclear Information System (INIS)

Scheuerer, Martina; Weis, Johannes

2012-01-01

Highlights: ► Pressurized thermal shocks are important phenomena for plant life extension and aging. ► The thermal-hydraulics of PTS have been studied experimentally and numerically. ► In the Large Scale Test Facility a loss of coolant accident was investigated. ► CFD software is validated to simulate the buoyancy driven flow after ECC injection. - Abstract: Within the framework of the European Nuclear Reactor Integrated Simulation Project (NURISP), computational fluid dynamics (CFD) software is validated for the simulation of the thermo-hydraulics of pressurized thermal shocks. A proposed validation experiment is the test series performed within the OECD ROSA V project in the Large Scale Test Facility (LSTF). The LSTF is a 1:48 volume-scaled model of a four-loop Westinghouse pressurized water reactor (PWR). ROSA V Test 1-1 investigates temperature stratification under natural circulation conditions. This paper describes calculations which were performed with the ANSYS CFD software for emergency core cooling injection into one loop at single-phase flow conditions. Following the OECD/NEA CFD Best Practice Guidelines (Mahaffy, 2007) the influence of grid resolution, discretisation schemes, and turbulence models (shear stress transport and Reynolds stress model) on the mixing in the cold leg were investigated. A half-model was used for these simulations. The transient calculations were started from a steady-state solution at natural circulation conditions. The final calculations were obtained in a complete model of the downcomer. The results are in good agreement with data.

Description of the small plastics fragments in marine sediments along the Alang-Sosiya ship-breaking yard, India

Science.gov (United States)

Srinivasa Reddy, M.; Basha, Shaik; Adimurthy, S.; Ramachandraiah, G.

2006-07-01

This study aimed to assess the accumulation of small plastic debris in the intertidal sediments of the world's largest ship-breaking yard at Alang-Sosiya, India. Small plastics fragments were collected by flotation and separated according to their basic polymer type under a microscope, and subsequently identified by FT-IR spectroscopy as polyurethane, nylon, polystyrene, polyester and glass wool. The morphology of these materials was also studied using a scanning electron microscope. Overall, there were on average 81 mg of small plastics fragments per kg of sediment. The described plastic fragments are believed to have resulted directly from the ship-breaking activities at the site.

ROSA-II test data report, 13

International Nuclear Information System (INIS)

1978-07-01

Results of the ROSA-II test simulating a loss-of-coolant accident (LOCA) in a PWR are presented, including test conditions and interpretations of phenomena observed in test runs 502, 505, 506 and 507. Development tests were performed to find a more effective ECCS injection method than the existing one based on cold leg injection. A combined injection of hot water into upper plenum in early stage of blowdown and subsequent cold water into lower plenum is the most effective method for a cold leg break. A hot leg injection of a low pressure injection system is effective for direct core cooling and early reflooding. The generalization for actual reactors will require analyses with a reliable code. (auth.)

Evaluation of post-LOCA long term cooling performance in Korean standard nuclear power plants

International Nuclear Information System (INIS)

Bang, Young Seok; Jung, Jae Won; Seul, Kwang Won; Kim, Hho Jung

2001-01-01

The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break loss-of-coolant accidents (LOCA) and large break LOCA at cold leg. The RELAP5/MOD3.2.2 beta code is used to calculate the LTC sequences based on the LTC plan of the Korean Standard Nuclear Power Plants (KSNPP). A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from 0.02 to 0.5 ft 2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important action including the safety injection tank (SIT) isolation and the simultaneous injection in LTC procedure is investigated. As a result, it is found that the sufficient margin is available in avoiding the boron precipitation in the core. It is also found that a further specific condition for the SIT isolation action need to be setup and it is recommended that the early initiation of the simultaneous injection be taken for larger break LTC sequences. (author)

Evaluation of PWR response to main-steamline break with concurrent steam-generator tube rupture and small-break LOCA

International Nuclear Information System (INIS)

Laaksonen, J.T.; Sheron, B.W.

1982-12-01

In 1980, the NRC staff raised a potential safety issue involving a coincident steamline break, steam generator tube rupture, and small-break loss-of-coolant accident (LOCA). The bases for this concern were that the system response, primarily the maintenance of core cooling, was unanalyzed and the adequacy of the present guidance to operators to respond to combination LOCAs was unknown. This report discusses the staff evaluations performed to assess the system response and the adequacy of the present emergency operator guidelines. In all of the analyzed cases the primary coolant shrinkage, caused by overcooling, and the simultaneous loss of coolant can be compensated by the high pressure emergency core cooling system. The core remains covered with liquid, and the primary coolant remains subcooled, except in the vessel upper head. If the steamline break is outside the containment and cannot be isolated, the radiological consequences could be more severe than in any accident currently analyzed in a typical plant Final Safety Evaluation Report (FSAR). To decrease the risk of elevated offsite releases, an early diagnosis of the tube rupture has to be ensured. This can be done by upgrading operator instructions. The appropriate mitigating actions are in the existing instructions

Duality after supersymmetry breaking

International Nuclear Information System (INIS)

Shadmi, Yael; Cheng, Hsin-Chia

1998-05-01

Starting with two supersymmetric dual theories, we imagine adding a chiral perturbation that breaks supersymmetry dynamically. At low energy we then get two theories with soft supersymmetry-breaking terms that are generated dynamically. With a canonical Kaehler potential, some of the scalars of the ''magnetic'' theory typically have negative mass-squared, and the vector-like symmetry is broken. Since for large supersymmetry breaking the ''electric'' theory becomes ordinary QCD, the two theories are then incompatible. For small supersymmetry breaking, if duality still holds, the magnetic theory analysis implies specific patterns of chiral symmetry breaking in supersymmetric QCD with small soft masses

"Non-cold" dark matter at small scales: a general approach

Science.gov (United States)

Murgia, R.; Merle, A.; Viel, M.; Totzauer, M.; Schneider, A.

2017-11-01

Structure formation at small cosmological scales provides an important frontier for dark matter (DM) research. Scenarios with small DM particle masses, large momenta or hidden interactions tend to suppress the gravitational clustering at small scales. The details of this suppression depend on the DM particle nature, allowing for a direct link between DM models and astrophysical observations. However, most of the astrophysical constraints obtained so far refer to a very specific shape of the power suppression, corresponding to thermal warm dark matter (WDM), i.e., candidates with a Fermi-Dirac or Bose-Einstein momentum distribution. In this work we introduce a new analytical fitting formula for the power spectrum, which is simple yet flexible enough to reproduce the clustering signal of large classes of non-thermal DM models, which are not at all adequately described by the oversimplified notion of WDM . We show that the formula is able to fully cover the parameter space of sterile neutrinos (whether resonantly produced or from particle decay), mixed cold and warm models, fuzzy dark matter, as well as other models suggested by effective theory of structure formation (ETHOS). Based on this fitting formula, we perform a large suite of N-body simulations and we extract important nonlinear statistics, such as the matter power spectrum and the halo mass function. Finally, we present first preliminary astrophysical constraints, based on linear theory, from both the number of Milky Way satellites and the Lyman-α forest. This paper is a first step towards a general and comprehensive modeling of small-scale departures from the standard cold DM model.

Increasing trunk flexion transforms human leg function into that of birds despite different leg morphology.

Science.gov (United States)

Aminiaghdam, Soran; Rode, Christian; Müller, Roy; Blickhan, Reinhard

2017-02-01

Pronograde trunk orientation in small birds causes prominent intra-limb asymmetries in the leg function. As yet, it is not clear whether these asymmetries induced by the trunk reflect general constraints on the leg function regardless of the specific leg architecture or size of the species. To address this, we instructed 12 human volunteers to walk at a self-selected velocity with four postures: regular erect, or with 30 deg, 50 deg and maximal trunk flexion. In addition, we simulated the axial leg force (along the line connecting hip and centre of pressure) using two simple models: spring and damper in series, and parallel spring and damper. As trunk flexion increases, lower limb joints become more flexed during stance. Similar to birds, the associated posterior shift of the hip relative to the centre of mass leads to a shorter leg at toe-off than at touchdown, and to a flatter angle of attack and a steeper leg angle at toe-off. Furthermore, walking with maximal trunk flexion induces right-skewed vertical and horizontal ground reaction force profiles comparable to those in birds. Interestingly, the spring and damper in series model provides a superior prediction of the axial leg force across trunk-flexed gaits compared with the parallel spring and damper model; in regular erect gait, the damper does not substantially improve the reproduction of the human axial leg force. In conclusion, mimicking the pronograde locomotion of birds by bending the trunk forward in humans causes a leg function similar to that of birds despite the different morphology of the segmented legs. © 2017. Published by The Company of Biologists Ltd.

Roll type conducting polymer legs for rigid-flexible thermoelectric generator

Directory of Open Access Journals (Sweden)

Teahoon Park

2017-07-01

Full Text Available A roll-type conducting polymer film was explored as a flexible organic p-type thermoelectric leg using poly(3,4-ethylenedioxythiophene (PEDOT doped with tosylate. The PEDOT films were prepared through solution casting polymerization and rolled up for a roll-type leg. Due to the high flexibility, the roll-type PEDOT leg enabled easy contact to both top and bottom electrodes. Simulation on the dynamic heat transfer and convective cooling for a vertically roosted rod- and roll-type PEDOT leg showed that the temperature difference (ΔT between the hot and cold sides of the leg was much higher in the roll than that of the rod. The PEDOT legs were integrated with n-type Bi2Te3 blocks, to give a 36-couple rigid-flexible thermoelectric generator (RF-TEG. The maximum output voltage from the 36-couple RF-TEG under a ΔT of 7.9 K was determined as 36.7 mV along with a high output power of 115 nW. A wearable RF-TEG was prepared upon the combination of the 36-couple RF-TEG with an arm warmer, to afford an output voltage of 10.6 mV, which was generated constantly and steadily from human wrist heat.

Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

OpenAIRE

Hwang Bae; Dong Eok Kim; Sung-Uk Ryu; Sung-Jae Yi; Hyun-Sik Park

2017-01-01

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are s...

Real-time graphic display system for ROSA-V Large Scale Test Facility

International Nuclear Information System (INIS)

Kondo, Masaya; Anoda, Yoshinari; Osaki, Hideki; Kukita, Yutaka; Takigawa, Yoshio.

1993-11-01

A real-time graphic display system was developed for the ROSA-V Large Scale Test Facility (LSTF) experiments simulating accident management measures for prevention of severe core damage in pressurized water reactors (PWRs). The system works on an IBM workstation (Power Station RS/6000 model 560) and accommodates 512 channels out of about 2500 total measurements in the LSTF. It has three major functions: (a) displaying the coolant inventory distribution in the facility primary and secondary systems; (b) displaying the measured quantities at desired locations in the facility; and (c) displaying the time histories of measured quantities. The coolant inventory distribution is derived from differential pressure measurements along vertical sections and gamma-ray densitometer measurements for horizontal legs. The color display indicates liquid subcooling calculated from pressure and temperature at individual locations. (author)

Detection and control of potential core damage during a small-break LOCA

International Nuclear Information System (INIS)

Thomas, G.R.; Zebroski, E.L.

1981-01-01

A refreshing development in small-break LOCA analysis and testing is the recognition that this work can be of real value to a plant operator. Event-trees, or safety sequence diagrams, are being made increasingly realistic and are being used to develop and to test the abnormal transient operating guidelines (ATOGs) which provide a basis for operator response, training, and simulator work now under way. Perhaps the most monumental lesson of the TMI-2 accident is that the tradition of extreme worst-case analysis and data gathering can provide a directly negative contribution to safety if it is used as the basis for designing procedures and training of operator response. The event-trees for guiding operator response to abnormal conditions, ATOGs, must be based on physically realistic, best-estimate models. Possibly, the most dramatic risk reduction achieved will be attained through the use of realistic accident analysis, which leads to realistic operator guidelines and training, improved display of the critical information to the operator, and improved management structure. Given the dominant contribution of small-break LOCAs to the overall public risk envelope, realism should be the primary banner for future work in this field

On-line pressurizer surveillance system design to prevent small break LOCA through PORV using micro-computer

International Nuclear Information System (INIS)

Lee, Jong-Ho; Chang, Soon-Heung

1986-01-01

Small break LOCA caused by a stuck-open PORV is one of the important contributors to nuclear power plant risk. This paper deals with the design of a pressurizer surveillance system using micro-computer to prevent the malfunction of system and has assessed the effect of this improvement through Probabilistic Risk Assessment (PRA) method. Micro-computer diagnoses the malfunction of system by a process checking method and performs automatically backup action related to each malfunction. Owing to this improvement, we can correctly diagnose ''Spurious Opening'', ''Fail to Reclose'' and ''Small break LOCA'' which are difficult for operator to diagnose quickly and correctly and reduce the probability of a human error by an automatic backup action. (author)

Application of RELAP5/MOD3.1 code to the LOFT test L3-6

International Nuclear Information System (INIS)

Pylev, S.S.; Roginskaja, V.L.

1998-02-01

A calculation of LOFT Experiment L3-6, a small break equivalent to a 4-in diameter rupture in the cold leg of a four-loop commercial pressurized water reactor, has been performed to help validate RELAP5/MOD3.1 for this application. The version of the code to be used is SCDAP/RELAP5/MOD3.1.8d0. Three calculations were carried out in order to study the sensitivity to change break nozzle superheated discharge coefficient. Conducted comparative analysis of the LOFT L3-6 experiment shows on the whole a reasonable agreement between calculated data. Some discrepancies in the system pressure do not distort a picture of the transient. 6 refs

Application of RELAP5/MOD3.1 code to the LOFT test L3-6

Energy Technology Data Exchange (ETDEWEB)

Pylev, S.S.; Roginskaja, V.L.

1998-02-01

A calculation of LOFT Experiment L3-6, a small break equivalent to a 4-in diameter rupture in the cold leg of a four-loop commercial pressurized water reactor, has been performed to help validate RELAP5/MOD3.1 for this application. The version of the code to be used is SCDAP/RELAP5/MOD3.1.8d0. Three calculations were carried out in order to study the sensitivity to change break nozzle superheated discharge coefficient. Conducted comparative analysis of the LOFT L3-6 experiment shows on the whole a reasonable agreement between calculated data. Some discrepancies in the system pressure do not distort a picture of the transient. 6 refs.

Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

International Nuclear Information System (INIS)

Wilson, G.E.; Fletcher, C.D.; Davis, C.B.

1997-06-01

This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked ampersand influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs

Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

Energy Technology Data Exchange (ETDEWEB)

Wilson, G.E.; Fletcher, C.D.; Davis, C.B. [and others

1997-06-01

This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

SB LOCA analyses for Krsko Full Scope Simulator verification

International Nuclear Information System (INIS)

Prosek, A.; Parzer, I.; Mavko, B.

2000-01-01

Nuclear power plant simulators are intended to be used for training and maintaining competence to ensure safe, reliable operation of nuclear power plants throughout the world. The simulator shall be specified to a reference unit and its performance validation testing shall be provided. In this study a small-break loss-of-coolant accident (SB LOCA) response of Krsko nuclear power plant (NPP) was calculated for full scope simulator verification. The investigation included five cases with varying the break size in the cold leg of reactor coolant system. The plant specific and verified RELAP5/MOD2 model of Krsko nuclear power plant (NPP), developed in the past for 1882 MWt power, was adapted for 2000 MWt power (cycle 17) including the model for replacement steam generators. The results showed that the plant system response to breaks with small break area was slower compared to breaks with larger break area. The core heatup occurred in most of the cases analyzed. The acceptance criteria for emergency core cooling system were also met. The predicted results of the SB LOCA analysis for Krsko NPP suggest that they may be used for verification of the Krsko Full Scope Simulator performance. (author)

Attributing Changing Rates of Temperature Record Breaking to Anthropogenic Influences

Science.gov (United States)

King, Andrew D.

2017-11-01

Record-breaking temperatures attract attention from the media, so understanding how and why the rate of record breaking is changing may be useful in communicating the effects of climate change. A simple methodology designed for estimating the anthropogenic influence on rates of record breaking in a given time series is proposed here. The frequency of hot and cold record-breaking temperature occurrences is shown to be changing due to the anthropogenic influence on the climate. Using ensembles of model simulations with and without human-induced forcings, it is demonstrated that the effect of climate change on global record-breaking temperatures can be detected as far back as the 1930s. On local scales, a climate change signal is detected more recently at most locations. The anthropogenic influence on the increased occurrence of hot record-breaking temperatures is clearer than it is for the decreased occurrence of cold records. The approach proposed here could be applied in rapid attribution studies of record extremes to quantify the influence of climate change on the rate of record breaking in addition to the climate anomaly being studied. This application is demonstrated for the global temperature record of 2016 and the Central England temperature record in 2014.

Scaling criteria and an assessment of Semiscale Mod-3 scaling for small-break loss-of-coolant transients

International Nuclear Information System (INIS)

Larson, T.K.; Anderson, J.L.; Shimeck, D.J.

1982-01-01

Various methods of scaling fluid thermal-hydraulic test facilities and their relative merits and disadvantages are examined in light of nuclear reactor safety considerations. Particular emphasis is placed on examination of the scaling of the Semiscale Mod-3 system and determination of thermal-hydraulic phenomena thought to be important during a small break loss-of-coolant accident in a pressurized water nuclear reactor. The influence of geometric and dynamic scaling concerns in the Mod-3 system on small break behavior are addressed from an engineering viewpoint and corrective measures contemplated or required to make results from Semiscale tests more meaningful relative to expected PWR response are discussed

Protection against high intravascular pressure in giraffe legs

DEFF Research Database (Denmark)

Petersen, Karin K; Hørlyck, Arne; Østergaard, Kristine Hovkjær

2013-01-01

The high blood pressure in giraffe leg arteries renders giraffes vulnerable to edema. We investigated in 11 giraffes whether large and small arteries in the legs and the tight fascia protect leg capillaries. Ultrasound imaging of foreleg arteries in anesthetized giraffes and ex vivo examination....... All three findings can contribute to protection of the capillaries in giraffe legs from a high arterial pressure....... revealed abrupt thickening of the arterial wall and a reduction of its internal diameter just below the elbow. At and distal to this narrowing, the artery constricted spontaneously and in response to norepinephrine and intravascular pressure recordings revealed a dynamic, viscous pressure drop along...

Recovery From Exercise-Induced Muscle Damage: Cold-Water Immersion Versus Whole-Body Cryotherapy.

Science.gov (United States)

Abaïdia, Abd-Elbasset; Lamblin, Julien; Delecroix, Barthélémy; Leduc, Cédric; McCall, Alan; Nédélec, Mathieu; Dawson, Brian; Baquet, Georges; Dupont, Grégory

2017-03-01

To compare the effects of cold-water immersion (CWI) and whole-body cryotherapy (WBC) on recovery kinetics after exercise-induced muscle damage. Ten physically active men performed single-leg hamstring eccentric exercise comprising 5 sets of 15 repetitions. Immediately postexercise, subjects were exposed in a randomized crossover design to CWI (10 min at 10°C) or WBC (3 min at -110°C) recovery. Creatine kinase concentrations, knee-flexor eccentric (60°/s) and posterior lower-limb isometric (60°) strength, single-leg and 2-leg countermovement jumps, muscle soreness, and perception of recovery were measured. The tests were performed before and immediately, 24, 48, and 72 h after exercise. Results showed a very likely moderate effect in favor of CWI for single-leg (effect size [ES] = 0.63; 90% confidence interval [CI] = -0.13 to 1.38) and 2-leg countermovement jump (ES = 0.68; 90% CI = -0.08 to 1.43) 72 h after exercise. Soreness was moderately lower 48 h after exercise after CWI (ES = -0.68; 90% CI = -1.44 to 0.07). Perception of recovery was moderately enhanced 24 h after exercise for CWI (ES = -0.62; 90% CI = -1.38 to 0.13). Trivial and small effects of condition were found for the other outcomes. CWI was more effective than WBC in accelerating recovery kinetics for countermovement-jump performance at 72 h postexercise. CWI also demonstrated lower soreness and higher perceived recovery levels across 24-48 h postexercise.

Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2

International Nuclear Information System (INIS)

Ireland, J.R.

1982-01-01

A 200% double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops and cold-leg emergency-core-cooling systems (ECCS). The calculated peak cladding temperature of 950 K occurred during blowdown and the cladding temperature excursion was terminated at 175 s when complete core quenching occurred. Accumulator flows were initiated at 10 s when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated. Top quenching was caused by entrainment from the lower plenum and lower core regions. The entrained liquid was sufficient to form a small, saturated pool (0.3 m deep) above the upper core support plate. Also, some of the entrained liquid was carried out the hot legs and vaporized in the steam generators. Strong multidimensional effects were calculated in the reactor vessel, particularly with respect to rod quenching

Transient computational fluid dynamics analysis of emergency core cooling injection at natural circulation conditions

Energy Technology Data Exchange (ETDEWEB)

Scheuerer, Martina, E-mail: Martina.Scheuerer@grs.de [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Forschungsinstitute, 85748 Garching (Germany); Weis, Johannes, E-mail: Johannes.Weis@grs.de [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Forschungsinstitute, 85748 Garching (Germany)

2012-12-15

Highlights: Black-Right-Pointing-Pointer Pressurized thermal shocks are important phenomena for plant life extension and aging. Black-Right-Pointing-Pointer The thermal-hydraulics of PTS have been studied experimentally and numerically. Black-Right-Pointing-Pointer In the Large Scale Test Facility a loss of coolant accident was investigated. Black-Right-Pointing-Pointer CFD software is validated to simulate the buoyancy driven flow after ECC injection. - Abstract: Within the framework of the European Nuclear Reactor Integrated Simulation Project (NURISP), computational fluid dynamics (CFD) software is validated for the simulation of the thermo-hydraulics of pressurized thermal shocks. A proposed validation experiment is the test series performed within the OECD ROSA V project in the Large Scale Test Facility (LSTF). The LSTF is a 1:48 volume-scaled model of a four-loop Westinghouse pressurized water reactor (PWR). ROSA V Test 1-1 investigates temperature stratification under natural circulation conditions. This paper describes calculations which were performed with the ANSYS CFD software for emergency core cooling injection into one loop at single-phase flow conditions. Following the OECD/NEA CFD Best Practice Guidelines (Mahaffy, 2007) the influence of grid resolution, discretisation schemes, and turbulence models (shear stress transport and Reynolds stress model) on the mixing in the cold leg were investigated. A half-model was used for these simulations. The transient calculations were started from a steady-state solution at natural circulation conditions. The final calculations were obtained in a complete model of the downcomer. The results are in good agreement with data.

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

Energy Technology Data Exchange (ETDEWEB)

Liao, J.; Kucukboyaci, V. N.; Nguyen, L.; Frepoli, C. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using the WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)

Suggestion for a homogenizer installation in LOFT small break two-phase measurement

International Nuclear Information System (INIS)

Rieger, G.

1981-07-01

The purpose of this task, which was performed as an Austrian inkind contribution for the INEL research program is a) the evaluation of literature concerning homogenizers to improve two phase flow measurements for the LOFT small break test series, b) design of a homogenizer and c) recommandation of the location of a homogenizer in the LOFT piping system. To optimize the location of the homogenizer LTSF-tests should be performed according to the suggestions in this paper. (author)

Integrated analysis for a small break LOCA in the IRIS reactor using MELCOR and RELAP5 codes

International Nuclear Information System (INIS)

Del Nevo, A.; Manfredini, A.; Oriolo, F.; Paci, S.; Oriani, L.

2004-01-01

The pressurized light water cooled, medium power (1000 MWt) IRIS (International Reactor Innovative and Secure) has been under development for four years by an international consortium of over 21 organizations from ten countries. The plant conceptual design was completed in 2001 and the preliminary design is nearing completion. The pre-application licensing process with NRC started in October, 2002 and IRIS is one of the designs considered by US utilities as part of the ESP (Early Site Permit) process. This paper's focus is on the use of well known computer codes for integrated (reactor vessel and containment) calculations of the IRIS response to a small break loss of coolant accident (LOCA). In IRIS, large break LOCA events are eliminated by the use of a layout configuration in which the reactor vessel contains all the reactor coolant system components including the core, control rod drive mechanisms, pressurizer, steam generators, and coolant pumps. Thus the IRIS configuration has no large loop piping; also, no pipes with a diameter greater than 0.1 meters are part of the reactor coolant system boundary. For small break LOCAs, IRIS features an innovative mitigation approach that is based on maintaining coolant inventory rather than designing high and low pressure injection systems to provide makeup coolant to the reactor to maintain core cooling. The novel IRIS approach requires development of evaluation models that are different from those used for the current generation of pressurized water reactors. An analysis of small break LOCAs for IRIS is documented in two companion papers, and has been developed using a preliminary evaluation model based on the explicit coupling of the RELAP5 and GOTHIC codes. The objective of this paper is to compare the results obtained via the coupled RELAP/GOTHIC code with different computational tools. A reference case from the preliminary IRIS safety assessment was selected, and the same small break LOCA sequence is analyzed using

Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

Energy Technology Data Exchange (ETDEWEB)

Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

2017-08-15

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

PCTRAN enhancement for large break loss of coolant accident concurrent with loss of offsite power in VVER-1000 simulation

Energy Technology Data Exchange (ETDEWEB)

Hadad, Kamal; Esmaeili-Sanjavanmareh, Mansour [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering

2017-05-15

PCTRAN capability to simulate a large break loss of coolant accident concurrent with the loss of offsite power in Bushehr Nuclear Power Plant is enhanced and investigated. Following the correction of the accident scenario for Bushehr nuclear power plant in PCTRAN, simulation results are compared with the final safety assessment report of that plant. As a result, the primary loop thermal hydraulics parameters including pressure, total flow rates, leakage flow rates and reactor power are in a good agreement with the reference data. Hot and cold leg temperature variations have the same trends as reference data but have a maximum of 80 C disagreement at the transient initiation. The reason for this disagreement is explained and its adjustment is discussed. Improvements of PCTRAN simulator are mainly due to enhancing user control for atmospheric steam dump valve, containment pressure and emergency core cooling systems which are thoroughly described in this paper.

Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

Directory of Open Access Journals (Sweden)

Hwang Bae

2017-08-01

Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

Study of Dam-break Due to Overtopping of Four Small Dams in the Czech Republic

Directory of Open Access Journals (Sweden)

Zakaraya Alhasan

2015-01-01

Full Text Available Dam-break due to overtopping is one of the most common types of embankment dam failures. During the floods in August 2002 in the Czech Republic, several small dams collapsed due to overtopping. In this paper, an analysis of the dam break process at the Luh, Velký Bělčický, Melín, and Metelský dams breached during the 2002 flood is presented. Comprehensive identification and analysis of the dam shape, properties of dam material and failure scenarios were carried out after the flood event to assemble data for the calibration of a numerical dam break model. A simple one-dimensional mathematical model was proposed for use in dam breach simulation, and a computer code was compiled. The model was calibrated using the field data mentioned above. Comparison of the erodibility parameters gained from the model showed reasonable agreement with the results of other authors.

Assessment of TRAC-PF1/MOD1 code for large break LOCA in PWR

International Nuclear Information System (INIS)

Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio; Abe, Yutaka.

1993-03-01

As the first step of the REFLA/TRAC code development, the TRAC/PF1/MOD1 code has been assessed for various experiments that simulate postulated large-break loss-of-coolant accident (LBLOCA) in PWR to understand the predictive capability and to identify the problem areas of the code. The assessment calculations were performed for separate effect tests for critical flow, counter current flow, condensation at cold leg and reflood as well as integral tests to understand predictability for individual phenomena. This report summarizes results from the assessment calculations of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The assessment calculations made clear the predictive capability and problem areas of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The areas, listed below, should be improved for more realistic and effective simulation of LBLOCA in PWR: (1) core heat transfer model during blowdown, (2) ECC bypass model at downcomer during refill, (3) condensation model during accumulator injection, and (4) core thermal hydraulic model during reflood. (author) 57 refs

Estimation of regional cutaneous cold sensitivity by analysis of the gasping response.

Science.gov (United States)

Burke, W E; Mekjavić, I B

1991-11-01

Regional cutaneous sensitivity to cooling was assessed in males by separately immersing four discrete skin regions in cold water (15 degrees C) during head-out immersion. The response measured was gasping at the onset of immersion; the gasping response appears to be the result of a nonthermoregulatory neurogenic drive from cutaneous cold receptors. Subjects of similar body proportions wore a neoprene "dry" suit modified to allow exposure to the water of either the arms, upper torso, lower torso, or legs, while keeping the unexposed skin regions thermoneutral. Each subject was immersed to the sternal notch in all four conditions of partial exposure plus one condition of whole body exposure. The five cold water conditions were matched by control immersions in lukewarm (34 degrees C) water, and trials were randomized. The magnitude of the gasping response was determined by mouth occlusion pressure (P0.1). For each subject, P0.1 values for the 1st min of immersion were integrated, and control trial values, although minimal, were subtracted from their cold water counterpart to account for any gasping due to the experimental design. Results were averaged and showed that the highest P0.1 values were elicited from whole body exposure, followed in descending order by exposures of the upper torso, legs, lower torso, and arms. Correction of the P0.1 response for differences in exposed surface area (A) and cooling stimulus (delta T) between regions gave a cold sensitivity index [CSI, P0.1/(A.delta T)] for each region and showed that the index for the upper torso was significantly higher than that for the arms or legs; no significant difference was observed between the indexes for the upper and lower torso.(ABSTRACT TRUNCATED AT 250 WORDS)

PWR small-break analysis using a PDP-11/AD10 computer system

International Nuclear Information System (INIS)

Venhuizen, J.R.; Hyer, F.K.

1983-01-01

A simulation of a pressurized water test reactor was developed to predict the dynamic response of the primary coolant system to gradual voiding caused by an anticipated transient or a small break. Comparison of the simulation results with data from the LOFT test reactor at the Idaho National Engineering Laboratory was performed to verify the models. The simulation, designed to operate on a PDP-11/55 minicomputer and Applied Dynamic AD10 synchronous digital computer, was used interactively to do scoping analysis prior to running the transient at the test reactor

The relationship between leg preference and knee mechanics during sidestepping in collegiate female footballers.

Science.gov (United States)

Brown, Scott R; Wang, Henry; Dickin, D Clark; Weiss, Kaitlyn J

2014-11-01

This study examined the relationship between leg preference and knee mechanics in females during sidestepping. Three-dimensional data were recorded on 16 female collegiate footballers during a planned 45° sidestep manoeuvre with their preferred and non-preferred kicking leg. Knee kinematics and kinetics during initial contact, weight acceptance, peak push-off, and final push-off phases of sidestepping were analysed in both legs. The preferred leg showed trivial to small increases (ES = 0.19-0.36) in knee flexion angle at initial contact, weight acceptance, and peak push-off, and small increases (ES = 0.21-0.34) in peak power production and peak knee extension velocity. The non-preferred leg showed a trivial increase (ES = 0.10) in knee abduction angle during weight acceptance; small to moderate increases (ES = 0.22-0.64) in knee internal rotation angle at weight acceptance, peak push-off, and final push-off; a small increase (ES = 0.22) in knee abductor moment; and trivial increases (ES = 0.09-0.14) in peak power absorption and peak knee flexion velocity. The results of this study show that differences do exist between the preferred and non-preferred leg in females. The findings of this study will increase the knowledge base of anterior cruciate ligament injury in females and can aid in the design of more appropriate neuromuscular, plyometric, and strength training protocols for injury prevention.

Study of the effect of slight variants to a 3-loop pressurized water nuclear reactor design in order to improve the reactor safety

International Nuclear Information System (INIS)

Castiglia, F.; Oliveri, E.; Taibi, S.; Vella, G.

1992-01-01

In order to improve the safety features of a 3-loop pressurized water nuclear reactor we propose a slight design variant consisting in the introduction of a bypass hole in the divider plate of the coolant chambers of the steam generators. The aim is to reduce both the extent and the duration of the core exposure and thus the maximum value of the peak cladding temperature, in case of a hypothetical cold leg small break loss of coolant accident. The proposal, as attested by a preliminary RELAP5/MOD3 analysis, seems to deserve some attention. (6 figures) (Author)

Review of boiling water reactor small break loss of coolant accidents

International Nuclear Information System (INIS)

Gururaj, P.M.; Dua, S.S.; Rao, A.S.

1981-01-01

This paper presents a review of the analytical and the experimental work performed by the General Electric Company to determine the performance of boiling water reactors (BWR) following postulated small break accidents (SBA). This review paper addresses the following issues: (1) the response of the BWR following small loss of inventory events; (2) methods of analysis and their justification; (3) necessity, if any, of operator action and the length of time available in which such action can be performed; and (4) operator interface following the SBA event. The results from these SBA studies for different BWR product lines show that even with the multiple system failures assumed, the BWR can successfully withstand an SBA. For a typical BWR/6, it takes the failure of 13 water delivery pumps to cause any significant core heatup. The only operator actions determined to be necessary are simple ones and ample time is available to the operator to perform these actions, if needed

High performance p-type segmented leg of misfit-layered cobaltite and half-Heusler alloy

International Nuclear Information System (INIS)

Hung, Le Thanh; Van Nong, Ngo; Snyder, G. Jeffrey; Viet, Man Hoang; Balke, Benjamin; Han, Li; Stamate, Eugen; Linderoth, Søren; Pryds, Nini

2015-01-01

Highlights: • p-type segmented leg of oxide and half-Heusler was for the first time demonstrated. • The maximum conversion efficiency reached a value of about 5%. • The results are among the highest reported values so far for oxide-based legs. • Oxide-based segmented leg is very promising for generating electricity. - Abstract: In this study, a segmented p-type leg of doped misfit-layered cobaltite Ca 2.8 Lu 0.15 Ag 0.05 Co 4 O 9+δ and half-Heusler Ti 0.3 Zr 0.35 Hf 0.35 CoSb 0.8 Sn 0.2 alloy was fabricated and characterized. The thermoelectric properties of single components, segmented leg, and the electrical contact resistance of the joint part were measured as a function of temperature. The output power generation characteristics of segmented legs were characterized in air under various temperature gradients, ΔT, with the hot side temperature up to 1153 K. At ΔT ≈ 756 K, the maximum conversion efficiency reached a value of ∼5%, which is about 65% of that expected from the materials without parasitic losses. The long-term stability investigation for two weeks at the hot and cold side temperatures of 1153/397 K shows that the segmented leg has good durability as a result of stable and low electrical resistance contacts

Axion cold dark matter in nonstandard cosmologies

International Nuclear Information System (INIS)

Visinelli, Luca; Gondolo, Paolo

2010-01-01

We study the parameter space of cold dark matter axions in two cosmological scenarios with nonstandard thermal histories before big bang nucleosynthesis: the low-temperature reheating (LTR) cosmology and the kination cosmology. If the Peccei-Quinn symmetry breaks during inflation, we find more allowed parameter space in the LTR cosmology than in the standard cosmology and less in the kination cosmology. On the contrary, if the Peccei-Quinn symmetry breaks after inflation, the Peccei-Quinn scale is orders of magnitude higher than standard in the LTR cosmology and lower in the kination cosmology. We show that the axion velocity dispersion may be used to distinguish some of these nonstandard cosmologies. Thus, axion cold dark matter may be a good probe of the history of the Universe before big bang nucleosynthesis.

Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

International Nuclear Information System (INIS)

Abass, N. M. N.

2012-02-01

A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

MELCOR based severe accident simulation for WWER-440 type nuclear power plants

International Nuclear Information System (INIS)

Vegh, E.; Buerger, L.; Gacs, A.; Gyenes, F.G.; Hozer, Z.; Makovi, P.

1997-01-01

SUBA is a MELCOR based severe accident simulator, installed this summer at the Hungarian Nuclear Safety Directorate. In this simulator the thermohydraulics, chemical reactions and material transport in the primary and secondary systems are calculated by the MELCOR code, but the containment, except the cavity, is modelled by the HERMET code, developed in our Institute. The instrumentation and control, the safety systems and the plant logic, are calculated by our models. This paper describes the main features of the used models and presents three different test transients. The presented transients are as follows: a small break LOCA, a cold leg large break LOCA, and the station blackout, without Diesel generators. In each treated transients the most important parameters are presented as time functions and the most significant events are analysed. (author)

Small extra dimensions from the interplay of gauge and supersymmetry breaking

International Nuclear Information System (INIS)

Buchmueller, W.; Catena, R.; Schmidt-Hoberg, K.

2008-03-01

Higher-dimensional theories provide a promising framework for unified extensions of the supersymmetric standard model. Compactifications to four dimensions often lead to U(1) symmetries beyond the standard model gauge group, whose breaking scale is classically undetermined. Without supersymmetry breaking, this is also the case for the size of the compact dimensions. Fayet-Iliopoulos terms generically fix the scale M of gauge symmetry breaking. The interplay with supersymmetry breaking can then stabilize the compact dimensions at a size 1/M, much smaller than the inverse supersymmetry breaking scale 1/μ. We illustrate this mechanism with an SO(10) model in six dimensions, compactified on an orbifold. (orig.)

Results of small break LOCA analysis for Kuosheng nuclear power plant using the RELAP5YA computer code

International Nuclear Information System (INIS)

Wang, L.C.; Jeng, S.C.; Chung, N.M.

2004-01-01

One lesson learned from the Three Mile Island (TMI) accident was the analysis methods used by Nuclear Steam Supply System (NSSS) vendors and/or nuclear fuel suppliers for small break Loss Of Coolant Accident (LOCA) analysis for compliance with appendix K to 10CFR50 should be revised, documented and submitted for USNRC approval and the plant-specific calculations using NRC-approved models for small-break LOCA to show compliance with 10CFR50.46 should be submitted for NRC approval. A study by Taiwan Power Company (TPC) under the guidance of Yankee Atomic Electric Company (YAEC) has been undertaken to perform this analysis for Kuosheng nuclear power plant. This paper presents the results of the analysis that are useful in satisfying the same requirements of the Republic Of China Atomic Energy Commission (ROCAEC). (author)

CATHARE2 calculation of SPE-3 test small break loca on PMK facility

Energy Technology Data Exchange (ETDEWEB)

Laugier, E.; Radet, J. [Institut de Protection et de Surete Nucleaire, Cadarache (France)

1995-09-01

Bind and post test calculations with CATHARE2 have been performed concerning the SPE-4 exercise organized under the auspices of IAEA on the hungarian PMK-2 facility, a one loop scaled model of VVER 440/213 Nuclear Power Plant. The SPE-4 test is a cold leg SBLOCA associated to a {open_quotes}bleed and feed{close_quotes} procedure applied in the secondary circuit. The present paper is devoted to the analysis of the post test calculation. For the first part of the transient (until the end of the SIT activations), the primary and secondary pressures are rather well predicted, leading to a good agreement with the experimental trips, as scram, flow coast down, SIT beginning and end of activation. Nevertheless, some discrepancy with the experiment may be due to an over prediction of the thermal exchanges from the primary to the secondary circuits. For the second part of the transient, the predicted primary circuit repressurization is shifted after the SITs are off, while in the experiment this event immediately follows the end of SIT activation. The delay in the calculation leads to underpredict primary and secondary pressures, thus anticipating the timing of events, such as LPIS and emergency feedwater activation.

Preliminary Analysis of Severe Accident Progression Initiated from Small Break LOCA of a SMART Reactor

International Nuclear Information System (INIS)

Jin, Young Ho; Park, Jong Hwa; Kim, Dong Ha; Cho, Seong Won

2010-01-01

SMART (System integrated Modular Advanced ReacTor), is under the development at Korea Atomic Energy Research Institute (KAERI). SMART is an integral type pressurized water reactor which contains a pressurizer, 4 reactor coolant pumps (RCPs), and 8 steam generator cassettes(S/Gs) in a single reactor vessel. This reactor has substantially enhanced its safety with an integral layout of its major components, 4 trains of safety injection systems (SISs), and an adoption of 4 trains of passive residual heat removal systems (PRHRS) instead of an active auxiliary feedwater system . The thermal power is 330 MWth. During the conceptual design stage, a preliminary PSA was performed. PSA results identified that a small break loss of coolant accident (SLOCA) with all safety injections unavailable is one of important severe core damage sequences. Clear understanding of this sequence helps in the developing accident mitigation strategies. MIDAS/SMR computer code is used to simulate the severe accident progression initiated from a small break LOCA in SMART reactor. This code has capability to model a helical steam generator which is adopted in SMART reactor. The important accident progression results for SMART reactor are then compared with the typical pressurized water reactor (PWR) result

Anatomic and functional leg-length inequality: A review and recommendation for clinical decision-making. Part I, anatomic leg-length inequality: prevalence, magnitude, effects and clinical significance

Directory of Open Access Journals (Sweden)

Knutson Gary A

2005-07-01

Full Text Available Abstract Background Leg-length inequality is most often divided into two groups: anatomic and functional. Part I of this review analyses data collected on anatomic leg-length inequality relative to prevalence, magnitude, effects and clinical significance. Part II examines the functional "short leg" including anatomic-functional relationships, and provides an outline for clinical decision-making. Methods Online database – Medline, CINAHL and MANTIS – and library searches for the time frame of 1970–2005 were done using the term "leg-length inequality". Results and Discussion Using data on leg-length inequality obtained by accurate and reliable x-ray methods, the prevalence of anatomic inequality was found to be 90%, the mean magnitude of anatomic inequality was 5.2 mm (SD 4.1. The evidence suggests that, for most people, anatomic leg-length inequality does not appear to be clinically significant until the magnitude reaches ~ 20 mm (~3/4". Conclusion Anatomic leg-length inequality is near universal, but the average magnitude is small and not likely to be clinically significant.

Comparison of steam-generator liquid holdup and core uncovery in two facilities of differing scale

International Nuclear Information System (INIS)

Motley, F.; Schultz, R.

1987-01-01

This paper reports on Run SB-CL-05, a test similar to Semiscale Run S-UT-8. The test results show that the core was uncovered briefly during the accident and that the rods overheated at certain core locations. Liquid holdup on the upflow side of the steam-generator tubes was observed. After the loop seal cleared, the core refilled and the rods cooled. These behaviors were similar to those observed in the Semiscale run. The Large-Scale Test Facility (LSTF) Run SB-CL-06 is a counterpart test to Semiscale Run S-LH-01. The comparison of the results of both tests shows similar phenomena. The similarity of phenomena in these two facilities build confidence that these results can be expected to occur in a PWR. Similar holdup has now been observed in the 6 tubes of Semiscale and in the 141 tubes of LSTF. It is now more believable that holdup may occur in a full-scale steam generator with 3000 or more tubes. These results confirm the scaling of these phenomena from Semiscale (1/1705) to LSTF (1/48). The TRAC results for SB-CL-05 are in reasonable agreement with the test data. TRAC predicted the core uncovery and resulting rod heatup. The liquid holdup on the upflow side of the steam-generator tubes was also correctly predicted. The clearing of the loop seal allowed core recovery and cooled the overheated rods just as it had in the data. The TRAC analysis results of Run SB-CL-05 are similar to those from Semiscale Run S-UT-8. The ability of the TRAC code to calculate the phenomena equally well in the two experiments of different scales confirms the scalability of the many models in the code that are important in calculating this small break

Effects of high temperature ECC injection on small and large break BWR LOCA simulation tests in ROSA-III program (RUNs 940 and 941)

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Nakamura, Hideo; Kumamaru, Hiroshige; Anoda, Yoshinari; Yonomoto, Taisuke; Murata, Hideo; Tasaka, Kanji

1990-03-01

The ROSA-III program, of which principal results are summarized in a report of JAERI 1307, conducted small and large-break loss-of-coolant experiments (RUNs 940 and 941) with high water temperature of the emergency core cooling system (ECCS) are one of the parametric study with respect to the ECCS effect on core cooling. This report presents all the experiment results of these two tests and describes additional finding with respect to the hot ECC effects on core cooling phenomena. By comparing these two tests (water temperature of 393 K) with the standard ECC tests of RUNs 922 and 926 (water temperature of 313 K), it was found that the ECC subcooling variation had a small influence on the core cooling phenomena in 5 % small break tests but had larger influence on them in 200 % break tests. The ECC subcooling effects described in the previous report are reviewed and the temperature distribution in the pressure vessel is investigated for these four tests. (author)

Simulation experiments for hot-leg U-bend two-phase flow phenomena

International Nuclear Information System (INIS)

Ishii, M.; Hsu, J.T.; Tucholke, D.; Lambert, G.; Kataoka, I.

1986-01-01

In order to study the two-phase natural circulation and flow termination during a small break loss of coolant accident in LWR, simulation experiments have been performed. Based on the two-phase flow scaling criteria developed under this program, an adiabatic hot leg U-bend simulation loop using nitrogen gas and water and a Freon 113 boiling and condensation loop were built. The nitrogen-water system has been used to isolate key hydrodynamic phenomena from heat transfer problems, whereas the Freon loop has been used to study the effect of phase changes and fluid properties. Various tests were carried out to establish the basic mechanism of the flow termination and reestablishment as well as to obtain essential information on scale effects of parameters such as the loop frictional resistance, thermal center, U-bend curvature and inlet geometry. In addition to the above experimental study, a preliminary modeling study has been carried out for two-phase flow in a large vertical pipe at relatively low gas fluxes typical of natural circulation conditions

Cold dark matter: Controversies on small scales.

Science.gov (United States)

Weinberg, David H; Bullock, James S; Governato, Fabio; Kuzio de Naray, Rachel; Peter, Annika H G

2015-10-06

The cold dark matter (CDM) cosmological model has been remarkably successful in explaining cosmic structure over an enormous span of redshift, but it has faced persistent challenges from observations that probe the innermost regions of dark matter halos and the properties of the Milky Way's dwarf galaxy satellites. We review the current observational and theoretical status of these "small-scale controversies." Cosmological simulations that incorporate only gravity and collisionless CDM predict halos with abundant substructure and central densities that are too high to match constraints from galaxy dynamics. The solution could lie in baryonic physics: Recent numerical simulations and analytical models suggest that gravitational potential fluctuations tied to efficient supernova feedback can flatten the central cusps of halos in massive galaxies, and a combination of feedback and low star formation efficiency could explain why most of the dark matter subhalos orbiting the Milky Way do not host visible galaxies. However, it is not clear that this solution can work in the lowest mass galaxies, where discrepancies are observed. Alternatively, the small-scale conflicts could be evidence of more complex physics in the dark sector itself. For example, elastic scattering from strong dark matter self-interactions can alter predicted halo mass profiles, leading to good agreement with observations across a wide range of galaxy mass. Gravitational lensing and dynamical perturbations of tidal streams in the stellar halo provide evidence for an abundant population of low-mass subhalos in accord with CDM predictions. These observational approaches will get more powerful over the next few years.

Simulation of a loss of primary coolant accident due to a large break in Angra 2 Nuclear Power Plant with RELAP5/MOD3.2.2G code

International Nuclear Information System (INIS)

Sabunddjian, Gaiane; Andrade, Delvonei Alves de

2003-01-01

This work presents the simulation, with RELAP5/MOD.3.2.2G code, of the postulate accident with loss of coolant in the primary circuit for large break (LBLOCA), which is described in Chapter 15 of the Final Safety Analysis Report of Angra 2 FSAR. The accident consists basically of the total break of the cold leg (Loop 20) of Angra 2 Plant. The rupture area considered is 4418 cm 2 , which represents 100% of the primary circuit pipe flow area. The Emergency Core Cooling System (ECCS) efficiency is also verified for this accident. In this simulation, failure and repair criteria are adopted for the ECCS components, in order to verify the system operation, in carrying out its function as expected by the project to preserve the integrity of the reactor core and to guarantee its cooling. LBLOCA accidents are characterized by a fast blowdown in the primary circuit to values that the low pressure injection system is activated and then, followed by the water injection by the accumulators. The thermal-hydraulic processes inherent to the accident phenomenon, such as hot leg vaporization and consequently core vaporization causing an inappropriate flow distribution in the reactor core, can lead to a reduction in the core liquid level, until the ECCS is capable to reflood it. It is important to point out that the results do not represent an independent calculation for the licensing process, but a calculation to give support to the qualification process of Angra 2 transient basic nodalization (author)

Angra 2 small break LOCA flow regime identification through RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Rocha, Marcelo da Silva; Sabundjian, Gaiane; Belchior Junior, Antonio; Andrade, Delvonei Alves de; Torres, Walmir Maximo; Conti, Thadeu das Neves; Macedo, Luiz Alberto; Umbehaun, Pedro Ernesto; Mesquita, Roberto Navarro de; Masotti, Paulo Henrique Ferraz, E-mail: msrocha@ipen.br, E-mail: gdjian@ipen.br, E-mail: abelchior@ipen.br, E-mail: delvonei@ipen.br, E-mail: wmtorres@ipen.br, E-mail: tnconti@ipen.br, E-mail: lamacedo@ipen.br, E-mail: umbehaun@ipen.br, E-mail: s, E-mail: rnavarro@ipen.br, E-mail: pmasotti@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2012-07-01

The purpose of this paper is to identify the flow regimes in the core of Angra 2 nuclear reactor with RELAP5/MOD3.2.gamma code (RELAP5, 2001). The postulated accident is the loss of coolant through a small break in the primary circuit (SBLOCA), which is described in Chapter 15 of the Final Safety Analysis Report of Angra 2 - FSAR (ETN, 2006). As the primary circuit pressure decreases due to the loss of coolant, several alternating two phase flow regimes are established in the primary circuit. This paper analyses the coolant two-phase flow behavior in the nuclear reactor core during the postulated accident. (author)

Security in Northern Europe after the Cold War

National Research Council Canada - National Science Library

Besserudhagen, Svein

1995-01-01

The end of the Cold War came with dramatic changes in Europe. NATO is searching for its future in a Europe threatened by instability and break down of government control and law and order in Russia...

An analysis of the influence of logistics activities on the export cold chain of temperature sensitive fruit through the Port of Cape Town

Directory of Open Access Journals (Sweden)

Leila L. Goedhals-Gerber

2015-09-01

Full Text Available Background: South Africa exports a large variety of different fruit types and cultivars worldwide. Yet, there is concern in the South African fruit industry that too much fruit and money is lost each year due to breaks along the fresh fruit export cold chain. Objective: The objective of this article was to identify the influence of logistics activities on breaks along the South African fruit export cold chain. The focus is specifically on temperature sensitive fruit, exported in refrigerated containers to Europe and the United Kingdom through the Port of Cape Town. This supply chain was selected as this was the most accessible supply chain in terms of retrieving the necessary temperature data. Method: The cold chain was investigated from the cold store, through all segments, until the Port of Cape Town. Temperature data collected with temperature monitoring devices from different fruit export supply chains of grapes, plums and pome fruit (apples and pears were analysed to identify the percentage of temperature breaks and the length of temperature breaks that occur at each segment of the cold chain. Results: The results show that a large number of breaks are experienced along South Africa’s fruit export cold chain, specifically at the interface between the cold store and the truck. In addition, the findings also show that there has been an improvement in the number of breaks experienced in the Port of Cape Town following the implementation of the NAVIS and Refcon systems. Conclusion: This article concludes by providing the fruit industry with areas that require addressing to improve operational procedures along the fruit export cold chain to help ensure that the fruit arrives at its final destination at optimal quality.

A through calculation of 1,100 MWe PWR large break LOCA by THYDE-P1 EM model

International Nuclear Information System (INIS)

Kanazawa, Masayuki; Asahi, Yoshiro; Hirano, Masashi

1984-07-01

THYDE-P1 is a code to analyze both the blowdown and refill-reflood phases of loss-of-coolant accidents (LOCAs) of pressurized water reactors (PWRs). Up to now, THYDE-P1 has been applied to various experiment analyses, which show its high capability to analyze LOCAs as a best estimate (BE) calculation code. In this report, evaluation model (EM) calculation method, especialy in the blowdown and refill phases, is established equivalently to WREM/J2 which is regarded as appropriate for an EM calculation code, and the results of them are compared and discussed. The present calculation was the first executed by THYDE-P1-EM, and was performed as Sample Calculation Run 80 which was a part of a series of THYDE-P sample calculations. The calculation was carried out from the LOCA initiation till 400 seconds for a guillotine break at the cold leg of a commercial 1,100 MWe PWR plant. The calculated results agreed well to that of the WREM/J2 code. (author)

Experiment on performance of upper head injection system with ROSA-II

International Nuclear Information System (INIS)

1976-09-01

Thermo-hydraulic behavior in the primary cooling system of a pressurized water reactor with an upper head injection system (UHI) in a postulated loss-of-coolant accident (LOCA) has been studied with ROSA-II test facility. Simulated UHI and internal structures of the pressure vessel were installed to the facility for the experiment. Nine maximum-sized double-ended break tests and one medium-sized split break test were performed for the cold-leg break condition. The results are as follows: (1) Fluid mixing in the upper head is not perfect. (2) Cold water injection into the steam or two-phase fluid causes violent depressurization due to the condensation. Flow pattern in the primary cooling system is largely influenced by the above two. (auth.)

Task driven optimal leg trajectories in insect-scale legged microrobots

Science.gov (United States)

Doshi, Neel; Goldberg, Benjamin; Jayaram, Kaushik; Wood, Robert

Origami inspired layered manufacturing techniques and 3D-printing have enabled the development of highly articulated legged robots at the insect-scale, including the 1.43g Harvard Ambulatory MicroRobot (HAMR). Research on these platforms has expanded its focus from manufacturing aspects to include design optimization and control for application-driven tasks. Consequently, the choice of gait selection, body morphology, leg trajectory, foot design, etc. have become areas of active research. HAMR has two controlled degrees-of-freedom per leg, making it an ideal candidate for exploring leg trajectory. We will discuss our work towards optimizing HAMR's leg trajectories for two different tasks: climbing using electroadhesives and level ground running (5-10 BL/s). These tasks demonstrate the ability of single platform to adapt to vastly different locomotive scenarios: quasi-static climbing with controlled ground contact, and dynamic running with un-controlled ground contact. We will utilize trajectory optimization methods informed by existing models and experimental studies to determine leg trajectories for each task. We also plan to discuss how task specifications and choice of objective function have contributed to the shape of these optimal leg trajectories.

Experiment data report for semiscale Mod-1 Test S-06-5. (LOFT counterpart test)

International Nuclear Information System (INIS)

1977-06-01

Recorded test data are presented for Test S-06-5 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-5 was conducted from initial conditions of 2272 psia and 536 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold legs of the intact and broken loops to simulate emergency core coolant injection in a PWR. The purpose of Test S-06-5 was to assess the influence of the break nozzle geometry on core thermal and system response and on the subcooled and low quality mass flow rates at the break locations

MARS-KS code validation activity through the atlas domestic standard problem

International Nuclear Information System (INIS)

Choi, K. Y.; Kim, Y. S.; Kang, K. H.; Park, H. S.; Cho, S.

2012-01-01

The 2 nd Domestic Standard Problem (DSP-02) exercise using the ATLAS integral effect test data was executed to transfer the integral effect test data to domestic nuclear industries and to contribute to improving the safety analysis methodology for PWRs. A small break loss of coolant accident of a 6-inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating interests from participants. Ten calculation results using MARS-KS code were collected, major prediction results were described qualitatively and code prediction accuracy was assessed quantitatively using the FFTBM. In addition, special code assessment activities were carried out to find out the area where the model improvement is required in the MARS-KS code. The lessons from this DSP-02 and recommendations to code developers are described in this paper. (authors)

Experimental research on pressurized water reactor(PWR) safety

International Nuclear Information System (INIS)

Kim, Dong Su; Chae, Sung Ki; Chang, Won Pyo

1991-12-01

The objective of this research is to analyze the experimental results already performed in BETHSY facility of CEA France and to establish essential technologies for the future implementation of both such an experiment and computer code assessment, which are not undergoing in Korea so far. The contents of the present study are divided into 2 categories; namely, analysis of the BETHSY experimental data received from CEA, and CATHARE computer code simulation for the selected experiments, i.e. 'Natural Circulation(Test 4.3a)' and '2 Cold Leg Break'. The later studies are performed under the aims of CATHARE assessment as well as qualification of KOSAC code developing at KAERI, which is the subject in the next year and will concern an adequacy of KOSAC for the prediction of low flow natural circulation and a small break transients. (Author)

Protein Determinants of Meiotic DNA Break Hotspots

Science.gov (United States)

Fowler, Kyle R.; Gutiérrez-Velasco, Susana

2013-01-01

SUMMARY Meiotic recombination, crucial for proper chromosome segregation and genome evolution, is initiated by programmed DNA double-strand breaks (DSBs) in yeasts and likely all sexually reproducing species. In fission yeast, DSBs occur up to hundreds of times more frequently at special sites, called hotspots, than in other regions of the genome. What distinguishes hotspots from cold regions is an unsolved problem, although transcription factors determine some hotspots. We report the discovery that three coiled-coil proteins – Rec25, Rec27, and Mug20 – bind essentially all hotspots with unprecedented specificity even without DSB formation. These small proteins are components of linear elements, are related to synaptonemal complex proteins, and are essential for nearly all DSBs at most hotspots. Our results indicate these hotspot determinants activate or stabilize the DSB-forming protein Rec12 (Spo11 homolog) rather than promote its binding to hotspots. We propose a new paradigm for hotspot determination and crossover control by linear element proteins. PMID:23395004

Flux rope breaking and formation of a rotating blowout jet

Science.gov (United States)

Joshi, Navin Chandra; Nishizuka, Naoto; Filippov, Boris; Magara, Tetsuya; Tlatov, Andrey G.

2018-05-01

We analysed a small flux rope eruption converted into a helical blowout jet in a fan-spine configuration using multiwavelength observations taken by Solar Dynamics Observatory, which occurred near the limb on 2016 January 9. In our study, first, we estimated the fan-spine magnetic configuration with the potential-field calculation and found a sinistral small filament inside it. The filament along with the flux rope erupted upwards and interacted with the surrounding fan-spine magnetic configuration, where the flux rope breaks in the middle section. We observed compact brightening, flare ribbons, and post-flare loops underneath the erupting filament. The northern section of the flux rope reconnected with the surrounding positive polarity, while the southern section straightened. Next, we observed the untwisting motion of the southern leg, which was transformed into a rotating helical blowout jet. The sign of the helicity of the mini-filament matches the one of the rotating jets. This is consistent with recent jet models presented by Adams et al. and Sterling et al. We focused on the fine thread structure of the rotating jet and traced three blobs with the speed of 60-120 km s- 1, while the radial speed of the jet is ˜400 km s- 1. The untwisting motion of the jet accelerated plasma upwards along the collimated outer spine field lines, and it finally evolved into a narrow coronal mass ejection at the height of ˜9Rsun. On the basis of detailed analysis, we discussed clear evidence of the scenario of the breaking of the flux rope and the formation of the helical blowout jet in the fan-spine magnetic configuration.

THE EFFECTS OF SINGLE LEG HOP PROGRESSION AND DOUBLE LEGS HOP PROGRESSION EXERCISE TO INCREASE SPEED AND EXPLOSIVE POWER OF LEG MUSCLE

Directory of Open Access Journals (Sweden)

Nining W. Kusnanik

2015-05-01

Full Text Available The main purpose of this study was to determine the effect of single leg hop progression and double legs hop progression exercise to increase speed and explosive power of leg muscles. Plyometric is one of the training methods that can increase explosive power. There are many models of plyometric training including single leg hop progression and double leg hop progression. This research was experimental using match subject design techniques. The subjects of this study were 39 students who joined basketball school club. There were 3 groups in this study: Group 1 were 13 students who given sin¬gle leg hop progression exercise, Group 2 were 13 students who given double legs hop progression exercise, Group 3 were 13 students who given conventional exercise. The data was collected during pre test and post test by testing 30m speed running and vertical jump. The data was analyzed using Analysis of Varians (Anova. It was found that there were significantly increased on speed and explosive power of leg muscles of Group 1 and Group 2. It can be stated that single leg hop progression exercise was more effective than double leg hop progression exercise. The recent findings supported the hypothesis that single leg hop progression and double legs hop progression exercise can increase speed and explosive power of leg muscles. These finding were supported by some previous studies (Singh, et al, 2011; Shallaby, H.K., 2010. The single leg hop progression is more effective than double legs hop progression. This finding was consistent with some previous evidences (McCurdy, et al, 2005; Makaruk et al, 2011.

Impact of UK NICE clinical guidelines 168 on referrals to a specialist academic leg ulcer service.

Science.gov (United States)

Davies, Huw Ob; Popplewell, Matthew; Bate, Gareth; Kelly, Lisa; Darvall, Katy; Bradbury, Andrew W

2018-03-01

Background Leg ulcers are a common cause of morbidity and disability and result in significant health and social care expenditure. The UK National Institute for Health and Care Excellence (NICE) Clinical Guideline (CG)168, published in July 2013, sought to improve care of patients with leg ulcers, recommending that patients with a break in the skin below the knee that had not healed within two weeks be referred to a specialist vascular service for diagnosis and management. Aim Determine the impact of CG168 on referrals to a leg ulcer service. Methods Patients referred with leg ulceration during an 18-month period prior to CG168 (January 2012-June 2013) and an 18-month period commencing six months after (January 2014-June 2015) publication of CG168 were compared. Results There was a two-fold increase in referrals (181 patients, 220 legs vs. 385 patients, 453 legs) but no change in mean age, gender or median-duration of ulcer at referral (16.6 vs. 16.2 weeks). Mean-time from referral to specialist appointment increased (4.8 vs. 6 weeks, p = 0.0001), as did legs with superficial venous insufficiency (SVI) (36% vs. 44%, p = 0.05). There was a trend towards more SVI endovenous interventions (32% vs. 39%, p = 0.271) with an increase in endothermal (2 vs. 32 legs, p = 0.001) but no change in sclerotherapy (24 vs. 51 legs) treatments. In both groups, 62% legs had compression. There was a reduction in legs treated conservatively with simple dressings (26% vs. 15%, p = 0.0006). Conclusions Since CG168, there has been a considerable increase in leg ulcer referrals. However, patients are still not referred until ulceration has been present for many months. Although many ulcers are multi-factorial and the mainstay of treatment remains compression, there has been an increase in SVI endovenous intervention. Further efforts are required to persuade community practitioners to refer patients earlier, to educate patients and encourage further investment in

Responses to Small Break Loss of Coolant Accidents for SMART

International Nuclear Information System (INIS)

Bae, Kyoo Hwan; Kim, Hee C.; Chang, Moon H.; Zee, Sung Q.; Kim, Si-Hwan; Lee, Un-Chul

2004-01-01

The SMART NSSS adopts the design characteristics of containing most of the primary circuit components, such as the reactor core, main coolant pumps (MCPs), steam generators (SGs), and N 2 gas pressurizer (PZR) in a single leak-tight Reactor Pressure Vessel (RPV) with a relatively large ratio of the primary coolant inventory to the core power compared to the conventional loop-type PWR. Due to these design characteristics, the SMART can fundamentally eliminate the possibility of Large Break Loss of Coolant Accidents (LBLOCAs), improve the natural circulation capability, and assure a sufficient time to mitigate the possibility of core uncover. Also, SMART adopts inherent safety improving features and passive engineered safety systems such as the substantially large negative moderator temperature coefficients, passive residual heat removal system, emergency core cooling system, and a steel-made leak-tight Safeguard Vessel (SV) housing the RPV. This paper presents the results of the safety analyses using a MARS/SMR code for the instantaneous guillotine ruptures of the major pipelines penetrating the RPV. The analysis results, employing conservative initial/boundary conditions and assumptions, show that the safety systems of the SMART basic design adequately remove the core decay heat without causing core uncover for all the cases of the Small Break Loss of Coolant Accidents (SBLOCAs). The sensitivity study results with variable SV conditions show that the reduced SV net free volume can shorten the time for reaching the thermal and mechanical equilibrium condition between the RPV and SV. Under these boundary conditions, the primary system inventory loss can be minimized and the core remains covered for a longer period of time without any makeup of the coolant. (authors)

Frequency probabilistic analysis of a small break LOCA due to a power operated relief valve (PORV) for Angra-1 pre-TMI and post-TMI

International Nuclear Information System (INIS)

Onusic Junior, J.

1986-01-01

After the TMI event efforts were aimed towards improvements in the operational and administrative procedures related to the power operated relief valves (PORVs) in order to decrease the probability of a small-break loss-of-coolant accident (LOCA) caused by stuck-open power operated relief valve. This paper presents a frequency probabilistic analysis of a small break LOCA due to a stuck open PORV and safety valve to the Angra I nuclear power plant in operating conditions pre-TMI and post-TMI. (Author) [pt

Simulation experiments of small break LOCA in upper plenum joint pipe for 5 MW heating test reactor

International Nuclear Information System (INIS)

Bo Jinhai; Jiang Shengyao; Zhang Youjie; Tong Yunxian; Sun Shusen; Yao Meisheng

1988-12-01

A simulation experiment of small break LOCA is introduced, which was caused by the breakdown of a small size or middle size pipe located at upper plenum, or by unexpected opening the safety valve. In the tests, the system pressure, temperature, void fraction and total loss of water were studied. The results showed that the total loss of water was nearly 20% of initial loading water. It means under this condition the 5MW low temperature heating reactor being built in Institute of Nnclear Engergy Technology of Tsinghua University is safe

ASTEC and MELCOR comparison for a VVER-1000 60 mm small break LOCA

International Nuclear Information System (INIS)

Georgieva, J.; Stefanova, A.; Groudev, P.; Tusheva, P.; Mladenov, I.; Dimov, D.; Passalacqua, R.

2005-01-01

In this paper a comparison between severe accident calculations performed for a WWER 1000 with the ASTEC1.1v0 and MELCOR 1.8.5 computer codes for a small break LOCA (ID 60 mm) without intervention of hydro accumulators is presented. This investigation has been performed in the framework of the SARNET project under the EURATOM 6th framework program. Once the accident sequence scenario is specified, both codes (MELCORE and ASTEC) are able to determine the core and containment damaged states, to estimate the release of radionuclides from the fuel as well as from the primary circuit and containment. Theses results are used to estimate the maximum period of the time during which the personnel could still take particular decisions in order to mitigate such an accident. The aim of the performed analysis is to estimate the discrepancy between ASTEC and MELCORE 1.8.5 calculations. Such discrepancies will be studied, if the case, proposal for ASTEC improvements will be made. Also the ASTEC capability to simulate specific reactor accident scenarios and/or particular safety systems will be tested. The final target is to propose severe accident management procedure for WWER 1000 reactors. In conclusions, the analysis for a small break LOCA (ID 60 mm without hydroelectricities) has shown some discrepancies between ASTEC and MELCORE especially during the degradation of the core. Further analyses are planed in which the MELCORE temperature 'set point' for core degradation (2520 K) will be progressively increased to approach the ASTEC one (which has been estimated to be about 3200 K). The comparison of the new results will allow a better evaluation of the in-vessel models implemented in ASTEC

Topological phases in frustrated synthetic ladders with an odd number of legs

Science.gov (United States)

Barbarino, Simone; Dalmonte, Marcello; Fazio, Rosario; Santoro, Giuseppe E.

2018-01-01

The realization of the Hofstadter model in a strongly anisotropic ladder geometry has now become possible in one-dimensional optical lattices with a synthetic dimension. In this work, we show how the Hofstadter Hamiltonian in such ladder configurations hosts a topological phase of matter which is radically different from its two-dimensional counterpart. This topological phase stems directly from the hybrid nature of the ladder geometry and is protected by a properly defined inversion symmetry. We start our analysis by considering the paradigmatic case of a three-leg ladder which supports a topological phase exhibiting the typical features of topological states in one dimension: robust fermionic edge modes, a degenerate entanglement spectrum, and a nonzero Zak phase; then, we generalize our findings—addressable in the state-of-the-art cold-atom experiments—to ladders with a higher number of legs.

LOCA analysis of SCWR-M with passive safety system

Energy Technology Data Exchange (ETDEWEB)

Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Fu, S.W. [Navy University of Engineering, Wuhan, Hubei (China); Xu, Z.H. [Shanghai Nuclear Engineering Research and Design Institute, Shanghai (China); Yang, Y.H. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Nuclear Technology, Karlsruhe Institute of Technology (KIT), Kaiserstr. 12, 76131 Karlsruhe (Germany)

2013-06-15

Highlights: • Application of the ATHLET-SC code to the trans-critical analysis for SCWR. • Development of a passive safety system for SCWR-M. • Analysis of hot/cold leg LOCA behaviour with different break size. • Introduction of some mitigation measures for SCWR-M -- Abstract: A new SCWR conceptual design (mixed spectrum supercritical water cooled reactor: SCWR-M) is proposed by Shanghai Jiao Tong University (SJTU). R and D activities covering core design, safety system design and code development of SCWR-M are launched at SJTU. Safety system design and analysis is one of the key tasks during the development of SCWR-M. Considering the current advanced reactor design, a new passive safety system for SCWR-M including isolation cooling system (ICS), accumulator injection system (ACC), gravity driven cooling system (GDCS) and automatic depressurization system (ADS) is proposed. Based on the modified and preliminarily assessed system code ATHLET-SC, loss of coolant accident (LOCA) analysis for hot and cold leg is performed in this paper. Three different break sizes are analyzed to clarify the hot and cold LOCA characteristics of the SCWR-M. The influence of the break location and break size on the safety performance of SCWR-M is also concluded. Several measures to induce the core coolant flow and to mitigate core heating up are also discussed. The results achieved so far demonstrate the feasibility of the proposed passive safety system to keep the SCWR-M core at safety condition during loss of coolant accident.

Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

Energy Technology Data Exchange (ETDEWEB)

Al-Falahi, A. [Helsinki Univ. of Technology, Espoo (Finland); Haennine, M. [VTT Energy, Espoo (Finland); Porkholm, K. [IVO International, Ltd., Vantaa (Finland)

1995-09-01

The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

Statistical Analysis of Hie (Cold Sensation and Hiesho (Cold Disorder in Kampo Clinic

Directory of Open Access Journals (Sweden)

Tetsuhiro Yoshino

2013-01-01

Full Text Available A cold sensation (hie is common in Japanese women and is an important treatment target in Kampo medicine. Physicians diagnose patients as having hiesho (cold disorder when hie disturbs their daily activity. However, differences between hie and hiesho in men and women are not well described. Hie can be of three types depending on body part where patients feel hie. We aimed to clarify the characteristics of patients with hie and hiesho by analyzing data from new patients seen at the Kampo Clinic at Keio University Hospital between 2008 and 2013. We collected information about patients’ subjective symptoms and their severity using visual analogue scales. Of 4,016 new patients, 2,344 complained about hie and 524 of those were diagnosed with hiesho. Hie was most common in legs/feet and combined with hands or lower back, rather than the whole body. Almost 30% of patients with hie felt upper body heat symptoms like hot flushes. Cold sensation was stronger in hiesho than non-hiesho patients. Patients with hie had more complaints. Men with hiesho had the same distribution of hie and had symptoms similar to women. The results of our study may increase awareness of hiesho and help doctors treat hie and other symptoms.

The Relationship among Leg Strength, Leg Power and Alpine Skiing Success.

Science.gov (United States)

Gettman, Larry R.; Huckel, Jack R.

The purpose of this study was to relate leg strength and power to alpine skiing success as measured by FIS points. Isometric leg strength was represented by the knee extension test described by Clarke. Leg power was measured by the vertical jump test and the Margaria-Kalamen stair run. Results in the strength and power tests were correlated with…

Support for cold neutron utilization

International Nuclear Information System (INIS)

Lee, Kye Hong; Han, Young Soo; Choi, Sungmin; Choi, Yong; Kwon, Hoon; Lee, Kwang Hee

2012-06-01

- Support for experiments by users of cold neutron scattering instrument - Short-term training of current and potential users of cold neutron scattering instrument for their effective use of the instrument - International collaboration for advanced utilization of cold neutron scattering instruments - Selection and training of qualified instrument scientists for vigorous research endeavors and outstanding achievements in experiments with cold neutron - Research on nano/bio materials using cold neutron scattering instruments - Bulk nano structure measurement using small angle neutron scattering and development of analysis technique

Small break loss of coolant accident analysis of advanced PWR plant designs utilizing DVI line venturis

International Nuclear Information System (INIS)

Kemper, Robert M.; Gagnon, Andre F.; McNamee, Kevin; Cheung, Augustine C.

1995-01-01

The Westinghouse Advanced Passive and evolutionary Pressurizer Water Reactors (i.e. AP600 and APWR) incorporate direct vessel injection (DVI) of emergency core coolant as a means of minimizing the potential spilling of emergency core cooling water during a loss of coolant accident (LOCA). As a result, the most limiting small break LOCA (SBLOCA) event for these designs, with respect core inventory makeup capability, is a postulated double ended rupture of one of the DVI lines. This paper presents the results of a design optimization study that examines the installation of a venturi in the DVI line as a means of limiting the reactor coolant lost from the reactor vessel. The comparison results demonstrate that by incorporating a properly sized venturi in the DVI line, core uncovery concerns as a result of a DVI line break can be eliminated for both the AP600 and APWR plants. (author)

Small break LOCA RELAP5/MOD3 uncertainty quantification: Bias and uncertainty evaluation for important phenomena

International Nuclear Information System (INIS)

Ortiz, M.G.; Ghan, L.S.; Vogl, J.

1991-01-01

The Nuclear Regulatory Commission (NRC) revised the Emergency Core Cooling System (ECCS) licensing rule to allow the use of Best Estimate (BE) computer codes, provided the uncertainty of the calculations are quantified and used in the licensing and regulation process. The NRC developed a generic methodology called Code Scaling, Applicability and Uncertainty (CSAU) to evaluate BE code uncertainties. The CSAU methodology was demonstrated with a specific application to a pressurized water reactor (PWR), experiencing a postulated large break loss-of-coolant accident (LBLOCA). The current work is part of an effort to adapt and demonstrate the CSAU methodology to a small break (SB) LOCA in a PWR of B and W design using RELAP5/MOD3 as the simulation tool. The subject of this paper is the Assessment and Ranging of Parameters (Element 2 of the CSAU methodology), which determines the contribution to uncertainty of specific models in the code

Radiative breaking of cosmologically acceptable grand unified theories

International Nuclear Information System (INIS)

Gato, B.; Leon, J.; Quiros, M.

1984-01-01

We present a cosmologically acceptable grand unified model where the breaking of SU(5) proceeds through radiative corrections induced by supergravity soft-breaking terms. The breaking scale is determined by dimensional transmutation. The model is compatible with the radiative breaking of SU(2)sub(L)xU(1)sub(Y) which provides an experimentally accessible low energy particle spectrum and small top quark mass. (orig.)

The one-leg standing radiograph

OpenAIRE

Pinsornsak, P.; Naratrikun, K.; Kanitnate, S.; Sangkomkamhang, T.

2016-01-01

Objectives The purpose of this study was to compare the joint space width between one-leg and both-legs standing radiographs in order to diagnose a primary osteoarthritis of the knee. Methods Digital radiographs of 100 medial osteoarthritic knees in 50 patients were performed. The patients had undergone one-leg standing anteroposterior (AP) views by standing on the affected leg while a both-legs standing AP view was undertaken while standing on both legs. The severity of the osteoarthritis wa...

Evaluation of steam generator U-tube integrity during PWR station blackout with secondary system depressurization

International Nuclear Information System (INIS)

Hidaka, Akihide; Asaka, Hideaki; Sugimoto, Jun; Ueno, Shingo; Yoshino, Takehito

1999-12-01

In PWR severe accidents such as station blackout, the integrity of steam generator U-tube would be threatened early at the transient among the pipes of primary system. This is due to the hot leg countercurrent natural circulation (CCNC) flow which delivers the decay heat of the core to the structures of primary system if the core temperature increases after the secondary system depressurization. From a view point of accident mitigation, this steam generator tube rupture (SGTR) is not preferable because it results in the direct release of primary coolant including fission products (FP) to the environment. Recent SCDAP/RELAP5 analyses by USNRC showed that the creep failure of pressurizer surge line which results in release of the coolant into containment would occur earlier than SGTR during the secondary system depressurization. However, the analyses did not consider the decay heat from deposited FP on the steam generator U-tube surface. In order to investigate the effect of decay heat on the steam generator U-tube integrity, the hot leg CCNC flow model used in the USNRC's calculation was, at first, validated through the analysis for JAERI's LSTF experiment. The CCNC model reproduced well the thermohydraulics observed in the LSTF experiment and thus the model is mostly reliable. An analytical study was then performed with SCDAP/RELAP5 for TMLB' sequence of Surry plant with and without secondary system depressurization. The decay heat from deposited FP was calculated by JAERI's FP aerosol behavior analysis code, ART. The ART analysis showed that relatively large amount of FPs may deposit on steam generator U-tube inlet mainly by thermophoresis. The SCDAP/RELAP5 analyses considering the FP decay heat predicted small safety margin for steam generator U-tube integrity during secondary system depressurization. Considering associated uncertainties in the analyses, the potential for SGTR cannot be ignored. Accordingly, this should be considered in the evaluation of merits

Single-leg squats can predict leg alignment in dancers performing ballet movements in "turnout".

Science.gov (United States)

Hopper, Luke S; Sato, Nahoko; Weidemann, Andries L

2016-01-01

The physical assessments used in dance injury surveillance programs are often adapted from the sports and exercise domain. Bespoke physical assessments may be required for dance, particularly when ballet movements involve "turning out" or external rotation of the legs beyond that typically used in sports. This study evaluated the ability of the traditional single-leg squat to predict the leg alignment of dancers performing ballet movements with turnout. Three-dimensional kinematic data of dancers performing the single-leg squat and five ballet movements were recorded and analyzed. Reduction of the three-dimensional data into a one-dimensional variable incorporating the ankle, knee, and hip joint center positions provided the strongest predictive model between the single-leg squat and the ballet movements. The single-leg squat can predict leg alignment in dancers performing ballet movements, even in "turned out" postures. Clinicians should pay careful attention to observational positioning and rating criteria when assessing dancers performing the single-leg squat.

Movement of the sacroiliac joint during the Active Straight Leg Raise test in patients with long-lasting severe sacroiliac joint pain.

Science.gov (United States)

Kibsgård, Thomas J; Röhrl, Stephan M; Røise, Olav; Sturesson, Bengt; Stuge, Britt

2017-08-01

The Active Straight Leg Raise is a functional test used in the assessment of pelvic girdle pain, and has shown to have good validity, reliability and responsiveness. The Active Straight Leg Raise is considered to examine the patients' ability to transfer load through the pelvis. It has been hypothesized that patients with pelvic girdle pain lack the ability to stabilize the pelvic girdle, probably due to instability or increased movement of the sacroiliac joint. This study examines the movement of the sacroiliac joints during the Active Straight Leg Raise in patients with pelvic girdle pain. Tantalum markers were inserted in the dorsal sacrum and ilium of 12 patients with long-lasting pelvic girdle pain scheduled for sacroiliac joint fusion surgery. Two to three weeks later movement of the sacroiliac joints during the Active Straight Leg Raise was measured with radiostereometric analysis. Small movements were detected. There was larger movement of the sacroiliac joint of the rested leg's sacroiliac joint compared to the lifted leg's side. A mean backward rotation of 0.8° and inward tilt of 0.3° were seen in the rested leg's sacroiliac joint. The movements of the sacroiliac joints during the Active Straight Leg Raise are small. There was a small backward rotation of the innominate bone relative to sacrum on the rested leg's side. Our findings contradict an earlier understanding that a forward rotation of the lifted leg's innominate occur while performing the Active Straight Leg Raise. Copyright © 2017. Published by Elsevier Ltd.

Performance Evaluation of SMART Passive Safety System for Small Break LOCA Using MARS Code

International Nuclear Information System (INIS)

Chun, Ji Han; Lee, Guy Hyung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo

2013-01-01

SMART has significantly enhanced safety by reducing its core damage frequency to 1/10 that of a conventional nuclear power plant. KAERI is developing a passive safety injection system to replace the active safety injection pump in SMART. It consists of four trains, each of which includes gravity-driven core makeup tank (CMT) and safety injection tank (SIT). This system is required to meet the passive safety performance requirements, i.e., the capability to maintain a safe shutdown condition for a minimum of 72 hours without an AC power supply or operator action in the case of design basis accidents (DBAs). The CMT isolation valve is opened by the low pressurizer pressure signal, and the SIT isolation valve is opened at 2 MPa. Additionally, two stages of automatic depressurization systems are used for rapid depressurization. Preliminary safety analysis of SMART passive safety system in the event of a small-break loss-of-coolant accident (SBLOCA) was performed using MARS code. In this study, the safety analysis results of a guillotine break of safety injection line which was identified as the limiting SBLOCA in SMART are given. The preliminary safety analysis of a SBLOCA for the SMART passive safety system was performed using the MARS code. The analysis results of the most limiting SI line guillotine break showed that the collapsed liquid level inside the core support barrel was maintained sufficiently high above the top of core throughout the transient. This means that the passive safety injection flow from the CMT and SIT causes no core uncovery during the 72 hours following the break with no AC power supply or operator action, which in turn results in a consistent decrease in the fuel cladding temperature. Therefore, the SMART passive safety system can meet the passive safety performance requirement of maintaining the plant at a safe shutdown condition for a minimum of 72 hours without AC power or operator action for a representing accident of SBLOCA

A review of the relationship between leg power and selected chronic disease in older adults

DEFF Research Database (Denmark)

Strollo, S. E.; Caserotti, Paolo; Ward, R. E.

2015-01-01

characterized. Importantly, individuals with these conditions have shown improved leg power with training. METHODS: A search was performed using PubMed to identify original studies published in English from January 1998 to August 2013. Leg power studies, among older adults ≥ 50 years of age, which assessed......OBJECTIVE: This review investigates the relationship between leg muscle power and the chronic conditions of osteoarthritis, diabetes mellitus, and cardiovascular disease among older adults. Current literature assessing the impact of chronic disease on leg power has not yet been comprehensively......), diabetes mellitus (n=5), and cardiovascular disease (n=6). Studies generally supported associations of lower leg power among older adults with chronic disease, although small sample sizes, cross-sectional data, homogenous populations, varied disease definitions, and inconsistent leg power methods limited...

Strongly correlated states of a small cold-atom cloud from geometric gauge fields

International Nuclear Information System (INIS)

Julia-Diaz, B.; Dagnino, D.; Barberan, N.; Guenter, K. J.; Dalibard, J.; Grass, T.; Lewenstein, M.

2011-01-01

Using exact diagonalization for a small system of cold bosonic atoms, we analyze the emergence of strongly correlated states in the presence of an artificial magnetic field. This gauge field is generated by a laser beam that couples two internal atomic states, and it is related to Berry's geometrical phase that emerges when an atom follows adiabatically one of the two eigenstates of the atom-laser coupling. Our approach allows us to go beyond the adiabatic approximation, and to characterize the generalized Laughlin wave functions that appear in the strong magnetic-field limit.

Strongly correlated states of a small cold-atom cloud from geometric gauge fields

Energy Technology Data Exchange (ETDEWEB)

Julia-Diaz, B. [Dept. ECM, Facultat de Fisica, U. Barcelona, E-08028 Barcelona (Spain); ICFO-Institut de Ciencies Fotoniques, Parc Mediterrani de la Tecnologia, E-08860 Barcelona (Spain); Dagnino, D.; Barberan, N. [Dept. ECM, Facultat de Fisica, U. Barcelona, E-08028 Barcelona (Spain); Guenter, K. J.; Dalibard, J. [Laboratoire Kastler Brossel, CNRS, UPMC, Ecole Normale Superieure, 24 rue Lhomond, F-75005 Paris (France); Grass, T. [ICFO-Institut de Ciencies Fotoniques, Parc Mediterrani de la Tecnologia, E-08860 Barcelona (Spain); Lewenstein, M. [ICFO-Institut de Ciencies Fotoniques, Parc Mediterrani de la Tecnologia, E-08860 Barcelona (Spain); ICREA-Institucio Catalana de Recerca i Estudis Avancats, E-08010 Barcelona (Spain)

2011-11-15

Using exact diagonalization for a small system of cold bosonic atoms, we analyze the emergence of strongly correlated states in the presence of an artificial magnetic field. This gauge field is generated by a laser beam that couples two internal atomic states, and it is related to Berry's geometrical phase that emerges when an atom follows adiabatically one of the two eigenstates of the atom-laser coupling. Our approach allows us to go beyond the adiabatic approximation, and to characterize the generalized Laughlin wave functions that appear in the strong magnetic-field limit.

Classification of a Supersolid: Trial Wavefunctions, Symmetry Breakings and Excitation Spectra

Science.gov (United States)

Chen, Yu; Ye, Jinwu; Tian, Guangshan

2012-11-01

A state of matter is characterized by its symmetry breaking and elementary excitations. A supersolid is a state which breaks both translational symmetry and internal U(1) symmetry. Here, we review some past and recent works in phenomenological Ginsburg-Landau theories, ground state trial wavefunctions and microscopic numerical calculations. We also write down a new effective supersolid Hamiltonian on a lattice. The eigenstates of the Hamiltonian contains both the ground state wavefunction and all the excited states (supersolidon) wavefunctions. We contrast various kinds of supersolids in both continuous systems and on lattices, both condensed matter and cold atom systems. We provide additional new insights in studying their order parameters, symmetry breaking patterns, the excitation spectra and detection methods.

Foot, leg, and ankle swelling

Science.gov (United States)

Swelling of the ankles - feet - legs; Ankle swelling; Foot swelling; Leg swelling; Edema - peripheral; Peripheral edema ... Foot, leg, and ankle swelling is common when the person also: Is overweight Has a blood clot in the leg Is older Has ...

Effects of water temperature on breeding phenology, growth and timing of metamorphosis of foothill yellow-legged frogs (Rana boylii) on the mainstem and selected tributaries of California's Trinity River - 2004-2009.

Science.gov (United States)

Clara Wheeler; James Bettaso; Donald Ashton; Hartwell Welsh

2013-01-01

The cold temperatures maintained in the Trinity River are beneficial to fish but may be problematic for foothill yellow-legged frogs. We examined the timing of breeding, reproductive output, and growth and development of tadpoles for populations of foothill yellow-legged frogs on the mainstem and six tributaries of the Trinity River. On the colder mainstem, onset of...

Development of the MARS input model for Ulchin 1/2 transient analyzer

International Nuclear Information System (INIS)

Jeong, J. J.; Kim, K. D.; Lee, S. W.; Lee, Y. J.; Chung, B. D.; Hwang, M.

2003-03-01

KAERI has been developing the NSSS transient analyzer based on best-estimate codes for Ulchin 1/2 plants. The MARS and RETRAN code are used as the best-estimate codes for the NSSS transient analyzer. Among the two codes, the MARS code is to be used for realistic analysis of small- and large-break loss-of-coolant accidents, of which break size is greater than 2 inch diameter. This report includes the input model requirements and the calculation note for the Ulchin 1/2 MARS input data generation (see the Appendix). In order to confirm the validity of the input data, we performed the calculations for a steady state at 100 % power operation condition and a double-ended cold leg break LOCA. The results of the steady-state calculation agree well with the design data. The results of the LOCA calculation seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the MARS input data can be used as a base input deck for the MARS transient analyzer for Ulchin 1/2

Development of the MARS input model for Ulchin 3/4 transient analyzer

International Nuclear Information System (INIS)

Jeong, J. J.; Kim, K. D.; Lee, S. W.; Lee, Y. J.; Lee, W. J.; Chung, B. D.; Hwang, M. G.

2003-12-01

KAERI has been developing the NSSS transient analyzer based on best-estimate codes.The MARS and RETRAN code are adopted as the best-estimate codes for the NSSS transient analyzer. Among these two codes, the MARS code is to be used for realistic analysis of small- and large-break loss-of-coolant accidents, of which break size is greater than 2 inch diameter. This report includes the MARS input model requirements and the calculation note for the MARS input data generation (see the Appendix) for Ulchin 3/4 plant analyzer. In order to confirm the validity of the input data, we performed the calculations for a steady state at 100 % power operation condition and a double-ended cold leg break LOCA. The results of the steady-state calculation agree well with the design data. The results of the LOCA calculation seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the MARS input data can be used as a base input deck for the MARS transient analyzer for Ulchin 3/4

Herpes Simplex Virus (Cold Sores)

Science.gov (United States)

... Print Share Cold Sores in Children: About the Herpes Simplex Virus Page Content ​A child's toddler and ... Cold sores (also called fever blisters or oral herpes) start as small blisters that form around the ...

Effect of cold indoor environment on physical performance of older women living in the community.

Science.gov (United States)

Lindemann, Ulrich; Oksa, Juha; Skelton, Dawn A; Beyer, Nina; Klenk, Jochen; Zscheile, Julia; Becker, Clemens

2014-07-01

the effects of cold on older persons' body and mind are not well documented, but with an increased number of older people with decreasing physical performance, these possible effects need to be understood. to investigate the effect of cold indoor environment on physical performance of older women. cross-sectional experimental study with two test conditions. movement laboratory in a climate chamber. eighty-eight community-dwelling, cognitively unimpaired older women (mean age 78 years). participants were exposed to moderately cold (15°C) and warm/normal (25°C) temperature in a climate chamber in random order with an interval of 1 week. The assessment protocol included leg extensor power (Nottingham Power Rig), sit-to-stand performance velocity (linear encoder), gait speed, walk-ratio (i.e. step length/cadence on an instrumented walk way), maximal quadriceps and hand grip strength. physical performance was lower in 15°C room temperature compared with 25°C room temperature for leg extensor power (P environment decreased important physical performance measures necessary for independent living. © The Author 2014. Published by Oxford University Press on behalf of the British Geriatrics Society. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Breaking the Paradox of Innovation in Small Businesses through Sustaining and Disruptive Reinvention

Directory of Open Access Journals (Sweden)

Vicki Baard

2007-06-01

Full Text Available In 2005 Deloitte Research released a paper examining the phenomenon they refer to asthe ‘innovation paradox’: the inability or reluctance of manufacturing firms to pursuestrategies that build the operational capabilities necessary for innovation that willprovide both profitability and growth. The report claims that this is due to the rapidlyincreasing complexity of global markets and the lack of synchronising innovation effortsacross their value chain, thus positioning the problem as an important contemporaryissue. While the research did not specifically target small and medium enterprises, theimplications for this business sector are considerable given their substantial contributionto global economies and their high failure rates in the first three to five years ofoperation. While not questioning the data in the Deloitte research, this paper doesquestion the assumption that the phenomenon is irreversible and the apparentunderlying self-fulfilling prophecy with respect to small to medium enterprises. Todemonstrate this the authors draw on a case study of a small manufacturing company inrural New South Wales, Australia, which operated between 1889 and 1983, to show thatthe breaking of the innovation paradox was successfully achieved by this firm in the latenineteenth and early twentieth century. Applying the case study to the Deloitte modelthe study demonstrates contemporary similarities by overlaying the Laycock history onthe successes / failures identified by Deloitte.

VVER-1000 small-medium break LOCAs predictions by ASTEC

International Nuclear Information System (INIS)

Georgieva, J.; Stefanova, A.; Atanasova, B.; Groudev, P.; Tusheva, P.; Mladenov, I.; Dimov, D.; Passalacqua, R.

2005-01-01

This paper deals with an assessment of ASTEC1.1v0 code in the simulation of small and medium break LOCAs (ranging from 30mm up to 70mm equivalent diameter). The reference power plant for this analysis is a VVER-1000/V320 (e.g. Units 5 and 6 at Kozloduy NPP). A preliminary comparison with MELCOR and RELAP-SCDAP severe accident codes will be discussed. This investigation has been performed in the framework of the SARNET project (under the Euratom 6 th framework program) by the FoBAUs group (Forum of Bulgarian ASTEC users). The FoBAUs group aims at the validation of the ASTEC code in the field of severe accidents. Future activities will target the ASTEC capability (as a PSA-level 2 tool) to simulate a large range of reactor accident scenarios with intervention of safety systems (either passive systems or operated by operators). The final target is to assess Severe Accident Management (SAM) procedures for VVER-1000 reactors. The ASTEC1.1v0 code version here used is the one released in June 2004 by the French IRSN (Institut de Radioprotection et de Surete Nucleaire) and the German GRS (Gesellschaft ReactorSicherheit mbH). (author)

Effects of moist cold stratification on germination, plant growth regulators, metabolites and embryo ultrastructure in seeds of Acer morrisonense (Sapindaceae).

Science.gov (United States)

Chen, Shun-Ying; Chou, Shih-Han; Tsai, Ching-Chu; Hsu, Wen-Yu; Baskin, Carol C; Baskin, Jerry M; Chien, Ching-Te; Kuo-Huang, Ling-Long

2015-09-01

Breaking of seed dormancy by moist cold stratification involves complex interactions in cells. To assess the effect of moist cold stratification on dormancy break in seeds of Acer morrisonense, we monitored percentages and rates of germination and changes in plant growth regulators, sugars, amino acids and embryo ultrastructure after various periods of cold stratification. Fresh seeds incubated at 25/15 °C for 24 weeks germinated to 61%, while those cold stratified at 5 °C for 12 weeks germinated to 87% in 1 week. Neither exogenous GA3 nor GA4 pretreatment significantly increased final seed germination percentage. Total ABA content of seeds cold stratified for 12 weeks was reduced about 3.3-fold, to a concentration similar to that in germinated seeds (radicle emergence). Endogenous GA3 and GA7 were detected in 8-week and 12-week cold stratified seeds but not in fresh seeds. Numerous protein and lipid bodies were present in the plumule, first true leaves and cotyledons of fresh seeds. Protein and lipid bodies decreased greatly during cold stratification, and concentrations of total soluble sugars and amino acids increased. The major non-polar sugars in fresh seeds were sucrose and fructose, but sucrose increased and fructose decreased significantly during cold stratification. The major free amino acids were proline and tryptophan in fresh seeds, and proline increased and tryptophan decreased during cold stratification. Thus, as dormancy break occurs during cold stratification seeds of A. morrisonense undergo changes in plant growth regulators, proteins, lipids, sugars, amino acids and cell ultrastructure. Copyright © 2015 Elsevier Masson SAS. All rights reserved.

On cold spots in tumor subvolumes

International Nuclear Information System (INIS)

Tome, Wolfgang A.; Fowler, Jack F.

2002-01-01

Losses in tumor control are estimated for cold spots of various 'sizes' and degrees of 'cold dose'. This question is important in the context of intensity modulated radiotherapy where differential dose-volume histograms (DVHs) for targets that abut a critical structure often exhibit a cold dose tail. This can be detrimental to tumor control probability (TCP) for fractions of cold volumes even as small as 1%, if the cold dose is lower than the prescribed dose by substantially more than 10%. The Niemierko-Goitein linear-quadratic algorithm with γ 50 slope 1-3 was used to study the effect of cold spots of various degrees (dose deficit below the prescription dose) and size (fractional volume of the cold dose). A two-bin model DVH has been constructed in which the cold dose bin is allowed to vary from a dose deficit of 1%-50% below prescription dose and to have volumes varying from 1% to 90%. In order to study and quantify the effect of a small volume of cold dose on TCP and effective uniform dose (EUD), a four-bin DVH model has been constructed in which the lowest dose bin, which has a fractional volume of 1%, is allowed to vary from 10% to 45% dose deficit below prescription dose. The highest dose bin represents a simultaneous boost. For fixed size of the cold spot the calculated values of TCP decreased rapidly with increasing degrees of cold dose for any size of the cold spot, even as small as 1% fractional volume. For the four-subvolume model, in which the highest dose bin has a fractional volume of 80% and is set at a boost dose of 10% above prescription dose, it is found that the loss in TCP and EUD is moderate as long as the cold 1% subvolume has a deficit less than approximately 20%. However, as the dose deficit in the 1% subvolume bin increases further it drives TCP and EUD rapidly down and can lead to a serious loss in TCP and EUD. Since a dose deficit to a 1% volume of the target that is larger than 20% of the prescription dose may lead to serious loss of

Optimisation of the flow path in a conceptual pool type reactor under natural circulation with lead coolant

International Nuclear Information System (INIS)

Thiele, R.; Anglart, H.

2014-01-01

This contribution investigates the effects of a bypass flow blocking bottom plate and the influence of the heat transfer between the hot and cold leg in a small pool type reactor cooled through natural convection with lead coolant. The computations are carried out using 3D computational fluid dynamics, where small-detail parts, such as the core and heat exchangers are modeled using a porous media approach. The introduction of full conjugate heat transfer shows that the heat transfer between the hot and cold leg can deteriorate flow in the cold leg and lead to recirculation zones. These zones become even more pronounced with the introduction of a bottom plate, which on the other hand also increases the flow through the core and lowers the maximum temperature in the core by approximately 150 K. Based on the results, redesign suggestions for the bottom plate and the internal wall are made. (author)

Development of a qualified nodalization for small-break LOCA transient analysis in PSB-VVER integral test facility by RELAP5 system code

Energy Technology Data Exchange (ETDEWEB)

Shahedi, S. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); Jafari, J., E-mail: jalil_jafari@yahoo.co [Reactors and Accelerators R and D School, Nuclear Science and Technology Research Institute, North Kargar Street, Tehran (Iran, Islamic Republic of); Boroushaki, M. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); D' Auria, F. [DIMNP, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

2010-10-15

This paper deals with development and qualification of a nodalization for modeling of the PSB-VVER integral test facility (ITF) by RELAP5/MOD3.2 code and prediction of its primary and secondary systems behaviors at steady state and transient conditions. The PSB-VVER is a full-height, 1/300 volume and power scale representation of a VVER-1000 NPP. A RELAP5 nodalization has been developed for PSB-VVER modeling and a nodalization qualification process has been applied for the developed nodalization at steady state and transient levels and a qualified nodalization has been proposed for modeling of the PSB ITF. The 11% small-break loss-of-coolant-accident (SBLOCA), i.e. rupture of one of the hydroaccumulators (HA) injection lines in the upper plenum (UP) region of reactor pressure vessel (RPV) below the hot legs (HL), inlets has been considered for nodalization qualification process. The influence of the different steam generator (SG) nodalizations on the RELAP5 results and on the nodalization qualification process has been examined. The 'steady state' qualification level includes checking the correctness of the initial and boundary conditions and geometrical fidelity. In the 'transient' qualification level, the time dependent results of the code calculation are compared with the experimental time trends from both the qualitative and quantitative point of view. For quantitative assessment of the results, a Fast Fourier Transform Based Method (FFTBM) has been used. The FFTBM was used to establish a range in which the steam generators nodalizations can vary.

Hydrogen and steam distribution following a small-break LOCA in large dry containment

Institute of Scientific and Technical Information of China (English)

DENG Jian; CAO Xuewu

2007-01-01

The hydrogen deflagration is one of the major risk contributors to threaten the integrity of the containment in a nuclear power plant, and hydrogen control in the case of severe accidents is required by nuclear regulations.Based on the large dry containment model developed with the integral severe-accident analysis tool, a small-break loss-of-coolant-accident (LOCA) without HPI, LPI, AFW and containment sprays, leading to the core degradation and large hydrogen generation, is calculated. Hydrogen and steam distribution in containment compartments is investigated. The analysis results show that significant hydrogen deflagration risk exits in the reactor coolant pump (RCP)compartment and the cavity during the early period, if no actions are taken to mitigate the effects of hydrogen accumulation.

Single-leg squats can predict leg alignment in dancers performing ballet movements in “turnout”

Science.gov (United States)

Hopper, Luke S; Sato, Nahoko; Weidemann, Andries L

2016-01-01

The physical assessments used in dance injury surveillance programs are often adapted from the sports and exercise domain. Bespoke physical assessments may be required for dance, particularly when ballet movements involve “turning out” or external rotation of the legs beyond that typically used in sports. This study evaluated the ability of the traditional single-leg squat to predict the leg alignment of dancers performing ballet movements with turnout. Three-dimensional kinematic data of dancers performing the single-leg squat and five ballet movements were recorded and analyzed. Reduction of the three-dimensional data into a one-dimensional variable incorporating the ankle, knee, and hip joint center positions provided the strongest predictive model between the single-leg squat and the ballet movements. The single-leg squat can predict leg alignment in dancers performing ballet movements, even in “turned out” postures. Clinicians should pay careful attention to observational positioning and rating criteria when assessing dancers performing the single-leg squat. PMID:27895518

Gravitational waves in cold dark matter

Science.gov (United States)

Flauger, Raphael; Weinberg, Steven

2018-06-01

We study the effects of cold dark matter on the propagation of gravitational waves of astrophysical and primordial origin. We show that the dominant effect of cold dark matter on gravitational waves from astrophysical sources is a small frequency dependent modification of the propagation speed of gravitational waves. However, the magnitude of the effect is too small to be detected in the near future. We furthermore show that the spectrum of primordial gravitational waves in principle contains detailed information about the properties of dark matter. However, depending on the wavelength, the effects are either suppressed because the dark matter is highly nonrelativistic or because it contributes a small fraction of the energy density of the universe. As a consequence, the effects of cold dark matter on primordial gravitational waves in practice also appear too small to be detectable.

An on-line pressurizer surveillance system design to prevent small-break loss-of-coolant accidents through power-operated relief valves using a microcomputer

International Nuclear Information System (INIS)

Lee, J.H.; Chang, S.H.

1987-01-01

A small-break loss-of-coolant accident (LOCA) caused by a stuck-open power-operated relief valve is one of the important contributors to nuclear power plant risk. A pressurizer surveillance system was designed to use a microcomputer to prevent the malfunction of the system; the effect of this improvement has been assessed through probabilistic risk assessment. The microcomputer diagnoses the malfunction of the system by a process-checking method and automatically performs the backup action related to each malfunction. This improvement means that we can correctly diagnose ''spurious opening,'' ''failure to reclose,'' and ''small-break LOCA,'' which are difficult for operators to diagnose quickly and correctly, and by taking automatic backup action one can reduce the probability of human error

Strongly coupled semidirect mediation of supersymmetry breaking

International Nuclear Information System (INIS)

Ibe, M.; Izawa, K.-I.; Nakai, Y.

2009-01-01

Strongly coupled semidirect gauge mediation models of supersymmetry breaking through massive mediators with standard-model charges are investigated by means of composite degrees of freedom. Sizable mediation is realized to generate the standard-model gaugino masses for a small mediator mass without breaking the standard-model symmetries.

Break a Leg

Science.gov (United States)

Gisolfi, Peter A.

2005-01-01

The basic dilemma in creating performance spaces in secondary schools is that such spaces often must serve multiple functions. Ideally, there should be three different spaces: one to present dramas; another for musical performances; and a third for assemblies--lectures, movies, political and community events. Often, the unsatisfactory solution to…

Genetic parameters for claw and leg health, foot and leg conformation, and locomotion in Danish Holsteins

DEFF Research Database (Denmark)

Laursen, M. V.; Boelling, D.; Mark, Thomas

2009-01-01

was defined as absence of hock infection, swollen hock, and bruising. The potential indicators were locomotion and foot and leg conformation, represented by rear leg side view, rear leg rear view, foot angle, and apparent hock quality and bone structure. The study was conducted using records from 429......,877 Danish Holstein cows in first lactation. Binary health traits were divided into 3 subcategories: claw health, leg health, and absence of all claw and leg disorders. Genetic (r(g)) and phenotypic correlations were estimated using a bivariate linear sire model and REML. Estimated heritabilities were 0.......01 for all 3 combined claw and leg health traits (on the observed binary scale), 0.09 for locomotion, 0.14 for rear leg rear view, 0.19 for rear leg side view, 0.13 for foot angle, 0.22 for apparent hock quality, and 0.27 for apparent bone structure. Heritabilities were 0.06 and 0.01 for claw health and leg...

Setting parameters in the cold chain

Directory of Open Access Journals (Sweden)

Victoria Rodríguez

2011-12-01

Full Text Available Breaks in the cold chain are important economic losses in food and pharmaceutical companies. Many of the failures in the cold chain are due to improper adjustment of equipment parameters such as setting the parameters for theoretical conditions, without a corresponding check in normal operation. The companies that transport refrigeratedproducts must be able to adjust the parameters of the equipment in an easy and quick to adapt their functioning to changing environmental conditions. This article presents the results of a study carried out with a food distribution company. The main objective of the study is to verify the effectiveness of Six Sigma as a methodological toolto adjust the equipment in the cold chain. The second objective is more speciÞ c and is to study the impact of: reducing the volume of storage in the truck, the initial temperature of the storage areain the truck and the frequency of defrost in the transport of refrigerated products.

Stable walking with asymmetric legs

International Nuclear Information System (INIS)

Merker, Andreas; Rummel, Juergen; Seyfarth, Andre

2011-01-01

Asymmetric leg function is often an undesired side-effect in artificial legged systems and may reflect functional deficits or variations in the mechanical construction. It can also be found in legged locomotion in humans and animals such as after an accident or in specific gait patterns. So far, it is not clear to what extent differences in the leg function of contralateral limbs can be tolerated during walking or running. Here, we address this issue using a bipedal spring-mass model for simulating walking with compliant legs. With the help of the model, we show that considerable differences between contralateral legs can be tolerated and may even provide advantages to the robustness of the system dynamics. A better understanding of the mechanisms and potential benefits of asymmetric leg operation may help to guide the development of artificial limbs or the design novel therapeutic concepts and rehabilitation strategies.

LOFT/LP-02-6, Loss of Fluid Test, 1. OECD Large Break Experiment

International Nuclear Information System (INIS)

1989-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The fourth OECD LOFT experiment was conducted on 3 October 1983. This was the first OECD LOFT large break experiment. The initial and boundary conditions were chosen to be representative of USNRC licensing limits for commercial PWRs. This included loss of off-site power coincident with LOCA initiation. This experiment included the first use in LOFT of pressurized fuel rods in the center bundle. The experiment was initiated by opening the quick-opening blow-down valves in the broken hot and cold legs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

Simulation of a beyond design-basis-accident with RELAP5/MOD3.1

Energy Technology Data Exchange (ETDEWEB)

Banati, J. [Lappeenranta Univ. of Technology, Lappeenranta (Finland)

1995-09-01

This paper summarizes the results of analyses, parametric and sensitivity studies, performed using the RELAP5/MOD3.1 computer code for the 4th IAEA Standard Problem Exercise (SPE-4). The test, conducted on the PMK-2 facility in Budapest, involved simulation of a Small Break Loss Of Coolant Accident (SBLOCA) with a 7.4% break in the cold leg of a VVER-440 type pressurized water reactor. According to the scenario, the unavailability of the high pressure injection system led to a beyond design basis accident. For prevention of core damage, secondary side bleed-and-feed accident management measures were applied. A brief description of the PMK-2 integral type test facility is presented, together with the profile and some key phenomenological aspects of this particular experiment. Emphasis is placed on the ability of the code to predict the main trends observed in the test and thus, an assessment is given for the code capabilities to represent the system transient.

Sleep apnea in patients reporting insomnia or restless legs symptoms.

Science.gov (United States)

Bianchi, M T; Goparaju, B; Moro, M

2016-01-01

Insomnia and restless legs syndrome (RLS) are defined by self-reported symptoms, and polysomnography (PSG) is not routinely indicated. Occult obstructive sleep apnea (OSA), common even in asymptomatic adults, may complicate management of patients presenting with insomnia or restless legs. To this end, we investigated objective sleep apnea metrics in a large retrospective cohort according to self-reported symptom profiles. We compared sleep apnea findings in patients referred to our center according to self-reported symptoms associated with insomnia, sleep apnea, and restless legs. The cohort included over 1900 adults who underwent diagnostic (n = 1418) or split-night (n = 504) PSGs and completed a symptom and medical history questionnaire. More than 30% of patients who did not endorse any OSA symptoms, but did endorse insomnia or restless legs symptoms, were found to have OSA based on apnea-hypopnea index (AHI) >5 during overnight laboratory testing. Regression models of the full cohort showed that the risk of OSA was related, as expected, to older age, male sex, elevated body mass index, and presence of OSA symptoms. The presence of insomnia symptoms did not alter the risk of OSA. The presence of restless legs symptoms showed a small odds ratio for lowered OSA risk. Objective evidence of OSA occurs similarly in those with insomnia or restless legs symptoms, even among those without self-reported OSA symptoms. Providers should be aware of the potential for occult OSA in populations with insomnia and restless legs, which may complicate their management in addition to presenting an independent medical risk itself. © 2015 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

Analysis of Post-LOCA Core Inlet Blockage to Evaluate In-vessel Downstream Effect in APR1400

International Nuclear Information System (INIS)

Bang, Young Seok

2015-01-01

The method was developed to have a conservatism to cover the uncertainty of analysis and the acceptance is judged by the representative bounding estimation. However, the important safety parameters such as the available driving head need to be confirmed by the plant specific calculation. Also an interaction between the debris induced head loss and the core flow rate needs to be explained because the head loss induced by debris in actual condition may reduce the core inflow rate faster. To confirm the safety parameters, in this study, thermal-hydraulic response considering the core inlet blockage (CIB) by debris during LTCC process following a double-ended guillotine break of cold leg (CLB), one of hot leg (HLB) and one of intermediate leg (ILB) of the APR1400 were calculated, respectively. MARS-KS 1.3 code has been used. The CIB has been modeled by the closure of valves to the core in exponential manner with time to observe the behavior near the complete blockage. To understand the effect of core inlet blockage (CIB) during a long term core cooling (LTCC) phase following a loss-of-coolant accident (LOCA) in the light of in-vessel downstream effect (IDE) of Generic Safety Issue (GSI) 191, double-ended guillotine break of hot leg (HLB), one of cold leg (CLB) and one of intermediate leg (ILB) were calculated, respectively. And the important safety parameters such as the available driving head and the head loss due to debris were calculated using MARS-KS code and discussed in comparison with the WCAP method. As a result, a little delayed heatup behavior of the fuel cladding was found for all the cases, which due to the redistribution of flow within the core after blockage

Analysis of Post-LOCA Core Inlet Blockage to Evaluate In-vessel Downstream Effect in APR1400

Energy Technology Data Exchange (ETDEWEB)

Bang, Young Seok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-10-15

The method was developed to have a conservatism to cover the uncertainty of analysis and the acceptance is judged by the representative bounding estimation. However, the important safety parameters such as the available driving head need to be confirmed by the plant specific calculation. Also an interaction between the debris induced head loss and the core flow rate needs to be explained because the head loss induced by debris in actual condition may reduce the core inflow rate faster. To confirm the safety parameters, in this study, thermal-hydraulic response considering the core inlet blockage (CIB) by debris during LTCC process following a double-ended guillotine break of cold leg (CLB), one of hot leg (HLB) and one of intermediate leg (ILB) of the APR1400 were calculated, respectively. MARS-KS 1.3 code has been used. The CIB has been modeled by the closure of valves to the core in exponential manner with time to observe the behavior near the complete blockage. To understand the effect of core inlet blockage (CIB) during a long term core cooling (LTCC) phase following a loss-of-coolant accident (LOCA) in the light of in-vessel downstream effect (IDE) of Generic Safety Issue (GSI) 191, double-ended guillotine break of hot leg (HLB), one of cold leg (CLB) and one of intermediate leg (ILB) were calculated, respectively. And the important safety parameters such as the available driving head and the head loss due to debris were calculated using MARS-KS code and discussed in comparison with the WCAP method. As a result, a little delayed heatup behavior of the fuel cladding was found for all the cases, which due to the redistribution of flow within the core after blockage.

Venous leg ulcers

Science.gov (United States)

2008-01-01

Introduction Leg ulcers usually occur secondary to venous reflux or obstruction, but 20% of people with leg ulcers have arterial disease, with or without venous disorders. Between 1.5 and 3.0/1000 people have active leg ulcers. Prevalence increases with age to about 20/1000 in people aged over 80 years. Methods and outcomes We conducted a systematic review and aimed to answer the following clinical questions: What are the effects of standard treatments, adjuvant treatments, and organisational interventions for venous leg ulcers? What are the effects of interventions to prevent recurrence of venous leg ulcers? We searched: Medline, Embase, The Cochrane Library, and other important databases up to September 2007 (BMJ Clinical Evidence reviews are updated periodically, please check our website for the most up-to-date version of this review). We included harms alerts from relevant organisations such as the US Food and Drug Administration (FDA) and the UK Medicines and Healthcare products Regulatory Agency (MHRA). Results We found 80 systematic reviews, RCTs, or observational studies that met our inclusion criteria. We performed a GRADE evaluation of the quality of evidence for interventions. Conclusions In this systematic review we present information relating to the effectiveness and safety of the following interventions: compression bandages and stockings, cultured allogenic (single or bilayer) skin replacement, debriding agents, dressings (cellulose, collagen, film, foam, hyaluronic acid-derived, semi-occlusive alginate), hydrocolloid (occlusive) dressings in the presence of compression, intermittent pneumatic compression, intravenous prostaglandin E1, larval therapy, laser treatment (low-level), leg ulcer clinics, multilayer elastic system, multilayer elastomeric (or non-elastomeric) high-compression regimens or bandages, oral treatments (aspirin, flavonoids, pentoxifylline, rutosides, stanozolol, sulodexide, thromboxane alpha2 antagonists, zinc), peri

Cold weather effects on Dresden Unit 1

Energy Technology Data Exchange (ETDEWEB)

Anagnostopoulos, H. [Commonwealth Edison Co., Morris, IL (United States)

1995-03-01

Dresden Unit 1 is in the final stages of a decommissioning effort directed at preparing the unit to enter a SAFSTOR status. Following an extended sub-zero cold wave, about 55,000 gallons of water were discovered in the lowest elevation of the spherical reactor enclosure. Cold weather had caused the freezing and breaking of several service water lines that had not been completely isolated. Two days later, at a regularly scheduled decommissioning meeting, the event was communicated to the decommissioning team, who quickly recognized the potential for freezing of a 42 inches diameter Fuel Transfer Tube that connects the sphere to the Spent Fuel Pool. The team directed that the pool gates between the adjacent Spent Fuel Pool and the Fuel Transfer Pool be installed, and a portable source of heat was installed on the Fuel Transfer Tube. It was later determined that, with the fuel pool gates removed, and with a worst case freeze break at the 502 elevation on the Fuel Transfer Tube (in the Sphere), the fuel in the Spent Fuel Pool could be uncovered to a level 3 below the top of active fuel.

Universal treatment of plumes and stresses for pressurized thermal shock evaluations

International Nuclear Information System (INIS)

Theofanous, T.G.; Angelini, S.; Yan, H.

1991-01-01

Thermally-induced stresses in a reactor pressure vessel wall, as a result of high-pressure safety injection, are an essential component of integrated risk analyses of pressurized thermal shock transients. Limiting cooldowns arise when this injection occurs under stagnated loop conditions which, in turn, correspond to a rather narrow range (in size) of small-break loss-of-coolant accidents. Moreover, at these conditions, the flow is thermally stratified, and in addition to the global cooldown, one must be concerned about the additional cooling potential due to the downcomer plumes formed by the cold streams pouring out of the cold legs. In the Nuclear Regulatory Commission's Integrated Pressurized Thermal Shock (IPTS) study, this stratification was calculated with the codes REMIX/NEWMIX. A comprehensive comparison with all available experimental data has currently been compiled. The stress analysis using this input was carried out at Oak Ridge National Laboratory using a one-dimensional approximation with the intent to conservatively bound the magnitude of thermal stresses

Sensitivity of break-flow-partition on the containment pressure and temperature

International Nuclear Information System (INIS)

Kwon, Young Min; Song, Jin Ho; Lee, Sang Yong

1994-01-01

For the case of RCS blowdown into the vapor region of a containment at low pressure, the blowdown mixture will start to boil at the containment pressure and liquid will separate from the flow near the break location. The flashed steam is added to the containment atmosphere and liquid is falled to the sump. Analytically, the break flow can be divided into steam and liquid in a number of ways. Discussed in this study is three partition models and Instantaneous Mixing(IM) Model employed in different containment analysis computer codes. IM model is employed in the CONTRANS code developed by ABB-CE for containment thermodynamic analysis. The various partition models were applied to the double ended discharge leg slot break (DEDLS) LOCA which is containment design base accident (CDBA) for Ulchin 3 and 4 PSAR. It was shown that IM model is the most conservative for containment design pressure analysis. Results of the CONTRANS analyses are compared with those of UCN PSAR for which CONTEMPT-LT code was used

Verification of the code ATHLET by post-test analysis of two experiments performed at the CCTF integral test facility

International Nuclear Information System (INIS)

Krepper, E.; Schaefer, F.

2001-03-01

In the framework of the external validation of the thermohydraulic code ATHLET Mod 1.2 Cycle C, which has been developed by the GRS, post test analyses of two experiments were done, which were performed at the japanese test facility CCTF. The test facility CCTF is a 1:25 volume-scaled model of a 1000 MW pressurized water reactor. The tests simulate a double end break in the cold leg of the PWR with ECC injection into the cold leg and with combined ECC injection into the hot and cold legs. The evaluation of the calculated results shows, that the main phenomena can be calculated in a good agreement with the experiment. Especially the behaviour of the quench front and the core cooling are calculated very well. Applying a two-channel representation of the reactor model the radial behaviour of the quench front could be reproduced. Deviations between calculations and experiment can be observed simulating the emergency injection in the beginning of the transient. Very high condensation rates were calculated and the pressure decrease in this phase of the transient is overestimated. Besides that, the pressurization due to evaporation in the refill phase is underestimated by ATHLET. (orig.) [de

H:q ratios and bilateral leg strength in college field and court sports players.

Science.gov (United States)

Cheung, Roy T H; Smith, Andrew W; Wong, Del P

2012-06-01

One of the key components in sports injury prevention is the identification of imbalances in leg muscle strength. However, different leg muscle characteristics may occur in large playing area (field) sports and small playing area (court) sports, which should be considered in regular injury prevention assessment. This study examined the isokinetic hamstrings-to-quadriceps (H:Q) ratio and bilateral leg strength balance in 40 male college (age: 23.4 ± 2.5 yrs) team sport players (field sport = 23, soccer players; court sport = 17, volleyball and basketball players). Five repetitions of maximal knee concentric flexion and concentric extension were performed on an isokinetic dynamometer at two speeds (slow: 60°·s(-1) and fast: 300°·s(-1)) with 3 minutes rest between tests. Both legs were measured in counterbalanced order with the dominant leg being determined as the leg used to kick a ball. The highest concentric peak torque values (Nm) of the hamstrings and quadriceps of each leg were analyzed after body mass normalization (Nm·kg(-1)). Court sport players showed significantly weaker dominant leg hamstrings muscles at both contraction speeds (P Sport-specific leg muscle strength was evident in college players from field and court sports. These results suggest the need for different muscle strength training and rehabilitation protocols for college players according to the musculature requirements in their respective sports.

Analysis of two small break loss-of-coolant experiments in the BETHSY facility using RELAP5/MOD3

International Nuclear Information System (INIS)

Roth, P.A.; Schultz, R.R.; Choi, C.J.

1992-07-01

Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d'Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d'Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system depressurization, break flow rate, core heat-up, and effect of operator actions were analyzed. Based on the test 9.lb/ISP-27 and 6.2 TC data, an assessment study of the RELAP5/MOD3 version 7 code was performed which included a study of the above phenomena along with countercurrent flow limitation and vapor pull-through. The code provided a reasonable simulation of the various phenomena which occurred during the tests

Water striders adjust leg movement speed to optimize takeoff velocity for their morphology

Science.gov (United States)

Yang, Eunjin; Son, Jae Hak; Lee, Sang-Im; Jablonski, Piotr G.; Kim, Ho-Young

2016-12-01

Water striders are water-walking insects that can jump upwards from the water surface. Quick jumps allow striders to avoid sudden dangers such as predators' attacks, and therefore their jumping is expected to be shaped by natural selection for optimal performance. Related species with different morphological constraints could require different jumping mechanics to successfully avoid predation. Here we show that jumping striders tune their leg rotation speed to reach the maximum jumping speed that water surface allows. We find that the leg stroke speeds of water strider species with different leg morphologies correspond to mathematically calculated morphology-specific optima that maximize vertical takeoff velocity by fully exploiting the capillary force of water. These results improve the understanding of correlated evolution between morphology and leg movements in small jumping insects, and provide a theoretical basis to develop biomimetic technology in semi-aquatic environments.

Generating Small Numbers by Tunneling in Multi-Throat Compactifications

Energy Technology Data Exchange (ETDEWEB)

Silverstein, Eva M

2001-07-25

A generic F-theory compactification containing many D3 branes develops multiple brane throats. The interaction of observers residing inside different throats involves tunneling suppression and, as a result, is very weak. This suggests a new mechanism for generating small numbers in Nature. One application is to the hierarchy problem: large supersymmetry breaking near the unification scale inside a shallow throat causes TeV-scale SUSY-breaking inside the standard-model throat. Another application, inspired by nuclear-decay, is in designing naturally long-lived particles: a cold dark matter particle residing near the standard model brane decays to an approximate CFT-state of a longer throat within a Hubble time. This suggests that most of the mass of the universe today could consist of CFT-matter and may soften structure formation at sub-galactic scales. The tunneling calculation demonstrates that the coupling between two throats is dominated by higher dimensional modes and consequently is much larger than a naive application of holography might suggest.

Lyden-af-Leg

DEFF Research Database (Denmark)

Toft, Herdis

Præsentation af seniorforsker-projekt Lyden-af-Leg i et traderingsperspektiv og med indledende fokus på YouTube som traderings-platform.......Præsentation af seniorforsker-projekt Lyden-af-Leg i et traderingsperspektiv og med indledende fokus på YouTube som traderings-platform....

Break the Kill Chain, Not the Budget: How to Avoid U.S. Strategic Retrenchment

Science.gov (United States)

2016-06-10

mass quantities. The United States maintained a Cold War offset strategy focused on technology and kill chain annihilation with an understanding of...Annihilating the kill chain with advanced instruments of war is more the norm than the exception in the post- Cold War era U.S. military. The U.S. Air...Masters Thesis 3. DATES COVERED (From - To) 27-07-2015 to 10-06-2016 4. TITLE AND SUBTITLE 5a. CONTRACT NUMBER BREAK THE KILL CHAIN

Yukawa unification in moduli-dominant SUSY breaking

International Nuclear Information System (INIS)

Khalil, S.; Tatsuo Kobayashi

1997-07-01

We study Yukawa in string models with moduli-dominant SUSY breaking. This type of SUSY breaking in general leads to non-universal soft masses, i.e. soft scalar masses and gaugino masses. Such non-universality is important for phenomenological aspects of Yukawa unification, i.e., successful electroweak breaking, SUSY corrections to the bottom mass and the branching ratio of b → sγ. We show three regions in the whole parameter space which lead to successful electroweak breaking and allow small SUSY corrections to the bottom mass. For these three regions we investigated the b → sγ decay and mass spectra. (author). 26 refs, 6 figs

Single-leg squats can predict leg alignment in dancers performing ballet movements in “turnout”

Directory of Open Access Journals (Sweden)

Hopper LS

2016-11-01

Full Text Available Luke S Hopper,1 Nahoko Sato,2 Andries L Weidemann1 1Western Australian Academy of Performing Arts, Edith Cowan University, Mt Lawley, WA, Australia; 2Department of Physical Therapy, Nagoya Gakuin University, Seto, Japan Abstract: The physical assessments used in dance injury surveillance programs are often adapted from the sports and exercise domain. Bespoke physical assessments may be required for dance, particularly when ballet movements involve “turning out” or external rotation of the legs beyond that typically used in sports. This study evaluated the ability of the traditional single-leg squat to predict the leg alignment of dancers performing ballet movements with turnout. Three-dimensional kinematic data of dancers performing the single-leg squat and five ballet movements were recorded and analyzed. Reduction of the three-dimensional data into a one-dimensional variable incorporating the ankle, knee, and hip joint center positions provided the strongest predictive model between the single-leg squat and the ballet movements. The single-leg squat can predict leg alignment in dancers performing ballet movements, even in “turned out” postures. Clinicians should pay careful attention to observational positioning and rating criteria when assessing dancers performing the single-leg squat. Keywords: injury, motion capture, clinical assessment

Initial heating in cold cars

NARCIS (Netherlands)

Daanen, H.A.M.; Teunissen, L.P.J.; Hoogh, I.M. de

2012-01-01

During the initial minutes after entering a cold car, people feel uncomfortably cold. Six different warming systems were investigated in a small car in order to find out how to improve the feeling of comfort using 16 volunteers. The methods were: no additional warming next to a standard heating

TLTA/6431, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: The Two-Loop-Test-Apparatus (TLTA) is a 1:624 volume scaled BWR/6 simulator. It was the predecessor of the better-scaled FIST facility. The facility is capable of full BWR system pressure and has a simulated core with a full size 8 x 8, full power single bundle of indirect electrically heated rods. All major BWR systems are simulated including lower plenum, guide tube, core region (bundle and bypass), upper plenum, steam separator, steam dome, annular downcomer, recirculation loops and ECC injection systems. The fundamental scaling consideration was to achieve real-time response. A number of the scaling compromises present in TLTA were corrected in the FIST configuration. These compromises include a number of regional volumes and component elevations. 2 - Description of test: 64.45 sqcm small break LOCA with activation of the full emergency core cooling system, but without activation of the automatic decompression system

PIPER-ONE: an experimental apparatus to evaluate thermal-hydraulic transients in BWRs after small breaks

International Nuclear Information System (INIS)

Mazzini, M.; D'Auria, F.; Vigni, P.

1981-01-01

This paper deals with the state of art of the research performed at the Instituto di Impianti Nucleari of Pisa University, aiming at construction of PIPER-ONE experimental facility. PIPER-ONE program is devoted to acquire direct experience on some basic phenomena, arising in BWR plants subsequently to small breaks, and on the use of the main thermal-hydraulic codes. The research has been planned taking into consideration recent trends of the studies all over the world of small LOCA thermal-hydraulics and particular needs of nuclear safety in Italy. Cost limitations and availability of some components, already installed at the Institute Laboratory, have influenced the design of the loop. The development steps of PIPER-ONE project are presented. Particularly, the overall flowsheet of the apparatus is reported. Some results of preliminary calculation, executed by RELAP4-Mod 6 code concerning both the experimental loop and the reference BWR are shown, too. A comparison with similar facilities in the world closes the paper

Leg Injuries and Disorders

Science.gov (United States)

... are important for motion and standing. Playing sports, running, falling, or having an accident can damage your legs. Common leg injuries include sprains and strains, joint dislocations, and fractures. ...

Spontaneous De-Icing Phenomena on Extremely Cold Surfaces

Science.gov (United States)

Song, Dong; Choi, Chang-Hwan

2017-11-01

Freezing of droplets on cold surfaces is universal phenomenon, while the mechanisms are still inadequately understood. Here we report spontaneous de-icing phenomena of an impacting droplet which occur on extreme cold surfaces. When a droplet impacts on cold surfaces lower than -80°, it takes more than two times longer for the droplet to freeze than the ones at -50°. Moreover, the frozen droplet below -80° breaks up into several large parts spontaneously in the end. When a droplet impacts on the extreme cold surfaces, evaporation and condensation occur immediately as the droplet approaches the substrate. A thick layer of frost forms between the droplet and substrate, decreasing the contact area of the droplet with substrate. It leads to impede the heat transfer and hence extends the freezing time significantly. On the extremely cold substrate, the droplet freezes from the center to the edge area, in contrast to a typical case freezing from the bottom to the top. This novel from-center-to-edge freezing process changes the internal tension of the frozen droplet and results in the instantaneous breakup and release eventually, which can be taken advantage of for effective deicing mechanisms.

Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

Energy Technology Data Exchange (ETDEWEB)

Boyack, B.E.; Steiner, J.L.; Harmony, S.C. [Los Alamos National Lab., Albuquerque, NM (United States)] [and others

1995-09-01

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

Long-term recovery of pressurized water reactors following a large break loss-of-coolant accident

International Nuclear Information System (INIS)

Fletcher, C.D.; Callow, R.A.

1989-01-01

The USNRC recently identified a possible safety concern for PWR's. Following the reflood phase of a large break loss-of-coolant accident, long-term cooling of the reactor core may not be ensured. Specifically, the concern is that, for a pump discharge cold leg break, the loop seals in the reactor coolant pump suction piping will refill with liquid and the post-reflood steam production may depress the liquid levels in the downflow sides of the loop seals. A loop seal depression would cause a corresponding depression of the core liquid levels and possibly a fuel rod heatup in the upper core region. This paper is intended as an introduction of the safety issue that: 1) describes the important aspects of the problem, 2) provides an initial analysis of the consequences, and 3) discusses ongoing work in this area. Because the elevation of the loop seals is near the mid-core elevation in plants of WE design, the concern is greatest for those plants. There is less concern for most plants of CE design, and likely no concern for plants of BW design. This issue was addressed by employing both steady-state and transient systems analysis approaches. Two approaches were used because of uncertainties regarding actual reactor coolant system behavior during the post-reflood period. The steady-state approach involved the development and application of a simple computer program to investigate reactor coolant system behavior assuming quiescent post-reflood conditions. The transient systems approach involved investigating this behavior using the RELAP5/MOD2 computer code and a comprehensive RELAP5 model of a WE PWR. The steady-state analysis indicated only a moderate fuel rod heatup is possible. The transient systems analysis indicated boiling and condensation-induced flow oscillations are sufficient to prevent fuel rod heatup. Analysis uncertainties are discussed. (orig./HP)

Exertional rhabdomyolysis in a collegiate american football player after preventive cold-water immersion: a case report.

Science.gov (United States)

Kahanov, Leamor; Eberman, Lindsey E; Wasik, Mitchell; Alvey, Thurman

2012-01-01

To describe a case of exertional rhabdomyolysis in a collegiate American football player after preventive coldwater immersion. A healthy man (19 years old) participated in full-contact football practice followed by conditioning (2.5 hours). After practice, he entered a coach-mandated postpractice cold-water immersion and had no signs of heat illness before developing leg cramps, for which he presented to the athletic training staff. After 10 minutes of repeated stretching, massage, and replacement of electrolyte-filled fluids, he was transported to the emergency room. Laboratory tests indicated a creatine kinase (CK) level of 2545 IU/L (normal range, 45-260 IU/L), CK-myoglobin fraction of 8.5 ng/mL (normal football practice as tolerated. Two months after the incident, his CK level remained high (1900 IU/L). The athlete demonstrated no signs of heat illness upon entering the cold-water immersion but experienced severe leg cramping after immersion, resulting in a diagnosis of exertional rhabdomyolysis. Previously described cases have not linked cold-water immersion with the pathogenesis of rhabdomyolysis. In this football player, CK levels appeared to be a poor indicator of rhabdomyolysis. Our patient demonstrated no other signs of the illness weeks after the incident, yet his elevated CK levels persisted. Cold-water immersion immediately after exercise should be monitored by the athletic training staff and may not be appropriate to prevent muscle damage, given the lack of supporting evidence.

Generic evaluation of feedwater transients and small break loss-of-coolant accidents in combustion engineering designed operating plants

International Nuclear Information System (INIS)

1980-01-01

The purpose of this report is to summarize the results of a generic evaluation of feedwater transients, small break loss-of-coolant accidents (LOCAs), and other TMI-2-related events in the Combustion Engineering (CE)-designed operating plants and to establish or confirm the bases for their continued operation. The results of this evaluation are presented in this report in the form of a set of findings and recommendations in each of the principal review areas

Lifshitz-sector mediated SUSY breaking

International Nuclear Information System (INIS)

Pospelov, Maxim; Tamarit, Carlos

2014-01-01

We propose a novel mechanism of SUSY breaking by coupling a Lorentz-invariant supersymmetric matter sector to non-supersymmetric gravitational interactions with Lifshitz scaling. The improved UV properties of Lifshitz propagators moderate the otherwise uncontrollable ultraviolet divergences induced by gravitational loops. This ensures that both the amount of induced Lorentz violation and SUSY breaking in the matter sector are controlled by Λ HL 2 /M P 2 , the ratio of the Hořava-Lifshitz cross-over scale Λ HL to the Planck scale M P . This ratio can be kept very small, providing a novel way of explicitly breaking supersymmetry without reintroducing fine-tuning. We illustrate our idea by considering a model of scalar gravity with Hořava-Lifshitz scaling coupled to a supersymmetric Wess-Zumino matter sector, in which we compute the two-loop SUSY breaking corrections to the masses of the light scalars due to the gravitational interactions and the heavy fields

HISTOMORPHOMETRIC ANALYSIS OF THE KNEE ARTICULAR CARTILAGE AND SYNOVIUM FOR METADIAPHYSEAL LEG LENGTHENING (EXPERIMENTAL-AND-MORPHOLOGICAL STUDY)

OpenAIRE

T. A. Stupina; N. A. Schoudlo; N. V. Petrovskaya; M. A. Stepanov

2013-01-01

The knee articular cartilage and synovial membrane have been studied for metadiaphyseal leg lengthening using the methods of light miscroscopy, computer morpho- and stereometry. The manner of bone integrity breaking, the rate and rhythm of distraction conformed to the lengthening technique most often used in the clinic. The results of the histomorphometric analysis have demonstrated that when osteotomy at the level of metadiaphysis and manual distraction by 1 mm a day for 4 times is performed...

Gravitino dark matter in R-parity breaking vacua

International Nuclear Information System (INIS)

Buchmueller, W.; Covi, L.; Ibarra, A.; Hamaguchi, K.; Yanagida, T.T.

2007-02-01

We show that in the case of small R-parity and lepton number breaking couplings, primordial nucleosynthesis, thermal leptogenesis and gravitino dark matter are naturally consistent for gravitino masses m 3/2 >or similar 5 GeV. We present a model where R-parity breaking is tied to B-L breaking, which predicts the needed small couplings. The metastable next-to-lightest superparticle has a decay length that is typically larger than a few centimeters, with characteristic signatures at the LHC. The photon flux produced by relic gravitino decays may be part of the apparent excess in the extragalactic diffuse gamma-ray flux obtained from the EGRET data for a gravitino mass m 3/2 ∝10 GeV. In this case, a clear signal can be expected from GLAST in the near future. (orig.)

Book Review: A History of the Czechoslovak Ocean Shipping Company, 1948–1989: How a Small, Landlocked Country Ran Maritime Business During the Cold War

DEFF Research Database (Denmark)

Taudal Poulsen, René

2016-01-01

Review of: Lenka Krátká: A History of the Czechoslovak Ocean Shipping Company, 1948–1989: How a Small, Landlocked Country Ran Maritime Business During the Cold War. Stuttgart: Ibidem Verlag, 2015. x + 271 pp., tables, notes, bibliography. ISBN: 978-3-8382-0666-0, £23.90 (pbk).......Review of: Lenka Krátká: A History of the Czechoslovak Ocean Shipping Company, 1948–1989: How a Small, Landlocked Country Ran Maritime Business During the Cold War. Stuttgart: Ibidem Verlag, 2015. x + 271 pp., tables, notes, bibliography. ISBN: 978-3-8382-0666-0, £23.90 (pbk)....

Deep sequencing of Brachypodium small RNAs at the global genome level identifies microRNAs involved in cold stress response

Directory of Open Access Journals (Sweden)

Chong Kang

2009-09-01

Full Text Available Abstract Background MicroRNAs (miRNAs are endogenous small RNAs having large-scale regulatory effects on plant development and stress responses. Extensive studies of miRNAs have only been performed in a few model plants. Although miRNAs are proved to be involved in plant cold stress responses, little is known for winter-habit monocots. Brachypodium distachyon, with close evolutionary relationship to cool-season cereals, has recently emerged as a novel model plant. There are few reports of Brachypodium miRNAs. Results High-throughput sequencing and whole-genome-wide data mining led to the identification of 27 conserved miRNAs, as well as 129 predicted miRNAs in Brachypodium. For multiple-member conserved miRNA families, their sizes in Brachypodium were much smaller than those in rice and Populus. The genome organization of miR395 family in Brachypodium was quite different from that in rice. The expression of 3 conserved miRNAs and 25 predicted miRNAs showed significant changes in response to cold stress. Among these miRNAs, some were cold-induced and some were cold-suppressed, but all the conserved miRNAs were up-regulated under cold stress condition. Conclusion Our results suggest that Brachypodium miRNAs are composed of a set of conserved miRNAs and a large proportion of non-conserved miRNAs with low expression levels. Both kinds of miRNAs were involved in cold stress response, but all the conserved miRNAs were up-regulated, implying an important role for cold-induced miRNAs. The different size and genome organization of miRNA families in Brachypodium and rice suggest that the frequency of duplication events or the selection pressure on duplicated miRNAs are different between these two closely related plant species.

Coleman-Weinberg symmetry breaking in an anisotropic spacetime

International Nuclear Information System (INIS)

Futamase, T.

1984-01-01

The Coleman-Weinberg mechanism of symmetry breaking in a Bianchi type-I universe is investigated. The one-loop effective potential for a phi 4 theory and for scalar electrodynamics is calculated by the zeta-function method. The result indicates that the symmetry of the theory will be restored in the highly anisotropic, cold, early universe, irrespective of the coupling between the scalar field and the spacetime curvature scalar. This mechanism of the phase transition explains the isotropy of our universe

Are the hamstrings from the drive leg or landing leg more active in baseball pitchers? An electromyographic study.

Science.gov (United States)

Erickson, Brandon J; Zaferiou, Antonia; Chalmers, Peter N; Ruby, Deana; Malloy, Phillip; Luchetti, Timothy J; Verma, Nikhil N; Romeo, Anthony A

2017-11-01

Ulnar collateral ligament reconstruction (UCLR) has become a common procedure among baseball players of all levels. There are several graft choices in performing UCLR, one of which is a hamstring (gracilis or semitendinosus) autograft. It is unclear whether the hamstring muscle from a pitcher's drive leg (ipsilateral side of the UCLR) or landing leg (contralateral side of the UCLR) is more active during the pitching motion. We hypothesized that the landing leg semitendinosus will be more electromyographically active than the drive leg. Healthy, elite male pitchers aged 16-21 years were recruited. Sixteen pitchers (average age, 17.6 ± 1.6 years; 67% threw right handed) underwent electromyographic analysis. Pitchers threw 5 fastballs at 100% effort from the wind-up with electromyographic analysis of every pitch. Activation of the semitendinosus and biceps femoris in both legs was compared within pitchers and between pitchers. Hamstring activity was higher in the drive leg than in the landing leg during each phase and in sum, although the difference was significant only during the double support phase (P = .021). On within-pitcher analysis, 10 of 16 pitchers had significantly more sum hamstring activity in the drive leg than in the landing leg, while only 4 of 16 had more activity in the landing leg (P = .043). During the baseball pitch, muscle activity of the semitendinosus was higher in the drive leg than in the landing leg in most pitchers. Surgeons performing UCLR using hamstring autograft should consider harvesting the graft from the pitcher's landing leg to minimize disruption to the athlete's pitching motion. Copyright © 2017 Journal of Shoulder and Elbow Surgery Board of Trustees. Published by Elsevier Inc. All rights reserved.

Preliminary analyses of AP600 using RELAP5

International Nuclear Information System (INIS)

Modro, S.M.; Beelman, R.J.; Fisher, J.E.

1991-01-01

This paper presents results of preliminary analyses of the proposed Westinghouse Electric Corporation AP600 design. AP600 is a two loop, 600 MW (e) pressurized water reactor (PWR) arranged in a two hot leg, four cold leg nuclear steam supply system (NSSS) configuration. In contrast to the present generation of PWRs it is equipped with passive emergency core coolant (ECC) systems. Also, the containment and the safety systems of the AP600 interact with the reactor coolant system and each other in a more integral fashion than present day PWRs. The containment in this design is the ultimate heat sink for removal of decay heat to the environment. Idaho National Engineering Laboratory (INEL) has studied applicability of the RELAP5 code to AP600 safety analysis and has developed a model of the AP600 for the Nuclear Regulatory Commission. The model incorporates integral modeling of the containment, NSSS and passive safety systems. Best available preliminary design data were used. Nodalization sensitivity studies were conducted to gain experience in modeling of systems and conditions which are beyond the applicability of previously established RELAP5 modeling guidelines or experience. Exploratory analyses were then undertaken to investigate AP600 system response during postulated accident conditions. Four small break LOCA calculations and two large break LOCA calculations were conducted

Radiosteriometric analysis of movement in the sacroiliac joint during a single-leg stance in patients with long-lasting pelvic girdle pain.

Science.gov (United States)

Kibsgård, Thomas J; Røise, Olav; Sturesson, Bengt; Röhrl, Stephan M; Stuge, Britt

2014-04-01

Chamberlain's projections (anterior-posterior X-ray of the pubic symphysis) have been used to diagnose sacroiliac joint mobility during the single-leg stance test. This study examined the movement in the sacroiliac joint during the single-leg stance test with precise radiostereometric analysis. Under general anesthesia, tantalum markers were inserted into the dorsal sacrum and the ilium of 11 patients with long-lasting and severe pelvic girdle pain. After two to three weeks, a radiostereometric analysis was conducted while the subjects performed a single-leg stance. Small movements were detected in the sacroiliac joint during the single-leg stance. In both the standing- and hanging-leg sacroiliac join, a total of 0.5 degree rotation was observed; however, no translations were detected. There were no differences in total movement between the standing- and hanging-leg sacroiliac joint. The movement in the sacroiliac joint during the single-leg stance is small and almost undetectable by the precise radiostereometric analysis. A complex movement pattern was seen during the test, with a combination of movements in the two joints. The interpretation of the results of this study is that, the Chamberlain examination likely is inadequate in the examination of sacroiliac joint movement in patients with pelvic girdle pain. Copyright © 2014 Elsevier Ltd. All rights reserved.

Climate and floods still govern California levee breaks

Science.gov (United States)

Florsheim, J.L.; Dettinger, M.D.

2007-01-01

Even in heavily engineered river systems, climate still governs flood variability and thus still drives many levee breaks and geomorphic changes. We assemble a 155-year record of levee breaks for a major California river system to find that breaks occurred in 25% of years during the 20th Century. A relation between levee breaks and river discharge is present that sets a discharge threshold above which most levee breaks occurred. That threshold corresponds to small floods with recurrence intervals of ???2-3 years. Statistical analysis illustrates that levee breaks and peak discharges cycle (broadly) on a 12-15 year time scale, in time with warm-wet storm patterns in California, but more slowly or more quickly than ENSO and PDO climate phenomena, respectively. Notably, these variations and thresholds persist through the 20th Century, suggesting that historical flood-control effects have not reduced the occurrence or frequency of levee breaks. Copyright 2007 by the American Geophysical Union.

Preliminary test conditions for KNGR SBLOCA DVI ECCS performance test

International Nuclear Information System (INIS)

Bae, Kyoo Whan; Song, Jin Ho; Chung, Young Jong; Sim, Suk Ku; Park, Jong Kyun

1999-03-01

The Korean Next Generation Reactor (KNGR) adopts 4-train Direct Vessel Injection (DVI) configuration and injects the safety injection water directly into the downcomer through the 8.5'' DVI nozzle. Thus, the thermal hydraulic phenomena such as ECCS mixing and bypass are expected to be different from those observed in the cold leg injection. In order to investigate the realistic injection phenomena and modify the analysis code developed in the basis of cold leg injection, thermal hydraulic test with the performance evaluation is required. Preliminarily, the sequence of events and major thermal hydraulic phenomena during the small break LOCA for KNGR are identified from the analysis results calculated by the CEFLASH-4AS/REM. It is shown from the analysis results that the major transient behaviors including the core mixture level are largely affected by the downcomer modeling. Therefore, to investigate the proper thermal hydraulic phenomena occurring in the downcomer with limited budget and time, the separate effects test focusing on this region is considered to be effective and the conceptual test facility based on this recommended. For this test facility the test initial and boundary conditions are developed using the CEFLASH-4AS/REM analysis results that will be used as input for the preliminary test requirements. The final test requirements will be developed through the further discussions with the test performance group. (Author). 10 refs., 18 tabs., 4 figs

Gravitino dark matter in R-parity breaking vacua

Energy Technology Data Exchange (ETDEWEB)

Buchmueller, W.; Covi, L.; Ibarra, A. [Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany); Hamaguchi, K.; Yanagida, T.T. [Tokyo Univ. (Japan). Dept. of Physics

2007-02-15

We show that in the case of small R-parity and lepton number breaking couplings, primordial nucleosynthesis, thermal leptogenesis and gravitino dark matter are naturally consistent for gravitino masses m{sub 3/2} >or similar 5 GeV. We present a model where R-parity breaking is tied to B-L breaking, which predicts the needed small couplings. The metastable next-to-lightest superparticle has a decay length that is typically larger than a few centimeters, with characteristic signatures at the LHC. The photon flux produced by relic gravitino decays may be part of the apparent excess in the extragalactic diffuse gamma-ray flux obtained from the EGRET data for a gravitino mass m{sub 3/2}{proportional_to}10 GeV. In this case, a clear signal can be expected from GLAST in the near future. (orig.)

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

International Nuclear Information System (INIS)

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

1977-02-01

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

Iris small break loca phenomena identification and ranking table (PIRT)

International Nuclear Information System (INIS)

Larson, T.K.; Moody, F.J.; Wilson, G.E.; Brown, W.L.; Frepoli, C.; Hartz, J.; Woods, B.G.; Oriani, L.

2007-01-01

The international reactor innovative and secure (IRIS) is a modular pressurized water reactor with an integral configuration (all primary system components - reactor core, internals, pumps, steam generators, pressurizer, and control rod drive mechanisms - are inside the reactor vessel). The IRIS plant conceptual design was completed in 2001 and the preliminary design is currently underway. The pre-application licensing process with the United States Nuclear Regulatory Commission (USNRC) started in October 2002. The first line of defense in IRIS is to eliminate event initiators that could potentially lead to core damage. If it is not possible to eliminate certain accidents altogether, then the design inherently reduces their consequences and/or decreases their probability of occurring. One of the most obvious advantages of the IRIS Safety-by-Design TM approach is the elimination of large break loss-of-coolant accidents (LBLOCAs), since no large primary penetrations of the reactor vessel or large loop piping exist. While the IRIS Safety-by-Design TM approach is a logical step in the effort to produce advanced reactors, the desired advances in safety must still be demonstrated in the licensing arena. With the elimination of LBLOCA, an important next consideration is to show the IRIS design fulfills the promise of increased safety also for small break LOCAs (SBLOCAs). Accordingly, the SBLOCA phenomena identification and ranking table (PIRT) project was established. The primary objective of the IRIS SBLOCA PIRT project was to identify the relative importance of phenomena in the IRIS response to SBLOCAs. This relative importance, coupled with the current relative state of knowledge for the phenomena, provides a framework for the planning of the continued experimental and analytical efforts. To satisfy the SBLOCA PIRT project objectives, Westinghouse organized an expert panel whose members were carefully selected to insure that the PIRT results reflect internationally

Results of two-phase natural circulation in hot-leg U-bend simulation experiments

International Nuclear Information System (INIS)

Ishii, M.; Lee, S.Y.; Abou El-Seoud, S.

1987-01-01

In order to study the two-phase natural circulation and flow termination during a small break loss of coolant accident in LWR, simulation experiments have been performed using two different thermal-hydraulic loops. The main focus of the experiment was the two-phase flow behavior in the hot-leg U-bend typical of BandW LWR systems. The first group of experiments was carried out in the nitrogen gas-water adiabatic simulation loop and the second in the Freon 113 boiling and condensation loop. Both of the loops have been designed as a flow visualization facility and built according to the two-phase flow scaling criteria developed under this program. The nitrogen gas-water system has been used to isolate key hydrodynamic phenomena such as the phase distribution, relative velocity between phases, two-phase flow regimes and flow termination mechanisms, whereas the Freon loop has been used to study the effect of fluid properties, phase changes and coupling between hydrodynamic and heat transfer phenomena. Significantly different behaviors have been observed due to the non-equilibrium phase change phenomena such as the flashing and condensation in the Freon loop. The phenomena created much more unstable hydrodynamic conditions which lead to cyclic or oscillatory flow behaviors

Development of a compound energy system for cold region houses using small-scale natural gas cogeneration and a gas hydrate battery

International Nuclear Information System (INIS)

Obara, Shin'ya; Kikuchi, Yoshinobu; Ishikawa, Kyosuke; Kawai, Masahito; Yoshiaki, Kashiwaya

2015-01-01

In this study, an independent energy system for houses in cold regions was developed using a small-scale natural gas CGS (cogeneration), air-source heat pump, heat storage tank, and GHB (gas hydrate battery). Heat sources for the GHB were the ambient air and geothermal resources of the cold region. The heat cycle of CO 2 hydrate as a source of energy was also experimentally investigated. To increase the formation speed of CO 2 hydrates, a ferrous oxide–graphite system catalyst was used. The ambient air of cold regions was used as a heat source for the formation process (electric charge) of the GHB, and the heat supplied by a geothermal heat exchanger was used for the dissociation process (electric discharge). Using a geothermal heat source, fuel consumption was halved because of an increased capacity for hydrate formation in the GHB, a shortening of the charging and discharging cycle, and a decrease in the freeze rate of hydrate formation space. Furthermore, when the GHB was introduced into a cold region house, the application rate of renewable energy was 47–71% in winter. The spread of the GHB can greatly reduce fossil fuel consumption and the associated greenhouse gases released from houses in cold regions. - Highlights: • Compound energy system for cold region houses by a gas hydrate battery was proposed. • Heat sources of a gas hydrate battery are exhaust heat of the CGS and geothermal. • Drastic reduction of the fossil fuel consumption in a cold region is realized

Spectral distortion due to scattered cold neutrons in beryllium filter

International Nuclear Information System (INIS)

Sakamoto, Yukio; Inoue, Kazuhiko

1980-01-01

Polycrystalline beryllium filters are used to discriminate the cold neutrons from the thermal neutrons with energies above Bragg cut-off energy. The cold neutron scattering cross section is very small, but the remaining cross section is not zero. Then the neutrons scattered once from the filter in the cold neutron energy region have chance of impinging on the outlet of filter. Those neutrons are almost upscattered and develop into thermal neutrons; thus the discriminated cold neutrons include a small spectral distortion due to the thermal neutrons. In the present work we have evaluated the effect on the cold neutron spectrum due to the repeatedly scattered and transmitted neutrons by using a Monte Carlo calculation method. (author)

Exertional Rhabdomyolysis in a Collegiate American Football Player After Preventive Cold-Water Immersion: A Case Report

Science.gov (United States)

Kahanov, Leamor; Eberman, Lindsey E.; Wasik, Mitchell; Alvey, Thurman

2012-01-01

Objective: To describe a case of exertional rhabdomyolysis in a collegiate American football player after preventive cold-water immersion. Background: A healthy man (19 years old) participated in full-contact football practice followed by conditioning (2.5 hours). After practice, he entered a coach-mandated post-practice cold-water immersion and had no signs of heat illness before developing leg cramps, for which he presented to the athletic training staff. After 10 minutes of repeated stretching, massage, and replacement of electrolyte-filled fluids, he was transported to the emergency room. Laboratory tests indicated a creatine kinase (CK) level of 2545 IU/L (normal range, 45–260 IU/L), CK-myoglobin fraction of 8.5 ng/mL (normal rhabdomyolysis. Treatment: The patient was treated with rest and rehydration. One week after the incident, he began biking and swimming. Eighteen days later, the patient continued to demonstrate elevated CK levels (527 IU/L) but described no other symptoms and was allowed to return to football practice as tolerated. Two months after the incident, his CK level remained high (1900 IU/L). Uniqueness: The athlete demonstrated no signs of heat illness upon entering the cold-water immersion but experienced severe leg cramping after immersion, resulting in a diagnosis of exertional rhabdomyolysis. Previously described cases have not linked cold-water immersion with the pathogenesis of rhabdomyolysis. Conclusions: In this football player, CK levels appeared to be a poor indicator of rhabdomyolysis. Our patient demonstrated no other signs of the illness weeks after the incident, yet his elevated CK levels persisted. Cold-water immersion immediately after exercise should be monitored by the athletic training staff and may not be appropriate to prevent muscle damage, given the lack of supporting evidence. PMID:22488291

An application of RELAP5/MOD3 to the post-LOCA long term cooling performance evaluation

International Nuclear Information System (INIS)

Bang, Young Seok; Jung, Jae Won; Seul, Kwang Won; Kim, Hho Jung

1998-01-01

A realistic long-term calculation to be used in the post-LOCA long term cooling (LTC) analysis is described in this study, which was required to resolve the post-LOCA LTC issues including the concern on boric acid precipitation in the reactor core. The analysis scope is defined according to the LTC plan of UCN Units 3/4 and the plant calculation model are developed suitable to the LTC procedure. The LTC sequences following the cold leg small break LOCAs of 0.02 ft2 to 0.5 ft2 are calculated by RELAP5/ MOD3.2.2. Based on the calculation results, the establishment of shutdown cooling system entry condition and the behavior of boron transport are evaluated. The effect of model simplification is also investigated

Analysis of a large-break LOCA at lower operational modes

Energy Technology Data Exchange (ETDEWEB)

Jeong, H.Y.; Jun, H.Y.; Lee, K. [Korea Electric Power Corporation, Taejon (Korea)

2000-10-01

To improve Technical Specifications and Emergency Operating Guidelines (EOGs) applicable at lower operational modes it is required to perform the safety analysis reflecting the operational characteristics in those modes. Because the component availability and system configurations at lower modes are different from those of power mode, the plant safety at lower modes should be confirmed through independent analyses. In the present study, a large-break loss-of-coolant accident is analyzed to evaluate the containment pressure and temperature control function for the preparation of EOGs applicable at lower modes. To reach the required shutdown condition, the plant cool-down is controlled by the secondary steam flow and auxiliary feedwater. The mass and energy releases from primary system are obtained from RELAP5/MOD3.1 calculation and the containment pressure and temperature are evaluated with CONTEMPT-LT code. The reference plant is Korean Next Generation Reactor having 4,000 MW thermal power. Two cases of cold leg LOCA initiated at Mode 3 with and without SIT operation are calculated. At the given plant conditions, all safety injection pumps are still available. The calculation at the condition of maximum mass and energy release shows that the containment pressure and temperature can be controlled within acceptable criteria, which means the operations of 2 or 4 fan coolers are the possible success paths to achieve the containment P/T control safety function. The peak cladding temperature with minimum safety injection flow does not show remarkable excursion, which implies the lower mode LOCA at Mode 3 can be bounded by the results obtained at full power from the viewpoint of ECCS performance. (author)

Controlled fragmentation of multimaterial fibres and films via polymer cold-drawing.

Science.gov (United States)

Shabahang, Soroush; Tao, Guangming; Kaufman, Joshua J; Qiao, Yangyang; Wei, Lei; Bouchenot, Thomas; Gordon, Ali P; Fink, Yoel; Bai, Yuanli; Hoy, Robert S; Abouraddy, Ayman F

2016-06-23

Polymer cold-drawing is a process in which tensile stress reduces the diameter of a drawn fibre (or thickness of a drawn film) and orients the polymeric chains. Cold-drawing has long been used in industrial applications, including the production of flexible fibres with high tensile strength such as polyester and nylon. However, cold-drawing of a composite structure has been less studied. Here we show that in a multimaterial fibre composed of a brittle core embedded in a ductile polymer cladding, cold-drawing results in a surprising phenomenon: controllable and sequential fragmentation of the core to produce uniformly sized rods along metres of fibre, rather than the expected random or chaotic fragmentation. These embedded structures arise from mechanical-geometric instabilities associated with 'neck' propagation. Embedded, structured multimaterial threads with complex transverse geometry are thus fragmented into a periodic train of rods held stationary in the polymer cladding. These rods can then be easily extracted via selective dissolution of the cladding, or can self-heal by thermal restoration to re-form the brittle thread. Our method is also applicable to composites with flat rather than cylindrical geometries, in which case cold-drawing leads to the break-up of an embedded or coated brittle film into narrow parallel strips that are aligned normally to the drawing axis. A range of materials was explored to establish the universality of this effect, including silicon, germanium, gold, glasses, silk, polystyrene, biodegradable polymers and ice. We observe, and verify through nonlinear finite-element simulations, a linear relationship between the smallest transverse scale and the longitudinal break-up period. These results may lead to the development of dynamical and thermoreversible camouflaging via a nanoscale Venetian-blind effect, and the fabrication of large-area structured surfaces that facilitate high-sensitivity bio-detection.

Leg og dannelse

DEFF Research Database (Denmark)

Skovbjerg, Helle Marie

2017-01-01

lederen i det pædagogiske tidskrift Asterisk: ”Leg i skolen, leg i klasserummet, ja legende læring i skolen udgør derimod en enorm, seriøs og ubrugt læringsressource – ikke alene med effekter på kreativiteten, men også på den faglige læring” (Holm, 2015, p. 2). Legens værdi gøres altså først og fremmest...

The Motor and the Brake of the Trailing Leg in Human Walking: Leg Force Control Through Ankle Modulation and Knee Covariance

Science.gov (United States)

Toney, Megan E.; Chang, Young-Hui

2016-01-01

Human walking is a complex task, and we lack a complete understanding of how the neuromuscular system organizes its numerous muscles and joints to achieve consistent and efficient walking mechanics. Focused control of select influential task-level variables may simplify the higher-level control of steady state walking and reduce demand on the neuromuscular system. As trailing leg power generation and force application can affect the mechanical efficiency of step-to-step transitions, we investigated how joint torques are organized to control leg force and leg power during human walking. We tested whether timing of trailing leg force control corresponded with timing of peak leg power generation. We also applied a modified uncontrolled manifold analysis to test whether individual or coordinated joint torque strategies most contributed to leg force control. We found that leg force magnitude was adjusted from step-to-step to maintain consistent leg power generation. Leg force modulation was primarily determined by adjustments in the timing of peak ankle plantar-flexion torque, while knee torque was simultaneously covaried to dampen the effect of ankle torque on leg force. We propose a coordinated joint torque control strategy in which the trailing leg ankle acts as a motor to drive leg power production while trailing leg knee torque acts as a brake to refine leg power production. PMID:27334888

The Occurrence of Cold Spells in the Alps Related to ClimateChange

Directory of Open Access Journals (Sweden)

Seon Ki Park

2010-08-01

Full Text Available Climate change is not only a likely prospect for the end of this century, but it is already occurring. Part of the changes will include global warming and increasing temperature variability, both at global and regional scales. This increased variability was investigated in this paper from the point of view of the occurrence of cold spells in the Alps in the future climate (2071–2100, compared with the present climate (1961–1990. For this purpose, a regionalisation of the climate change effects was performed within the Alps. To avoid possible errors in the estimate of the 2m air temperature, the analysis was performed on the soil surface temperature. To get realistic values for this variable, a land surface scheme, UTOPIA, has been run on the selected domain, using the output of the Regional Climate Model (RegCM3 simulations as the driving force. The results show that, in general, the number of cold breaks is decreasing over the Alps, due to the temperature increment. However, there are certain zones where the behaviour is more complicated. The analysis of the model output also allowed a relationship to be found between the number of cold breaks and their duration. The significance of these results over the whole area was assessed.

A Hydroxyurea-induced Leg Ulcer

OpenAIRE

Hwang, Seon-Wook; Hong, Soon-Kwon; Kim, Sang-Hyun; Seo, Jong-Keun; Lee, Deborah; Sung, Ho-Suk

2009-01-01

Hydroxyurea is a cytostatic agent that has recently become the drug of choice in the treatment of various myeloproliferative diseases. The cutaneous side effects of hydroxyurea include xerosis, hyperpigmentation, nail discoloration, and scaling. Leg ulcers have only rarely been reported in association with hydroxyurea treatment. A 75-year-old woman presented with leg ulcers, nail discoloration, and xerosis. The leg ulcers were refractory to conventional treatment. She had been taking oral hyd...

Duality, exchange-degeneracy breaking, and exotic states

International Nuclear Information System (INIS)

Goldstein, G.R.; Haridas, P.

1979-01-01

We study the connection between exchange-degeneracy breaking and multiquark states within the framework of a highly constrained dual approach. We show that M 4 (baryonium) states emerge at the daughter trajectory level as a consequence of small exchange-degeneracy breaking in the meson-meson system (approx.delta) and larger exchange-degeneracy breaking of the baryon trajectories in the meson-baryon system (approx.epsilon). The M 4 states are coupled weakly to external mesons in proportion to the breaking parameter delta. Assuming M 4 couplings to B-barB channels are strong, as determined by duality with normal mesons in the B-barB system, consistency requires epsilon approx. √delta-bar, thereby relating the larger breaking of baryon trajectories to the violation of the Okubo-Zweig-Iizuka-type rule for M 4 . It is shown that exotic baryon states, B 5 , also emerge from this scheme at the daughter level and that dibaryons will appear at the second daughter level

Chiral symmetry breaking and cooling in lattice QCD

International Nuclear Information System (INIS)

Woloshyn, R.M.; Lee, F.X.

1995-08-01

Chiral symmetry breaking is calculated as a function of cooling in quenched lattice QCD. A non-zero signal is found for the chiral condensate beyond one hundred cooling steps, suggesting that there is chiral symmetry breaking associated with instantons. Quantitatively, the chiral condensate in cooled gauge field configurations is small compared to the value without cooling. (author) 7 refs., 1 tab., 3 figs

Dissipative - free jumps for the magnetoacoustic branch of cold plasma motions

International Nuclear Information System (INIS)

Bakholdin, I.B.

2000-01-01

Dissipative-free jumps were studied in hydrodynamic model of cold plasma moving with the rate close to magnetoacoustic one. The jumps for the generalized Korteweg-de Vries equation with similar nonlinear and dispersion properties were studied. Among them there were jumps with emission and solution type jumps. Furthermore, the numerical investigation into the initial break decomposition in cold plasma confirmed the validity of assumption that in the given type of jumps as in case of the generalized Korteweg-de Vries equation. Paper describes the analytical method enabling to forecast the structure nature of such jumps in the general case [ru

Multinode analysis of small breaks for B and W's 205-fuel-assembly nuclear plants with internals vent valves

International Nuclear Information System (INIS)

Jones, R.C.; Dunn, B.M.; Parks, C.E.

1976-03-01

Multinode analyses were conducted for several small breaks in the reactor coolant system of B and W's 205-fuel-assembly nuclear plants with internals vent valves. The multinode blowdown code CRAFT was used to evaluate the hydrodynamics and transient water inventories of the reactor coolant system. The FOAM code was used to compute a swell level history for the core, and the THETA1-B code was used to perform transient fuel pin thermal calculations. Curves showing the parameters of interest are presented. The results are well within the Final Acceptance Criteria

Multinode analysis of small breaks for B and W's 145-fuel-assembly nuclear plants with internals vent valves

International Nuclear Information System (INIS)

Parks, C.E.; Allen, R.J.; Cartin, L.R.

1976-03-01

Multinode analyses were conducted for several small breaks in the reactor coolant system of B and W's 145 fuel-assembly nuclear plants with internals vent valves. The multinode blowdown code CRAFT was used to evaluate the hydrodynamics and transient water inventories of the reactor coolant system. The FOAM code was used to compute a swell level history for the core, and the THETA1-B code was used to perform transient fuel pin thermal calculations. Curves showing the parameters of interest are presented. These results are well within the Final Acceptance Criteria

Assessment of BETHSY Test 9.1.b using RELAP5/MOD3

International Nuclear Information System (INIS)

Lee, S.; Chung, B.D.; Kim, H.J.

1993-06-01

The 2'' cold leg break test 9.l.b, conducted at the BETHSY facility was analyzed using the RELAP5/MOD3 Version 5m5 code. The test 9.l.b was conducted with the main objective being the investigation of the thermal-hydraulic mechanisms responsible for the large core uncovery and fuel heat-up, requiring the implementation of an ultimate procedure. The present analysis demonstrates the code's capability to predict, with sufficient accuracy, the main phenomena occurring in the depressurization transient, both from a qualitative and quantitative point of view. Nevertheless, several differences regarding the evolution of phenomena and affecting the timing order have to be pointed out in the base calculation. Three calculations were carried out to study the sensitivity to change of the nodalization in the components of the loop seal cross-over legs, and of the auxiliary feedwater control logics, and of the break discharge coefficient

The experimental investigation of supersymmetry breaking

International Nuclear Information System (INIS)

Peskin, M.E.

1996-04-01

If Nature is supersymmetric at the weak interaction scale, what can we hope to learn from experiments on supersymmetric particles? The most mysterious aspect of phenomenological supersymmetry is the mechanism of spontaneous supersymmetry breaking. This mechanism ties the observable pattern of supersymmetric particle masses to aspects of the underlying unified theory at very small distance scales. In this article, I will discuss a systematic experimental program to determine the mechanism of supersymmetry breaking. Both pp and e + e - colliders of the next generation play an essential role

Determinants and Politics of German Military Transformation in the Post-Cold War Era

Science.gov (United States)

2011-06-01

Cold War. Additionally, the prevalent antimilitarism called for armed forces that had to break with their historic record of authoritarianism and... paternalism ‖ in NATO affairs. In light of these diverging perceptions, the new Strategic Concept (SC 99), approved at the anniversary summit in

Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2

International Nuclear Information System (INIS)

Ireland, J.R.

1982-01-01

A 200-percent double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops (three intact, one broken) and cold-leg emergency-core-cooling systems (ECCS). The finely noded TRAC model employed 440 three dimensional (r, THETA, z) vessel cells along with approximately 300 one-dimensional cells that modeled the primary system loops. The calculated peak-clad temperature of 950 0 K occurred during blowdown and the clad temperature excursion was terminated at 175 s, when complete core quenching occurred. Accumulator flows were initiated at 10 s, when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated

Pipe crawler with extendable legs

International Nuclear Information System (INIS)

Zollinger, W.T.

1992-01-01

A pipe crawler for moving through a pipe in inchworm fashion having front and rear leg assemblies separated by air cylinders to increase and decrease the spacing between assemblies. Each leg of the four legs of an assembly is moved between a wall-engaging, extended position and a retracted position by a separate air cylinder. The air cylinders of the leg assemblies are preferably arranged in pairs of oppositely directed cylinders with no pair lying in the same axial plane as another pair. Therefore, the cylinders can be as long as a leg assembly is wide and the crawler can crawl through sections of pipes where the diameter is twice that of other sections. The crawler carries a valving system, a manifold to distribute air supplied by a single umbilical air hose to the various air cylinders in a sequence controlled electrically by a controller. The crawler also utilizes a rolling mechanism, casters in this case, to reduce friction between the crawler and pipe wall thereby further extending the range of the pipe crawler. 8 figs

Pipe crawler with extendable legs

Science.gov (United States)

Zollinger, W.T.

1992-06-16

A pipe crawler for moving through a pipe in inchworm fashion having front and rear leg assemblies separated by air cylinders to increase and decrease the spacing between assemblies. Each leg of the four legs of an assembly is moved between a wall-engaging, extended position and a retracted position by a separate air cylinder. The air cylinders of the leg assemblies are preferably arranged in pairs of oppositely directed cylinders with no pair lying in the same axial plane as another pair. Therefore, the cylinders can be as long as a leg assembly is wide and the crawler can crawl through sections of pipes where the diameter is twice that of other sections. The crawler carries a valving system, a manifold to distribute air supplied by a single umbilical air hose to the various air cylinders in a sequence controlled electrically by a controller. The crawler also utilizes a rolling mechanism, casters in this case, to reduce friction between the crawler and pipe wall thereby further extending the range of the pipe crawler. 8 figs.

Homework particularities for small school children.

Science.gov (United States)

Beiusanu, Corina; Vlaicu, Brigitha

2013-01-01

The present study was centered on the particularities of the duration of preparing homework, taking breaks during homework preparation, and the way the breaks should take place for small school children. The study has been done on a sample of 235 small school children from Oradea, 114 boys and 121 girls, between the ages 7 and 10 years old, using an anonymous questioner, with 41 items, which investigates the lifestyle of the small school children. The duration of homework preparation it is significantly more reduced for the school children in 1st grade in comparison with the ones in 3 grade (p lunch. Half of the children from grades I-IV prepare their homework with no break. A very small number of children spend their homework break time in a healthy manner, while the rest prefer to play computer games (46.95%) or to watch television (46.08%). More than half of the schoolchildren need 1-2 hours at home to prepare their homework. Most of the school children prepare their homework after lunch, in an optimal interval of time. Half of the questioned children prepare their homework with no break. Those who are taking breaks prefer activities which get the children even more tired, therefore being non-hygienic methods of spending homework breaks.

The Advanced Neutron Source liquid deuterium cold source

International Nuclear Information System (INIS)

Lucas, A.T.

1995-08-01

The Advanced Neutron Source will employ two cold sources to moderate neutrons to low energy (<10 meV). The cold neutrons produced are then passed through beam guides to various experiment stations. Each cold source moderator is a sphere of 410-mm internal diameter. The moderator material is liquid deuterium flowing at a rate of 1 kg/s and maintained at subcooled temperatures at all points of the circuit, to prevent boiling. Nuclear beat deposited within the liquid deuterium and its containment structure totals more than 30 kW. All of this heat is removed by the liquid deuterium, which raises its temperature by 5 K. The liquid prime mover is a cryogenic circulator that is situated in the return leg of the flow loop. This arrangement minimizes the heat added to the liquid between the heat exchanger and the moderator vessel, allowing the moderator to be operated at the minimum practical temperature. This report describes the latest thinking at the time of project termination. It also includes the status of various systems at that time and outlines anticipated directions in which the design would have progressed. In this regard, some detail differences between this report and official design documents reflect ideas that were not approved at the time of closure but are considered noteworthy

WIMPless dark matter from non-Abelian hidden sectors with anomaly-mediated supersymmetry breaking

International Nuclear Information System (INIS)

Feng, Jonathan L.; Shadmi, Yael

2011-01-01

In anomaly-mediated supersymmetry breaking models, superpartner masses are proportional to couplings squared. Their hidden sectors therefore naturally contain WIMPless dark matter, particles whose thermal relic abundance is guaranteed to be of the correct size, even though they are not weakly interacting massive particles. We study viable dark matter candidates in WIMPless anomaly-mediated supersymmetry breaking models with non-Abelian hidden sectors and highlight unusual possibilities that emerge in even the simplest models. In one example with a pure SU(N) hidden sector, stable hidden gluinos freeze out with the correct relic density, but have an extremely low, but natural, confinement scale, providing a framework for self-interacting dark matter. In another simple scenario, hidden gluinos freeze out and decay to visible Winos with the correct relic density, and hidden glueballs may either be stable, providing a natural framework for mixed cold-hot dark matter, or may decay, yielding astrophysical signals. Last, we present a model with light hidden pions that may be tested with improved constraints on the number of nonrelativistic degrees of freedom. All of these scenarios are defined by a small number of parameters, are consistent with gauge coupling unification, preserve the beautiful connection between the weak scale and the observed dark matter relic density, and are natural, with relatively light visible superpartners. We conclude with comments on interesting future directions.

Comparison of Critical Flow Models' Evaluations for SBLOCA Tests

International Nuclear Information System (INIS)

Kim, Yeon Sik; Park, Hyun Sik; Cho, Seok

2016-01-01

A comparison of critical flow models between the Trapp-Ransom and Henry-Fauske models for all SBLOCA (small break loss of coolant accident) scenarios of the ATLAS (Advanced thermal-hydraulic test loop for accident simulation) facility was performed using the MARS-KS code. For the comparison of the two critical models, the accumulated break mass was selected as the main parameter for the comparison between the analyses and tests. Four cases showed the same respective discharge coefficients between the two critical models, e.g., 6' CL (cold leg) break and 25%, 50%, and 100% DVI (direct vessel injection) breaks. In the case of the 4' CL break, no reasonable results were obtained with any possible Cd values. In addition, typical system behaviors, e.g., PZR (pressurizer) pressure and collapsed core water level, were also compared between the two critical models. Four cases showed the same respective discharge coefficients between the two critical models, e.g., 6' CL break and 25%, 50%, and 100% DVI breaks. In the case of the 4' CL break, no reasonable results were obtained with any possible Cd values. In addition, typical system behaviors, e.g., PZR pressure and collapsed core water level, were also compared between the two critical models. From the comparison between the two critical models for the CL breaks, the Trapp-Ransom model predicted quite well with respect to the other model for the smallest and larger breaks, e.g., 2', 6', and 8.5' CL breaks. In addition, from the comparison between the two critical models for the DVI breaks, the Trapp-Ransom model predicted quite well with respect to the other model for the smallest and larger breaks, e.g., 5%, 50%, and 100% DVI breaks. In the case of the 50% and 100% breaks, the two critical models predicted the test data quite well.

GPS tracking devices reveal foraging strategies of black-legged kittiwakes

Science.gov (United States)

Kotzerka, Jana; Garthe, Stefan; Hatch, Scott A.

2010-01-01

The Black-legged Kittiwake Rissa tridactyla is the most abundant gull species in the world, but some populations have declined in recent years, apparently due to food shortage. Kittiwakes are surface feeders and thus can compensate for low food availability only by increasing their foraging range and/or devoting more time to foraging. The species is widely studied in many respects, but long-distance foraging and the limitations of conventional radio telemetry have kept its foraging behavior largely out of view. The development of Global Positioning System (GPS) loggers is advancing rapidly. With devices as small as 8 g now available, it is possible to use this technology for tracking relatively small species of oceanic birds like kittiwakes. Here we present the first results of GPS telemetry applied to Black-legged Kittiwakes in 2007 in the North Pacific. All but one individual foraged in the neritic zone north of the island. Three birds performed foraging trips only close to the colony (within 13 km), while six birds had foraging ranges averaging about 40 km. The maximum foraging range was 59 km, and the maximum distance traveled was 165 km. Maximum trip duration was 17 h (mean 8 h). An apparently bimodal distribution of foraging ranges affords new insight on the variable foraging behaviour of Black-legged Kittiwakes. Our successful deployment of GPS loggers on kittiwakes holds much promise for telemetry studies on many other bird species of similar size and provides an incentive for applying this new approach in future studies.

Leg som ustyrlig deltagelseskultur

DEFF Research Database (Denmark)

Toft, Herdis

2017-01-01

- og spilteoretikere Johan Huizinga og Roger Caillois. Deres teorier og begrebsdannelser har været brugt til at påpege leg dels som et æstetisk baseret betydningssystem, dels som et affektivt og stemningsbaseret oplevelsessystem samt endelig som et socialt baseret relationssystem. I artiklen vælger vi...... at fokusere på leg som et socialt baseret relationssystem og yderligere zoome ind på et af legens systemiske væsenstræk, nemlig brugen af regulerbare regelsæt, som legerne uden ’politi’ forhandler sig frem til før, under og efter legen. Fælles for Huizinga og Caillois er, at de knytter leg uløseligt sammen...

Reactor cold neutron source facility, the first in Japan

International Nuclear Information System (INIS)

Utsuro, Masahiko; Maeda, Yutaka; Kawai, Takeshi; Tashiro, Tameyoshi; Sakakibara, Shoji; Katada, Minoru.

1986-01-01

In the Research Reactor Institute, Kyoto University, the first cold neutron source facility for the reactor in Japan was installed, and various tests are carried out outside the reactor. Nippon Sanso K.K. had manufactured it. After the prescribed tests outside the reactor, this facility will be installed soon in the reactor, and its outline is described on this occasion. Cold neutrons are those having very small energy by being cooled to about-250 deg C. Since the wavelength of the material waves of cold neutrons is long, and their energy is small, they are very advantageous as an experimental means for clarifying the structure of living body molecules and polymers, the atom configuration in alloys, and atomic and molecular movements by neutron scattering and neutron diffraction. The basic principle of the cold neutron source facility is to irradiate thermal neutrons on a cold moderator kept around 20 K, and to moderate and cool the neutrons by nuclear scattering to convert to cold neutrons. The preparatory research on cold neutrons and hydrogen liquefaction, the basic design to put the cold neutron source facility in the graphite moderator facility, the safety countermeasures, the manufacture and quality control, the operation outside the reactor and the performance are reported. The cold neutron source facility comprises a cold moderator tank and other main parts, a deuterium gas tank, a helium refrigerator and instrumentation. (Kako, I.)

Thermo-mechanical Design Methodology for ITER Cryodistribution cold boxes

Science.gov (United States)

Shukla, Vinit; Patel, Pratik; Das, Jotirmoy; Vaghela, Hitensinh; Bhattacharya, Ritendra; Shah, Nitin; Choukekar, Ketan; Chang, Hyun-Sik; Sarkar, Biswanath

2017-04-01

The ITER cryo-distribution (CD) system is in charge of proper distribution of the cryogen at required mass flow rate, pressure and temperature level to the users; namely the superconducting (SC) magnets and cryopumps (CPs). The CD system is also capable to use the magnet structures as a thermal buffer in order to operate the cryo-plant as much as possible at a steady state condition. A typical CD cold box is equipped with mainly liquid helium (LHe) bath, heat exchangers (HX’s), cryogenic valves, filter, heaters, cold circulator, cold compressor and process piping. The various load combinations which are likely to occur during the life cycle of the CD cold boxes are imposed on the representative model and impacts on the system are analyzed. This study shows that break of insulation vacuum during nominal operation (NO) along with seismic event (Seismic Level-2) is the most stringent load combination having maximum stress of 224 MPa. However, NO+SMHV (Séismes Maximaux Historiquement Vraisemblables = Maximum Historically Probable Earthquakes) load combination is having the least safety margin and will lead the basis of the design of the CD system and its sub components. This paper presents and compares the results of different load combinations which are likely to occur on a typical CD cold box.

Legāti

OpenAIRE

Segliņa, Aiga

2010-01-01

Autore teorētiski analizē legāta jēdzienu testamentārās mantošanas ietvaros un atspoguļo praktiska pētījuma rezultātus. Teorētiskā daļa apskata legāta nodibināšanas formu un spēkā esamību, tā iegūšanu un atraidīšanu, izpildi un zaudēšanu, novēlējuma robežas un aprobežojumus. Pētījums veikts aptaujas veidā ar mērķi noskaidrot, cik liela Latvijas iedzīvotāju daļa apzinās legāta nodrošinātās priekšrocības testamentārajā mantošanā. Apskatīts notāra neitralitātes jautājums attiecībā pret mantošana...

Nonlinear mode interaction in equal-leg angle struts susceptible to cellular buckling.

Science.gov (United States)

Bai, L; Wang, F; Wadee, M A; Yang, J

2017-11-01

A variational model that describes the interactive buckling of a thin-walled equal-leg angle strut under pure axial compression is presented. A formulation combining the Rayleigh-Ritz method and continuous displacement functions is used to derive a system of differential and integral equilibrium equations for the structural component. Solving the equations using numerical continuation reveals progressive cellular buckling (or snaking) arising from the nonlinear interaction between the weak-axis flexural buckling mode and the strong-axis flexural-torsional buckling mode for the first time-the resulting behaviour being highly unstable. Physical experiments conducted on 10 cold-formed steel specimens are presented and the results show good agreement with the variational model.

String breaking with Wilson loops?

CERN Document Server

Kratochvila, S; Kratochvila, Slavo; Forcrand, Philippe de

2003-01-01

A convincing, uncontroversial observation of string breaking, when the static potential is extracted from Wilson loops only, is still missing. This failure can be understood if the overlap of the Wilson loop with the broken string is exponentially small. In that case, the broken string ground state will only be seen if the Wilson loop is long enough. Our preliminary results show string breaking in the context of the 3d SU(2) adjoint static potential, using the L\\"uscher-Weisz exponential variance reduction approach. As a by-product, we measure the fundamental SU(2) static potential with improved accuracy and see clear deviations from Casimir scaling.

Immediate effects of the trunk stabilizing exercise on static balance parameters in double-leg and one-leg stances

OpenAIRE

Kim, Jwa-jun; Park, Se-yeon

2016-01-01

[Purpose] The purpose of this study was to evaluate the immediate effect of stabilizing exercise using the PNF technique on standing balance in one-leg and double-leg stances. [Subjects and Methods] The present study recruited 34 healthy participants from a local university. The Participants performed four balance tests (double-leg stance with and without vision, one-leg stance with and without vision), before and after exercise. The exercise consisted of exercises performed using PNF techniq...

Artificial Leg Design and Control Research of a Biped Robot with Heterogeneous Legs Based on PID Control Algorithm

Directory of Open Access Journals (Sweden)

Hualong Xie

2015-04-01

Full Text Available A biped robot with heterogeneous legs (BRHL is proposed to provide an ideal test-bed for intelligent bionic legs (IBL. To make artificial leg gait better suited to a human, a four-bar mechanism is used as its knee joint, and a pneumatic artificial muscle (PAM is used as its driving source. The static mathematical model of PAM is established and the mechanical model of a single degree of freedom of a knee joint driven by PAM is analyzed. A control simulation of an artificial leg based on PID control algorithm is carried out and the simulation results indicate that the artificial leg can simulate precisely a normal human walking gait.

Symmetry Breaking in MILP Formulations for Unit Commitment Problems

KAUST Repository

Lima, Ricardo

2015-12-11

This paper addresses the study of symmetry in Unit Commitment (UC) problems solved by Mixed Integer Linear Programming (MILP) formulations, and using Linear Programming based Branch & Bound MILP solvers. We propose three sets of symmetry breaking constraints for UC MILP formulations exhibiting symmetry, and its impact on three UC MILP models are studied. The case studies involve the solution of 24 instances by three widely used models in the literature, with and without symmetry breaking constraints. The results show that problems that could not be solved to optimality within hours can be solved with a relatively small computational burden if the symmetry breaking constraints are assumed. The proposed symmetry breaking constraints are also compared with the symmetry breaking methods included in two MILP solvers, and the symmetry breaking constraints derived in this work have a distinct advantage over the methods in the MILP solvers.

Symmetry Breaking in MILP Formulations for Unit Commitment Problems

KAUST Repository

Lima, Ricardo; Novais, Augusto Q.

2015-01-01

This paper addresses the study of symmetry in Unit Commitment (UC) problems solved by Mixed Integer Linear Programming (MILP) formulations, and using Linear Programming based Branch & Bound MILP solvers. We propose three sets of symmetry breaking constraints for UC MILP formulations exhibiting symmetry, and its impact on three UC MILP models are studied. The case studies involve the solution of 24 instances by three widely used models in the literature, with and without symmetry breaking constraints. The results show that problems that could not be solved to optimality within hours can be solved with a relatively small computational burden if the symmetry breaking constraints are assumed. The proposed symmetry breaking constraints are also compared with the symmetry breaking methods included in two MILP solvers, and the symmetry breaking constraints derived in this work have a distinct advantage over the methods in the MILP solvers.

A legged anchoring mechanism for capsule endoscopes using micropatterned adhesives.

Science.gov (United States)

Glass, Paul; Cheung, Eugene; Sitti, Metin

2008-12-01

This paper presents a new concept for an anchoring mechanism to enhance existing capsule endoscopes. The mechanism consists of three actuated legs with compliant feet lined with micropillar adhesives to be pressed into the intestine wall to anchor the device at a fixed location. These adhesive systems are inspired by gecko and beetle foot hairs. Single-leg and full capsule mathematical models of the forces generated by the legs are analyzed to understand capsule performance. Empirical friction models for the interaction of the adhesives with an intestinal substrate were experimentally determined in vitro using dry and oil-coated elastomer micropillar arrays with 140 microm pillar diameter, 105 microm spacing between pillars, and an aspect ratio of 1:1 on fresh porcine small intestine specimens. Capsule prototypes were also tested in a simulated intestine environment and compared with predicted peristaltic loads to assess the viability of the proposed design. The experimental results showed that a deployed 10 gr capsule robot can withstand axial peristaltic loads and anchor reliably when actuation forces are greater than 0.27 N using dry micropillars. Required actuation forces may be reduced significantly by using micropillars coated with a thin silicone oil layer.

Spontaneous Symmetry-Breaking in a Network Model for Quadruped Locomotion

Science.gov (United States)

Stewart, Ian

2017-12-01

Spontaneous symmetry-breaking proves a mechanism for pattern generation in legged locomotion of animals. The basic timing patterns of animal gaits are produced by a network of spinal neurons known as a Central Pattern Generator (CPG). Animal gaits are primarily characterized by phase differences between leg movements in a periodic gait cycle. Many different gaits occur, often having spatial or spatiotemporal symmetries. A natural way to explain gait patterns is to assume that the CPG is symmetric, and to classify the possible symmetry-breaking periodic motions. Pinto and Golubitsky have discussed a four-node model CPG network for biped gaits with ℤ2 × ℤ2 symmetry, classifying the possible periodic states that can arise. A more specific rate model with this structure has been analyzed in detail by Stewart. Here we extend these methods to quadruped gaits, using an eight-node network with ℤ4 × ℤ2 symmetry proposed by Golubitsky and coworkers. We formulate a rate model and calculate how the first steady or Hopf bifurcation depends on its parameters, which represent four connection strengths. The calculations involve a distinction between “real” gaits with one or two phase shifts (pronk, bound, pace, trot) and “complex” gaits with four phase shifts (forward and reverse walk, forward and reverse buck). The former correspond to real eigenvalues of the connection matrix, the latter to complex conjugate pairs. The partition of parameter space according to the first bifurcation, ignoring complex gaits, is described explicitly. The complex gaits introduce further complications, not yet fully understood. All eight gaits can occur as the first bifurcation from a fully synchronous equilibrium, for suitable parameters, and numerical simulations indicate that they can be asymptotically stable.

Robust and efficient walking with spring-like legs

Energy Technology Data Exchange (ETDEWEB)

Rummel, J; Blum, Y; Seyfarth, A, E-mail: juergen.rummel@uni-jena.d, E-mail: andre.seyfarth@uni-jena.d [Lauflabor Locomotion Laboratory, University of Jena, Dornburger Strasse 23, 07743 Jena (Germany)

2010-12-15

The development of bipedal walking robots is inspired by human walking. A way of implementing walking could be performed by mimicking human leg dynamics. A fundamental model, representing human leg dynamics during walking and running, is the bipedal spring-mass model which is the basis for this paper. The aim of this study is the identification of leg parameters leading to a compromise between robustness and energy efficiency in walking. It is found that, compared to asymmetric walking, symmetric walking with flatter angles of attack reveals such a compromise. With increasing leg stiffness, energy efficiency increases continuously. However, robustness is the maximum at moderate leg stiffness and decreases slightly with increasing stiffness. Hence, an adjustable leg compliance would be preferred, which is adaptable to the environment. If the ground is even, a high leg stiffness leads to energy efficient walking. However, if external perturbations are expected, e.g. when the robot walks on uneven terrain, the leg should be softer and the angle of attack flatter. In the case of underactuated robots with constant physical springs, the leg stiffness should be larger than k-tilde = 14 in order to use the most robust gait. Soft legs, however, lack in both robustness and efficiency.

Robust and efficient walking with spring-like legs

International Nuclear Information System (INIS)

Rummel, J; Blum, Y; Seyfarth, A

2010-01-01

The development of bipedal walking robots is inspired by human walking. A way of implementing walking could be performed by mimicking human leg dynamics. A fundamental model, representing human leg dynamics during walking and running, is the bipedal spring-mass model which is the basis for this paper. The aim of this study is the identification of leg parameters leading to a compromise between robustness and energy efficiency in walking. It is found that, compared to asymmetric walking, symmetric walking with flatter angles of attack reveals such a compromise. With increasing leg stiffness, energy efficiency increases continuously. However, robustness is the maximum at moderate leg stiffness and decreases slightly with increasing stiffness. Hence, an adjustable leg compliance would be preferred, which is adaptable to the environment. If the ground is even, a high leg stiffness leads to energy efficient walking. However, if external perturbations are expected, e.g. when the robot walks on uneven terrain, the leg should be softer and the angle of attack flatter. In the case of underactuated robots with constant physical springs, the leg stiffness should be larger than k-tilde = 14 in order to use the most robust gait. Soft legs, however, lack in both robustness and efficiency.

Application of fast fourier transform method to evaluate the accuracy of sbloca data base

International Nuclear Information System (INIS)

D'Auria, F.; Galassi, G.M.; Leonardi, M.; Galetti, M.R.

1997-01-01

The purpose of this paper is to perform the quantitative accuracy evaluation of a small break LOCA data base and then evaluate the accuracy of RELAP5/MOD2 code i.e. of the ensemble constituted by the code itself, the user, the nodalization and the selected code options, in predicting this kind of transient. In order to achieve this objective, qualitative accuracy evaluation results from several tests performed in 4 facilities (LOBI, SPES, BETHSY and LSTF) are used. The quantitative evaluation is achieved adopting a method developed at University of Pisa, which has capabilities in quantifying the errors in code predictions with respect to the measured experimental signal, using the Fast Fourier Transform; this allows an integral representation of code discrepancies in the frequency domain. The RELAP5/MOD2 code has been extensively used at the University of Pisa and the nodalizations of the 4 facilities have been qualified through the application to several experiments performed in the same facilities. (author)

Does a crouched leg posture enhance running stability and robustness?

Science.gov (United States)

Blum, Yvonne; Birn-Jeffery, Aleksandra; Daley, Monica A; Seyfarth, Andre

2011-07-21

Humans and birds both walk and run bipedally on compliant legs. However, differences in leg architecture may result in species-specific leg control strategies as indicated by the observed gait patterns. In this work, control strategies for stable running are derived based on a conceptual model and compared with experimental data on running humans and pheasants (Phasianus colchicus). From a model perspective, running with compliant legs can be represented by the planar spring mass model and stabilized by applying swing leg control. Here, linear adaptations of the three leg parameters, leg angle, leg length and leg stiffness during late swing phase are assumed. Experimentally observed kinematic control parameters (leg rotation and leg length change) of human and avian running are compared, and interpreted within the context of this model, with specific focus on stability and robustness characteristics. The results suggest differences in stability characteristics and applied control strategies of human and avian running, which may relate to differences in leg posture (straight leg posture in humans, and crouched leg posture in birds). It has been suggested that crouched leg postures may improve stability. However, as the system of control strategies is overdetermined, our model findings suggest that a crouched leg posture does not necessarily enhance running stability. The model also predicts different leg stiffness adaptation rates for human and avian running, and suggests that a crouched avian leg posture, which is capable of both leg shortening and lengthening, allows for stable running without adjusting leg stiffness. In contrast, in straight-legged human running, the preparation of the ground contact seems to be more critical, requiring leg stiffness adjustment to remain stable. Finally, analysis of a simple robustness measure, the normalized maximum drop, suggests that the crouched leg posture may provide greater robustness to changes in terrain height

Large LOCA-earthquake combination probability assessment - Load combination program. Project 1 summary report

Energy Technology Data Exchange (ETDEWEB)

Lu, S; Streit, R D; Chou, C K

1980-01-01

This report summarizes work performed for the U.S. Nuclear Regulatory Commission (NRC) by the Load Combination Program at the Lawrence Livermore National Laboratory to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR-1, is used for this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated with a deterministic fracture mechanics model that incorporates stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without an earthquake, is very small (on the order of 10{sup -12}). The probability of a leak was found to be several orders of magnitude greater than that of a complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported. (author)

Large LOCA-earthquake combination probability assessment - Load combination program. Project 1 summary report

International Nuclear Information System (INIS)

Lu, S.; Streit, R.D.; Chou, C.K.

1980-01-01

This report summarizes work performed for the U.S. Nuclear Regulatory Commission (NRC) by the Load Combination Program at the Lawrence Livermore National Laboratory to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR-1, is used for this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated with a deterministic fracture mechanics model that incorporates stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without an earthquake, is very small (on the order of 10 -12 ). The probability of a leak was found to be several orders of magnitude greater than that of a complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported. (author)

Leg intravenous pressure during head-up tilt.

NARCIS (Netherlands)

Groothuis, J.T.; Poelkens, F.; Wouters, C.W.; Kooijman, H.M.; Hopman, M.T.E.

2008-01-01

Leg vascular resistance is calculated as the arterial-venous pressure gradient divided by blood flow. During orthostatic challenges it is assumed that the hydrostatic pressure contributes equally to leg arterial, as well as to leg venous pressure. Because of venous valves, one may question whether,

THYDE-B1/MOD1: a computer code for analysis of small-break loss-of-coolant accident of boiling water reactors

International Nuclear Information System (INIS)

Muramatsu, Ken; Akimoto, Masayuki

1982-08-01

THYDE-B1/MOD1 is a computer code to analyze thermo-hydraulic transients of the reactor cooling system of a BWR, mainly during a small-break loss-of-coolant accidnet (SB-LOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions, i.e., subcooled liquid, saturated mixture and saturated steam regions are allowed to exist. The regions are separated by moving boundaries, tracked by mass and energy balances for each region. The interior of the pressure vessel is represented by two volumes with three regions: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SB-LOCAs in which the hydraulic transient is relatively slow and the cooling of the core strongly depends on the mixture level behavior in the vessel. In order to simulate the system behavior, THYDE-B1 is provided with analytical models for reactor kinetics, heat generation and conduction in fuel rods and structures, heat transfer between coolant and solid surfaces, coolant injection systems, breaks and discharge systems, jet pumps, recirculation pumps, and so on. The verification of the code has been conducted. A good predictability of the code has been indicated through the comparison of calculated results with experimental data provided by ROSA-III small-break tests. This report presents the analytical models, solution method, and input data requirements of the THYDE-B1/MOD1 code. (author)

Doppler ultrasound exam of an arm or leg

Science.gov (United States)

Peripheral vascular disease - Doppler; PVD - Doppler; PAD - Doppler; Blockage of leg arteries - Doppler; Intermittent claudication - Doppler; Arterial insufficiency of the legs - Doppler; Leg pain and ...

Study on the Break Accidents of the HTR-PM Primary Loop

International Nuclear Information System (INIS)

Lang Minggang; Sun Ximing; Zheng Yanhua

2014-01-01

In thermal hydraulics design and safety analysis of the HTR-PM, the THERMIX code was used to study the behavior of the helium in the primary system. Once the helium leaks from the primary loop through a break or a relief valve, it is hard to simulate the states of the leakage room with THERMIX. In this paper, the latest version of RELAP5/MOD4, was used to simulate the behavior of the helium released to the containment rooms. A RELAP5/MOD4 model of the HTR-PM, including the core, the primary system, the secondary loop and the containment, were developed and evaluated in this paper. Based on the model, this paper studied the accidents consequences of a large break in the pressure relief room and a small break in the instrument room of the HTR-PM reactor building. The simulating results illustrate that the temperature in the pressure relief room was no more than 200℃ after a un-isolating large break, and the temperature in the instrument room is less than 130 ℃ after a small un-isolating break. The analysis shows that the scram function and the ability to monitor the reactor temperature and pressure after accidents would not be affected by the break. (author)

Assessment of the vibration on the foam legged and sheet metal-legged passenger seat

Directory of Open Access Journals (Sweden)

L. Dahil

2015-10-01

Full Text Available In this study, it was aim ed to decrease the vibration reaching to passenger from the legs of vehicle seats. In order to determine the levels of vibrations reaching at passengers, a test pad placed under the passenger seat was used, and HVM100 device was used for digitizing the information obtained. By transferring the vibration data to system by using HVM100 device, the acceleration graphics were prepared with Blaze software. As a result, it was determined that the acceleration values of seat legs made of foam material were lower than that of seat legs made of 2 mm thick sheet metal, so they damped the vibration better.

Development of the MARS input model for Kori nuclear units 1 transient analyzer

International Nuclear Information System (INIS)

Hwang, M.; Kim, K. D.; Lee, S. W.; Lee, Y. J.; Lee, W. J.; Chung, B. D.; Jeong, J. J.

2004-11-01

KAERI has been developing the 'NSSS transient analyzer' based on best-estimate codes for Kori Nuclear Units 1 plants. The MARS and RETRAN codes have been used as the best-estimate codes for the NSSS transient analyzer. Among these codes, the MARS code is adopted for realistic analysis of small- and large-break loss-of-coolant accidents, of which break size is greater than 2 inch diameter. So it is necessary to develop the MARS input model for Kori Nuclear Units 1 plants. This report includes the input model (hydrodynamic component and heat structure models) requirements and the calculation note for the MARS input data generation for Kori Nuclear Units 1 plant analyzer (see the Appendix). In order to confirm the validity of the input data, we performed the calculations for a steady state at 100 % power operation condition and a double-ended cold leg break LOCA. The results of the steady-state calculation agree well with the design data. The results of the LOCA calculation seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the MARS input data can be used as a base input deck for the MARS transient analyzer for Kori Nuclear Units 1

Frustrated S = 1/2 Two-Leg Ladder with Different Leg Interactions

Science.gov (United States)

Tonegawa, Takashi; Okamoto, Kiyomi; Hikihara, Toshiya; Sakai, Tôru

2017-04-01

We explore the ground-state phase diagram of the S = 1/2 two-leg ladder. The isotropic leg interactions J1,a and J1,b between nearest neighbor spins in the legs a and b, respectively, are different from each other. The xy and z components of the uniform rung interactions are denoted by Jr and ΔJr, respectively, where Δ is the XXZ anisotropy parameter. This system has a frustration when J1,aJ1,b employ the physical consideration, the level spectroscopy analysis of the results obtained by the exact diagonalization method and also the density-matrix renormalization-group method. It is found that the non-collinear ferrimagnetic (NCFR) state appears as the ground state in the frustrated region of the parameters. Furthermore, the direct-product triplet-dimer (TD) state in which all rungs form the TD pair is the exact ground state, when J1,a + J1,b = 0 and 0≤ Δ ≲ 0.83. The obtained phase diagrams consist of the TD, XY and Haldane phases as well as the NCFR phase.

Cold in-place recycling using solventless emulsion - phase IV (emulsion qualification and long-term field performance).

Science.gov (United States)

2016-05-01

This report looks into how a successful Cold In-Place solventless emulsion behaves and how the emulsion : break test developed in Phase III of this project demonstrates that behavior. Modifications to the test have been : made to improve the consiste...

RELAP/MOD3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility

Energy Technology Data Exchange (ETDEWEB)

Bayless, P.D.

2003-01-17

The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. This work was performed as part of the U.S. Department of Energy's International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation in the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior.

RELAP/MOD3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility

International Nuclear Information System (INIS)

Bayless, P.D.

2003-01-01

The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. This work was performed as part of the U.S. Department of Energy's International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation in the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior

Effects of cold temperatures on the excitability of rat trigeminal ganglion neurons that are not for cold-sensing

Science.gov (United States)

Kanda, Hirosato; Gu, Jianguo G.

2016-01-01

Except a small population of primary afferent neurons for sensing cold to generate the sensations of innocuous and noxious cold, it is generally believed that cold temperatures suppress the excitability of other primary afferent neurons that are not for cold-sensing. These not-for-cold-sensing neurons include the majority of non-nociceptive and nociceptive afferent neurons. In the present study we have found that not-for-cold-sensing neurons of rat trigeminal ganglia (TG) change their excitability in several ways at cooling temperatures. In nearly 70% of not-for-cold-sensing TG neurons, the cooling temperature of 15°C increases their membrane excitability. We regard these neurons as cold-active neurons. For the remaining 30% of not-for-cold-sensing TG neurons, the cooling temperature of 15°C either has no effect (regarded as cold-ineffective neurons) or suppress (regarded as cold-suppressive neurons) their membrane excitability. For cold-active neurons, the cold temperature of 15°C increases their excitability as is evidenced by the increases in action potential (AP) firing numbers and/or reduction of AP rheobase when these neurons are depolarized electrically. The cold temperature of 15°C significantly inhibits M-currents and increases membrane input resistance of cold-active neurons. Retigabine, an M-current activator, abolishes the effect of cold temperatures on AP firing but not the effect of cold temperature on AP rheobase levels. The inhibition of M-currents and the increases of membrane input resistance are likely two mechanisms by which cooling temperatures increase the excitability of not-for-cold-sensing TG neurons. PMID:26709732

Approach to leg edema

Directory of Open Access Journals (Sweden)

Fulvio Pomero

2017-09-01

Full Text Available Edema is defined as a palpable swelling caused by an increase in interstitial fluid volume. Leg edema is a common problem with a wide range of possible causes and is the result of an imbalance in the filtration system between the capillary and interstitial spaces. Major causes of edema include venous obstruction, increased capillary permeability and increased plasma volume secondary to sodium and water retention. In both hospital and general practice, the patient with a swollen leg presents a common dilemma in diagnosis and treatment. The cause may be trivial or life-threatening and it is often difficult to determine the clinical pathway. The diagnosis can be narrowed by categorizing the edema according to its duration, distribution (unilateral or bilateral and accompanying symptoms. This work provides clinically oriented recommendations for the management of leg edema in adults.

Fabrication of a Microtubular La0.6Sr0.4Ti0.2Fe0.8O3−δ Membrane by Electrophoretic Deposition for Hydrogen Production

Directory of Open Access Journals (Sweden)

Kyoung-Jin Lee

2015-01-01

Full Text Available Microtubular type La0.6Sr0.4Ti0.2Fe0.8O3−δ (LSTF membranes were prepared by electrophoretic deposition (EPD. The oxygen permeation and hydrogen production behavior of the membranes were investigated under various conditions. LSTF green layer was successfully coated onto a carbon rod and, after heat treatment at 1400°C in air, a dense LSTF tubular membrane with a thickness of 250 mm can be obtained. The oxygen permeation and hydrogen production rate were enhanced by CH4 in the permeate side, and the hydrogen production rate by water splitting was 0.22 mL/min·cm2 at 1000°C. It is believed that hydrogen production via water splitting using these tubular LSTF membranes is possible.

Consistency of Trend Break Point Estimator with Underspecified Break Number

Directory of Open Access Journals (Sweden)

Jingjing Yang

2017-01-01

Full Text Available This paper discusses the consistency of trend break point estimators when the number of breaks is underspecified. The consistency of break point estimators in a simple location model with level shifts has been well documented by researchers under various settings, including extensions such as allowing a time trend in the model. Despite the consistency of break point estimators of level shifts, there are few papers on the consistency of trend shift break point estimators in the presence of an underspecified break number. The simulation study and asymptotic analysis in this paper show that the trend shift break point estimator does not converge to the true break points when the break number is underspecified. In the case of two trend shifts, the inconsistency problem worsens if the magnitudes of the breaks are similar and the breaks are either both positive or both negative. The limiting distribution for the trend break point estimator is developed and closely approximates the finite sample performance.

Composite quarks and leptons from dynamical supersymmetry breaking without messengers

International Nuclear Information System (INIS)

Arkani-Hamed, N.; Luty, M.A.; Terning, J.

1998-01-01

We present new theories of dynamical supersymmetry breaking in which the strong interactions that break supersymmetry also give rise to composite quarks and leptons with naturally small Yukawa couplings. In these models, supersymmetry breaking is communicated directly to the composite fields without open-quotes messengerclose quotes interactions. The compositeness scale can be anywhere between 10thinspTeV and the Planck scale. These models can naturally solve the supersymmetric flavor problem, and generically predict sfermion mass unification independent from gauge unification. copyright 1998 The American Physical Society

Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

Energy Technology Data Exchange (ETDEWEB)

Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

1995-09-01

There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

Revisiting LOFT L2-5 large break test in BEMUSE project context. Sensitivity studies

International Nuclear Information System (INIS)

Perez, Marina; Batet, Lluis; Pretel, Carme; Reventos, Francesc

2005-01-01

Full text of publication follows: Best estimate codes simulate NPPs behavior in principle without any special conservative assumptions. Due to several factors like code solution methods or user effects, the output parameters calculated have an uncertainty associated. The quantification of the these uncertainties becomes crucial when a safety statement is to be made. It is in this scope that GAMA group from CSNI (OECD/NEA) proposed the international BEMUSE project (Best Estimate - Uncertainty and Sensitivity Evaluation) having as main objective the evaluation of different methodologies for the uncertainty and sensitivity analysis of best-estimate code calculations. A number of methodologies prepared in different countries are used in the development of the project activities. The program work consists of 6 phases and currently the first two have already been concluded. Phase II consists in revisiting the ISP-13, the LOFT loss of coolant experiment L2-5 which simulated a double ended 200% cold leg break of a commercial PWR simultaneous with a loss of site power. In order to connect phase II with phase III, in which the uncertainty analysis will be carried out, quite a large number of sensitivity analysis have been performed by simulating system failures and varying fuel elements parameters among others. The presentation will focus on the results of the sensitivity analysis as well as its importance with regards to the uncertainty studies. The methodology used by UPC team was developed by ENUSA and the work is supported by the Spanish regulatory organization. (authors)

Evaluation of arm-leg coordination in flat breaststroke.

Science.gov (United States)

Chollet, D; Seifert, L; Leblanc, H; Boulesteix, L; Carter, M

2004-10-01

This study proposes a new method to evaluate arm-leg coordination in flat breaststroke. Five arm and leg stroke phases were defined with a velocity-video system. Five time gaps quantified the time between arm and leg actions during three paces of a race (200 m, 100 m and 50 m) in 16 top level swimmers. Based on these time gaps, effective glide, effective propulsion, effective leg insweep and effective recovery were used to identify the different stroke phases of the body. A faster pace corresponded to increased stroke rate, decreased stroke length, increased propulsive phases, shorter glide phases, and a shorter T1 time gap, which measured the effective body glide. The top level swimmers showed short time gaps (T2, T3, T4, measuring the timing of arm-leg recoveries), which reflected the continuity in arm and leg actions. The measurement of these time gaps thus provides a pertinent evaluation of swimmers' skill in adapting their arm-leg coordination to biomechanical constraints.

Preliminary analyses on hydrogen diffusion through small break of thermo-chemical IS process hydrogen plant

International Nuclear Information System (INIS)

Somolova, Marketa; Terada, Atsuhiko; Takegami, Hiroaki; Iwatsuki, Jin

2008-12-01

Japan Atomic Energy Agency has been conducting a conceptual design study of nuclear hydrogen demonstration plant, that is, a thermal-chemical IS process hydrogen plant coupled with the High temperature Engineering Test Reactor (HTTR-IS), which will be planed to produce a large amount of hydrogen up to 1000m 3 /h. As part of the conceptual design work of the HTTR-IS system, preliminary analyses on small break of a hydrogen pipeline in the IS process hydrogen plant was carried out as a first step of the safety analyses. This report presents analytical results of hydrogen diffusion behaviors predicted with a CFD code, in which a diffusion model focused on the turbulent Schmidt number was incorporated. By modifying diffusion model, especially a constant accompanying the turbulent Schmidt number in the diffusion term, analytical results was made agreed well with the experimental results. (author)

Leg contracture in mice after single and multifractionated 137Cs exposure

International Nuclear Information System (INIS)

Masuda, K.; Hunter, N.; Stone, H.B.; Withers, H.R.

1987-01-01

This is a report of studies of time-dose relationships for post-irradiation leg contractures in mice. The isoeffect doses for various degrees of contracture, measured 250 days after irradiation, increased with the number of fractions, but not with the overall treatment times, throughout 30 days. The isoeffect curves relating the total doses for given levels of responses to the doses per fraction were steeper for leg contractures than for acute skin reactions. The alpha/beta ratios ranged from 1.4 to 5.0 Gy, depending on the degrees of contracture. They were less than the 7.5 to 50 Gy for acute skin reactions as determined in previous experiments using the same animals and irradiation systems. Thus, the data resembled those from other slowly-responding normal tissues such as the spinal cord, kidney and lung. The leg contracture consisted of dermatogenic, myogenic, and arthrogenic components; after the mice were sacrificed there was residual contracture following removal of the skin and muscle. Inhibition of bone growth accounted for only a small proportion of the contracture. The overall response reflected responses of several tissue types

Leg contracture in mice after single and multifractionated 137Cs exposure

Energy Technology Data Exchange (ETDEWEB)

Masuda, K.; Hunter, N.; Stone, H.B.; Withers, H.R.

1987-08-01

This is a report of studies of time-dose relationships for post-irradiation leg contractures in mice. The isoeffect doses for various degrees of contracture, measured 250 days after irradiation, increased with the number of fractions, but not with the overall treatment times, throughout 30 days. The isoeffect curves relating the total doses for given levels of responses to the doses per fraction were steeper for leg contractures than for acute skin reactions. The alpha/beta ratios ranged from 1.4 to 5.0 Gy, depending on the degrees of contracture. They were less than the 7.5 to 50 Gy for acute skin reactions as determined in previous experiments using the same animals and irradiation systems. Thus, the data resembled those from other slowly-responding normal tissues such as the spinal cord, kidney and lung. The leg contracture consisted of dermatogenic, myogenic, and arthrogenic components; after the mice were sacrificed there was residual contracture following removal of the skin and muscle. Inhibition of bone growth accounted for only a small proportion of the contracture. The overall response reflected responses of several tissue types.

Relationship between Leg Mass, Leg Composition and Foot Velocity on Kicking Accuracy in Australian Football.

Science.gov (United States)

Hart, Nicolas H; Nimphius, Sophia; Spiteri, Tania; Cochrane, Jodie L; Newton, Robert U

2016-06-01

Kicking a ball accurately over a desired distance to an intended target is arguably the most important skill to acquire in Australian Football. Therefore, understanding the potential mechanisms which underpin kicking accuracy is warranted. The aim of this study was to examine the relationship between leg mass, leg composition and foot velocity on kicking accuracy in Australian Football. Thirty-one Australian Footballers (n = 31; age: 22.1 ± 2.8 years; height: 1.81 ± 0.07 m; weight: 85.1 ± 13.0 kg; BMI: 25.9 ± 3.2) each performed ten drop punt kicks over twenty metres to a player target. Athletes were separated into accurate (n = 15) and inaccurate (n = 16) kicking groups. Leg mass characteristics were assessed using whole body DXA scans. Foot velocity was determined using a ten-camera optoelectronic, three-dimensional motion capture system. Interactions between leg mass and foot velocity evident within accurate kickers only (r = -0.670 to -0.701). Relative lean mass was positively correlated with kicking accuracy (r = 0.631), while no relationship between foot velocity and kicking accuracy was evident in isolation (r = -0.047 to -0.083). Given the evident importance of lean mass, and its interaction with foot velocity for accurate kickers; future research should explore speed-accuracy, impulse-variability, limb co-ordination and foot-ball interaction constructs in kicking using controlled with-in subject studies to examine the effects of resistance training and skill acquisition programs on the development of kicking accuracy. Key pointsAccurate kickers expressed a very strong inverse relationship between leg mass and foot velocity. Inaccurate kickers were unable to replicate this, with greater volatility in their performance, indicating an ability of accurate kickers to mediate foot velocity to compensate for leg mass in order to deliver the ball over the required distance.Accurate kickers exhibited larger quantities of relative lean mass and lower quantities

Microfluidic droplet generator with controlled break-up mechanism

KAUST Repository

Gonzalez, David Conchouso

2017-04-13

Droplet generation devices and systems that parallelize droplet generation devices are provided. The droplet generation devices can include a symmetric block-and-break system and a tapered droplet generation zone. The symmetric block-and-break system can include a pair of break channels and a pair of bypass channels symmetrically arranged with respect to the dispersed-phase input channel and the output channel. The droplet generation devices can generate monodisperse droplets with a predefined volume over a range of flow rates, pressures, and fluid properties. The droplet generation devices are therefore capable of parallelization to achieve large-capacity droplet generation, e.g. greater than 1 L/hr, with small overall coefficients of variation.

Unification of SUSY breaking and GUT breaking

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, Tatsuo [Department of Physics, Hokkaido University,Sapporo 060-0810 (Japan); Omura, Yuji [Department of Physics, Nagoya University,Nagoya 464-8602 (Japan)

2015-02-18

We build explicit supersymmetric unification models where grand unified gauge symmetry breaking and supersymmetry (SUSY) breaking are caused by the same sector. Besides, the SM-charged particles are also predicted by the symmetry breaking sector, and they give the soft SUSY breaking terms through the so-called gauge mediation. We investigate the mass spectrums in an explicit model with SU(5) and additional gauge groups, and discuss its phenomenological aspects. Especially, nonzero A-term and B-term are generated at one-loop level according to the mediation via the vector superfields, so that the electro-weak symmetry breaking and 125 GeV Higgs mass may be achieved by the large B-term and A-term even if the stop mass is around 1 TeV.

Duplex sonography of the near-surface leg veins

International Nuclear Information System (INIS)

Mendoza, E.

2007-01-01

The book contains the following contributions: The ultrasonograph, selection of the ultrasonic transducer, anatomy of the near-surface vein system, physiology of the near-surface vein system, varicose status classification, systematics of the duplex sonography of near-surface leg veins, provocational maneuver for the duplex sonographic varicose diagnostics, exploration of vena saphena parva, perforans veins, side branches, phlebitis, sonography for varicose therapy, postsurgical sonography, deep leg veins, examination of near-surface leg veins for the pathology of the deep vein system, differential diagnostic clarification of leg oedema from the phlebologic-lymphological view, diagnostic side features along the near-surface leg veins

A cold mass support system based on the use of oriented fiberglass epoxy rods in bending

International Nuclear Information System (INIS)

Green, Michael A.; Corradi, Carol A.; LaMantia, Roberto F.; Zbasnik, Jon P.

2002-01-01

This report describes a cold mass support system that uses oriented fiberglass epoxy (other low heat leak oriented fiber material can also be used) rods. In the direction of the rods, where forces are carried in tension or compression, the support system is very stiff. In the other directions, the rods are subjected to bending stresses. When the support rods are put in bending the cold mass support is quite compliant. This type of support system can be used in situation where space for a cold mass support system is limited and where compliance can be tolerated in at least one direction. Break test data for 15.9-mm and 19.1-mm diameter oriented fiberglass rods is presented in this report. The cold mass supports for the DFBX distribution boxes are presented as an example of this type of cold mass support system

Application of force-length curve for determination of leg stiffness during a vertical jump.

Science.gov (United States)

Struzik, Artur; Zawadzki, Jerzy

2016-01-01

The aim of this study was to present the methodology for estimation of a leg stiffness during a countermovement jump. The question was asked whether leg stiffness in the countermovement and take-off phases are similar to each other as demonstrated in previous reports. It was also examined whether the stiffness in left lower limb is similar to the one in right lower limb. The research was conducted on 35 basketball players. Each participant performed three countermovement jumps with arm swing to the maximum height. Measurements employed a Kistlerforce plate and a BTS SMART system for motion analysis. Leg stiffness (understood as an inclination of the curve of ground reaction forces vs. length) was computed for these parts of countermovement and take-off phases where its value was relatively constant and F(Δl) relationship was similar to linear. Mean value (±SD) of total stiffness of both lower limbs in the countermovement phase was 7.1 ± 2.3 kN/m, whereas this value in the take-off phase was 7.5 ± 1 kN/m. No statistically significant differences were found between the leg stiffness in the countermovement and the take-off phases. No statistically significant differences were found during the comparison of the stiffness in the right and left lower limb. The calculation methodology allows us to estimate the value of leg stiffness based on the actual shape of F(Δl) curve rather than on extreme values of ΔF and Δl. Despite different tasks of the countermovement and the take-off phases, leg stiffness in these phases is very similar. Leg stiffness during a single vertical jump maintains a relatively constant value in the parts with a small value of acceleration.

Electrical phenomena in breaking the adhesive contact and in the destruction of solids. Development stages: From the gas discharge to cold nuclear fusion (on the 90th anniversary of the Academician of the Russian Academy of Sciences, B. V. Deryagin)

International Nuclear Information System (INIS)

Khrustalev, Yu.A.

1993-01-01

A brief review is presented of the research performed mainly by B. V. Deryagin and his coworkers into the electrophysical phenomena that accompany the breaking of an adhesive bond and of solids themselves by a mechanical action (the charging of fresh surfaces and gas discharge processes, caused by an excess charge, the emission of electrons and of x-ray radiation, as well as cold nuclear fusion). The relationship between the emission of electrons and mechanochemical processes in solids is pointed out. 118 refs

Validation of cold chain during distribution of parenteral nutrition

Directory of Open Access Journals (Sweden)

Federico Tuan

2015-09-01

Full Text Available Objective: this study aims to demonstrate the suitability of the process used to condition the extemporaneous mixtures of parenteral nutrition for distribution, considering the objective of preserving the cold chain during transport until it reaches the patient, necessary to ensure stability, effectiveness and safety of these mixtures. Method: concurrent validation, design and implementation of a protocol for evaluating the process of packaging and distribution of MNPE developed by a pharmaceutical laboratory. Running tests, according to predefined acceptance criteria. It is performed twice, in summer and on routes that require longer transfer time. Evaluation of conservation of temperature by monitoring the internal temperature values of each type of packaging, recorded by data loggers calibrated equipment. Results: the different tests meet the established criteria. The collected data ensure the maintenance of the cold chain for longer than the transfer time to the most distant points. Conclusions: this study establishes the suitability of the processes to maintaining the cold chain for transfer from the laboratory to the patient pharmacist. Whereas the breaking of cold chain can cause changes of compatibility and stability of parenteral nutrition and failures nutritional support, this study contributes to patient safety, one of the relevant dimensions of quality of care the health.

The Versatility of Perforator-Based Propeller Flap for Reconstruction of Distal Leg and Ankle Defects

Directory of Open Access Journals (Sweden)

Durga Karki

2012-01-01

Full Text Available Introduction. Soft tissue coverage of distal leg and ankle region represents a challenge and such defect usually requires a free flap. However, this may lead to considerable donor site morbidity, is time consuming, and needs facility of microsurgery. With the introduction of perforator flap, management of small- and medium-size defects of distal leg and ankle region is convenient, less time consuming, and with minimal donor site morbidity. When local perforator flap is designed as propeller and rotated to 180 degree, donor site is closed primarily and increases reach of flap, thus increasing versatility. Material and Methods. From June 2008 to May 2011, 20 patients were treated with perforator-based propeller flap for distal leg and ankle defects. Flap was based on single perforator of posterior tibial and peroneal artery rotated to 180 degrees. Defect size was from 4 cm × 3.5 cm to 7 cm × 5 cm. Results. One patient developed partial flap necrosis, which was managed with skin grafting. Two patients developed venous congestion, which subsided spontaneously without complications. Small wound dehiscence was present in one patient. Donor site was closed primarily in all patients. Rest of the flaps survived well with good aesthetic results. Conclusion. The perforator-based propeller flap for distal leg and ankle defects is a good option. This flap design is safe and reliable in achieving goals of reconstruction. The technique is convenient, less time consuming, and with minimal donor site morbidity. It provides aesthetically good result.

Interstitial water studies on small core samples, Deep Sea Drilling Project, Leg 5

Science.gov (United States)

Manheim, F. T.; Chan, K.M.; Sayles, F.L.

1970-01-01

Leg 5 samples fall into two categories with respect to interstitial water composition: 1) rapidly deposited terrigenous or appreciably terrigenous deposits, such as in Hole 35 (western Escanaba trough, off Cape Mendocino, California); and, 2) slowly deposited pelagic clays and biogenic muds and oozes. Interstitial waters in the former show modest to slight variations in chloride and sodium, but drastic changes in non-conservative ions such as magnesium and sulfate. The pelagic deposits show only relatively minor changes in both conservative and non-conservative pore fluid constituents. As was pointed out in earlier Leg Reports, it is believed that much of the variation in chloride in pore fluids within individual holes is attributable to the manipulation of samples on board ship and in the laboratory. On the other hand, the scatter in sodium is due in part to analytical error (on the order of 2 to 3 per cent, in terms of a standard deviation), and it probably accounts for most of the discrepancies in total anion and cation balance. All constituents reported here, with the exception of bulk water content, were analyzed on water samples which were sealed in plastic tubes aboard ship and were subsequently opened and divided into weighed aliquots in the laboratory. Analytical methods follow the atomic absorption, wet chemical and emission spectrochemical techniques briefly summarized in previous reports, e.g. Manheim et al., 1969, and Chan and Manheim, 1970. The authors acknowledge assistance from W. Sunda, D. Kerr, C. Lawson and H. Richards, and thank D. Spencer, P. Brewer and E. Degens for allowing the use of equipment and laboratory facilities.

Collisional drag may lead to disappearance of wave-breaking phenomenon of lower hybrid oscillations

International Nuclear Information System (INIS)

Maity, Chandan; Chakrabarti, Nikhil

2013-01-01

The inhomogeneity in the magnetic field in a cold electron-ion non-dissipative homogeneous plasma leads to the breaking of lower hybrid modes via phase mixing phenomenon [Maity et al. Phys. Plasmas 19, 102302 (2012)]. In this work, we show that an inclusion of collisional drag force in fluid equations may lead to the disappearance of the wave-breaking phenomenon of lower hybrid oscillations. The nonlinear analysis in Lagrangian variables provides an expression for a critical value of damping rate, above which spikes in the plasma density profile may disappear. The critical damping rate depends on the perturbation and magnetic field inhomogeneity amplitudes as well as the ratio of the magnetic field inhomogeneity and perturbation scale lengths.

Comet assay as a cold chain control tool

International Nuclear Information System (INIS)

Duarte, Renato Cesar

2009-01-01

Bearing in mind an ever more demanding market regarding the quality of food, it has been necessary to develop processes that meet the demands of consumers. Within the existing processes the cold chain and irradiation stand out. The cold chain comprises all the stages of conserving food from production, cooling, freezing, storing and transportation to the final consumer. Irradiation, as a means of conserving food, prolongs the shelf life, inhibits budding and reduces pathogenic contamination among other benefits. Is very important the identification of food degradation in function of failure on the processes which they were subjected. The comet assay is a screening test widely studied, considerate fast and with low cost. By the fact of the test identify breaks on the DNA, may be possible use the comet test on the control of cold chain failures that degrade de food. The labels and stamp, do not consider the previous food situation and indicate failures from the moment where they be placed in contact with the product. With the comet assay is possible to check the degradation that has occurred in liver chicken samples until the moment of comet's test realization. (author)

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