Depressurization as an accident management strategy to minimize the consequences of direct containment heating

Energy Technology Data Exchange (ETDEWEB)

Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P.; Dobbe, C.A. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

1990-10-01

Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of {approximately}922 K (1200{degree}F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs.

Depressurization study of CAREM 25 reactor considering the structures heat transfer

International Nuclear Information System (INIS)

Doval, A.

1990-01-01

This work presents the CAREM 25 reactor depressurization analysis results as an alternative form of accidents mitigation. Such results will help to determine design pressure valves for the emergency injection system as well as the depressurization valve diameter. Calculations were made with BLOW.MOD2 program. (Author) [es

Study on depressurization measurements and effect in PWR

International Nuclear Information System (INIS)

Ji Duan; Cao Xuewu

2006-01-01

Implementation of new regulations on nuclear powered plant design and operation raise new design and management requirement for plants, and the operational plants also need accident management to enhance the reactor operation safety. Thus, for sake of reducing risk of high-pressure and mitigating the consequence, depressurization is a measure carried out to reduce primary pressure. With SCDAP/RELAP5 this paper studies the depressurization measurements and effect factors in pressurized water reactor under the important severe accident sequences induced by very small break lost of coolant accident (VSBLOCA), anticipated transient without scram (ATWS) and station blackout (SBO) plus auxiliary feedwater failure. (author)

Experimental Investigation on the Behavior of Supercritical CO2 during Reservoir Depressurization.

Science.gov (United States)

Li, Rong; Jiang, Peixue; He, Di; Chen, Xue; Xu, Ruina

2017-08-01

CO 2 sequestration in saline aquifers is a promising way to address climate change. However, the pressure of the sequestration reservoir may decrease in practice, which induces CO 2 exsolution and expansion in the reservoir. In this study, we conducted a core-scale experimental investigation on the depressurization of CO 2 -containing sandstone using NMR equipment. Three different series of experiments were designed to investigate the influence of the depressurization rate and the initial CO2 states on the dynamics of different trapping mechanisms. The pressure range of the depressurization was from 10.5 to 4.0 MPa, which covered the supercritical and gaseous states of the CO 2 (named as CO 2 (sc) and CO 2 (g), respectively). It was found that when the aqueous phase saturated initially, the exsolution behavior strongly depended on the depressurization rate. When the CO 2 and aqueous phase coexisting initially, the expansion of the CO 2 (sc/g) contributed to the incremental CO 2 saturation in the core only when the CO 2 occurred as residually trapped. It indicates that the reservoir depressurization has the possibility to convert the solubility trapping to the residual trapping phase, and/or convert the residual trapping to mobile CO 2 .

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)

Characterization of liquid entrainment in the AP1000 automatic depressurization system from APEX tests

International Nuclear Information System (INIS)

Richard F Wright; Terry L Schulz; Jose N Reyes; John Groome

2005-01-01

Full text of publication follows: The AP1000 is a 1000 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 relies heavily on the 600 MWe AP600 which received design certification in 1999. A critical part of the AP600 design certification process involved the testing of the passive safety systems. A one-fourth height, one-fourth pressure test facility, APEX-600, was constructed at the Oregon State University to study design basis events, and to provide a body of data to be used to validate the computer models used to analyze the AP600. This facility was extensively modified to reflect the design changes for AP1000 including higher power in the electrically heated rods representing the reactor core, and changes in the size of the pressurizer, core makeup tanks and automatic depressurization system. The APEX-1000 test facility was used to perform design basis accident simulations and separate effects tests to support the AP1000 design certification process. In the event of a LOCA, the AP1000 passive core cooling system provides sources of core makeup water along with an automatic depressurization system (ADS) consisting of several stages of valves which reduce the reactor coolant system pressure in a controlled manner. The final stage of this system, ADS-4, consists of four large valves that open off the hot legs, reducing the pressure to allow gravity injection from the in-containment refueling water storage tank (IRWST) and eventually the containment sump. The 67% increase in power from AP600 to AP1000 results in proportionally larger steam velocities exiting the core. Higher steam velocities could increases the potential for significant liquid entrainment out the ADS-4 lines, affecting the liquid inventory in the reactor. Tests were performed in APEX-1000 to characterize the two

Mobile Test Capabilities

Data.gov (United States)

Federal Laboratory Consortium — The Electrical Power Mobile Test capabilities are utilized to conduct electrical power quality testing on aircraft and helicopters. This capability allows that the...

Study on primary coolant system depressurization effect factor in pressurized water reactor

International Nuclear Information System (INIS)

Ji Duan; Cao Xuewu

2006-01-01

The progression of high-pressure core melting severe accident induced by very small break loss of coolant accident plus the loss of main feed water and auxiliary feed water failure is studied, and the entry condition and modes of primary cooling system depressurization during the severe accident are also estimated. The results show that the temperature below 650 degree C is preferable depressurization input temperature allowing recovery of core cooling, and the available and effective way to depressurize reactor cooling system and to arrest very small break loss of coolant accident sequences is activating pressurizer relief valves initially, then restoring the auxiliary feedwater and opening the steam generator relief valves. It can adequately reduce the primary pressure and keep the capacity loop of long-term core cooling. (authors)

The effect of the rate of hydrostatic pressure depressurization on cells in culture.

Science.gov (United States)

Tworkoski, Ellen; Glucksberg, Matthew R; Johnson, Mark

2018-01-01

Changes in hydrostatic pressure, at levels as low as 10 mm Hg, have been reported in some studies to alter cell function in vitro; however, other studies have found no detectable changes using similar methodologies. We here investigate the hypothesis that the rate of depressurization, rather than elevated hydrostatic pressure itself, may be responsible for these reported changes. Hydrostatic pressure (100 mm Hg above atmospheric pressure) was applied to bovine aortic endothelial cells (BAECs) and PC12 neuronal cells using pressurized gas for periods ranging from 3 hours to 9 days, and then the system was either slowly (~30 minutes) or rapidly (~5 seconds) depressurized. Cell viability, apoptosis, proliferation, and F-actin distribution were then assayed. Our results did not show significant differences between rapidly and slowly depressurized cells that would explain differences previously reported in the literature. Moreover, we found no detectable effect of elevated hydrostatic pressure (with slow depressurization) on any measured variables. Our results do not confirm the findings of other groups that modest increases in hydrostatic pressure affect cell function, but we are not able to explain their findings.

Experimental and theoretical investigation on the depressurization of a vessel with internals

International Nuclear Information System (INIS)

Vigni, P.; Oriolo, F.; Rosa, U.

1978-01-01

This paper is about some blow-down experiments performed at the Scalbatraio Center of the University of Pisa. The blow-down tests have been made to investigate the depressurization of a vessel with internal structures, reproducing the geometry of a BWR. The experimental data have been compared with calculations performed by the RELAP program, in order to evaluate the scaling effects related to their application to large scale units. (author)

RELAP5 Prediction of Transient Tests in the RD-14 Test Facility

International Nuclear Information System (INIS)

Lee, Sukho; Kim, Manwoong; Kim, Hho-Jung; Lee, John C.

2005-01-01

Although the RELAP5 computer code has been developed for best-estimate transient simulation of a pressurized water reactor and its associated systems, it could not assess the thermal-hydraulic behavior of a Canada deuterium uranium (CANDU) reactor adequately. However, some studies have been initiated to explore the applicability for simulating a large-break loss-of-coolant accident in CANDU reactors. In the present study, the small-reactor inlet header break test and the steam generator secondary-side depressurization test conducted in the RD-14 test facility were simulated with the RELAP5/MOD3.2.2 code to examine its extended capability for all the postulated transients and accidents in CANDU reactors. The results were compared with experimental data and those of the CATHENA code performed by Atomic Energy of Canada Limited.In the RELAP5 analyses, the heated sections in the facility were simulated as a multichannel with five pipe models, which have identical flow areas and hydraulic elevations, as well as a single-pipe model.The results of the small-reactor inlet header break and the steam generator secondary-side depressurization simulations predicted experimental data reasonably well. However, some discrepancies in the depressurization of the primary heat transport system after the header break and consequent time delay of the major phenomena were observed in the simulation of the small-reactor inlet header break test

Application case study of AP1000 automatic depressurization system (ADS) for reliability evaluation by GO-FLOW methodology

Energy Technology Data Exchange (ETDEWEB)

Hashim, Muhammad, E-mail: hashimsajid@yahoo.com; Hidekazu, Yoshikawa, E-mail: yosikawa@kib.biglobe.ne.jp; Takeshi, Matsuoka, E-mail: mats@cc.utsunomiya-u.ac.jp; Ming, Yang, E-mail: myang.heu@gmail.com

2014-10-15

Highlights: • Discussion on reasons why AP1000 equipped with ADS system comparatively to PWR. • Clarification of full and partial depressurization of reactor coolant system by ADS system. • Application case study of four stages ADS system for reliability evaluation in LBLOCA. • GO-FLOW tool is capable to evaluate dynamic reliability of passive safety systems. • Calculated ADS reliability result significantly increased dynamic reliability of PXS. - Abstract: AP1000 nuclear power plant (NPP) utilized passive means for the safety systems to ensure its safety in events of transient or severe accidents. One of the unique safety systems of AP1000 to be compared with conventional PWR is the “four stages Automatic Depressurization System (ADS)”, and ADS system originally works as an active safety system. In the present study, authors first discussed the reasons of why four stages ADS system is added in AP1000 plant to be compared with conventional PWR in the aspect of reliability. And then explained the full and partial depressurization of RCS system by four stages ADS in events of transient and loss of coolant accidents (LOCAs). Lastly, the application case study of four stages ADS system of AP1000 has been conducted in the aspect of reliability evaluation of ADS system under postulated conditions of full RCS depressurization during large break loss of a coolant accident (LBLOCA) in one of the RCS cold legs. In this case study, the reliability evaluation is made by GO-FLOW methodology to determinate the influence of ADS system in dynamic reliability of passive core cooling system (PXS) of AP1000, i.e. what will happen if ADS system fails or successfully actuate. The GO-FLOW is success-oriented reliability analysis tool and is capable to evaluating the systems reliability/unavailability alternatively to Fault Tree Analysis (FTA) and Event Tree Analysis (ETA) tools. Under these specific conditions of LBLOCA, the GO-FLOW calculated reliability results indicated

Rapid depressurization of a compressible fluid

International Nuclear Information System (INIS)

Dang, M.; Dupont, J.F.; Weber, H.

1978-08-01

The rapid depressurization of a plenum is a situation frequently encountered in the dynamical analysis of nuclear gas cycles of the HHT type. Various methods of numerical analyses for a 1-dimensional flow model are examined: finite difference method; control volume method; method of characteristics. Based on the shallow water analogy to compressible flow, the numerical results are compared with those from a water table set up to simulate a standard problem. (Auth.)

Depressurization as a means of leak checking large vacuum vessels

International Nuclear Information System (INIS)

Callis, R.W.; Langhorn, A.; Petersen, P.I.; Ward, C.; Wesley, J.

1985-01-01

A common problem associated with large vacuum vessels used in magnetic confinement fusion experiments is that leak checking is hampered by the inaccessibility to most of the vacuum vessel surface. This inaccessibility is caused by the close proximity of magnetic coils, diagnostics and, for those vessels that are baked, the need to completely surround the vessel with a thermal insulation blanket. These obstructions reduce the effectiveness of the standard leak checking method of using a mass spectrometer and spraying a search gas such as helium on the vessel exterior. Even when the presence of helium is detected, its entry point into the vessel cannot always be pinpointed. This paper will describe a method of overcoming this problem. By slightly depressurizing the vessel, an influx of helium through the leak is created. The leak site can then be identified by personnel within the vessel using standard sniffing procedures. There are two conditions which make this method of leak checking practical. First, the vessel need only be depressurized 2 psi, thus allowing personnel inside to perform the sniffing operation. Second, the sniffing probe used (Leybold--Heraus ''Quick Test'') could detect a change in helium concentration as small as 100 ppb, which allows for faster scanning of the vessel inferior. Use of this technique to find an elusive 10 -3 Torrxl/s leak in the Doublet III tokamak vacuum vessel will be presented

ACTIVE SOIL DEPRESSURIZATION (ASD) DEMONSTRATION IN A LARGE BUILDING

Science.gov (United States)

The report gives results of an evaluation of the feasibility of implementing radon resistant construction techniques -- especially active soil depressurization (ASD) -- in new large buildings in Florida. Indoor radon concentrations and radon entry were monitored in a finished bui...

Maximum Recoverable Gas from Hydrate Bearing Sediments by Depressurization

KAUST Repository

Terzariol, Marco

2017-11-13

The estimation of gas production rates from hydrate bearing sediments requires complex numerical simulations. This manuscript presents a set of simple and robust analytical solutions to estimate the maximum depressurization-driven recoverable gas. These limiting-equilibrium solutions are established when the dissociation front reaches steady state conditions and ceases to expand further. Analytical solutions show the relevance of (1) relative permeabilities between the hydrate free sediment, the hydrate bearing sediment, and the aquitard layers, and (2) the extent of depressurization in terms of the fluid pressures at the well, at the phase boundary, and in the far field. Close form solutions for the size of the produced zone allow for expeditious financial analyses; results highlight the need for innovative production strategies in order to make hydrate accumulations an economically-viable energy resource. Horizontal directional drilling and multi-wellpoint seafloor dewatering installations may lead to advantageous production strategies in shallow seafloor reservoirs.

Vacuum horizontal drainage for depressurization of uranium tailings

International Nuclear Information System (INIS)

Pakalnis, R.; Chedsey, G.; Robertson, A.M.; Follin, S.

1985-01-01

A recent advance in tailings slope depressurization is the application of vacuum assist horizontal drainage. Horizontal drains have been used for several decades to reduce water pressures in slopes in order to improve stability. The benefit from vacuum assist arises from an increased hydraulic gradient caused by induced negative atmospheric pressures. The vacuum assist system has, since its inception in 1982, been successfully employed at two soil and four rock slope projects located in Western Canada. This paper describes the first application of this system in the United States. The technical feasibility of employing vacuum assisted horizontal drains to depressurize a uranium tailings dam near Riverton, Wyoming has been evaluated. Two horizontal drains (300 ft.) were installed and their effect monitored by nine piezometers. The study was conducted over a three-week internal with vacuum being applied for three and four day periods. The drawdowns achieved through vacuum drainage was found to be approximately double that obtained by gravity alone. The volume of water exhausted under vacuum during the seven day interval was approximately double that obtained by gravity alone

Experimental investigation of iodine removal and containment depressurization in containment spray system test facility of 700 MWe Indian pressurized heavy water reactors

Energy Technology Data Exchange (ETDEWEB)

Jain, Manish [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Kandar, T.K.; Vhora, S.F.; Mohan, Nalini [Directorate of Technology Development, Nuclear Power Corporation of India Limited, Mumbai (India); Iyer, K.N. [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India)

2017-05-15

Highlights: • Depressurization rate in a scaled down vessel filled with air and steam is studied. • Iodine removal rate in a scaled down vessel filled with steam/air is investigated. • Effect of SMD and vessel pressure on depressurization rate is studied. • Depressurization rate decreases with the increase in the droplet size (590 μm – 1 mm) • Decrease in pressure and iodine concentration with time follow exponential trend. - Abstract: As an additional safety measure in the new 700 MWe Indian pressurized heavy water reactors, the first of a kind system called containment Spray System is introduced. The system is designed to cater/mitigate the conditions after design basis accidents i.e., loss of coolant accident and main steam line break. As a contribution to the safety analysis of condition following loss-of-coolant accidents, experiments are carried out to establish the performance of the system. The loss of coolant is simulated by injecting saturated steam and iodine vapors into the containment vessel in which air is enclosed at atmospheric and room temperature, and then the steam-air mixture is cooled by sprays of water. The effect of water spray on the containment vessel pressure and the iodine scrubbing in a scaled down facility is investigated for the containment spray system of Indian pressurized heavy water reactors. The experiments are carried out in the scaled down vessel of the diameter of 2.0 m and height of 3.5 m respectively. Experiments are conducted with water at room temperature as the spray medium. Two different initial vessel pressure i.e. 0.7 bar and 1.0 bar are chosen for the studies as they are nearing the loss of coolant accident & main steam line break pressures in Indian pressurized heavy water reactors. These pressures are chosen based on the containment resultant pressures after a design basis accident. The transient temperature and pressure distribution of the steam in the vessel are measured during the depressurization

Gest-sip1 experiments and post-test calculations with the relap5 code

International Nuclear Information System (INIS)

Achilli, A.; Cattadori, G.; Ferri, R.; Gandolfi, S.; Bianchi, F.; Meloni, P.

2001-01-01

The SIP-1 apparatus (Sistema di Iniezione Passiva) was conceived, designed, numerically simulated and tested by the SIET company as an innovative depressurization and make-up device for the New Generation LWRs. In particular it is suitable to cope with those accidents where pressure in the circuit must be dumped to allow low pressure injection systems to intervene. The main peculiarity of SIP-1 is the capability of de-pressurizing a system by cold water injection, rather than by discharging mass to the outlet, as in the common depressurization systems. ENEA sponsored all the research activity, starting from the SIP-1 design, its numerical simulation with the Relap5 code, the realisation of an experimental facility up to the test execution and post-test calculations. An experimental campaign on the GEST-SIP1 facility was performed in July 2000. The facility is mainly constituted by a U-tube Steam Generator which a proper model of SIP-1 apparatus is connected to. A series of Small Break LOCAs was simulated by varying the break size and different steady conditions were investigated to verify the stability of SIP-1, the lack of unexpected interventions and the actuation modalities. This paper deals with the description of the GEST-SIP1 experimental facility, the SIP-1 operating principles, the most meaningful results of the tests and the capability of the Relap5 code in reproducing phenomena and events. (author)

Analytical method and result of radiation exposure for depressurization accident of HTTR

International Nuclear Information System (INIS)

Sawa, K.; Shiozawa, S.; Mikami, H.

1990-01-01

The Japan Atomic Energy Research Institute (JAERI) is now proceeding with the construction design of the High Temperature Engineering Test Reactor (HTTR). Since the HTTR has some characteristics different from LWRs, analytical method of radiation exposure in accidents provided for LWRs can not be applied directly. This paper describes the analytical method of radiation exposure developed by JAERI for the depressurization accident, which is the severest accident in respect to radiation exposure among the design basis accidents of the HTTR. The result is also described in this paper

Perspectives on Severe Accident Management by Depressurization and External Water Injection under Extended SBO Conditions

International Nuclear Information System (INIS)

Seol, Wookcheol; Park, Jongwoon

2014-01-01

Three major issues of severe accident management guideline (SAMG) after this sort of extended SBO would be depressurization of the primary system, external water injection and hydrogen management inside a containment. Under this situation, typical SAM actions would be depressurization and external water delivery into the core. However, limited amount of external water would necessitate optimization between core cooling, containment integrity and fission product removal. In this paper, effects of SAM actions such as depressurization and external water injection on the reactor and containment conditions after extended SBO are analyzed using MAAP4 code. Positive and negative aspects are discussed with respect to core cooling and fission product retention inside a primary system. Conclusions are made as following: Firstly, early depressurization action itself has two-faces: positive with respect to delay of the reactor vessel failure but negative with respect to the containment failure and fission product retention inside the primary system. Secondly, in order to prevent containment overpressure failure after external water injection, re-closing of PORV later should be considered in SAM, which has never been considered in the previous SAMG. Finally, in case of external water injection, the flow rate should be optimized considering not only the cooling effect but also the long term fission product retention inside the primary system

Analysis of design and operational effects of filtered containment venting on depressurization and fission product release

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Woon; Seol, Wook-Cheol; Kim, Jisu [Dongguk Univ., Gyeongbuk (Korea, Republic of)

2017-03-15

Effects of design and operational parameters of filtered containment venting system during a specified containment depressurization and relative aero sol release amount are analyzed. The analyses is performed by using the MAAP4 code for the APR1400 reactor. Major results uniquely identified from the analyses can be noted as following: Even though containment depressurization is accelerated as the pipe size increases, the venting system solution is also depleted earlier. Elapsed times to reach lower end pressure of 2 bar are nearly identical regardless of the vent initiation pressure and thus early venting is not much beneficial than late venting. Stroke time of the isolation valves has no effect on the depressurization performance and thus slow opening is beneficial for load reduction from the vent effluent.

Automatic depressurization system of BWR type nuclear power plant

International Nuclear Information System (INIS)

Fujii, Masahiko.

1993-01-01

In the present invention, depressurization is conducted while keeping versatility and retardancy of a water injection system so that safety is improved. That is, a means that judges whether a turbine driving water injection system is operated or not by the following conditions. (1) a discharging pressure of the turbine driving pump is greater than a set value, (2) a flow rate of the turbine driving water injection system is greater than a set value, (3) an injection valve of the turbine driving water injection system into a reactor is opened, or combination of (1) to (3). With such procedures, when an automatic depressurization system is necessary during operation of the turbine driving water injection system, reactor pressure is decreased till a low pressure water injection system is operated, but pressure is not decreased to such a level that the turbine driving water injection system is isolated. Therefore, versatility and retardancy of the water injection system are ensured. As a result, reliability of a reactor cooling means is improved. (I.S.)

Mechanical Properties of Porous Titanium Structure Fabricated by Investment Casting with Pressurization/Depressurization System

International Nuclear Information System (INIS)

Kang, San; Lee, Ji-Woon; Hyun, Soong-Keun; Lee, Byong-Pil; Kim, Myoung-Gyun; Kim, Young-Jig

2014-01-01

A porous titanium structure was fabricated by investment casting with a pressurization/depressurization system, and its mechanical properties were studied. A Micro-Vickers hardness profile revealed that hardness gradually increased from the matrix to the metal/mold interface. A compression test was conducted on a single cell of the porous Ti structure. The theoretical and experimental values of yield strength were in good agreement. Such agreement suggested that the reaction layer did not affect the macro-mechanical properties of the porous Ti structure.

A study on design enhancement of automatic depressurization system in a passive PWR

International Nuclear Information System (INIS)

Yu, Sung Sik

1993-02-01

In a Passive PWR, the successful actuation of the Automatic Depressurization System is essentially required so that no core damage is occurred following small LOCA. But it has been shown in the previous studies that Core Damage Frequency form small LOCA is significantly caused by unavailability of the ADS. In this study, the design vulnerabilities impacting the ADS unavailability are identified through the reliability assessment using the fault tree methodology and then the design enhancements towards improving the system reliability are developed. A series of small LOCA analyses using RELAP5 code are performed to validate the system requirements for the successful depressurization and to study the thermal-hydraulic feasibility of the proposed design enhancements. The impact on CDF according to the change of system unavailability is also analyzed. In addition, aqualitative analysis is performed to reduce the inadvertent opening of the ADS valves. From the results of the analyses, the ADS is understood to have less incentive on the reliability improvement through system simplification. It is found that based on system characteristics, the major contributor to the system unavailability is the first stage. A series-parallel configuration with two trains of eight valves, which shows a higher reliability compared to the base ADS design, is recommended as an alternative first stage of the ADS. In addition, establishment of the appropriate ADS operation strategy is proposed such as allowing manual operation of the first stage and allowing the forced depressurization using the normal residual heat removal system connected to the RCS following the successful depressurization up to the 3rd stage and the failure of the 4th stage

A depressurization assistance control based on the posture of a seated patient on a wheelchair.

Science.gov (United States)

Chugo, Daisuke; Fujita, Kazuya; Sakaida, Yuki; Yokota, Sho; Takase, Kunikatsu

2011-01-01

For reducing the risk of pressure sore caused by long period sitting on a wheelchair, we develop a depressurization motion assistance system which is low cost and suitable for practical use. Our developing system consists of a seating cushion which the patient sits on and four air cells which can lift or incline the seating cushion. Each air cell is actuated by small air compressor, which can drive using batteries on the wheelchair respectively, and each compressor has a pressure sensor on its body. In this paper, our key ideas are two topics. One topic is mechanical design for practical use. We realize thin mechanism which enables easy implementation to the general wheelchair. For realizing this thinly design, we develop the tilt mechanism using elasticity of acrylic resin and the controller which uses only pressure sensors for estimating its lifting height and inclination. The other topic is assistance control scheme based on the patient's depressurization operation for increasing a rehabilitation performance. For realizing the proposed control scheme, we analyze the hip depressurization operation by the nursing specialists and use its results for estimating the patient's condition. Using our system, the patient can depressurize by his own will on the general wheelchair easily. The performance of our system is verified by experiments using our prototype. © 2011 IEEE

Ensuring US National Aeronautics Test Capabilities

Science.gov (United States)

Marshall, Timothy J.

2010-01-01

U.S. leadership in aeronautics depends on ready access to technologically advanced, efficient, and affordable aeronautics test capabilities. These systems include major wind tunnels and propulsion test facilities and flight test capabilities. The federal government owns the majority of the major aeronautics test capabilities in the United States, primarily through the National Aeronautics and Space Administration (NASA) and the Department of Defense (DoD). However, changes in the Aerospace landscape, primarily the decrease in demand for testing over the last 20 years required an overarching strategy for management of these national assets. Therefore, NASA established the Aeronautics Test Program (ATP) as a two-pronged strategic initiative to: (1) retain and invest in NASA aeronautics test capabilities considered strategically important to the agency and the nation, and (2) establish a strong, high level partnership with the DoD. Test facility utilization is a critical factor for ATP because it relies on user occupancy fees to recover a substantial part of the operations costs for its facilities. Decreasing utilization is an indicator of excess capacity and in some cases low-risk redundancy (i.e., several facilities with basically the same capability and overall low utilization). However, low utilization does not necessarily translate to lack of strategic importance. Some facilities with relatively low utilization are nonetheless vitally important because of the unique nature of the capability and the foreseeable aeronautics testing needs. Unfortunately, since its inception, the customer base for ATP has continued to shrink. Utilization of ATP wind tunnels has declined by more than 50% from the FY 2006 levels. This significant decrease in customer usage is attributable to several factors, including the overall decline in new programs and projects in the aerospace sector; the impact of computational fluid dynamics (CFD) on the design, development, and research

DESIGN AND TESTING OF SUB-SLAB DEPRESSURIZATION FOR RADON MITIGATION IN NORTH FLORIDA HOUSES - PART I. PERFORMANCE AND DURABILITY - VOLUME 1. TECHNICAL REPORT

Science.gov (United States)

The report gives results of a demonstration/research project to evaluate sub-slab depressurization (SSD) techniques for radon mitigation in North Florida where the housing stock is primarily slab-on-grade and the sub-slab medium typically consists of native soil and sand. Objecti...

DESIGN AND TESTING OF SUB-SLAB DEPRESSURIZATION FOR RADON MITIGATION IN NORTH FLORIDA HOUSES - PART I. PERFORMANCE AND DURABILITY - VOLUME 2. DATA APPENDICES

Science.gov (United States)

The report gives results of a demonstration/research project to evaluate sub-slab depressurization (SSD) techniques for radon mitigation in North Florida where the housing stock is primarily slab-on-grade and the sub-slab medium typically consists of native soil and sand. Objecti...

Sandia Laboratories technical capabilities: testing

International Nuclear Information System (INIS)

Lundergan, C.D.

1975-12-01

The testing capabilities at Sandia Laboratories are characterized. Selected applications of these capabilities are presented to illustrate the extent to which they can be applied in research and development programs

Effect of Acute Intermittent CPAP Depressurization during Sleep in Obese Patients.

Science.gov (United States)

Jun, Jonathan C; Unnikrishnan, Dileep; Schneider, Hartmut; Kirkness, Jason; Schwartz, Alan R; Smith, Philip L; Polotsky, Vsevolod Y

2016-01-01

Obstructive Sleep Apnea (OSA) describes intermittent collapse of the airway during sleep, for which continuous positive airway pressure (CPAP) is often prescribed for treatment. Prior studies suggest that discontinuation of CPAP leads to a gradual, rather than immediate return of baseline severity of OSA. The objective of this study was to determine the extent of OSA recurrence during short intervals of CPAP depressurization during sleep. Nine obese (BMI = 40.4 ± 3.5) subjects with severe OSA (AHI = 88.9 ± 6.8) adherent to CPAP were studied during one night in the sleep laboratory. Nasal CPAP was delivered at therapeutic (11.1 ± 0.6 cm H20) or atmospheric pressure, in alternating fashion for 1-hour periods during the night. We compared sleep architecture and metrics of OSA during CPAP-on and CPAP-off periods. 8/9 subjects tolerated CPAP withdrawal. The average AHI during CPAP-on and CPAP-off periods was 3.6 ± 0.6 and 15.8 ± 3.6 respectively (p<0.05). The average 3% ODI during CPAP-on and CPAP-off was 4.7 ± 2 and 20.4 ± 4.7 respectively (p<0.05). CPAP depressurization also induced more awake (p<0.05) and stage N1 (p<0.01) sleep, and less stage REM (p<0.05) with a trend towards decreased stage N3 (p = 0.064). Acute intermittent depressurization of CPAP during sleep led to deterioration of sleep architecture but only partial re-emergence of OSA. These observations suggest carryover effects of CPAP.

Effect of Acute Intermittent CPAP Depressurization during Sleep in Obese Patients.

Directory of Open Access Journals (Sweden)

Jonathan C Jun

Full Text Available Obstructive Sleep Apnea (OSA describes intermittent collapse of the airway during sleep, for which continuous positive airway pressure (CPAP is often prescribed for treatment. Prior studies suggest that discontinuation of CPAP leads to a gradual, rather than immediate return of baseline severity of OSA. The objective of this study was to determine the extent of OSA recurrence during short intervals of CPAP depressurization during sleep.Nine obese (BMI = 40.4 ± 3.5 subjects with severe OSA (AHI = 88.9 ± 6.8 adherent to CPAP were studied during one night in the sleep laboratory. Nasal CPAP was delivered at therapeutic (11.1 ± 0.6 cm H20 or atmospheric pressure, in alternating fashion for 1-hour periods during the night. We compared sleep architecture and metrics of OSA during CPAP-on and CPAP-off periods.8/9 subjects tolerated CPAP withdrawal. The average AHI during CPAP-on and CPAP-off periods was 3.6 ± 0.6 and 15.8 ± 3.6 respectively (p<0.05. The average 3% ODI during CPAP-on and CPAP-off was 4.7 ± 2 and 20.4 ± 4.7 respectively (p<0.05. CPAP depressurization also induced more awake (p<0.05 and stage N1 (p<0.01 sleep, and less stage REM (p<0.05 with a trend towards decreased stage N3 (p = 0.064.Acute intermittent depressurization of CPAP during sleep led to deterioration of sleep architecture but only partial re-emergence of OSA. These observations suggest carryover effects of CPAP.

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

Data report of ROSA/LSTF experiment SB-CL-32. 1% cold leg break LOCA with SG depressurization and no gas inflow

International Nuclear Information System (INIS)

Takeda, Takeshi

2014-11-01

An experiment SB-CL-32 was conducted on May 28, 1996 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-CL-32 simulated a 1% cold leg small-break loss-of-coolant accident in a pressurized water reactor under assumptions of total failure of high pressure injection system and no inflow of non-condensable gas from accumulator (ACC) tanks of emergency core cooling system. Secondary-side depressurization of both steam generators (SGs) as an accident management (AM) action to achieve the depressurization rate of 200 K/h in the primary system was initiated 10 min after the break. After the initiation of AM action, auxiliary feedwater injection into the SG secondary-side was started with some delay. After the onset of AM action, the primary pressure decreased following the SG secondary-side pressure. Core uncovery by core boil-off started with liquid level drop in crossover leg downflow-side. The core liquid level recovered rapidly after first loop seal clearing (LSC). The surface temperature of simulated fuel rod then increased up to 669 K. Core uncovery by core boil-off took place before second LSC induced by steam condensation on ACC coolant injected into cold legs following the primary depressurization. The core liquid level recovered rapidly after the second LSC. The observed maximum fuel rod surface temperature was 772 K. The experiment was terminated when the continuous core cooling was confirmed because of the coolant injection by low pressure injection system after the isolation of ACC system. The obtained data would be useful to define the conditions for counterpart testing of other integral test facilities to address scaling problems through thermal-hydraulic phenomena. This report summarizes the test procedures, conditions and major observation in the ROSA/LSTF experiment SB-CL-32. (author)

Numerical modeling of the waves evolution generated by the depressurization of the vessels containing a supercritical parameters coolant

Science.gov (United States)

Alekseev, Maksim V.; Vozhakov, Ivan S.; Lezhnin, Sergey I.; Pribaturin, Nikolay A.

2017-10-01

The development of power plants focuses on increasing the parameters of water coolants up to a supercritical level. Depressurization of the unit circuits with such a coolant leads to emergency situations. Their scenarios can change significantly with the variation of initial pressure and temperature before the start of depressurization. When the pressure drops from the supercritical single-phase region of the initial thermodynamic parameters of the coolant, either the liquid boils up, or the vapor is condensed. Because of the rapid pressure decrease, the phase transition can be non-equilibrium that must be taken into account in the simulation. In the present study, an axisymmetric problem of the outflow of a water coolant from the pipe butt-end is considered. The equations of continuity, momentum and energy for a two-phase homogeneous mixture are solved numerically. The vapor and liquid properties are calculated using the TTSE software package (The Tabular Taylor Series Expansion Method). On the basis of the computer complex LCPFCT (The Flux-Corrected Transport Algorithm) the program code was developed for solving numerous problems on the depressurization of vessels or pipelines, containing superheated water or gas under high pressure. Different variants of outflow in the external model atmosphere and generation of waves are analyzed. The calculated data on the interaction of pressure waves with a barrier are calculated. To describe phase transitions, an asymptotic relaxation model of nonequilibrium evaporation and condensation has been created and tested.

Evaluation of a coolant injection into the in-vessel with a RCS depressurization by using SCDAP/RELAP5

International Nuclear Information System (INIS)

Rae-Joon, Park; Sang-Baik, Kim; Hee-Dong, Kim

2007-01-01

As part of the evaluations of a severe accident management strategy, a coolant injection in the vessel with a reactor coolant system (RCS) depressurization has been evaluated by using the SCDAP/RELAP5 computer code. Two high pressure sequences of a small break loss of coolant accident (LOCA) without safety injection (SI) and a total loss of feed water (LOFW) accident have been analyzed in optimized power reactor OPR-1000. The SCDAP/RELAP5 results have shown that only one train operation of a high pressure safety injection at 30,000 seconds with a RCS depressurization by using one condenser dump valve at 6 minutes after an entrance of the severe accident management guidance prevents a reactor vessel failure for the small break LOCA without SI. In this case, only train operation of the low pressure safety injection (LPSI) without the high pressure safety injection (HPSI) does not prevent a reactor vessel failure. Only one train operation of the HPSI at 20,208 seconds with a RCS depressurization by using two safety depressurization system valves at 40 minutes after an initial opening of the safety relief valve prevents a reactor vessel failure for the total LOFW. (authors)

Depressurization accidents in a medium-sized high-temperature gas reactor

International Nuclear Information System (INIS)

Ron, S.; Tzoref, J.; Gal, D.

1992-01-01

The amount of fission product release during a core heatup accident in a medium-sized high-temperature gas reactor depends on the size of the inadvertent opening in the primary circuit; this dependence is assessed. The opening triggers a depressurization event that is assumed to be coupled with the failure of the forced circulation in both decay-heat removal systems. The scenario investigated is a beyond-design-base accident. The DSNP modular simulation code is used. This paper reports that a two-dimensional model is developed to simulate the HTR-500 design. The study shows that the depressurization process does not contribute significantly to the sweeping out (from the primary circuit) of fission products released from the fuel during the core heatup. There is also no significant variation in the results when the opening size is >33 cm 2 , and only a slight sensitivity is found when the rupture size is between 3.3 and 33 cm 2 . The fission product release decreases considerably in the range from 1 to 3.3 cm 2 . The small-sized rupture is of major significance, as the failure of the relief valves to reclose increases the frequency of the event

Experimental study on secondary depressurization action for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V/LSTF test SB-PV-03)

International Nuclear Information System (INIS)

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

2005-06-01

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 is important in case of high pressure injection (HPI) system failure during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-03, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. Total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and operator AM actions on steam generator (SG) secondary depressurization at a rate of -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes were assumed as experiment conditions. It is clarified that the AM actions are effective on primary system depressurization until the end of AIS injection at 1.6 MPa, but thereafter become less effective due to inflow of the non-condensable gas, resulting in delay of low pressure injection (LPI) actuation and whole core heatup under continuous water discharge through the bottom break. The report describes these thermohydraulic phenomena related with transient primary coolant mass and AM actions in addition to estimation of non-condensable gas behavior which affected primary-to-secondary heat transfer. (author)

Study on the experimental VHTR safety with analysis for a hypothetical rapid depressurization accident

International Nuclear Information System (INIS)

Mitake, S.; Suzuki, K.; Ohno, T.; Okada, T.

1982-01-01

A hypothetical rapid depressurization accident of the experimental VHTR has been analyzed, including all phenomena in the accident, from its initiating depressurization of the coolant to consequential radiological hazard. Based on reliability analysis of the engineered safety features, all possible sequences, in which the safety systems are in success or in failure, have been investigated with event tree analysis. The result shows the inherent safety characteristics of the reactor and the effectiveness of the engineered safety features. And through the analysis, it has been indicated that further investigations on some phenomena in the accident, e.g., air ingress by natural circulation flow and fission product transport in the plant, will bring forth more reasonable and sufficient safety of the reactor

Geomechanical response of permafrost-associated hydrate deposits to depressurization-induced gas production

Science.gov (United States)

Rutqvist, J.; Moridis, G.J.; Grover, T.; Collett, T.

2009-01-01

In this simulation study, we analyzed the geomechanical response during depressurization production from two known hydrate-bearing permafrost deposits: the Mallik (Northwest Territories, Canada) deposit and Mount Elbert (Alaska, USA) deposit. Gas was produced from these deposits at constant pressure using horizontal wells placed at the top of a hydrate layer (HL), located at a depth of about 900??m at the Mallik site and 600??m at the Mount Elbert site. The simulation results show that general thermodynamic and geomechanical responses are similar for the two sites, but with substantially higher production and more intensive geomechanical responses at the deeper Mallik deposit. The depressurization-induced dissociation begins at the well bore and then spreads laterally, mainly along the top of the HL. The depressurization results in an increased shear stress within the body of the receding hydrate and causes a vertical compaction of the reservoir. However, its effects are partially mitigated by the relatively stiff permafrost overburden, and compaction of the HL is limited to less than 0.4%. The increased shear stress may lead to shear failure in the hydrate-free zone bounded by the HL overburden and the downward-receding upper dissociation interface. This zone undergoes complete hydrate dissociation, and the cohesive strength of the sediment is low. We determined that the likelihood of shear failure depends on the initial stress state as well as on the geomechanical properties of the reservoir. The Poisson's ratio of the hydrate-bearing formation is a particularly important parameter that determines whether the evolution of the reservoir stresses will increase or decrease the likelihood of shear failure.

Rapid depressurization event analysis in BWR/6 using RELAP5 and contain

Energy Technology Data Exchange (ETDEWEB)

Mueftueoglu, A.K.; Feltus, M.A. [Pennsylvania State Univ., University Park, PA (United States)

1995-09-01

Noncondensable gases may become dissolved in Boiling Water Reactor (BWR) water level instrumentation during normal operations. Any dissolved noncondensable gases inside these water columns may come out of solution during rapid depressurization events, and displace water from the reference leg piping resulting in a false high level. These water level errors may cause a delay or failure in actuation, or premature shutdown of the Emergency Core Cooling System. (ECCS). If a rapid depressurization causes an erroneously high water level, preventing automatic ECCS actuation, it becomes important to determine if there would be other adequate indications for operator response and other signals for automatic actuation such as high drywell pressure. It is also important to determine the effect of the level signal on ECCS operation after it is being actuated. The objective of this study is to determine the detailed coupled containment/NSSS response during this rapid depressurization events in BWR/6. The selected scenarios involve: (a) inadvertent opening of all ADS valves, (b) design basis (DB) large break loss of coolant accident (LOCA), and (c) main steam line break (MSLB). The transient behaviors are evaluated in terms of: (a) vessel pressure and collapsed water level response, (b) specific transient boundary conditions, (e.g., scram, MSIV closure timing, feedwater flow, and break blowdown rates), (c) ECCS initiation timing, (d) impact of operator actions, (e) whether indications besides low-low water level were available. The results of the analysis had shown that there would be signals to actuate ECCS other than low reactor level, such as high drywell pressure, low vessel pressure, high suppression pool temperature, and that the plant operators would have significant indications to actuate ECCS.

Fusion Materials Irradiation Test Facility: experimental capabilities and test matrix

International Nuclear Information System (INIS)

Opperman, E.K.

1982-01-01

This report describes the experimental capabilities of the Fusion Materials Irradiation Test Facility (FMIT) and reference material specimen test matrices. The description of the experimental capabilities and the test matrices has been updated to match the current single test cell facility ad assessed experimenter needs. Sufficient detail has been provided so that the user can plan irradiation experiments and conceptual hardware. The types of experiments, irradiation environment and support services that will be available in FMIT are discussed

Package testing capabilities at the Pacific Northwest Laboratory

International Nuclear Information System (INIS)

Taylor, J.M.

1993-01-01

The purpose of this paper is to describe the package testing capabilities at the Pacific Northwest Laboratory (PNL). In the past all of the package testing that was performed at PNL was done on prototype or mocked up radioactive material packaging. Presently, we are developing the capability to perform testing on non-radioactive material packaging. The testing on the non-radioactive material packaging will be done to satisfy the new performance oriented packaging requirements (DOT Docket HM-181, 1991). This paper describes the equipment used to perform the performance oriented packaging tests and also describes some testing capability for testing radioactive material packaging

Experimental Investigation Evaporation of Liquid Mixture Droplets during Depressurization into Air Stream

Science.gov (United States)

Liu, L.; Bi, Q. C.; Terekhov, Victor I.; Shishkin, Nikolay E.

2010-03-01

The objective of this study is to develop experimental method to study the evaporation process of liquid mixture droplets during depressurization and into air stream. During the experiment, a droplet was suspended on a thermocouple; an infrared thermal imager was used to measure the droplet surface temperature transition. Saltwater droplets were used to investigate the evaporation process during depressurization, and volatile liquid mixtures of ethanol, methanol and acetone in water were applied to experimentally research the evaporation into air stream. According to the results, the composition and concentration has a complex influence on the evaporation rate and the temperature transition. With an increase in the share of more volatile component, the evaporation rate increases. While, a higher salt concentration in water results in a lower evaporation rate. The shape variation of saltwater droplet also depends on the mass concentration in solution, whether it is higher or lower than the eutectic point (22.4%). The results provide important insight into the complex heat and mass transfer of liquid mixture during evaporation.

RIA testing capability of the transient reactor test facility

International Nuclear Information System (INIS)

Crawford, D.C.; Swanson, R.W.

1999-01-01

The advent of high-burnup fuel implementation in LWRs has generated international interest in high-burnup LWR fuel performance. Recent testing under simulated RIA conditions has demonstrated that certain fuel designs fail at peak fuel enthalpy values that are below existing regulatory criteria. Because many of these tests were performed with non-prototypically aggressive test conditions (i.e., with power pulse widths less than 10 msec FWHM and with non-protoypic coolant configurations), the results (although very informative) do not indisputably identify failure thresholds and fuel behavior. The capability of the TREAT facility to perform simulated RIA tests with prototypic test conditions is currently being evaluated by ANL personnel. TREAT was designed to accommodate test loops and vehicles installed for in-pile transient testing. During 40 years of TREAT operation and fuel testing and evaluation, experimenters have been able to demonstrate and determine the transient behavior of several types of fuel under a variety of test conditions. This experience led to an evolution of test methodology and techniques which can be employed to assess RIA behavior of LWR fuel. A pressurized water loop that will accommodate RIA testing of LWR and CANDU-type fuel has completed conceptual design. Preliminary calculations of transient characteristics and energy deposition into test rods during hypothetical TREAT RIA tests indicate that with the installation of a pressurized water loop, the facility is quite capable of performing prototypic RIA testing. Typical test scenarios indicate that a simulated RIA with a 72 msec FWHM pulse width and energy deposition of 1200 kJ/kg (290 cal/gm) is possible. Further control system enhancements would expand the capability to pulse widths as narrow as 40 msec. (author)

DSMC Simulations of Disturbance Torque to ISS During Airlock Depressurization

Science.gov (United States)

Lumpkin, F. E., III; Stewart, B. S.

2015-01-01

The primary attitude control system on the International Space Station (ISS) is part of the United States On-orbit Segment (USOS) and uses Control Moment Gyroscopes (CMG). The secondary system is part of the Russian On orbit Segment (RSOS) and uses a combination of gyroscopes and thrusters. Historically, events with significant disturbances such as the airlock depressurizations associated with extra-vehicular activity (EVA) have been performed using the RSOS attitude control system. This avoids excessive propulsive "de-saturations" of the CMGs. However, transfer of attitude control is labor intensive and requires significant propellant. Predictions employing NASA's DSMC Analysis Code (DAC) of the disturbance torque to the ISS for depressurization of the Pirs airlock on the RSOS will be presented [1]. These predictions were performed to assess the feasibility of using USOS control during these events. The ISS Pirs airlock is vented using a device known as a "T-vent" as shown in the inset in figure 1. By orienting two equal streams of gas in opposite directions, this device is intended to have no propulsive effect. However, disturbance force and torque to the ISS do occur due to plume impingement. The disturbance torque resulting from the Pirs depressurization during EVAs is estimated by using a loosely coupled CFD/DSMC technique [2]. CFD is used to simulate the flow field in the nozzle and the near field plume. DSMC is used to simulate the remaining flow field using the CFD results to create an in flow boundary to the DSMC simulation. Due to the highly continuum nature of flow field near the T-vent, two loosely coupled DSMC domains are employed. An 88.2 cubic meter inner domain contains the Pirs airlock and the T-vent. Inner domain results are used to create an in flow boundary for an outer domain containing the remaining portions of the ISS. Several orientations of the ISS solar arrays and radiators have been investigated to find cases that result in minimal

Dynamic solution of vessel depressuring; Simulacao dinamica de despressurizacao em vasos

Energy Technology Data Exchange (ETDEWEB)

Rodriguez, Marco Tulio; Silva Netto, Rafael [Chemtech Servicos de Engenharia e Software Ltda., Rio de Janeiro, RJ (Brazil); Aires, Joyce Stone S. [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil). Centro de Pesquisas (CENPES)

2004-07-01

Vessel depressuring is an important phenomenon on chemical and petrochemical processes, specially those related to oil and gas exploration, production and processing. The correct modeling of this phenomenon and prediction of the temperatures, mass and thermal rates involved is essential to the adequate design of the equipment, materials specification and safety standards assurance. To fulfill these requirements, we developed a method to simulate the phenomenon of depressuring. In this approach, the dynamic process is discretized along time, being calculated at each interval the conditions inside the vessel, material flow and heat exchanges with the environment, through mass and energy balances and thermodynamic equilibrium calculation. The method can include different models to calculate heat exchanges and flow through the relief valve, and new methods can be incorporated if necessary. The efficiency of the method was verified by comparing its results with the ones obtained by market-leaders process simulators, when it was proved the robustness of the method and precision of the results. To increase its flexibility of use, the method was incorporated to the PETROBRAS Process Simulator - PETROX (PETROBRAS, 2004), in a development made by Chemtech under a contract with PETROBRAS and already used in large scale by PETROBRAS to simulate its process units. (author)

Westinghouse Hanford Company package testing capabilities

International Nuclear Information System (INIS)

Hummer, J.H.; Mercado, M.S.

1993-07-01

The Department of Energy's Hanford Site is a 1,450-km 2 (560-mi 2 ) installation located in southeastern Washington State. Established in 1943 as a plutonium production facility, Hanford's role has evolved into one of environmental restoration and remediation. Many of these environmental restoration and remediation activities involve transportation of radioactive/hazardous materials. Packagings used for the transportation of radioactive/hazardous materials must be capable of meeting certain normal transport and hypothetical accident performance criteria. Evaluations of performance to these criteria typically involve a combination of analysis and testing. Required tests may include the free drop, puncture, penetration, compression, thermal, heat, cold, vibration, water spray, water immersion, reduced pressure, and increased pressure tests. The purpose of this paper is to outline the Hanford capabilities for performing each of these tests

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.)

The Advanced Test Reactor Irradiation Facilities and Capabilities

International Nuclear Information System (INIS)

S. Blaine Grover; Raymond V. Furstenau

2007-01-01

The Advanced Test Reactor (ATR) is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. The ATR is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These different capabilities include passive sealed capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. The ATR has enhanced capabilities in experiment monitoring and control systems for instrumented and/or temperature controlled experiments. The control systems utilize feedback from thermocouples in the experiment to provide a custom blended flowing inert gas mixture to control the temperature in the experiments. Monitoring systems have also been utilized on the exhaust gas lines from the experiment to monitor different parameters, such as fission gases for fuel experiments, during irradiation. ATR's unique control system provides axial flux profiles in the experiments, unperturbed by axially positioned control components, throughout each reactor operating cycle and over the duration of test programs requiring many years of irradiation. The ATR irradiation positions vary in diameter from 1.6 cm (0.625 inches) to 12.7 cm (5.0 inches) over an active core length of 122 cm (48.0 inches). Thermal and fast neutron fluxes can be adjusted radially across the core depending on the needs of individual test programs. This paper will discuss the different irradiation capabilities available and the cost/benefit issues related to each capability. Examples of different experiments will also be discussed to demonstrate the use of the capabilities and facilities at ATR for performing irradiation experiments

Effect of Acute Intermittent CPAP Depressurization during Sleep in Obese Patients

Science.gov (United States)

Jun, Jonathan C.; Unnikrishnan, Dileep; Schneider, Hartmut; Kirkness, Jason; Schwartz, Alan R.; Smith, Philip L.; Polotsky, Vsevolod Y.

2016-01-01

Background Obstructive Sleep Apnea (OSA) describes intermittent collapse of the airway during sleep, for which continuous positive airway pressure (CPAP) is often prescribed for treatment. Prior studies suggest that discontinuation of CPAP leads to a gradual, rather than immediate return of baseline severity of OSA. The objective of this study was to determine the extent of OSA recurrence during short intervals of CPAP depressurization during sleep. Methods Nine obese (BMI = 40.4 ± 3.5) subjects with severe OSA (AHI = 88.9 ± 6.8) adherent to CPAP were studied during one night in the sleep laboratory. Nasal CPAP was delivered at therapeutic (11.1 ± 0.6 cm H20) or atmospheric pressure, in alternating fashion for 1-hour periods during the night. We compared sleep architecture and metrics of OSA during CPAP-on and CPAP-off periods. Results 8/9 subjects tolerated CPAP withdrawal. The average AHI during CPAP-on and CPAP-off periods was 3.6 ± 0.6 and 15.8 ± 3.6 respectively (pCPAP-on and CPAP-off was 4.7 ± 2 and 20.4 ± 4.7 respectively (pCPAP depressurization also induced more awake (pCPAP during sleep led to deterioration of sleep architecture but only partial re-emergence of OSA. These observations suggest carryover effects of CPAP. PMID:26731735

MODELING THE INFLUENCE OF ACTIVE SUBSLAB DEPRESSURIZATION (ASD) SYSTEMS ON AIRFLOWS IN SUBSLAB AGGREGATE BEDS

Science.gov (United States)

A simple model is presented that allows the pressure difference in a subslab aggregate layer to be estimated as a function of radial distance from the central suction point of an active subslab depressurization system by knowing the average size, thickness, porosity, and permeabi...

BLOW.MOD2: program for a vessel depressurization calculation with the contribution of structures

International Nuclear Information System (INIS)

Doval, A.

1990-01-01

The BLOW.MOD2 program developed to calculate pressure vessels' depressurization is presented, considering heat contribution of the structures. The results are opposite to those obtained from other more complex numerical models, being the comparison extremely satisfactory. BLOW.MOD2 is a software of the 'Systems Sub-Branch', INVAP S.E. (Author) [es

Marshall Space Flight Center's Impact Testing Facility Capabilities

Science.gov (United States)

Finchum, Andy; Hubbs, Whitney; Evans, Steve

2008-01-01

Marshall Space Flight Center s (MSFC) Impact Testing Facility (ITF) serves as an important installation for space and missile related materials science research. The ITF was established and began its research in spacecraft debris shielding in the early 1960s, then played a major role in the International Space Station debris shield development. As NASA became more interested in launch debris and in-flight impact concerns, the ITF grew to include research in a variety of impact genres. Collaborative partnerships with the DoD led to a wider range of impact capabilities being relocated to MSFC as a result of the closure of Particle Impact Facilities in Santa Barbara, California. The Particle Impact Facility had a 30 year history in providing evaluations of aerospace materials and components during flights through rain, ice, and solid particle environments at subsonic through hypersonic velocities. The facility s unique capabilities were deemed a "National Asset" by the DoD. The ITF now has capabilities including environmental, ballistic, and hypervelocity impact testing utilizing an array of air, powder, and two-stage light gas guns to accommodate a variety of projectile and target types and sizes. Numerous upgrades including new instrumentation, triggering circuitry, high speed photography, and optimized sabot designs have been implemented. Other recent research has included rain drop demise characterization tests to obtain data for inclusion in on-going model development. The current and proposed ITF capabilities range from rain to micrometeoroids allowing the widest test parameter range possible for materials investigations in support of space, atmospheric, and ground environments. These test capabilities including hydrometeor, single/multi-particle, ballistic gas guns, exploding wire gun, and light gas guns combined with Smooth Particle Hydrodynamics Code (SPHC) simulations represent the widest range of impact test capabilities in the country.

Depressurization experiments on a plugged fibrous insulation in a horizontal pressure tube

International Nuclear Information System (INIS)

Lang, H.; Weise, H.J.; Ennen, P.

1977-08-01

Hot gas ducts for high-temperature reactors with a helium turbine are subject to additional operational loads not caused by the gas temperature. They include vibrations, caused by high gas velocities or by the sound fields emitted from the turbine, and stresses, originating from fast, short-time pressure changes. Such pressure changes occur as a rule if the generator coupled with the turbine has to be disconnected from the grid. In order to avoid no-load operation of the turbine a bypass between HP and LP side of the turbine is opened. As a consequence of this measure a sudden pressure drop occurs in the free flow cross-section causing differential pressures within the insulation. As the size of these differential pressures depends on the insulating material, the density of plugging, the kind of internals, and on the position and size of the depressurization borings, the pressure distributions in the insulation were measured on a test tube for the HP channel. (orig./RW) [de

Control rod ejection analysis during a depressurization accident and the development of a rod-ejection-preventing device

International Nuclear Information System (INIS)

Mitake, S.; Itoh, K.; Fukushima, H.; Inoue, T.

1982-01-01

The control rods used for the experimental VHTR are suspended in the core by means of flexible steel cables and it is conceivable that an accidental rod ejection could occur due to a depressurization accident. The computer code AFLADE was developed in order to analyze the possibility of accidental rod ejection, and several studies were performed. The parametric study results showed that the adopted design condition for the VHTR core will not cause a rod ejection accident. In parallel with these accident analyses, a rod-ejection-preventing device was developed in preparation for a hypothetical accident, and its function was verified by the component tests

Porflow Capabilities, Usage, History, and Testing

International Nuclear Information System (INIS)

Collard, L.B.

1998-05-01

To support closure of the Savannah River Site High Level Waste tanks, the PORFLOW computer program is being applied to predict long term movement of residual contaminants from the tanks. The PORFLOW program has greater capabilities than simpler programs that have been used previously, and PORFLOW results have been accepted by state and federal regulators throughout the United States. This document briefly discusses the PORFLOW capabilities and presents lists of reports showing PORFLOW's usage history and testing

Depressurization accident analyses for the Fort St. Vrain Reactor

International Nuclear Information System (INIS)

Paul, D.D.

1976-01-01

Design-basis depressurization accident analyses for the Fort St. Vrain reactor were performed using the FLODIS (Ref. 4) code. The FLODIS code models the active core, side reflector, gas annulus between the core barrel and the PCRV liner, and the PCRV cooling system. Results are presented for the Pelton circulators operating at 10,550, 8800, and 7000 rpm. Maximum temperatures of selected components are plotted as a function of time during the transient. None of the components studied exceeded the temperature at which failure or damage may occur. However, there must be sufficient mixing of the outlet gas in the lower plenum to insure the integrity of the steel liners of the steam generator inlet ducts

MHTGR inherent heat transfer capability

International Nuclear Information System (INIS)

Berkoe, J.M.

1992-01-01

This paper reports on the Commercial Modular High Temperature Gas-Cooled Reactor (MHTGR) which achieves improved reactor safety performance and reliability by utilizing a completely passive natural convection cooling system called the RCCS to remove decay heat in the event that all active cooling systems fail to operate. For the highly improbable condition that the RCCS were to become non-functional following a reactor depressurization event, the plant would be forced to rely upon its inherent thermo-physical characteristics to reject decay heat to the surrounding earth and ambient environment. A computational heat transfer model was created to simulate such a scenario. Plant component temperature histories were computed over a period of 20 days into the event. The results clearly demonstrate the capability of the MHTGR to maintain core integrity and provide substantial lead time for taking corrective measures

Mathematical simulation of the drying of suspensions and colloidal solutions by their depressurization

Science.gov (United States)

Lashkov, V. A.; Levashko, E. I.; Safin, R. G.

2006-05-01

The heat and mass transfer in the process of drying of high-humidity materials by their depressurization has been investigated. The results of experimental investigation and mathematical simulation of the indicated process are presented. They allow one to determine the regularities of this process and predict the quality of the finished product. A technological scheme and an engineering procedure for calculating the drying of the liquid base of a soap are presented.

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 Fission Product Liftoff Behavior During Depressurization Transients

International Nuclear Information System (INIS)

Tak, Nam-il; Yoon, Churl; Lee, Sung Nam

2016-01-01

As one of crucial technologies for the NHDD project, the development of the GAMMA-FP code is on-going. The GAMMA-FP code is targeted for fission product transport analysis under accident conditions. A well-known experiment named COMEDIE considered two important phenomena, i.e., fission product plateout and liftoff, for fission product transport within the primary circuit of a prismatic high temperature gas cooled reactor. The accumulated fission products on the structural material via the plateout can be liftoff during a blowdown phase after a pipe break accident. Since the fission product liftoff can increase a radioactivity risk, it is important to predict the amount of fission product liftoff during depressurization accidents. In this work, a model for fission product liftoff is implemented into the GAMMA-FP code and the GAMMA-FP code with the implemented model is validated using the COMEDIE blowdown test data. The results of GAMMA-FP show that the GAMMA-FP code can reliably simulate a pressure transient during blowdown phase after a pipe break accident. In addition, a reasonable amount of fission product liftoff was predicted by the GAMMA-FP code. The maximum difference between the measured and predicted liftoff fraction was less than a factor of 10. More in-depth study is required to increase the accuracy of prediction for a fission product liftoff

Space Environmental Effects Testing Capability at the Marshall Space Flight Center

Science.gov (United States)

DeWittBurns, H.; Craven, Paul; Finckenor, Miria; Nehls, Mary; Schneider, Todd; Vaughn, Jason

2012-01-01

Understanding the effects of the space environment on materials and systems is fundamental and essential for mission success. If not properly understood and designed for, the effects of the environment can lead to degradation of materials, reduction of functional lifetime, and system failure. In response to this need, the Marshall Space Flight Center has developed world class Space Environmental Effects (SEE) expertise and test facilities to simulate the space environment. Capabilities include multiple unique test systems comprising the most complete SEE testing capability available. These test capabilities include charged particle radiation (electrons, protons, ions), ultraviolet radiation (UV), vacuum ultraviolet radiation (VUV), atomic oxygen, plasma effects, space craft charging, lunar surface and planetary effects, vacuum effects, and hypervelocity impacts as well as the combination of these capabilities. In addition to the uniqueness of the individual test capabilities, MSFC is the only NASA facility where the effects of the different space environments can be tested in one location. Combined with additional analytical capabilities for pre- and post-test evaluation, MSFC is a one-stop shop for materials testing and analysis. The SEE testing and analysis are performed by a team of award winning experts nationally recognized for their contributions in the study of the effects of the space environment on materials and systems. With this broad expertise in space environmental effects and the variety of test systems and equipment available, MSFC is able to customize tests with a demonstrated ability to rapidly adapt and reconfigure systems to meet customers needs. Extensive flight experiment experience bolsters this simulation and analysis capability with a comprehensive understanding of space environmental effects.

Automated ultrasonic testing--capabilities, limitations and methods

International Nuclear Information System (INIS)

Beller, L.S.; Mikesell, C.R.

1977-01-01

The requirements for precision and reproducibility of ultrasonic testing during inservice inspection of nuclear reactors are both quantitatively and qualitatively more severe than most current practice in the field can provide. An automated ultrasonic testing (AUT) system, which provides a significant advancement in field examination capabilities, is described. Properties of the system, its application, and typical results are discussed

Feasibility study for the adoption of POSRV for KNGR safety depressurization system

International Nuclear Information System (INIS)

Kwon, Young Min; Lim, Hong Sik; Song, Jin Ho; Sim, Suk Ku; Park, Jong Kyun

1999-03-01

The Korean Next Generation Reactor (KNGR) adopted an advanced design feature of safety depressurization system(SDS) to rapidly de pressurize the reactor coolant system(RCS) in case of beyond design basis events of severe accidents, or a highly unlikely event of a total loss of feedwater (TLOFW) to both steam generators. Two design approaches were considered for the KNGR SDS design. The use of bleed valves similar to those of ABB-CE's system 80+ is design option 1, while in design option 2, the Power Operated Safety Relief valve (POSRV) is considered to provide the combined function of overpressure protection and rapid depressurization. The purpose of this report is to investigate the feasibility of adoption of French SebimPOSRVs for KNGR SDS (design option 2). This report provides the methodology to analyze the TLOFW event with Sebim valves and presents the results of thermal hydraulic analyses using a best-estimate version CEFLASH-4AS/REM for the TLOFW event with feed and bleed. The analyses were performed using a preliminary KNGR design data. For design option 2, if the operator opens two out of the three Sebim valves in conjunction with the four HPSI pumps before a hot leg saturation condition, the decay heat removal and core inventory make-up function can be successfully accomplished. The results of the present investigation demonstrate that the two design options are both feasible to mitigate the consequences of the TLOFW event with a sufficient margin. (Author). 22 refs., 3 tabs., 19 figs

ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

Energy Technology Data Exchange (ETDEWEB)

Takeda, Takeshi; Ohtsu, Iwao [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokaimura (Japan)

2017-08-15

An experiment using the Primaerkreislaeufe Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg small-break loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

Directory of Open Access Journals (Sweden)

Takeshi Takeda

2017-08-01

Full Text Available An experiment using the Primӓrkreislӓufe Versuchsanlage (PKL was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF on a cold leg small-break loss-of-coolant accident with an accident management (AM measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

CNTB program for the analysis of partially mixed containment atmospheres during depressurization events

International Nuclear Information System (INIS)

Landoni, J.A.

1979-07-01

This program describes the analytical models for the CNTB computer program, which is permanently filed in the archive library of the General Atomic (GA) San Diego Data Center under reference number THSD-2699. Developed during the last four years, this computer program has been successfully applied in its presented form to the type of containment atmosphere transients illustrated in this report. For example, the CNTB computer program is applicable (1) to the design basis depressurization accident (DBDA) to determine the effect of the partial mixing on the containment atmospheric peak pressure (known as nonmixing penalty) and (2) for Class 9 accidents, such as the loss of forced circulation (LOFC), for the AIPA Phase I studies. The capability of the CNTB computer program has been substantially improved over its precursor, the CONTEMPT-G computer program, to predict the thermodynamic behavior of the containment atmosphere during helium releases, assuming partial mixing of the original air with the effluent and to predict the amount of the environmental leaks under closed and open containment conditions. In addition, the CNTB computer program running times are considerably below the ones required for the CONTEMPT-G computer program. Computational solution of the variable parameters in the containment atmosphere is effected by an iterative technique, while the temperatures for its boundaries are obtained by finite differences. The CNTB computer program, written in FORTRAN V, has been implemented at GA on the UNIVAC 1110 computer

Pore network modelling of heavy oil depressurization : a parametric study of factors affecting critical gas saturation and three-phase relative permeabilities

Energy Technology Data Exchange (ETDEWEB)

Bondino, I.; McDougall, S.D. [Heriot-Watt Univ., Edinburgh, Scotland (United Kingdom); Hamon, G. [TotalFina Elf Exploration and Production (France)

2002-07-01

A review of how the bubble nucleation process affects the efficiency of heavy oil recovery was presented along with a discussion regarding a pore-scale simulator technique to depressurize heavy oil systems. A light oil depressurization simulation is also presented in which a straightforward instantaneous nucleation (IN) model and a more intricate progressive nucleation (PN) model have been used. Simulation results are compared to those derived from the heavy oil systems. The nucleation of bubbles, their growth by solute diffusion and expansion, plus the final stages of coalescence migration and production are the main steps in the depressurization process which were accounted for in a 3-phase simulator. The model can also determine the impact of bubble density and gas-oil diffusion coefficient on critical gas saturation and 3-phase relative permeability. The difference in results for light and heavy oils was also highlighted. In the first scenario, the evolution of gas was characterized by embryonic bubbles that are quickly and randomly nucleated once bubble-point pressure is reached. A stochastic algorithm was developed for PN from experimental observations. IN and PN observations were not necessarily contradictory. It was determined that the high interfacial tension of heavy oils leads to a more compact, capillary-dominated pattern of gas evolution compared to light oils, resulting in improved recoveries for heavy oil systems. 23 refs., 6 tabs., 23 figs.

Analysis of Depressurization Performance in Containment of Wolsong NPP Unit 1 through Containment Filtered Venting System

International Nuclear Information System (INIS)

Lee, Sunghan; Kim, Jinhyuck; Suh, Nam Duk; Cho, Songwon

2014-01-01

Containment filtered venting system (CFVS) is designed to open and to close isolation valves passively by an operator. CFVS is operated when the containment pressure exceeds the design pressure (225 kPa(a)) and is closed when the containment pressure decreases below 151 kPa(a). The aim of this study is to analyze the depressurization performance of Wolsong unit 1 through CFVS during SBO. The thermal-hydraulic behavior in containment of Wolsong unit 1 was evaluated using the MELCOR 1.8.6 code developed at Sandia National Laboratories (SNL) for the U.S. Nuclear Regulatory Commission (NRC). In addition, in order to evaluate the effects of the CFVS according to the venting area, a sensitivity study depending on different venting area of the CFVS was conducted. Finally, an analysis of the effects of filtering and scrubbing of radioactive material for CFVS is important but not treated in this paper. The SBO accident is chosen to analyze the thermal-hydraulic behavior of Wolsong unit 1. During SBO, the analysis of CFVS affecting on the depressurization of the containment was conducted using MELCOR 1.8.6 code. Also, a sensitivity study was carried out to evaluate the depressurization performance according to the venting area of CFVS. The results show that the containment pressure is considerably decreased and the integrity of the containment could be maintained in case of CFVS operating. Therefore, CFVS has the capacity to keep the containment pressure below the design pressure during SBO. In addition, there are large differences in the containment pressure depending on venting area. We found that the decreasing rate of the pressure in the containment and water level in CFVS depends on the venting area. In the future, a proper requirement for CFVS sizing criteria according to accident scenarios such as LBLOCA, SBLOCA and SGTR, etc. should be evaluated in order to review the licensing for CFVS. Finally, analyses of aerosols, fission product, and radioactive material

Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor

International Nuclear Information System (INIS)

Ott, Larry J.; Ellis, Ronald James; McDuffee, Joel Lee; Spellman, Donald J.; Bevard, Bruce Balkcom

2009-01-01

The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

Two-phase mixture level swell and liquid entrainment/off-take in a vessel during rapid depressurization

International Nuclear Information System (INIS)

Kim, Chang Hyun

2004-02-01

An experimental study has been performed to analyze the two-phase mixture level swell and the liquid entrainment/off-take through the break in a vessel, which are important phenomena to determine the bleed capacity of the Safety Depressurization System (SDS) of Korea Advanced Power Reactor 1400 (APR1400). Three separate experiments are performed in this study: (a) the depressurization and two-phase mixture level swell experiment: (b) the two-phase mixture level measurement experiment: (c) the liquid entrainment and off-take experiment. A series of experiments has been performed using a scaled pressurized vessel in various depressurization conditions to analyze the two-phase mixture level swell and the liquid entrainment/off-take phenomena from the two-phase mixture surface in the first experiment. The test parameters are the initial pressure (10 - 38.75bars), the initial water level (43.7% - 80.0% of full height), the orifice inner diameter (10mm, 17.5mm, and 20mm). The liquid off-take takes place in certain experimental conditions. The measured parameters in the present experiments are axial void fraction distributions, pressures, temperatures in the test vessel, and the mixture density and mass flowrate through the discharge pipe. An assessment of RELAP5/MOD3 code with the present experimental data has been performed. With appropriate nodalization and time step, RELAP5/MOD3 showed reasonable agreement with the present experimental data for the gradual depressurization without liquid off-take. In the case that the off-take takes place, however, RELAP5/MOD3 under-predicts the amount of liquid entrainment/off-take during depressurization. In the second experiment, an assessment of an ultrasonic sensor and a two-wire type capacitance probe for the two-phase mixture level measurement has been performed under the same experimental conditions to adopt an appropriate measurement method for the two-phase mixture level swell and to investigate pool void fraction by the

Commonwealth Edison Company pressure locking test report

Energy Technology Data Exchange (ETDEWEB)

Bunte, B.D.; Kelly, J.F.

1996-12-01

Pressure Locking is a phenomena which can cause the unseating thrust for a gate valve to increase dramatically from its typical static unseating thrust. This can result in the valve actuator having insufficient capability to open the valve. In addition, this can result in valve damage in cases where the actuator capability exceeds the valve structural limits. For these reasons, a proper understanding of the conditions which may cause pressure locking and thermal binding, as well as a methodology for predicting the unseating thrust for a pressure locked or thermally bound valve, are necessary. This report discusses the primary mechanisms which cause pressure locking. These include sudden depressurization of piping adjacent to the valve and pressurization of fluid trapped in the valve bonnet due to heat transfer. This report provides a methodology for calculating the unseating thrust for a valve which is pressure locked. This report provides test data which demonstrates the accuracy of the calculation methodology.

Commonwealth Edison Company pressure locking test report

International Nuclear Information System (INIS)

Bunte, B.D.; Kelly, J.F.

1996-01-01

Pressure Locking is a phenomena which can cause the unseating thrust for a gate valve to increase dramatically from its typical static unseating thrust. This can result in the valve actuator having insufficient capability to open the valve. In addition, this can result in valve damage in cases where the actuator capability exceeds the valve structural limits. For these reasons, a proper understanding of the conditions which may cause pressure locking and thermal binding, as well as a methodology for predicting the unseating thrust for a pressure locked or thermally bound valve, are necessary. This report discusses the primary mechanisms which cause pressure locking. These include sudden depressurization of piping adjacent to the valve and pressurization of fluid trapped in the valve bonnet due to heat transfer. This report provides a methodology for calculating the unseating thrust for a valve which is pressure locked. This report provides test data which demonstrates the accuracy of the calculation methodology

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

International Nuclear Information System (INIS)

Sudoh, Takashi; Murao, Yoshio.

1983-08-01

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

Hydro-geomechanical behaviour of gas-hydrate bearing soils during gas production through depressurization and CO2 injection

Science.gov (United States)

Deusner, C.; Gupta, S.; Kossel, E.; Bigalke, N.; Haeckel, M.

2015-12-01

Results from recent field trials suggest that natural gas could be produced from marine gas hydrate reservoirs at compatible yields and rates. It appears, from a current perspective, that gas production would essentially be based on depressurization and, when facing suitable conditions, be assisted by local thermal stimulation or gas hydrate conversion after injection of CO2-rich fluids. Both field trials, onshore in the Alaska permafrost and in the Nankai Trough offshore Japan, were accompanied by different technical issues, the most striking problems resulting from un-predicted geomechanical behaviour, sediment destabilization and catastrophic sand production. So far, there is a lack of experimental data which could help to understand relevant mechanisms and triggers for potential soil failure in gas hydrate production, to guide model development for simulation of soil behaviour in large-scale production, and to identify processes which drive or, further, mitigate sand production. We use high-pressure flow-through systems in combination with different online and in situ monitoring tools (e.g. Raman microscopy, MRI) to simulate relevant gas hydrate production scenarios. Key components for soil mechanical studies are triaxial systems with ERT (Electric resistivity tomography) and high-resolution local strain analysis. Sand production control and management is studied in a novel hollow-cylinder-type triaxial setup with a miniaturized borehole which allows fluid and particle transport at different fluid injection and flow conditions. Further, the development of a large-scale high-pressure flow-through triaxial test system equipped with μ-CT is ongoing. We will present results from high-pressure flow-through experiments on gas production through depressurization and injection of CO2-rich fluids. Experimental data are used to develop and parametrize numerical models which can simulate coupled process dynamics during gas-hydrate formation and gas production.

Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP

International Nuclear Information System (INIS)

Ramirez G, C.; Chavez M, C.

2012-10-01

The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

Energy penalties associated with the use of a sub-slab depressurization system

International Nuclear Information System (INIS)

Clarkin, M.; Brennan, T.; Osborne, M.C.

1990-01-01

One of the primary radon mitigation techniques used to reduce indoor radon concentrations in houses is a sub-slab depressurization system. In this type of system, a fan removes soil gases containing radon from beneath the floor slab and exhausts the gases to the outdoors by creating a pressure field beneath the slab that is negative relative to the basement air pressure. Because of this negative pressure, indoor conditioned air can be drawn through the floor penetrations and exhausted outdoors. In order to determine the amount of conditioned air that is being lost, a series of experiments utilizing tracer gases were performed in three houses. This paper presents the results of these experiments

AP1000 station blackout study with and without depressurization using RELAP5/SCDAPSIM

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)

2016-10-15

Highlights: • A representative RELAP5/SCDAPSIM model of AP1000 has been developed. • Core is modeled using SCDAP. • A SBO for the AP1000 has been simulated for high pressure (no depressurization) and low pressure (depressurization). • Significant differences in the damage progression have been observed for the two cases. • Results also reinforced the fact that surge line fails before vessel failure in case of high pressure scenario. - Abstract: Severe accidents like TMI-2, Chernobyl, Fukushima made it inevitable to analyze station blackout (SBO) for all the old as well as new designs although it is not a regulatory requirement in most of the countries. For such improbable accidents, a SBO for the AP1000 using RELAP5/SCDAPSIM has been simulated. Many improvements have been made in fuel damage progression models of SCDAP after the Fukushima accident which are now being tested for the new reactor designs. AP1000 is a 2-loop pressurized water reactor (PWR) with all the emergency core cooling systems based on natural circulation. Its core design is very similar to 3-loop PWR with 157 fuel assemblies. The primary circuit pumps, pressurizer and steam generators (with necessary secondary side) are modeled using RELAP5. The core has been divided into 20 axial nodes and 6 radial rings; the corresponding six groups of assemblies have been modeled as six pipe components with proportionate flow area. Fuel assemblies are modeled using SCDAP fuel and control components. SCDAP has 2d-heat conduction and radiative heat transfer, oxidation and complete severe fuel damage progression models. The final input deck achieved all the steady state thermal hydraulic conditions comparable to the design control document of AP1000. To quantify the core behavior, under unavailability of all safety systems, various time profiles for SBO simulations @ high pressure and low pressure have been compared. This analysis has been performed for 102% (3468 MWt) of the rated core power. The

Nuclear reactor steam depressurization valve

International Nuclear Information System (INIS)

Moore, G.L.

1991-01-01

This patent describes improvement in a nuclear reactor plant, an improved steam depressurization valve positioned intermediate along a steam discharge pipe for controlling the venting of steam pressure from the reactor through the pipe. The improvement comprises: a housing including a domed cover forming a chamber and having a partition plate dividing the chamber into a fluid pressure activation compartment and a steam flow control compartment, the valve housing being provided with an inlet connection and an outlet connection in the steam flow control compartment, and a fluid duct in communication with a source of fluid pressure for operating the valve; a valve set mounted within the fluid flow control compartment comprising a cylindrical section surrounding the inlet connection with one end adjoining the connection and having a radially projecting flange at the other end with a contoured extended valve sealing flange provided with an annular valve sealing member, and a valve cylinder traversing the partition plate and reciprocally movable within an opening in the partition plate with one terminal and extending into the fluid pressure activation compartment and the other terminal end extending into the steam flow control compartment coaxially aligned with the valve seat surrounding the inlet connection, the valve cylinder being surrounded by two bellow fluid seals and provided with guides to inhibit lateral movement, an end of the valve cylinder extending into the fluid flow control compartment having a radially projecting flange substantially conterminous with the valve seat flange and having a contoured surface facing and complimentary to the contoured valve seating surface whereby the two contoured valve surfaces can meet in matching relationship, thus providing a pressure actuated reciprocatable valve member for making closing contact with the valve seat and withdrawing therefrom for opening fluid flow through the valve

Space Environmental Effects (SEE) Testing Capability: NASA/Marshall Space Flight Center

Science.gov (United States)

DeWittBurns, H.; Crave, Paul; Finckenor, Miria; Finchum, Charles; Nehls, Mary; Schneider, Todd; Vaughn, Jason

2012-01-01

Understanding the effects of the space environment on materials and systems is fundamental and essential for mission success. If not properly understood and designed for, the space environment can lead to materials degradation, reduction of functional lifetime, and system failure. Ground based testing is critical in predicting performance NASA/MSFC's expertise and capabilities make up the most complete SEE testing capability available.

New Environmental Testing Capabilities at INTA

Science.gov (United States)

Olivo, Esperanza; Hernandez, Daniel; Garranzo, Daniel; Barandiaran, Javier; Reina, Manuel

2012-07-01

In this paper we aim to present and describe the facilities for aerospace environmental testing at INTA; the Spanish National Institute for Aerospace Technique with emphasis on the Thermal Vacuum testing facility with dimensions 4 m x 4 m x 4 m and a temperature range from +150oC to -175 oC and 10-6 vacuum conditions with the new Thermo Elastic Distortion (TED) measurement capability designed at INTA. It will be presented the validation data for the empty chamber, with specimens such a 3m diameter reflector and antenna towers for both, thermal cycling and TED measurements. For TED, it will be shown the feasibility study and the solution finally selected. Apart from those, it will be shown other complementary facilities for environmental testing such as 320 (2x160) kN dual shaker with a new 3 m x 3 m sliding table and other complementary facilities.

Radon remediation of a two-storey UK dwelling by active sub-slab depressurization: observations on hourly Radon concentration variations

International Nuclear Information System (INIS)

Denman, A.R.

2008-01-01

Radon concentration levels in a two-storey detached single-family dwelling in Northamptonshire, UK, were monitored at hourly intervals throughout a 5-week period during which sub-slab depressurization remediation measures, including an active sump system, were installed. Remediation of the property was accomplished successfully, with the mean radon levels upstairs and downstairs greatly reduced and the prominent diurnal variability in radon levels present prior to remediation almost completely removed. Following remediation, upstairs and downstairs radon concentrations were 32% and 16% of their pre-remediation values respectively. The mean downstairs radon concentration was lower than that upstairs, with pre-and post-remediation values of the upstairs/downstairs concentration ratio, R U/D , of 0.93 and 1.76 respectively. Cross-correlation between upstairs and downstairs radon concentration time-series indicates a time-lag of the order of 1 hour or less, suggesting that diffusion of soil-derived radon from downstairs to upstairs either occurs within that time frame or forms a relatively insignificant contribution to the upstairs radon level. Cross-correlation between radon concentration time-series and the corresponding time-series for local atmospheric parameters demonstrated correlation between radon concentrations and internal/external pressure-difference prior to remediation. This correlation disappears following remediation, confirming the effectiveness of the remediation procedure in mitigating radon ingress from the ground via the stack-effect. Overall, these observations provide further evidence that radon emanation from building materials makes a not insignificant contribution to radon concentration levels within the building. Furthermore, since this component remains essentially unaffected by sub-slab depressurization, its proportional contribution to the total radon levels in the home increases following remediation, leading to the conclusion that where

Droplet Impact on a Heated Surface under a Depressurized Environment

Science.gov (United States)

Hatakenaka, Ryuta; Tagawa, Yoshiyuki

2016-11-01

Behavior of a water droplet of the diameter 1-3mm impacting on a heated surface under depressurized environment (100kPa -1kPa) has been studied. A syringe pump for droplet generation and a heated plate are set into a transparent acrylic vacuum chamber. The internal pressure of the chamber is automatically controlled at a target pressure with a rotary pump, a pressure transducer, and an electrical valve. A silicon wafer of the thickness 0.28 mm is mounted on the heater plate, whose temperature is directly measured by attaching a thermocouple on the backside. The droplet behavior is captured using a high-speed camera in a direction perpendicular to droplet velocity. Some unique behaviors of droplet are observed by decreasing the environmental pressure, which are considered to be due to two basic elements: Enhancement of evaporation due to the lowered saturation temperature, and shortage of pneumatic spring effect between the droplet and heated wall due to the lowered pressure of the air.

Thermo-hydraulic characteristics of serpentine tubing in the boilers of gas cooled reactors under condition of rapid and slow depressurization

International Nuclear Information System (INIS)

Abouhadra, D.S.; Byrne, J.E.

2003-01-01

In nuclear reactors of the magnox or advanced gas cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accidents using two phase flow codes requires knowledge of the heat transfer behaviour of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear electric . The tests were carried out on the thermal hydraulics experimental research assembly (THERA) loop at manchester university. Depressurization from an initial pressure of 60 bar, with fluid subcooling of 5 k, 50 k, and 100 k was controlled by discharging the test section contents through suitably chosen orifices to produce blowdown to 10% of the initial pressure over a time scale of 30 s to 3600 s. pressures and temperatures in the serpentine were measured at average time intervals of approximately 1 s

Radioactive waste material testing capabilities in Romania

International Nuclear Information System (INIS)

Vieru, G.

1999-01-01

Radioactive material including wastes, generated by Romanian nuclear facilities are packaged in accordance with national and IAEA's Regulation for a safe transport to the disposal center. The evaluation and certification of packages is accomplished by subjecting these packages to normal and simulated test conditions in order to prove the package to technical performances. The standards provide to package designers the possibility to use analysis, testing or a combination of these. The paper describes the experimental and simulating qualification tests for type A packages used for transport and storage of radioactive wastes (low level). Testing are used to substantiate assumptions used in analytical models and to demonstrate package structural response. There are also presented testing capabilities which are used to perform and simulate the required qualification tests. By direct comparison of analysis and experimental results, the degree of reliability of analytical methods and admissibility of assumptions taken in package designing and in demonstrating its safety under conditions of INR - Pitesti, within the contract between the INR - Pitesti and IAEA - Vienna, were determined. (author)

Revisiting the Fundamentals and Capabilities of the Stack Compression Test

DEFF Research Database (Denmark)

Alves, L.M.; Nielsen, Chris Valentin; Martin, P.A.F.

2011-01-01

performance by comparing the flow curves obtained from its utilisation with those determined by means of compressive testing carried out on solid cylinder specimens of the same material. Results show that mechanical testing of materials by means of the stack compression test is capable of meeting...... the increasing demand of accurate and reliable flow curves for sheet metals....

SSC string test facility for superconducting magnets: Testing capabilities and program for collider magnets

International Nuclear Information System (INIS)

Kraushaar, P.; Burgett, W.; Dombeck, T.; McInturff, A.; Robinson, W.; Saladin, V.

1993-05-01

The Accelerator Systems String Test (ASST) R ampersand D Testing Facility has been established at the SSC Laboratory to test Collider and High Energy Booster (HEB) superconducting magnet strings. The facility is operational and has had two testing periods utilizing a half cell of collider prototypical magnets with the associated spool pieces and support systems. This paper presents a description of the testing capabilities of the facility with respect to components and supporting subsystems (cryogenic, power, quench protection, controls and instrumentation), the planned testing program for the collider magnets

Development of the ETOC: a European test of olfactory capabilities

NARCIS (Netherlands)

Thomas-Danguin, T.; Rouby, C.; Sicard, G.; Vigouroux, M.; Farget, V.; Johanson, A.; Bengtzon, A.; Hall, G.; Ormel, W.; Graaf, de C.; Rousseau, F.; Dumont, J.P.

2003-01-01

A number of smell tests designed to evaluate human olfactory capabilities have been published, but none have been validated cross-culturally. The aim of this study was therefore to develop a reliable and quick olfactory test that could be used to evaluate efficiently the olfactory abilities of a

NASA GRC's High Pressure Burner Rig Facility and Materials Test Capabilities

Science.gov (United States)

Robinson, R. Craig

1999-01-01

The High Pressure Burner Rig (HPBR) at NASA Glenn Research Center is a high-velocity. pressurized combustion test rig used for high-temperature environmental durability studies of advanced materials and components. The facility burns jet fuel and air in controlled ratios, simulating combustion gas chemistries and temperatures that are realistic to those in gas turbine engines. In addition, the test section is capable of simulating the pressures and gas velocities representative of today's aircraft. The HPBR provides a relatively inexpensive. yet sophisticated means for researchers to study the high-temperature oxidation of advanced materials. The facility has the unique capability of operating under both fuel-lean and fuel-rich gas mixtures. using a fume incinerator to eliminate any harmful byproduct emissions (CO, H2S) of rich-burn operation. Test samples are easily accessible for ongoing inspection and documentation of weight change, thickness, cracking, and other metrics. Temperature measurement is available in the form of both thermocouples and optical pyrometery. and the facility is equipped with quartz windows for observation and video taping. Operating conditions include: (1) 1.0 kg/sec (2.0 lbm/sec) combustion and secondary cooling airflow capability: (2) Equivalence ratios of 0.5- 1.0 (lean) to 1.5-2.0 (rich), with typically 10% H2O vapor pressure: (3) Gas temperatures ranging 700-1650 C (1300-3000 F): (4) Test pressures ranging 4-12 atmospheres: (5) Gas flow velocities ranging 10-30 m/s (50-100) ft/sec.: and (6) Cyclic and steady-state exposure capabilities. The facility has historically been used to test coupon-size materials. including metals and ceramics. However complex-shaped components have also been tested including cylinders, airfoils, and film-cooled end walls. The facility has also been used to develop thin-film temperature measurement sensors.

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

International Nuclear Information System (INIS)

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

1975-09-01

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

Demonstration of load rating capabilities through physical load testing : Sioux County bridge case study.

Science.gov (United States)

2013-08-01

The objective of this work, Pilot Project - Demonstration of Capabilities and Benefits of Bridge Load Rating through Physical Testing, was to demonstrate the capabilities for load testing and rating bridges in Iowa, study the economic benefit of perf...

Demonstration of load rating capabilities through physical load testing : Johnson County bridge case study.

Science.gov (United States)

2013-08-01

The objective of this work, Pilot Project - Demonstration of Capabilities and Benefits of Bridge Load Rating through Physical Testing, was to demonstrate the capabilities for load testing and rating bridges in Iowa, study the economic benefit of perf...

Demonstration of load rating capabilities through physical load testing : Ida County bridge case study.

Science.gov (United States)

2013-08-01

The objective of this work, Pilot Project - Demonstration of Capabilities and Benefits of Bridge Load Rating through Physical Testing, was to demonstrate the capabilities for load testing and rating bridges in Iowa, study the economic benefit of perf...

BWR recirculation loop discharge line break LOCA tests with break areas of 50 and 100% assuming HPCS failure at ROSA-III test facility

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Tasaka, Kanji; Yonomoto, Taisuke; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Murata, Hideo; Shiba, Masayoshi; Iriko, Masanori.

1985-03-01

This report presents the experimental results of RUN 962 and RUN 963 in ROSA-III program, which are 50 and 100 % break LOCA tests at the BWR recirculation pump discharge line, respectively. The ROSA-III test facility simulates a volumetrically scaled (1/424) BWR system and has four half-length electrically heated fuel bundles, two active recirculation loops, three types of ECCSs and steam and feedwater systems. The experimental data of RUN 962 and RUN 963 were compared with those of RUN 961, a 200 % discharge line break test to study the break area effects on the transient thermal hydraulic phenomena. The least flow areas at the jet pump drive nozzles and recirculation pump discharge nozzle in the broken recirculation loop limitted the discharge flows from the pressure vessel and the depressurization rate in the 100 and 200 % break tests, whereas the least flow area at break nozzle limitted the depressurization rate in the 50 % break test. The highest PCT was observed in the 50 % break test among the three tests. (author)

Experiment data report for Semiscale Mod-1 test S-02-5 (blowdown heat transfer test)

International Nuclear Information System (INIS)

1975-12-01

Recorded test data are presented for Test S-02-5 of the Semiscale Mod-1 blowdown heat transfer test series. Test S-02-5 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 water-cooled nuclear reactor system and to provide data for the assessment of the Loss-of-Fluid Test (LOFT) design basis. Test S-02-5 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,253 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient with full core power (1.6 MW). An electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core. System flow was set to achieve the full design core temperature differential of 66 0 F. The flow resistance of the intact loop was based on core area scaling. During system depressurization, core power was reduced from the initial level of 1.6 MW in such a manner as to simulate the surface heat flux response of the LOFT nuclear fuel rods until such time that departure from nucleate boiling occurs

Test Methods and Protocols for Environmental and Safety Hazards Associated with Home Energy Retrofits

Energy Technology Data Exchange (ETDEWEB)

Cautley, D. [NorthernSTAR Building America Partnership, St. Paul, MN (United States); Viner, J. [NorthernSTAR Building America Partnership, St. Paul, MN (United States); Lord, M. [NorthernSTAR Building America Partnership, St. Paul, MN (United States); Pearce, M. [NorthernSTAR Building America Partnership, St. Paul, MN (United States)

2012-12-01

A number of health hazards and hazards to the durability of homes may be associated with energy retrofitting and home renovation projects. Among the hazards associated with energy retrofit work, exposure to radon is thought to cause more than 15,000 deaths per year in the U.S., while carbon monoxide poisoning results in about 20,000 injuries and 450 deaths per year. Testing procedures have been developed for identifying and quantifying hazards during retrofitting. These procedures commonly include a battery of tests to screen combustion appliances for safe operation, including worst case depressurization measurement, backdrafting (spillage) under depressurized or normal conditions, and carbon monoxide production.

A cryogenic test stand for full length SSC magnets with superfluid capability

International Nuclear Information System (INIS)

Peterson, T.J.; Mazur, P.O.

1989-02-01

The Fermilab Magnet Test Facility performs testing of the full scale SSC magnets on test stands capable of simulating the cryogenic environment of the SSC main ring. One of these test stands, Stand 5, also has the ability to operate the magnet under test at temperatures from 1.8K to 4.5K with either supercritical helium or subcooled liquid, providing at least 25 Watts of refrigeration. At least 50 g/s flow is available from 2.3K to 4.5K, whereas superfluid operation occurs with zero flow. Cooldown time from 4.5K to 1.8K is 1.5 hours. A maximum current capability of 10,000 amps is provided, as is instrumentation to monitor and control the cryogenic conditions. This paper describes the cryogenic design of this test stand. 8 refs., 6 figs

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

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)

Fused Reality for Enhanced Flight Test Capabilities

Science.gov (United States)

Bachelder, Ed; Klyde, David

2011-01-01

The feasibility of using Fused Reality-based simulation technology to enhance flight test capabilities has been investigated. In terms of relevancy to piloted evaluation, there remains no substitute for actual flight tests, even when considering the fidelity and effectiveness of modern ground-based simulators. In addition to real-world cueing (vestibular, visual, aural, environmental, etc.), flight tests provide subtle but key intangibles that cannot be duplicated in a ground-based simulator. There is, however, a cost to be paid for the benefits of flight in terms of budget, mission complexity, and safety, including the need for ground and control-room personnel, additional aircraft, etc. A Fused Reality(tm) (FR) Flight system was developed that allows a virtual environment to be integrated with the test aircraft so that tasks such as aerial refueling, formation flying, or approach and landing can be accomplished without additional aircraft resources or the risk of operating in close proximity to the ground or other aircraft. Furthermore, the dynamic motions of the simulated objects can be directly correlated with the responses of the test aircraft. The FR Flight system will allow real-time observation of, and manual interaction with, the cockpit environment that serves as a frame for the virtual out-the-window scene.

Savannah River release: test of the new ARAC capability

International Nuclear Information System (INIS)

Dickerson, M.H.

1977-01-01

Working jointly from opposite sides of the nation Lawrence Livermore Laboratory (LLL) and the Savannah River Laboratory (SRL) quickly assessed the consequences of an early-morning tritium release in May 1974 from the Savannah River Plant, in South Carolina. Measurements confirmed the accuracy of the LLL predictions. Due to the small quantity involved and to the release location (well within the plant confines), the release was not dangerous to the public. The emergency provided a dramatic test of procedures and capabilities of the new Atmospheric Release Advisory Capability (ARAC) center at Livermore, which was not yet operational, demonstrating the capacity for quick response, and the feasibility of real-time data acquisition and transmittal across the continent

Experimental study and modelization of a propane storage tank depressurization

International Nuclear Information System (INIS)

Veneau, Tania

1995-01-01

The risks associated with the fast depressurization of propane storage tanks reveals the importance of the 'source term' determination. This term is directly linked, among others, to the characteristics of the jet developed downstream of the breach. The first aim of this work was to provide an original data bank concerning drop velocity and diameter distributions in a propane jet. For this purpose, a phase Doppler anemometer bas been implemented on an experimental set-up. Propane blowdowns have been performed with different breach sizes and several initial pressures in the storage tank. Drop diameter and velocity distributions have been investigated at different locations in the jet zone. These measurements exhibited the fragmentation and vaporisation trends in the jet. The second aim of this work concerned the 'source term'. lt required to study the coupling between the fluid behaviour inside the tank and the flow through the breach. This model took into account the phase exchange when flashing occurred in the tank. The flow at the breach was described with an homogeneous relaxation model. This coupled modelization has been successfully and exhaustively validated. lt originality lies on the application to propane flows. (author) [fr

An intermediate heat exchanging-depressurizing loop for nuclear hydrogen production

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Soo [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); No, Hee Cheon, E-mail: hcno@kaist.ac.k [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Yoon, Ho Joon; Lee, Jeong Ik [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

2010-10-15

Sulfur-iodine (SI) cycle should overcome many engineering challenges to commercialize and prove its feasibilities to compete other thermo-chemical cycles. Some critical issues such as structural material, harsh operating condition and high capital costs were considered obstacles to be actualized. Operating SI cycle at low-pressure is one of the solutions to actualize the cycle. The flash operation with over-azeotropic HI at low pressure does not require temperature and pressure as high as those in the existing methods as well as heating for separation. The operation in low pressure reduces corrosion problems and enables us to use flexible selection of structural material. We devised an intermediate heat exchanging-depressurizing loop to eliminate high operating pressure in the hydrogen side as well as a large pressure difference between the reactor side and the hydrogen side. Molten salts are adequate candidates as working fluids under the high-temperature condition with homogeneous phase during pressure changing process. Using molten salts, 2.20-4.65 MW of pumping work is required to change the pressure from 1 bar to 7 MPa. We selected BeF{sub 2}-containing salts as the possible candidates based on preliminary economic and thermal hydraulic consideration.

A test battery measuring auditory capabilities of listening panels

DEFF Research Database (Denmark)

Ghani, Jody; Ellermeier, Wolfgang; Zimmer, Karin

2005-01-01

a battery of tests covering a larger range of auditory capabilities in order to assess individual listeners. The format of all tests is kept as 'objective' as possible by using a three-alternative forced-choice paradigm in which the subject must choose which of the sound samples is different, thus keeping...... the instruction to the subjects simple and common for all tests. Both basic (e.g. frequency discrimination) and complex (e.g. profile analysis) psychoacoustic tests are covered in the battery and a threshold of discrimination or detection is obtained for each test. Data were collected on 24 listeners who had been...... recruited for participation in an expert listening panel for evaluating the sound quality of hi-fi audio systems. The test battery data were related to the actual performance of the listeners when judging the degradation in quality produced by audio codecs....

Cost/benefit analysis of adding a feed-and-bleed capability to Combustion Engineering pressurized-water reactors

International Nuclear Information System (INIS)

Gallup, D.R.; Gahan, E.; Cherdack, R.; Skala, G.

1983-08-01

This report presents the results of a cost/benefit analysis for the addition of a feed-and-bleed capability to the San Onofre Nuclear Generating Station, Unit 2, (SONGS 2). Two cases of feed-and-bleed capability were investigated: 1) adding power-operated relief valves (PORVs) to the pressurizer for depressurization and using the present high-pressure safety-injection (HPSI) system for reactor-coolant-system (RCS) inventory make-up and 2) adding an independent single-train feed-and-bleed system. For the first case, it is estimated that the core-melt frequency would be incrementally reduced by 4.0E-6 per year, a factor of 1.3, at a cost of $2.5 M to $4.3 M depending on when the equipment is installed. For the second case, it is estimated that the core-melt frequency would be incrementally reduced by 1.2E-5 per year, a factor of 3, at a cost of $7.0 M to $10.3 M

Recent Ground Hold and Rapid Depressurization Testing of Multilayer Systems

Science.gov (United States)

Johnson, Wesley L.

2014-01-01

In the development of flight insulation systems for large cryogenic orbital storage (spray on foam and multilayer insulation), testing need include all environments that are experienced during flight. While large efforts have been expended on studying, bounding, and modeling the orbital performance of the insulation systems, little effort has been expended on the ground hold and ascent phases of a mission. Historical cryogenic in-space systems that have flown have been able to ignore these phases of flight due to the insulation system being within a vacuum jacket. In the development phase of the Nuclear Mars Vehicle and the Shuttle Nuclear Vehicle, several insulation systems were evaluated for the full mission cycle. Since that time there had been minimal work on these phases of flight until the Constellation program began investigating cryogenic service modules and long duration upper stages. With the inception of the Cryogenic Propellant Storage and Transfer Technology Demonstration Mission, a specific need was seen for the data and as such, several tests were added to the Cryogenic Boil-off Reduction System liquid hydrogen test matrix to provide more data on a insulation system. Testing was attempted with both gaseous nitrogen (GN2) and gaseous helium (GHe) backfills. The initial tests with nitrogen backfill were not successfully completed due to nitrogen liquefaction and solidification preventing the rapid pumpdown of the vacuum chamber. Subsequent helium backfill tests were successful and showed minimal degradation. The results are compared to the historical data.

Development of a hybrid safety system: Actuation of the secondary automatic depressurization system at an early stage

International Nuclear Information System (INIS)

Nishimoto, Masae; Umezawa, Shigemitsu; Okabe, Kazuharu; Matsuoka, Tsuyoshi

1996-01-01

A Hybrid Safety System, which is an optimum combination of active and passive safety systems, has been developed in order to improve the safety, reliability and economic features of the next generation of PWRs. The passive safety systems include Automatic primary Depressurization System (ADS), Secondary Automatic Depressurization System (SADS), advanced accumulators, gravity injection system and so on. In this study the authors have improved the actuation logic of the passive safety systems. The original logic in the previous study actuates ADS at an early stage of an event such as a Loss-of-Coolant Accident (LOCA), and this is followed by the actuation of SADS. In this study they divide SADS into two systems. The first, small SADS, uses small valves corresponding to the relief valves of the conventional PWR plants. The second, large SADS, corresponds to the original SADS using multiple valves of large capacity. With the new logic, the passive systems are actuated during a typical small LOCA. Small LOCA analyses using several break areas were performed for a 1,400 MWe PWR plant with a Hybrid Safety System. The results predict that core uncovery does not occur in the case of a relatively small break area and that core heat removal during a small LOCA is improved in comparison with the analyses for conventional PWR plants, where the secondary pressure remains higher during the event. The results also predict that this new logic make it possible to reduce the ADS valve size and the actuation pressure setpoint of the passive safety systems

Radioactive material package testing capabilities at Sandia National Laboratories

International Nuclear Information System (INIS)

Uncapher, W.L.; Hohnstreiter, G.F.

1995-01-01

Evaluation and certification of radioactive and hazardous material transport packages can be accomplished by subjecting these packages to normal transport and hypothetical accident test conditions. The regulations allow package designers to certify packages using analysis, testing, or a combination of analysis and testing. Testing can be used to substantiate assumptions used in analytical models and to demonstrate package structural and thermal response. Regulatory test conditions include impact, puncture, crush, penetration, water spray, immersion, and thermal environments. Testing facilities are used to simulate the required test conditions and provide measurement response data. Over the past four decades, comprehensive testing facilities have been developed at Sandia National Laboratories to perform a broad range of verification and certification tests on hazardous and radioactive material packages or component sections. Sandia's facilities provide an experience base that has been established during the development and certification of many package designs. These unique facilities, along with innovative instrumentation data collection capabilities and techniques, simulate a broad range of testing environments. In certain package designs, package testing can be an economical alternative to complex analysis to resolve regulatory questions or concerns

Results from Operational Testing of the Siemens Smart Grid-Capable Electric Vehicle Supply Equipment

Energy Technology Data Exchange (ETDEWEB)

Bennett, Brion [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-05-01

The Idaho National Laboratory conducted testing and analysis of the Siemens smart grid capable electric vehicle supply equipment (EVSE), which was a deliverable from Siemens for the U.S. Department of Energy FOA-554. The Idaho National Laboratory has extensive knowledge and experience in testing advanced conductive and wireless charging systems though INL’s support of the U.S. Department of Energy’s Advanced Vehicle Testing Activity. This document details the findings from the EVSE operational testing conducted at the Idaho National Laboratory on the Siemens smart grid capable EVSE. The testing conducted on the EVSE included energy efficiency testing, SAE J1772 functionality testing, abnormal conditions testing, and charging of a plug-in vehicle.

Overview of the 2006-2008 JOGMEC/NRCan/Aurora Mallik Gas Hydrate Production Test Program

Science.gov (United States)

Yamamoto, K.; Dallimore, S. R.

2008-12-01

During the winters of 2007 and 2008 the Japan Oil, Gas and Metals National Corporation (JOGMEC) and Natural Resources Canada (NRCan), with Aurora Research Institute as the operator, carried out an on-shore gas hydrate production test program at the Mallik site, Mackenzie Delta, Northwest Territories, Canada. The prime objective of the program was to verify the feasibility of depressurization technique by drawing down the formation pressure across a 12m perforated gas hydrate bearing section. This project was the second full scale production test at this site following the 2002 Japex/JNOC/GSC et al Mallik research program in which seven participants organizatinos from five countries undertook a thermal test using hot water circulation Field work in 2007 was devoted to establishing a production test well, installing monitoring devices outside of casing, conducting base line geophysical studies and undertaking a short test to gain practical experience prior to longer term testing planned for 2008 . Hydrate-dissociated gas was produced to surface by depressurization achieved by lowering the fluid level with a dowhole pump. However, the operation was terminated 60 hours after the start of the pumping mainly due to sand production problems. In spite of the short period (12.5 hours of ellapsed pumping time), at least 830m3 of the gas was produced and accumulated in the borehole. Sand screens were installed across the perforated interval at the bottom hole for the 2008 program to overcome operational problems encountered in 2007 and achieve sustainable gas production. Stable bottom hole flowing pressures were successfully achieved during a 6 day test with continuous pump operation. Sustained gas production was achieved with rates between 2000- 4000m3/day and cummulative gas volume in the surface of approximately 13,000m3. Temperature and pressure data measured at the bottom hole and gas and water production rates gave positive evidence for the high efficiency of gas

Test plan for In Situ Vitrification Engineering-Scale Test No. 6, EG ampersand G Idaho, Inc., Job Number 318230

International Nuclear Information System (INIS)

1991-03-01

The objectives of the test included the effects of in situ vitrification on containerized sludge contained in a simulated randomly-disposed array. From this arrangement, the test results obtained the following data applicable to Idaho National Engineering Laboratory Large Field Testing: canister burst pressure and temperature, canister depressurization rate, melt encapsulation rate of the canister and the hood area plenum temperatures, pressures, compositional analyses, and flows as affected by gas releases. 10 figs., 1 tab

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

International Nuclear Information System (INIS)

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

1976-09-01

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

The enhancements and testing for the MCNPX depletion capability

International Nuclear Information System (INIS)

Fensin, M. L.; Hendricks, J. S.; Anghaie, S.

2008-01-01

depletion tools. MCNPX depletion results acceptably match the solutions of the benchmark calculations and therefore give confidence in the ability to model more complex fission systems. MCNPX depletion enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. Further capability enhancement and testing are under development in order to further improve the usefulness of the technology. (authors)

Non-Integrated Standalone Tests of APR1400 Simulator

International Nuclear Information System (INIS)

Hwang, Su Hyun; Lee, Jeong Ik; Hong, Soon Joon; Lee, Byung Chul; Seo, Jeong Gwan; Lee, Myung Soo

2007-01-01

APR1400 being developed for the construction of New Kori 3 and 4 Units has improved safety and more economical efficiency compared with previous PWR. The ESF(Engineered Safety Features) newly introduced to enhance safety are as follows: DVI (Direct Vessel Injection), Fluidic Device, IRWST (In-containment Refueling Water Storage Tank). So the transient pattern of anticipated accidents will show different characteristics from previous PWR. There are multidimensional flow phenomena like as emergency core cooling coolant bypass discharge in the downcomer, downcomer boiling, and different safety injection characteristics due to fluidic device during LBLOCA. Also there is the phenomenon of critical flow due to the open of pressurizer POSRV (Pilot Operated Safety Relief valve) connected to IRWST and safety depressurization system and the prediction of discharge flow is very important. KEPRI is developing APR1400 simulator using RELAP-RT . RELAP-RT was developed by DS and S (Data systems and Solutions) based on RELAP5/MOD3.2. The improved features of RELAP-RT to function as a simulator are as follows: Add simulator functionality - Control by simulator executive - IC snap and reset capability - Back-track snap and reset capability - Fast time capability. Fast time capability(examples) - The rate of condensation has been limited. - Fictional choking model has been developed for internal junctions. - Wall heat transfer coefficients and heat fluxes has been limited. In this study, various NISTs (Non-Integrated Standalone Tests) were performed to verify the capability of RELAP-RT as APR1400 simulator by the comparison with RELAP5/MOD3.3

Enhancing AP1000 reactor accident management capabilities for long term accidents

International Nuclear Information System (INIS)

Jiang Pingting; Liu Mengying; Duan Chengjie; Liao Yehong

2015-01-01

Passive safety actions are considered as main measures under severe accident in AP1000 power plant. However, risk is still existed. According to PSA, several probable scenarios for AP1000 nuclear power plant are analyzed in this paper with MAAP the severe accident analysis code. According to the analysis results, several deficiencies of AP1000 severe accident management are found. The long term cooling and containment depressurization capability for AP1000 power plant appear to be most important factors under such accidents. Then, several temporary strategies for AP1000 power plant are suggested, including PCCWST temporary water supply strategy after 72h, temporary injection strategy for IRWST, hydrogen relief action in fuel building, which would improve the safety of AP1000 power plant. At last, assessments of effectiveness for these strategies are performed, and the results are compared with analysis without these strategies. The comparisons showed that correct actions of these strategies would effectively prevent the accident process of AP1000 power plant. (author)

An Approach to the Flammability Testing of Aerospace Materials

Science.gov (United States)

Hirsch, David B.

2012-01-01

Presentation reviews: (1) Current approach to evaluation of spacecraft materials flammability (2) The need for and the approach to alternative routes (3) Examples of applications of the approach recommended a) Crew Module splash down b) Crew Module depressurization c) Applicability of NASA's flammability test data to other sample configurations d) Applicability of NASA's ground flammability test data to spacecraft environments

The Temperature of the Dimethylhydrazine Drops Moving in the Atmosphere after Depressurization of the Fuel Tank Rockets

Directory of Open Access Journals (Sweden)

Bulba Elena

2016-01-01

Full Text Available This work includes the results of the numerical modeling of temperature changes process of the dimethylhydrazine (DMH drops, taking into account the radial temperature gradient in the air after the depressurization of the fuel compartments rockets at high altitude. There is formulated a mathematical model describing the process of DMH drops thermal state modifying when it's moving to the Earth's surface. There is the evaluation of the influence of the characteristic size of heptyl drops on the temperature distribution. It's established that the temperatures of the small size droplets practically completely coincide with the distribution of temperature in the atmosphere at altitudes of up to 40 kilometers.

NASA Stennis Space Center Integrated System Health Management Test Bed and Development Capabilities

Science.gov (United States)

Figueroa, Fernando; Holland, Randy; Coote, David

2006-01-01

Integrated System Health Management (ISHM) is a capability that focuses on determining the condition (health) of every element in a complex System (detect anomalies, diagnose causes, prognosis of future anomalies), and provide data, information, and knowledge (DIaK)-not just data-to control systems for safe and effective operation. This capability is currently done by large teams of people, primarily from ground, but needs to be embedded on-board systems to a higher degree to enable NASA's new Exploration Mission (long term travel and stay in space), while increasing safety and decreasing life cycle costs of spacecraft (vehicles; platforms; bases or outposts; and ground test, launch, and processing operations). The topics related to this capability include: 1) ISHM Related News Articles; 2) ISHM Vision For Exploration; 3) Layers Representing How ISHM is Currently Performed; 4) ISHM Testbeds & Prototypes at NASA SSC; 5) ISHM Functional Capability Level (FCL); 6) ISHM Functional Capability Level (FCL) and Technology Readiness Level (TRL); 7) Core Elements: Capabilities Needed; 8) Core Elements; 9) Open Systems Architecture for Condition-Based Maintenance (OSA-CBM); 10) Core Elements: Architecture, taxonomy, and ontology (ATO) for DIaK management; 11) Core Elements: ATO for DIaK Management; 12) ISHM Architecture Physical Implementation; 13) Core Elements: Standards; 14) Systematic Implementation; 15) Sketch of Work Phasing; 16) Interrelationship Between Traditional Avionics Systems, Time Critical ISHM and Advanced ISHM; 17) Testbeds and On-Board ISHM; 18) Testbed Requirements: RETS AND ISS; 19) Sustainable Development and Validation Process; 20) Development of on-board ISHM; 21) Taxonomy/Ontology of Object Oriented Implementation; 22) ISHM Capability on the E1 Test Stand Hydraulic System; 23) Define Relationships to Embed Intelligence; 24) Intelligent Elements Physical and Virtual; 25) ISHM Testbeds and Prototypes at SSC Current Implementations; 26) Trailer

Determination of lubricating capabilities with a mechanical test device; Ermittlung des Verschleissschutzverhaltens mit der MPH-Apparatur

Energy Technology Data Exchange (ETDEWEB)

Krause, D.; Feldmann, D.G.; Schmidt, J. [Technische Univ. Hamburg-Harburg (Germany). Inst. fuer Produktentwicklung und Konstruktionstechnik; Padgurskas, J. [Lithuanian Univ. of Agriculture (Lithuania)

2006-02-15

This paper describes a friction and wear test in a newly developed test machine, which was developed at the TU Hamburg-Harburg to investigate the lubricating capability of hydraulic fluids. The aim of the development of the new test procedure is a better representation of the tribological contacts and effects in fluid power machinery. The investigation of the lubrication capabilities of hydraulic fluids using a line contact showed, that a distinction between different fluids regarding their lubrication capabilities can be made, using friction-, wear- and erosion tests (galling). The high reproducibility of the boundary conditions during different tests was achieved by steady design modifications of the test rig and the development of a computer program for fully-automatic control of the test procedure. The developed test machine fulfils the requirements of a simple test procedure and simple shape test specimen, which could be produced from principally every type of material and production machines, existing in every company that produce fluid power components. (orig.)

Measurement of iodine released in a blowdown accident in the HTR-Modul. Final report on flow tests

International Nuclear Information System (INIS)

Zentis, A.

1993-01-01

A passive measuring device has been designed which consists of several filter cartridges of differnt length, and which is placed into the depressurization channel of the reactor. The dependence of the rate of flow through the filter on the flow rate in the depressurization channel must be known in order to be able to derive from the radioactivity deposited and measured in the filters a value indicating the total amount of iodine released. The report explains the basic principles of design of the instrument and of the experiments, and gives an interpretation of results of the flow tests in the AVA (aerodynamic testing facility) at Interatom. These flow tests have shown that it is feasible to determine the order of magnitude of iodine emissions with the given method and instrument. (orig./HP) [de

Physical modelling of LNG rollover in a depressurized container filled with water

Science.gov (United States)

Maksim, Dadonau; Denissenko, Petr; Hubert, Antoine; Dembele, Siaka; Wen, Jennifer

2015-11-01

Stable density stratification of multi-component Liquefied Natural Gas causes it to form distinct layers, with upper layer having a higher fraction of the lighter components. Heat flux through the walls and base of the container results in buoyancy-driven convection accompanied by heat and mass transfer between the layers. The equilibration of densities of the top and bottom layers, normally caused by the preferential evaporation of Nitrogen, may induce an imbalance in the system and trigger a rapid mixing process, so-called rollover. Numerical simulation of the rollover is complicated and codes require validation. Physical modelling of the phenomenon has been performed in a water-filled depressurized vessel. Reducing gas pressure in the container to levels comparable to the hydrostatic pressure in the water column allows modelling of tens of meters industrial reservoirs using a 20 cm laboratory setup. Additionally, it allows to model superheating of the base fluid layer at temperatures close the room temperature. Flow visualizations and parametric studies are presented. Results are related to outcomes of numerical modelling.

Results from the Operational Testing of the Eaton Smart Grid Capable Electric Vehicle Supply Equipment

Energy Technology Data Exchange (ETDEWEB)

Bennett, Brion [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-10-01

The Idaho National Laboratory conducted testing and analysis of the Eaton smart grid capable electric vehicle supply equipment (EVSE), which was a deliverable from Eaton for the U.S. Department of Energy FOA-554. The Idaho National Laboratory has extensive knowledge and experience in testing advanced conductive and wireless charging systems though INL’s support of the U.S. Department of Energy’s Advanced Vehicle Testing Activity. This document details the findings from the EVSE operational testing conducted at the Idaho National Laboratory on the Eaton smart grid capable EVSE. The testing conducted on the EVSE included energy efficiency testing, SAE J1772 functionality testing, abnormal conditions testing, and charging of a plug-in vehicle.

Reusable LH2 tank technology demonstration through ground test

Science.gov (United States)

Bianca, C.; Greenberg, H. S.; Johnson, S. E.

1995-01-01

The paper presents the project plan to demonstrate, by March 1997, the reusability of an integrated composite LH2 tank structure, cryogenic insulation, and thermal protection system (TPS). The plan includes establishment of design requirements and a comprehensive trade study to select the most suitable Reusable Hydrogen Composite Tank system (RHCTS) within the most suitable of 4 candidate structural configurations. The 4 vehicles are winged body with the capability to deliver 25,000 lbs of payload to a circular 220 nm, 51.6 degree inclined orbit (also 40,000 lbs to a 28.5 inclined 150 nm orbit). A prototype design of the selected RHCTS is established to identify the construction, fabrication, and stress simulation and test requirements necessary in an 8 foot diameter tank structure/insulation/TPS test article. A comprehensive development test program supports the 8 foot test article development and involves the composite tank itself, cryogenic insulation, and integrated tank/insulation/TPS designs. The 8 foot diameter tank will contain the integrated cryogenic insulation and TPS designs resulting from this development and that of the concurrent lightweight durable TPS program. Tank ground testing will include 330 cycles of LH2 filling, pressurization, body loading, depressurization, draining, and entry heating.

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

State of the art report on the materials testing capabilities for IASCC susceptibility testing at SCK-CEN

Energy Technology Data Exchange (ETDEWEB)

Bosch, R.-W.; Boydens, P.; Vankeerbergen, R.; Van Nieuwenhove, R.; Van Dyck, S

1999-08-01

An overview of the current IASCC testing facilities at the Belgian Nuclear Research Centre SCK-CEN is given. The testing techniques are reviewed, and their capabilities as well as their limitations are discussed. Possible future developments in testing techniques are discussed. IASCC is caused by a complex interaction between materials, its environment and mechanical stresses. Characterisation techniques assessing mechanical stresses as well as electrochemical and microstructural characteristics are reported on.

Testing of indoor radon-reduction techniques in basement houses having adjoining wings. Final report, August 1988-September 1989

International Nuclear Information System (INIS)

Messing, M.

1990-11-01

The report gives results of tests of indoor radon reduction techniques in 12 existing Maryland houses, with the objective of determining when basement houses with adjoining wings require active soil depressurization (ASD) treatment of both wings, and when treatment of the basement alone is sufficient. In five basement houses with adjoining slabs on grade, ASD treatment of both wings provided an incremental additional radon reduction of 0 to 5.2 pCi/L, compared to ASD treatment of either one of the slabs alone. However, basement-only treatment reduced radon to <4 pCi/L in all five houses. In six basement houses having adjoining crawl spaces, ASD treatment of both wings (including sub-liner depressurization of the crawl space) provided little additional reduction compared to basement-only treatment, when sub-slab communication was good. When communication was not good, treatment of both wings was required to achieve <4 pCi/L. Tests of one fully slab-on-grade house showed that, when there is good aggregate under the slab, a one-pipe sub-slab depressurization system can achieve <1-2 pCi/L, even when there are forced-air supply ducts under the slab

Test methods for determining asphaltene stability in crude oils

Energy Technology Data Exchange (ETDEWEB)

Asomaning, S. [Baker Petrolite, Sugar Land, TX (United States)

2001-07-01

The development of test methods for the determination of the stability of asphaltenes in crude oils was rendered necessary, due to the high cost of remediating asphaltene deposition in harsh production environments, namely the underwater systems in offshore deepwater. The Oliensis Spot Test, two saturates, aromatics, resins and asphaltenes (SARA)-based screens (the Colloidal Instability Index and Asphaltene-Resin ratio), a solvent titration method with near infrared radiation (NIR) solids detection, and live oil depressurization were used for the purposes of this study, to predict the stability of asphaltenes in crude oils with different API gravity. A complete description of the test methods was provided, and the experimental data obtained as a result was presented. Correlation with data on the deposition histories of the oils was used to validate the experimental stability data. The author discussed the effectiveness of the different tests for the prediction of the stability of asphaltenes in crude oils. The prediction of a crude oil's propensity towards asphaltene precipitation was more accurate with the Colloidal Instability Index and the solvent titration method. Live oil depressurization proved to be very effective for the prediction of the stability of asphaltenes for light oils, where most stability tests fail. tabs., 31 figs.

A simple method for environmental cell depressurization for use with an electron microscope.

Science.gov (United States)

Ogawa, Naoki; Mizokawa, Ryo; Saito, Minoru; Ishikawa, Akira

2017-12-01

With the aid of the environmental cell (EC) in electron microscopy, hydrated specimens have been observed at high resolutions that optical microscopy cannot attain. Due to the ultra-high vacuum conditions of the inner column of the electron microscope, the EC requires sealing films that are sufficiently thin to allow electron transmission and that are sufficiently tough to withstand the pressure difference between the inside and outside of the EC. However, most hydrated specimens can be observed at low vacuum because the saturated vapor pressure of water is known to be 0.02 atm at room temperature. These concepts have been used in the differential pumping system, but it is complicated and relatively expensive. In this work, we propose a simple method for depressurization of the EC using a 'balloon structure' and demonstrate the theoretical benefits and practical improvement for specimen observations in low-vacuum conditions. © The Author 2017. Published by Oxford University Press on behalf of The Japanese Society of Microscopy. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

Liquefied Gaseous Fuels Spill Test Facility: Overview of STF capabilities

International Nuclear Information System (INIS)

Gray, H.E.

1993-01-01

The Liquefied Gaseous Fuels Spill Test Facility (STF) constructed at the Department of Energy's Nevada Test Site is a basic research tool for studying the dynamics of accidental releases of various hazardous liquids. This Facility is designed to (1) discharge, at a controlled rate, a measured volume of hazardous test liquid on a prepared surface of a dry lake bed (Frenchman Lake); (2) monitor and record process operating data, close-in and downwind meteorological data, and downwind gaseous concentration levels; and (3) provide a means to control and monitor these functions from a remote location. The STF will accommodate large and small-scale testing of hazardous test fluid release rates up to 28,000 gallons per minute. Spill volumes up to 52,800 gallons are achievable. Generic categories of fluids that can be tested are cryogenics, isothermals, aerosol-forming materials, and chemically reactive. The phenomena that can be studied include source definition, dispersion, and pool fire/vapor burning. Other capabilities available at the STF include large-scale wind tunnel testing, a small test cell for exposing personnel protective clothing, and an area for developing mitigation techniques

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

International Nuclear Information System (INIS)

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

1976-09-01

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

Y-12 defense programs. Nuclear Packaging Systems testing capabilities

International Nuclear Information System (INIS)

1995-06-01

The Nuclear Packaging Systems (NPS) Department can manage/accomplish any packaging task. The NPS organization is responsible for managing the design, testing, certification, procurement, operation, refurbishment, maintenance, and disposal of packaging used to transport radioactive materials, other hazardous materials, and general cargoes on public roads and within the Oak Ridge Y-12 Plant. Additionally, the NPS Department has developed a Quality Assurance plan for all packaging, design and procurement of nonweapon shipping containers for radioactive materials, and design and procurement of performance-oriented packaging for hazardous materials. Further, the NPS Department is responsible for preparation and submittal of Safety Analysis Reports for Packaging (SARP). The NPS Department coordinates shipping container procurement and safety certification activities that have lead-times of up to two years. A Packaging Testing Capabilities Table at the Oak Ridge complex is included as a table

Overview of US fast-neutron facilities and testing capabilities

International Nuclear Information System (INIS)

Evans, E.A.; Cox, C.M.; Jackson, R.J.

1982-01-01

Rather than attempt a cataloging of the various fast neutron facilities developed and used in this country over the last 30 years, this paper will focus on those facilities which have been used to develop, proof test, and explore safety issues of fuels, materials and components for the breeder and fusion program. This survey paper will attempt to relate the evolution of facility capabilities with the evolution of development program which use the facilities. The work horse facilities for the breeder program are EBR-II, FFTF and TREAT. For the fusion program, RTNS-II and FMIT were selected

Results from the Operational Testing of the General Electric Smart Grid Capable Electric Vehicle Supply Equipment (EVSE)

Energy Technology Data Exchange (ETDEWEB)

Carlson, Richard Barney [Idaho National Lab. (INL), Idaho Falls, ID (United States); Scoffield, Don [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bennett, Brion [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2013-12-01

The Idaho National Laboratory conducted testing and analysis of the General Electric (GE) smart grid capable electric vehicle supply equipment (EVSE), which was a deliverable from GE for the U.S. Department of Energy FOA-554. The Idaho National Laboratory has extensive knowledge and experience in testing advanced conductive and wireless charging systems though INL’s support of the U.S. Department of Energy’s Advanced Vehicle Testing Activity. This document details the findings from the EVSE operational testing conducted at the Idaho National Laboratory on the GE smart grid capable EVSE. The testing conducted on the EVSE included energy efficiency testing, SAE J1772 functionality testing, abnormal conditions testing, and charging of a plug-in vehicle.

SSE software test management STM capability: Using STM in the Ground Systems Development Environment (GSDE)

Science.gov (United States)

Church, Victor E.; Long, D.; Hartenstein, Ray; Perez-Davila, Alfredo

1992-01-01

This report is one of a series discussing configuration management (CM) topics for Space Station ground systems software development. It provides a description of the Software Support Environment (SSE)-developed Software Test Management (STM) capability, and discusses the possible use of this capability for management of developed software during testing performed on target platforms. This is intended to supplement the formal documentation of STM provided by the SEE Project. How STM can be used to integrate contractor CM and formal CM for software before delivery to operations is described. STM provides a level of control that is flexible enough to support integration and debugging, but sufficiently rigorous to insure the integrity of the testing process.

Large-Scale Testing and High-Fidelity Simulation Capabilities at Sandia National Laboratories to Support Space Power and Propulsion

International Nuclear Information System (INIS)

Dobranich, Dean; Blanchat, Thomas K.

2008-01-01

Sandia National Laboratories, as a Department of Energy, National Nuclear Security Agency, has major responsibility to ensure the safety and security needs of nuclear weapons. As such, with an experienced research staff, Sandia maintains a spectrum of modeling and simulation capabilities integrated with experimental and large-scale test capabilities. This expertise and these capabilities offer considerable resources for addressing issues of interest to the space power and propulsion communities. This paper presents Sandia's capability to perform thermal qualification (analysis, test, modeling and simulation) using a representative weapon system as an example demonstrating the potential to support NASA's Lunar Reactor System

Integrated corridor management initiative : demonstration phase evaluation - Dallas technical capability analysis test plan.

Science.gov (United States)

This report presents the test plan for conducting the Technical Capability Analysis for the United States : Department of Transportation (U.S. DOT) evaluation of the Dallas U.S. 75 Integrated Corridor : Management (ICM) Initiative Demonstration. The ...

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

International Nuclear Information System (INIS)

L'Homme, A.

1987-06-01

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

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

International Nuclear Information System (INIS)

Feldman, E.M.; Patton, M.L. Jr.; Sackett, K.E.

1977-02-01

Recorded test data are presented for Test S-05-1 of the Semiscale Mod-1 alternate ECC injection 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-05-1 was conducted from initial conditions of 2263 psia and 544 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 cold leg broken loop piping. During the test, cooling water was injected into the vessel lower plenum to simulate emergency core coolant injection in a PWR, with the flow rate based on system volume scaling

Mudstone depressurization behaviour in an open pit coal mine, Indonesia

Energy Technology Data Exchange (ETDEWEB)

Marchand, G.; Waterhouse, J. [Golder Associates, West Perth, WA (Australia); Crisostomo, J. [PT Adaro Indonesia, Jakarta (Indonesia)

2010-07-01

Mining activities in the Tutupan mine in Indonesia began in the mid-1990s. The open pit mine's coal seams are interbedded with fine-grained sandstones, mudstones, and carbonaceous mudstones. Slope stability analyses at the pit have integrated hydrogeology with geotechnical engineering analyses to optimize slope designs and reduce the risk of slope failure. This paper discussed the impact of mining and dewatering on mudstone depressurization. Sensors were placed at key points in the mine to obtain data related to the mudstone units. Reductions in pore pressure occurred as a result of groundwater flow away from the observed zones, increases in porosity, and increases in total porosity caused by an expansion of the rock mass as a result of drainage and hydrostatic unloading. Mudstone pore pressure trends with time were interpreted by determining the thickness of the mudstone unit, the presence or absence of known thin sandstone beds, unloading from overhead mining activities, and the position of the mudstone within the sedimentary sequence. The study showed that unloading activities have a significant impact on pore pressure in thick mudstone units, regardless of the depth, thickness, or properties of the unit. Pore pressure within high wall mudstone units typically decreased to values equivalent to the elevation of the unit where it was exposed to dips in a high wall. The dewatering of sandstone units in low walls caused a decline in pore pressure within the thick mudstone units located beneath the sandstones. Differences in primary permeabilities were attributed to greater fracturing in deeper and stronger rock units. 3 refs., 4 figs.

Risk Management Program Application for the Component Test Capability

International Nuclear Information System (INIS)

Stephanie L. Austad; Jeffrey D. Bryan

2009-01-01

This paper documents the application of the risk management program requirements to Component Test Capability (CTC) Project activities for each CTC alternative. In particular, DOE O 413.3A, 'Program and Project Management for the Acquisition of Capital Assets,' and DOE G 413.3-7, 'Risk Management Guide for Project Management,' will apply in the event that Alternative 4, Single, Standalone Component Test Facility (CTF), is selected and approved. As such, it is advisable to begin planning to meet the associated Department of Energy (DOE) requirements and guidance as early in the acquisition process as practicable. This white paper is intended to assist in this planning and to support associated decision-making activities. Nontechnical risks associated with each alternative will be identified to support the Next Generation Nuclear Plant (NGNP) CTC alternatives analysis. Technical risks are assumed to be addressed through the Technology Development Risk Management modeling process and are inherent to the alternatives

A New High-Speed, High-Cycle, Gear-Tooth Bending Fatigue Test Capability

Science.gov (United States)

Stringer, David B.; Dykas, Brian D.; LaBerge, Kelsen E.; Zakrajsek, Andrew J.; Handschuh, Robert F.

2011-01-01

A new high-speed test capability for determining the high cycle bending-fatigue characteristics of gear teeth has been developed. Experiments were performed in the test facility using a standard spur gear test specimens designed for use in NASA Glenn s drive system test facilities. These tests varied in load condition and cycle-rate. The cycle-rate varied from 50 to 1000 Hz. The loads varied from high-stress, low-cycle loads to near infinite life conditions. Over 100 tests were conducted using AISI 9310 steel spur gear specimen. These results were then compared to previous data in the literature for correlation. Additionally, a cycle-rate sensitivity analysis was conducted by grouping the results according to cycle-rate and comparing the data sets. Methods used to study and verify load-path and facility dynamics are also discussed.

Testing an integrated model of operations capabilities An empirical study of Australian airlines

NARCIS (Netherlands)

Nand, Alka Ashwini; Singh, Prakash J.; Power, Damien

2013-01-01

Purpose - The purpose of this paper is to test the integrated model of operations strategy as proposed by Schmenner and Swink to explain whether firms trade-off or accumulate capabilities, taking into account their positions relative to their asset and operating frontiers.

Sandia Laboratories technical capabilities. Auxiliary capabilities: environmental health information science

International Nuclear Information System (INIS)

1975-09-01

Sandia Laboratories is an engineering laboratory in which research, development, testing, and evaluation capabilities are integrated by program management for the generation of advanced designs. In fulfilling its primary responsibility to ERDA, Sandia Laboratories has acquired extensive research and development capabilities. The purpose of this series of documents is to catalog the many technical capabilities of the Laboratories. After the listing of capabilities, supporting information is provided in the form of highlights, which show applications. This document deals with auxiliary capabilities, in particular, environmental health and information science. (11 figures, 1 table) (RWR)

Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

Energy Technology Data Exchange (ETDEWEB)

Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

2012-02-01

The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

Thirteen years test experience with short-circuit withstand capability of large power transformers

NARCIS (Netherlands)

Smeets, R.P.P.; Paske, te L.H.; Leufkens, P.P.; Fogelberg, T.

2009-01-01

The ability to withstand a short circuit is recognised more and more as an essential characteristic of power transformers. IEC and IEEE Standards, as well as other national standards specify short-circuit testing and how to check the withstand capability. Unfortunately, however, there is extensive

Integrated corridor management initiative : demonstration phase evaluation, San Diego technical capability analysis test plan.

Science.gov (United States)

2012-08-01

This report presents the test plan for conducting the Technical Capability Analysis for the United States Department of Transportation (U.S. DOT) evaluation of the San Diego Integrated Corridor Management (ICM) Initiative Demonstration. The ICM proje...

Numerical simulation of Class 3 hydrate reservoirs exploiting using horizontal well by depressurization and thermal co-stimulation

International Nuclear Information System (INIS)

Yang, Shengwen; Lang, Xuemei; Wang, Yanhong; Wen, Yonggang; Fan, Shuanshi

2014-01-01

Highlights: • Depressurization and thermal co-stimulation using horizontal well were proposed. • 3D stimulation showed that gas release rate was 3 × 10 5 m 3 per day within 450 days. • 2D stimulation showed that Class 3 hydrates could be dissociated within 8500 days. • 2D Simulation showed that heat flow was 1620 W lasting 1500 days, and decreased fast. • 1.1× 10 5 kg water was collected within 2000 days and then no more water was produced. - Abstract: Class 3 hydrate reservoirs exploiting using horizontal well by depressurization and thermal co-stimulation was simulated using the HydarteResSim code. Results showed that more than 20% of hydrates in the reservoirs had been dissociated within 450 days at the well temperature of 42 °C and well pressure of 0.1P 0 , 0.2P 0 (P 0 is the initial pressure of the reservoirs, simplifying 42 °C and 0.1P 0 , 42 °C and 0.2P 0 ). While the production behavior of 42 °C and 0.5P 0 , 42 °C and 0.8P 0 were not so exciting. In order to understand the production character of the well in long term, the cross section of 1 m length reservoirs was simulated. Simulation results showed that 4.5 × 10 5 m 3 gas would be collected within 4500 days and 1.1 × 10 6 kg water could be produced within 1500 days in the well at 42 °C and 0.1P 0 . 3.5 × 10 5 m 3 gas would be collected within 8500 days and 1.1 × 10 6 kg water could be produced within 1500 days in the well at 42 °C and 0.2P 0 . The heat flow was 1620 W at the beginning and then decreased rapidly in the two cases. For reservoirs of 1495.2 m in length, about 6.7 × 10 8 m 3 and 5.3 × 10 8 m 3 gas would be collected in the well corresponding to conditions of 42 °C and 0.1P 0 , and 42 °C and 0.2P 0

Development and implementation of a propeller test capability for GL-10 "Greased Lightning" propeller design

Science.gov (United States)

Duvall, Brian Edward

Interest in small unmanned aerial vehicles has increased dramatically in recent years. Hybrid vehicles which allow forward flight as a fixed wing aircraft and a true vertical landing capability have always had applications. Management of the available energy and noise associated with electric propeller propulsion systems presents many challenges. NASA Langley has developed the Greased Lightning 10 (GL-10) vertical takeoff, unmanned aerial vehicle with ten individual motors and propellers. All are used for propulsion during takeoff and contribute to acoustic noise pollution which is an identified nuisance to the surrounding users. A propeller test capability was developed to gain an understanding of how the noise can be reduced while meeting minimum thrust requirements. The designed propeller test stand allowed for various commercially available propellers to be tested for potential direct replacement of the current GL-10 propellers and also supported testing of a newly designed propeller provided by the Georgia Institute of Technology. Results from the test program provided insight as to which factors affect the noise as well as performance characteristics. The outcome of the research effort showed that the current GL-10 propeller still represents the best choice of all the candidate propellers tested.

The mechanisms of transitions from natural convection and nucleate boiling to nucleate boiling or film boiling caused by rapid depressurization in highly subcooled water

International Nuclear Information System (INIS)

Sakurai, Akira; Shiotsu, Masahiro; Hata, Koichi; Fukuda, Katsuya

1999-01-01

The mechanisms of transient boiling process including the transitions to nucleate boiling or film boiling from initial heat fluxes, q in , in natural convection and nucleate boiling regimes caused by exponentially decreasing system pressure with various decreasing periods, τ p on a horizontal cylinder in a pool of highly subcooled water were clarified. The transient boiling processes with different characteristics were divided into three groups for low and intermediate q in in natural convection regime, and for high q in in nucleate boiling regime. The transitions at maximum heat fluxes from low q in in natural convection regime to stable nucleate boiling regime occurred independently of the τ p values. The transitions from intermediate and high q in values in natural convection and nucleate boiling to stable film boiling occurred for short τ p values, although those to stable nucleate boiling occurred for tong τ p values. The CHF and corresponding surface superheat values at which the transition to film boiling occurred were considerably lower and higher than the steady-state values at the corresponding pressure during the depressurization respectively. It was suggested that the transitions to stable film boiling at transient critical heat fluxes from intermediate q in in natural convection and from high q in in nucleate boiling for short τ p occur due to explosive-like heterogeneous spontaneous nucleation (HSN). The photographs of typical vapor behavior due to the HSN during depressurization from natural convection regime for short τ p were shown. (author)

Role of fission-reactor-testing capabilities in the development of fusion technology

International Nuclear Information System (INIS)

Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Schmunk, R.E.; Takata, M.L.; Watts, K.D.

1981-01-01

Testing of fusion materials and components in fission reactors will be increasingly important in the future due to the near-term lack of fusion engineering test devices, and the long-term high demand for testing when fusion reactors become available. Fission testing is capable of filling many gaps in fusion reactor design information, and thus should be aggressively pursued. EG and G Idaho has investigated the application of fission testing in three areas, which are discussed in this paper. First, we investigated radiation damage to magnet insulators. This work is now continuing with the use of an improved test capsule. Second, a study was performed which indicated that a fission-suppressed hybrid blanket module could be effectively tested in a reactor such as the Engineering Test Reactor (ETR), closely reproducing the predicted performance in a fusion environment. Finally, we explored a conceptual design for a fission-based Integrated Test Facility (ITF), which can accommodate entire First Wall/Blanket (FW/B) modules for testing in a nuclear environment, simultaneously satisfying many of the FW/B test requirements. This ITF can provide a cyclic neutron/gamma flux, as well as the necessary module support functions

New Cryogenic Optical Test Capability at Marshall Space Flight Center's Space Optics Manufacturing Technology Center

Science.gov (United States)

Kegley, Jeff; Burdine, Robert V. (Technical Monitor)

2002-01-01

A new cryogenic optical testing capability exists at Marshall Space Flight Center's Space Optics Manufacturing Technology Center (SOMTC). SOMTC has been performing optical wavefront testing at cryogenic temperatures since 1999 in the X-ray Cryogenic Test Facility's (XRCF's) large vacuum chamber. Recently the cryogenic optical testing capability has been extended to a smaller vacuum chamber. This smaller horizontal cylindrical vacuum chamber has been outfitted with a helium-cooled liner that can be connected to the facility's helium refrigeration system bringing the existing kilowatt of refrigeration capacity to bear on a 1 meter diameter x 2 meter long test envelope. Cryogenic environments to less than 20 Kelvin are now possible in only a few hours. SOMTC's existing instruments (the Instantaneous Phase-shifting Interferometer (IPI) from ADE Phase-Shift Technologies and the PhaseCam from 4D Vision Technologies) view the optic under test through a 150 mm clear aperture BK-7 window. Since activation and chamber characterization tests in September 2001, the new chamber has been used to perform a cryogenic (less than 30 Kelvin) optical test of a 22.5 cm diameter x 127 cm radius of curvature Si02 mirror, a cryogenic survival (less than 30 Kelvin) test of an adhesive, and a cryogenic cycle (less than 20 Kelvin) test of a ULE mirror. A vibration survey has also been performed on the test chamber. Chamber specifications and performance data, vibration environment data, and limited test results will be presented.

NGNP Component Test Capability Design Code of Record

Energy Technology Data Exchange (ETDEWEB)

S.L. Austad; D.S. Ferguson; L.E. Guillen; C.W. McKnight; P.J. Petersen

2009-09-01

The Next Generation Nuclear Plant Project is conducting a trade study to select a preferred approach for establishing a capability whereby NGNP technology development testing—through large-scale, integrated tests—can be performed for critical HTGR structures, systems, and components (SSCs). The mission of this capability includes enabling the validation of interfaces, interactions, and performance for critical systems and components prior to installation in the NGNP prototype.

Experiment data report for Semiscale Mod-1 Test S-05-5 (alternate ECC injection test)

International Nuclear Information System (INIS)

Collins, B.L.; Patton, M.L. Jr.; Sackett, K.E.

1977-04-01

Recorded test data are presented for Test S-05-5 of the Semiscale Mod-1 alternate ECC injection 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-05-5 was conducted from initial conditions of 2263 psia and 537 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 cold leg broken loop piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. The upper plenum was vented through a reflood bypass line interconnecting the hot and cold legs of the broken loop

Missile Defense: Ballistic Missile Defense System Testing Delays Affect Delivery of Capabilities

Science.gov (United States)

2016-04-28

Page 1 GAO-16-339R Ballistic Missile Defense 441 G St. N.W. Washington, DC 20548 April 28, 2016 Congressional Committees Missile Defense... Ballistic Missile Defense System Testing Delays Affect Delivery of Capabilities For over half a century, the Department of Defense (DOD) has been...funding efforts to develop a system to detect, track, and defeat enemy ballistic missiles. The current system—the Ballistic Missile Defense System

Experiment data report for semiscale Mod-1 test S-06-4 (LOFT counterpart test)

International Nuclear Information System (INIS)

Gillins, R.L.; Sackett, K.E.; Coppin, C.E.

1977-12-01

Recorded test data are presented for Test S-06-4 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-06-4 was conducted from initial conditions of 15,653 kPa and 564 K 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 leg of the intact loop to simulate emergency core coolant injection in a PWR. The heater rods in the electrically heated core were operated at an axial peak power density which was 100 percent of the maximum peak power density

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

International Nuclear Information System (INIS)

Collins, B.L.; Patton, M.L. Jr.; Sackett, K.E.

1977-07-01

Recorded test data are presented for Test S-06-1 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 an hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-1 was conducted from initial conditions of 15 568 kPa and 564 K 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 leg of the intact loop to simulate emergency core coolant injection in a PWR. The heater rods in the electrically heated core were operated at an axial peak power density which was 30% of the maximum peak power density

Test of 6-in.-thick pressure vessels. Series 3: intermediate test vessel V-7

International Nuclear Information System (INIS)

Merkle, J.G.; Robinson, G.C.; Holz, P.P.; Smith, J.E.; Bryan, R.H.

1976-08-01

The test of intermediate test vessel V-7 was a crack-initiation fracture test of a 152-mm-thick (6-in.), 990-mm-OD (39-in.) vessel of ASTM A533, grade B, class 1 steel plate with a sharp outside surface flaw 457 mm (18 in.) long and about 135 mm (5.3 in.) deep. The vessel was heated to 91 0 C (196 0 F) and pressurized hydraulically until leakage through the flaw terminated the test at a peak pressure of 147 MPa (21,350 psi). Fracture toughness data obtained by testing precracked Charpy-V and compact-tension specimens machined from a prolongation of the cylindrical test shell were used in pretest analyses of the flawed vessel. The vessel, as expected, did not burst. Upon depressurization, the ruptured ligament closed so as to maintain static pressure without leakage at about 129 MPa

Experiment data report for semiscale Mod-1 Test S-06-2 (LOFT counterpart test)

International Nuclear Information System (INIS)

Patton, M.L. Jr.; Collins, B.L.; Sackett, K.E.

1977-08-01

Recorded test data are presented for Test S-06-2 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 an hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-2 was conducted from initial conditions of 15 513 kPa and 563 K 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 leg of the intact loop to simulate emergency core coolant injection in a PWR. The heater rods in the electrically heated core were operated at an axial peak power density which was 50% of the maximum peak power density

Pre-test analysis of ATLAS SBO with RCP seal leakage scenario using MARS code

Energy Technology Data Exchange (ETDEWEB)

Pham, Quang Huy; Lee, Sang Young; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-10-15

This study presents a pre-test calculation for the Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) SBO experiment with RCP seal leakage scenario. Initially, turbine-driven auxfeed water pumps are used. Then, outside cooling water injection method is used for long term cooling. The analysis results would be useful for conducting the experiment to verify the APR 1400 extended SBO optimum mitigation strategy using outside cooling water injection in future. The pre-test calculation for ATLAS extended SBO with RCP seal leakage and outside cooling water injection scenario is performed. After Fukushima nuclear accident, the capability of coping with the extended station blackout (SBO) becomes important. Many NPPs are applying FLEX approach as main coping strategies for extended SBO scenarios. In FLEX strategies, outside cooling water injection to reactor cooling system (RCS) and steam generators (SGs) is considered as an effective method to remove residual heat and maintain the inventory of the systems during the accident. It is worthwhile to examine the soundness of outside cooling water injection method for extended SBO mitigation by both calculation and experimental demonstration. From the calculation results, outside cooling water injection into RCS and SGs is verified as an effective method during extended SBO when RCS and SGs depressurization is sufficiently performed.

Improvement in post test accident analysis results prediction for the test no. 2 in PSB test facility by applying UMAE methodology

International Nuclear Information System (INIS)

Dubey, S.K.; Petruzzi, A.; Giannotti, W.; D'Auria, F.

2006-01-01

This paper mainly deals with the improvement in the post test accident analysis results prediction for the test no. 2, 'Total loss of feed water with failure of HPIS pumps and operator actions on primary and secondary circuit depressurization', carried-out on PSB integral test facility in May 2005. This is one the most complicated test conducted in PSB test facility. The prime objective of this test is to provide support for the verification of the accident management strategies for NPPs and also to verify the correctness of some safety systems operating only during accident. The objective of this analysis is to assess the capability to reproduce the phenomena occurring during the selected tests and to quantify the accuracy of the code calculation qualitatively and quantitatively for the best estimate code Relap5/mod3.3 by systematically applying all the procedures lead by Uncertainty Methodology based on Accuracy Extrapolation (UMAE), developed at University of Pisa. In order to achieve these objectives test facility nodalisation qualification for both 'steady state level' and 'on transient level' are demonstrated. For the 'steady state level' qualification compliance to acceptance criteria established in UMAE has been checked for geometrical details and thermal hydraulic parameters. The following steps have been performed for evaluation of qualitative qualification of 'on transient level': visual comparisons between experimental and calculated relevant parameters time trends; list of comparison between experimental and code calculation resulting time sequence of significant events; identification/verification of CSNI phenomena validation matrix; use of the Phenomenological Windows (PhW), identification of Key Phenomena and Relevant Thermal-hydraulic Aspects (RTA). A successful application of the qualitative process constitutes a prerequisite to the application of the quantitative analysis. For quantitative accuracy of code prediction Fast Fourier Transform Based

Monitoring system of depressurization valves of migrated gas in annular space of flexible risers

Energy Technology Data Exchange (ETDEWEB)

Mesquita, Luiz A.; Santos, Joilson M.; Carvalho, Antonio L.; Loureiro, Patricia [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil)

2005-07-01

PETROBRAS Research and Development Center - CENPES developed an automatic system for monitoring pressure of annular space due to permeation of gas in flexible risers to inspect continuously integrity of such lines. To help maintaining physical integrity of flexible risers, two PSV's are installed to end fittings on top of riser, so that operation of any valve grants the maximum admissible gas pressure within the riser annular space, as overpressure might cause damages to external polymeric layer of flexible riser. Due to the fact that there is no mechanism allowing operation to verify correct PSV performance and frequency of valve's closings and openings, we felt to be necessary the development and implement an automatic instrumented system, integrated to platform's automation and control infrastructure. The objective of this instrumentation is to monitor and register pressure of annular space in flexible riser, as well as XV's depressurization frequency. Having such information registered and monitored, can infer some riser structural conditions, anticipating repairs and preventive maintenance. In this paper we present developed system details including instruments required, application, operation of associated screens that are used in the ECOS, with events, alarms and industrial automation services required (Application development and system integration). (author)

Experiment data report for Semiscale Mod-1 Test S-05-4 (alternate ECC injection test)

International Nuclear Information System (INIS)

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

1977-03-01

Recorded test data are presented for Test S-05-4 of the Semiscale Mod-1 alternate emergency core coolant injection 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-05-4 was conducted from initial conditions of 2266 psia and 543 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 cold leg broken loop piping. During the test, cooling water was injected into the cold leg of each loop and into the vessel upper plenum to simulate emergency core coolant injection in a PWR. The upper plenum coolant injection was scaled according to the heat stored in the metal mass of the upper plenum

Analysis of the ANO-2 turbine trip test

International Nuclear Information System (INIS)

McDonald, T.A.; Tessier, J.H.; Senda, Y.; Waterman, M.D.

1983-01-01

The start-up tests performed with the Arkansas Nuclear One-Unit Two (ANO-2) plant provided an opportunity for studying the validity of certain integral systems codes. In particular, the turbine trip from 98.2 percent full power test was investigated with the RELAP5/MOD1 (cycle 18) ode. A detailed plant model was developed and used to understand the test reports. The early depressurization portion of the transient was reproduced; however, the resultant repressurization was not well represented due to uncertainty in the data and plant response. As a result of these computations and detailed analyses of the test data considerable insight was drawn as to the best way to perform and gather data from such integral systems tests for use in code verification studies

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

Experiment data report for semiscale Mod-1 test S-28-2 (steam generator tube rupture test)

International Nuclear Information System (INIS)

Patton, M.L.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-2 of the Semiscale Mod-1 steam generator tube rupture 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-28-2 was conducted from initial conditions of 15 936 kPa and 558 K 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 leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. For Test S-28-2, accumulator injection into the intact loop hot leg was provided to simulate simulate the rupture of six steam generator tubes

PKL-tests, test series IIB (end of blowdown). Vol. 2

International Nuclear Information System (INIS)

Umminger, K.; Mandl, R.; Nopper, H.; Siemens AG Unternehmensbereich KWU, Erlangen

1987-01-01

As part of the federally subsidized research project 1500 287/A0, the system behavior of a 1300 MWe pressurized water reactor (PWR) was investigated during the depressurization phase (end-of-blowdown, EOB), as well as during the refill and reflood phases of a loss of coolant accident involving a large break in the reactor coolant loop. Appropriate modifications to the system and supplementary instrumentation have made it possible to simulate the EOB (as of 26 bar), the refill phase and reflood phase in sequence. This report includes a detailed description of the instrumentation and the data acquisition system used in Test Series PKL IIB. (orig.) With 6 refs., 2 tabs., 60 figs [de

Some applications of fission-based testing capabilities in the development of fusion technology

International Nuclear Information System (INIS)

Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.; Schmunk, R.E.; Takata, M.L.; Watts, K.D.

1981-10-01

The testing of fusion materials and components in fission reactors will be increasingly important in the future due to the near-term lack of fusion engineering test devices, and the long-term high demand for fusion testing when they do become available. Fission testing is capable of filling many gaps in fusion reactor design information, and should be aggressively pursued. EG and G Idaho has investigated the application of fission testing in three areas, which are discussed in this paper. First, work was performed on the irradiation of magnet insulators. This work is continuing with an improved test environment. Second, a study was performed which indicated that a fission-suppressed hybrid blanket module could be effectively tested in a reactor such as the Engineering Test Reactor (ETR), closely reproducing the predicted performance in a fusion environment. Finally, a conceptual design is presented for a fission-based Integrated Test Facility (ITF), which can accommodate entire wall/blanket (FW/B) modules for testing in a nuclear environment, simultaneously satisfying many of the FW/B test requirements. This ITF can provide a cyclic neutron/gamma flux, as well as the necessary module support functions

FMEF/experimental capabilities

International Nuclear Information System (INIS)

Burgess, C.A.; Dronen, V.R.

1981-01-01

The Fuels and Materials Examination Facility (FMEF), under construction at the Hanford site north of Richland, Washington, will be one of the most modern facilities offering irradiated fuels and materials examination capabilities and fuel fabrication development technologies. Scheduled for completion in 1984, the FMEF will provide examination capability for fuel assemblies, fuel pins and test pins irradiated in the FFTF. Various functions of the FMEF are described, with emphasis on experimental data-gathering capabilities in the facility's Nondestructive and Destructive examination cell complex

Mobile Landing Platform with Core Capability Set (MLP w/CCS): Combined Initial Operational Test and Evaluation and Live Fire Test and Evaluation Report

Science.gov (United States)

2015-07-01

SUBTITLE Mobile Landing Platform with Core Capability Set (MLP w/CCS) Combined Initial Operational Test and Evaluation ( IOT &E) and Live Fire Test and...based on data from a series of integrated test events, a dedicated end-to-end Initial Operational Test and Evaluation ( IOT &E), and two Marine Corps...Internally Transportable Vehicles (ITVs).   ii the LMSR to anchor within a few miles of the shore. Using MLP (CCS), the equipment is transported ashore

Experiment data report for Semiscale Mod-1 test S-28-3 (steam generator tube rupture test)

International Nuclear Information System (INIS)

Gillins, R.L.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-3 of the Semiscale Mod-1 steam generator tube rupture 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-28-3 was conducted from initial conditions of 15621 kPa and 555 K 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 leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Twelve steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

Experiment data report for semiscale Mod-1 test S-28-4 (steam generator tube rupture test)

International Nuclear Information System (INIS)

Esparza, V.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-4 of the Semiscale Mod-1 steam generator tube rupture 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-28-4 was conducted from initial conditions of 15 646 kPa and 557 K 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 leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Thirty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

Definition of Capabilities Needed for a Single Event Effects Test Facility

International Nuclear Information System (INIS)

Riemer, Bernie; Gallmeier, Franz X.

2014-01-01

The Federal Aviation Administration (FAA) is contemplating new regulations mandating testing of the vulnerability of flight-critical avionics to single event effects (SEE). A limited number of high-energy neutron test facilities currently serve the SEE industrial and institutional research community. The FAA recognizes that existing facilities have insufficient test capacity to meet new demand from such mandates; it desires more flexible irradiation capabilities to test complete, large systems and would like capabilities to address greater concerns for thermal neutrons. For this reason, the FAA funded this study by Spallation Neutron Source (SNS) staff with the ultimate aim of developing options for SEE test facilities using high-energy neutrons at the SNS complex. After an investigation of current SEE test practices and assessment of future testing requirements, three concepts were identified covering a range of test functionality, neutron flux levels, and fidelity to the atmospheric neutron spectrum. The costs and times required to complete each facility were also estimated. SEE testing is generally performed by accelerating the event rate to a point where the effects are still dominated by single events and double event causes of failures are negligible. In practice, acceleration factors of as high as 10 6 are applicable for component testing, whereas for systems testing acceleration factors of 10 4 seem to be the upper limit. It is strongly desirable that the irradiation facility be tunable over a large range of high-energy neutron fluxes of 10 2 - 10 4 n/cm 2 /s for systems testing and from 10 4 - 10 7 n/cm 2 /s for components testing. The most capable, most flexible, and highest-test-capacity option is a new stand-alone target station named the High-Energy neutron Test Station (HETS). It is also the most expensive option, with a cost to complete of approximately $100 million. Dual test enclosures would allow for simultaneous testing activity effectively

Definition of Capabilities Needed for a Single Event Effects Test Facility

Energy Technology Data Exchange (ETDEWEB)

Riemer, Bernie [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Spallation Neutron Source (SNS); Gallmeier, Franz X. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Spallation Neutron Source (SNS)

2014-12-01

The Federal Aviation Administration (FAA) is contemplating new regulations mandating testing of the vulnerability of flight-critical avionics to single event effects (SEE). A limited number of high-energy neutron test facilities currently serve the SEE industrial and institutional research community. The FAA recognizes that existing facilities have insufficient test capacity to meet new demand from such mandates; it desires more flexible irradiation capabilities to test complete, large systems and would like capabilities to address greater concerns for thermal neutrons. For this reason, the FAA funded this study by Spallation Neutron Source (SNS) staff with the ultimate aim of developing options for SEE test facilities using high-energy neutrons at the SNS complex. After an investigation of current SEE test practices and assessment of future testing requirements, three concepts were identified covering a range of test functionality, neutron flux levels, and fidelity to the atmospheric neutron spectrum. The costs and times required to complete each facility were also estimated. SEE testing is generally performed by accelerating the event rate to a point where the effects are still dominated by single events and double event causes of failures are negligible. In practice, acceleration factors of as high as 10⁶ are applicable for component testing, whereas for systems testing acceleration factors of 10⁴ seem to be the upper limit. It is strongly desirable that the irradiation facility be tunable over a large range of high-energy neutron fluxes of 10² - 10⁴ n/cm²/s for systems testing and from 10⁴ - 10⁷ n/cm²/s for components testing. The most capable, most flexible, and highest-test-capacity option is a new stand-alone target station named the High-Energy neutron Test Station (HETS). It is also the most expensive option, with a cost to complete of approximately $100 million. Dual test enclosures would

Experiment data report for Semiscale Mod-1 Test S-05-3 (alternate ECC injection test)

International Nuclear Information System (INIS)

Feldman, E.M.; Patton, M.L. Jr.; Sackett, K.E.

1977-03-01

Recorded test data are presented for Test S-05-3 of the Semiscale Mod-1 alternate ECC injection 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-05-3 was conducted from initial conditions of 2263 psia and 545 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 cold leg broken loop piping. During the test, cooling water was injected into the cold leg sides of the intact and broken loops and into the vessel upper plenum to simulate emergency core coolant injection in a PWR. For Test S-05-3, specifically the effects of upper plenum coolant injection on core thermal and system response were being investigated

Depressurization accident analysis of MPBR by PBRSIM with chemical reaction model

International Nuclear Information System (INIS)

No, Hee Cheon; Kadak, A. C.

2002-01-01

The simple model for natural circulation is implemented into PBR S IM to provide air inlet velocity from the containment air space. For the friction and form loss only the pebble region is considered conservatively modeling laminar flow through a packed bed. For the chemical reaction model of PBR S IM the oxidation rate is determined as the minimum value of three mechanisms estimated at each time step: oxygen mass flow rate entering the bottom of the reflector, oxidation rate by kinetics, and oxygen mass flow rate arriving at the graphite surface by diffusion. Oxygen mass flux arriving at the graphite surface by diffusion is estimated based on energy-mass analogy. Two types of exothermic chemical reaction are considered: (C + zO 2 → xCO + yCO 2 ) and (2CO + O 2 2CO 2 ). The heterogeneous and homogeneous chemical reaction rates by kinetics are determined by INEEL and Bruno correlations, respectively. The instantaneous depressurization accident of MPBR is simulated using PBR S IM with chemical model. The air inlet velocity is initially rapidly dropped within 10 hr and reaches a saturation value of about 1.5cm/s. The oxidation rate by the diffusion process becomes lower than that by the chemical kinetics above 600K. The maximum pebble bed temperatures without and with chemical reaction reach the peak values of 1560 and 1617 .deg. C at 80 hr and 92 hr, respectively. As the averaged temperatures in the bottom reflector and the pebble bed regions increase with time, (C+1/2O2 ->CO) reaction becomes dominant over (C+O 2 →CO 2 ) reaction. Also, the CO generated by (C+1/2O 2 →CO) reaction will be consumed by (2CO+O 2 →2CO 2 ) reaction and the energy homogeneously generated by this CO depletion reaction becomes dominant over the heterogeneous reaction

Thermal-hydraulic tests for reactor safety system

International Nuclear Information System (INIS)

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

2002-05-01

Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor, APR1400 are carried out for several geometries with the B and C (Blowdown and Condensation) facility in the condition of high temperature and pressure and with a small test facility in the condition of atmospheric temperature and pressure. Tests for the critical heat flux are performed with the RCS(Reactor Coolant System) facility as well as with the Freon CHF Loop in the condition of high temperature and pressure. The atmospheric temperature and pressure facility is utilized for development of the high standard thermal hydraulic measurement technology. The optical method is developed to measure the local thermal-hydraulic behavior for the single and two-phase boiling phenomena

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

International Nuclear Information System (INIS)

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

2003-02-01

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

Fuel elements and fuel element materials. Experimental facilities for fission products lift-off tests

International Nuclear Information System (INIS)

Blanchard, R.J.; Veyrat, J.F.

1978-01-01

One of the hypothetical accidents on the HTGR primary cooling circuits is the failure of a circuit resulting in a depressurization in the primary loops of the reactor. There is a risk of release of fission products in relation to the size of the failure. Experimental facilities for HTGR tests were developed: an in pile helium loop Comedie and an out of pile helium loop

Experiment data report for Semiscale Mod-1 Test S-28-1 (steam generator tube rupture test series)

International Nuclear Information System (INIS)

Collins, B.L.; Coppin, C.E.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-1 of the Semiscale Mod-1 steam generator tube rupture 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-28-1 was conducted from initial conditions of 15 767 kPa and 557 K 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 leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Sixty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

Overview of fuel testing capabilities at the OECD Halden reactor project

Energy Technology Data Exchange (ETDEWEB)

Wiesenack, W [Institutt for Atomenergi, Halden (Norway). OECD Halden Reaktor Projekt

1994-12-31

Fuel performance and reliability investigations at the OECD Haiden Reactor Project are described. They are supported by a variety of irradiation rigs, suitable irradiation techniques and a range of instrumentation. Testing capabilities and applications are mainly aimed at exploring mechanisms of fuel behaviour and high burnup. Examples of fuel performance taken from data provided by the Halden Project for the IAEA Co-ordinated Research Programme FUMEX are presented. A number of heavily instrumented rigs to suit different test objects have been developed: base irradiation rig, gas meter rig, diameter measurement rig, ramp rig, gas flow rig, instrumented fuel assembly. In core-measurements and variety of sensors as : fuel thermocouples, bellows pressure transducers, fuel stack elongation detectors, cladding diameter gauge and cladding elongation detectors have been used. Techniques which make it possible to obtain reliable data for all relevant burnups from beginning-of-life to ultra high exposure reaching 100 Mwd/kg UO{sub 2} are described. 7 figs., 3 refs.

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

COBALT: A GN&C Payload for Testing ALHAT Capabilities in Closed-Loop Terrestrial Rocket Flights

Science.gov (United States)

Carson, John M., III; Amzajerdian, Farzin; Hines, Glenn D.; O'Neal, Travis V.; Robertson, Edward A.; Seubert, Carl; Trawny, Nikolas

2016-01-01

The COBALT (CoOperative Blending of Autonomous Landing Technology) payload is being developed within NASA as a risk reduction activity to mature, integrate and test ALHAT (Autonomous precision Landing and Hazard Avoidance Technology) systems targeted for infusion into near-term robotic and future human space flight missions. The initial COBALT payload instantiation is integrating the third-generation ALHAT Navigation Doppler Lidar (NDL) sensor, for ultra high-precision velocity plus range measurements, with the passive-optical Lander Vision System (LVS) that provides Terrain Relative Navigation (TRN) global-position estimates. The COBALT payload will be integrated onboard a rocket-propulsive terrestrial testbed and will provide precise navigation estimates and guidance planning during two flight test campaigns in 2017 (one open-loop and closed- loop). The NDL is targeting performance capabilities desired for future Mars and Moon Entry, Descent and Landing (EDL). The LVS is already baselined for TRN on the Mars 2020 robotic lander mission. The COBALT platform will provide NASA with a new risk-reduction capability to test integrated EDL Guidance, Navigation and Control (GN&C) components in closed-loop flight demonstrations prior to the actual mission EDL.

Evaluation of food emergency response laboratories' capability for 210Po analysis using proficiency test material with verifiable traceability

International Nuclear Information System (INIS)

Zhongyu Wu; Zhichao Lin; Mackill, P.; Cong Wei; Noonan, J.; Cherniack, J.; Gillis-Landrum, D.

2009-01-01

Measurement capability and data comparability are essential for emergency response when analytical data from cooperative laboratories are used for risk assessment and post incident decision making. In this study, the current capability of food emergency response laboratories for the analysis of 210 Po in water was evaluated using a proficiency test scheme in compliance with ISO-43 and ILAC G13 guidelines, which comprises a test sample preparation and verification protocol and an insightful statistical data evaluation. The results of performance evaluations on relative bias, value trueness, precision, false positive detection, minimum detection limit, and limit of quantification, are presented. (author)

BWR Full Integral Simulation Test (FIST) Phase II test results and TRAC-BWR model qualification

International Nuclear Information System (INIS)

Sutherland, W.A.; Alamgir, M.; Findlay, J.A.; Hwang, W.S.

1985-10-01

Eight matrix tests were conducted in the FIST Phase I. These tests investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. There are nine tests in Phase II of the FIST program. They include the following LOCA tests: BWR/6 LPCI line break, BWR/6 intermediate size recirculation break, and a BWR/4 large break. Steady state natural circulation tests with feedwater makeup performed at high and low pressure, and at high pressure with HPCS makeup, are included. Simulation of a transient without rod insertion, and with controlled depressurization, was performed. Also included is a simulation of the Peach Bottom turbine trip test. The final two tests simulated a failure to maintain water level during a postulated accident. A FIST program objective is to assess the TRAC code by comparisons with test data. Two post-test predictions made with TRACB04 are compared with Phase II test data in this report. These are for the BWR/6 LPCI line break LOCA, and the Peach Bottom turbine trip test simulation

The assessment of RELAP5/MOD2 based on pressurizer transient experiments

International Nuclear Information System (INIS)

Xue Hanjun; Tanrikut, A.; Menzel, R.

1992-03-01

Two typical experiments have been performed in Chinese test facility under full pressure load corresponding to typical PWRs, 1) dynamic behavior of pressurizer due to relief valve operations (Case-I) is extremely important in transients and accident conditions regarding depressurization of PWR primary system; 2) Outsurge/Insurge operation is one of the transient which is often encountered and experienced in pressurizer systems due to pressure transients in primary system of PWRs. The simulation capability of RELAP5/MOD2 is good in comparison to experimental results. The physical models (such as interface model, stratification model), playing a major role in such simulation, seems to be realistic. The effect of realistic valve modeling in depressurization simulation is extremely important. Sufficient data for relief valve (the dynamic characteristics of valve) play a major role. The time dependent junction model and the trip valve model with a reduced discharge coefficient of 0.2 give better predictions in agreement with the experiment data while the trip valve models with discharge coefficient 1.0 yield overdepressurization. The simulation of outsurge/insurge transient yields satisfactory results. The thermal non-equilibrium model is important with respect to simulation of complicated physical phenomena in outsurge/insurge transient but has a negligible effect upon the depressurization simulation. (orig./HP)

Current Status of Aerosol Generation and Measurement Facilities for the Verification Test of Containment Filtered Venting System in KAERI

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung Il; An, Sang Mo; Ha, Kwang Soon; Kim, Hwan Yeol [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this study, the design of aerosol generation and measurement systems are explained and present circumstances are also described. In addition, the aerosol test plan is shown. Containment Filtered Venting System (FCVS) is one of the safety features to reduce the amount of released fission product into the environment by depressurizing the containment. Since Chernobyl accident, the regulatory agency in several countries in Europe such as France, Germany, Sweden, etc. have been demanded the installation of the CFVS. Moreover, the feasibility study on the CFVS was also performed in U.S. After the Fukushima accident, there is a need to improve a containment venting or installation of depressurizing facility in Korea. As a part of a Ministry of Trade, Industry and Energy (MOTIE) project, KAERI has been conducted the integrated performance verification test of CFVS. As a part of the test, aerosol generation system and measurement systems were designed to simulate the fission products behavior. To perform the integrated verification test of CFVS, aerosol generation and measurement system was designed and manufactured. The component operating condition is determined to consider the severe accident condition. The test will be performed in normal conditions at first, and will be conducted under severe condition, high pressure and high temperature. Undesirable difficulties which disturb the elaborate test are expected, such as thermophoresis on the pipe, vapor condensation on aerosol, etc.

Analyses of production tests and MDT tests conducted in Mallik and Alaska methane hydrate reservoirs : what can we learn from these well tests?

Energy Technology Data Exchange (ETDEWEB)

Kurihara, M.; Funatsu, K.; Ouchi, H. [Japan Oil Engineering Co., Tokyo (Japan); Masuda, Y. [Tokyo Univ., Tokyo (Japan). School of Engineering; Yamamoto, K. [Japan Oil, Gas and Metals National Corp., Tokyo (Japan); Narita, H. [National Inst. of Advanced Industrial Science and Technology, Tokyo (Japan); Dallimore, S.R. [Natural Resources Canada, Ottawa, ON (Canada). Geological Survey of Canada; Collett, T.S. [United States Geological Survey, Reston, VA (United States); Hancock, S.H. [APA Petroleum Engineering Ltd., Calgary, AB (Canada)

2008-07-01

This paper described a series of pressure drawdown tests conducted to evaluate a modular formation dynamics tester (MDT) wireline tool. The tests were conducted at the Mallik methane hydrate (MH) reservoir as well as in MH reservoirs in Alaska over a period of several years. Production tests were also conducted to evaluate depressurization methods, and measure production and bottomhole pressure (BHP) below known MH stability pressures in order to estimate permeability and MH dissociation radius properties. The results of the tests were then history-matched using a numerical simulator. An analysis of the simulation study showed that the MDT tests were useful in estimating initial effective permeability levels in the presence of MH. However, wellbore storage erased important data used to indicate the radius of MH dissociation and effective permeability after partial MH dissociation. The study also showed that steady flow conditions must be established before obtaining solutions from history-matched production tests. Parameters accurately estimated using the MDT and production tests were outlined, and suggestions for future designs and analyses for MH reservoirs were presented. 14 refs., 7 tabs., 17 figs.

Development of a test facility for analyzing supercritical fluid blowdown

International Nuclear Information System (INIS)

Roberto, Thiago D.; Alvim, Antonio C.M.

2015-01-01

The generation IV nuclear reactors under development mostly use supercritical fluids as the working fluid because higher temperatures improve the thermal efficiency. Supercritical fluids are used by modern nuclear power plants to achieve thermal efficiencies of around 45%. With water as the supercritical working fluid, these plants operate at a high temperature and pressure. However, experiments on supercritical water are limited by technical and financial difficulties. These difficulties can be overcome by using model fluids, which have more feasible supercritical conditions and exhibit a lower critical pressure and temperature. Experimental research is normally used to determine the conditions under which model fluids represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine model fluids that can represent supercritical fluids in a transient state. This paper presents an application of fractional scale analysis to determine the simulation parameters for a depressurization test facility. Carbon dioxide (CO 2 ) and R134a gas were considered as the model fluids because their critical point conditions are more feasible than those of water. The similarities of water (prototype), CO 2 (model) and R134a (model) for depressurization in a pressure vessel were analyzed. (author)

Capabilities of the Power Burst Facility

International Nuclear Information System (INIS)

Spencer, W.A.; Jensen, A.M.; McCardell, R.K.

1982-01-01

The unique and diverse test capabilities of the Power Burst Facility (PBF) are described in this paper. The PBF test reactor, located at the Idaho National Engineering Laboratory, simulates normal, off-normal, and accident operating conditions of light water reactor fuel rods. An overview description is given, with specific detail on design and operating characteristics of the driver core, experiment test loop, fission product detection system, test train assembly facility, and support equipment which make the testing capability of the PBF so versatile

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

International Nuclear Information System (INIS)

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

1975-04-01

Recorded test data are presented for Test S-01-1 of the semiscale Mod-1 isothermal blowdown test series. Test S-01-1 is one of several semiscale Mod-1 experiments which are counterparts of the planned Loss-of-Fluid Test (LOFT) nonnuclear experiments. System hardware is representative of the LOFT design, selected using volumetric scaling methods, and initial conditions duplicate those identified for the LOFT nonnuclear tests. Test S-01-1 was conducted from an initial temperature of 540 0 F and an initial pressure of 1596 psig. A simulated intermediate size double-ended hot leg break (0.00145 ft 2 break area on each end) was used to investigate the system response to a slow depressurization transient. An orificed structure was used in the pressure vessel to simulate the LOFT core simulator. During system depressurization, coolant was injected into the vessel downcomer inlet annulus to investigate the effectiveness of injection into the inlet annulus with respect to delivery of coolant to the lower plenum. Following the blowdown portion of Test S-01-1, coolant spray was introduced into the pressure suppression tank to determine the response of the pressure suppression system. The purpose of this report is to make available the uninterpreted data from Test S-01-1 for future data analysis and test results reporting activities. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to assure that they are reasonable and consistent. (U.S.)

Testing of indoor radon reduction techniques in central Ohio houses: Phase 2 (Winter 1988-1989). Final report, September 1988-May 1989

International Nuclear Information System (INIS)

Findlay, W.O.; Robertson, A.; Scott, A.G.

1990-05-01

The report gives results of tests of developmental indoor radon reduction techniques in nine slab-on-grade and four crawl-space houses near Dayton, Ohio. The slab-on-grade tests indicated that, when there is a good layer of aggregate under the slab, the sub-slab ventilation (SSV) mitigation technique, with only one or two suction pipes, can generally reduce indoor concentrations below 2 pCi/L (86 to 99% reduction). These reductions can be achieved even when: there are forced-air supply ducts under the slab; the slab is large (up to 2600 sq ft); and the foundation walls are hollow block. Operating the SSV system in suction always gave greater reductions than did operating in pressure. The crawl-space tests demonstrated that depressurizing under a plastic liner over the crawl-space floor was able to reduce living-area radon concentrations below 2 pCi/L (81 to 96% reduction). The performance of such sub-liner depressurization gave better reductions than did crawl-space ventilation (blowing air into, or out of, the crawl space). Completely covering the crawl-space floor with plastic sheeting was not always necessary to get adequate performance

Five Tubes Rupture at Cold Side of Steam Generator Simulation Test Report Using the ATLAS

International Nuclear Information System (INIS)

Kang, Kyoung Ho; Park, Hyun Sik; Cho, Seok

2010-12-01

In this study, a postulated SGTR event of the APR1400 was experimentally investigated with the ATLAS. In order to simulate a double-ended rupture of five U-tubes in the APR1400, the SGTR-CL-02 test was performed with the ATLAS. The main objectives of this test were not only to provide a physical insight into the system response of the APR1400 during the SGTR but also to produce integral effect experimental data to validate the safety analysis code. In the present report, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Compared to the case of a single U-tube rupture test, opening frequency of the MSSVs in the intact steam generator (SG-2) was highly reduced after 500 seconds in the present SGTR-CL-02 test. Large discharge of the primary inventory resulted in rapid depressurization of the primary system and consequently early injection of the SIP. Supply of cold ECC water by the SIPs reduced the energy transfer to the secondary side compared with the single U-tube rupture case. Less heat transfer to the secondary side had more influence on the secondary pressure of the affected steam generator than the break flow. 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

Five Tubes Rupture at Hot Side of Steam Generator Simulation Test Report Using the ATLAS

International Nuclear Information System (INIS)

Kang, Kyoung Ho; Park, Hyun Sik; Cho Seok

2010-12-01

In this study, a postulated SGTR event of the APR1400 was experimentally investigated with the ATLAS. In order to simulate a double-ended rupture of five U-tubes in the APR1400, the SGTR-HL-05 test was performed with the ATLAS. The main objectives of this test were not only to provide a physical insight into the system response of the APR1400 during the SGTR but also to produce integral effect experimental data to validate the SPACE code. In the present report, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. On the contrary to the case of a single U-tube rupture test, the MSSV of the intact steam generator was not opened any more after 1500 seconds in the present SGTR-HL-05 test. Large discharge of the primary inventory resulted in rapid depressurization of the primary system and consequently early injection of the SIP. Supply of cold ECC water by the SIPs reduced the energy transfer to the secondary side compared with the single U-tube rupture case. Less heat transfer to the secondary side had more influence on the secondary pressure of the affected steam generator than the break flow. 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

In-pile cladding tests at NRI Rez and PIE capabilities and experience

International Nuclear Information System (INIS)

Zmitko, M.

2002-01-01

In-pile cladding corrosion test facilities and relevant post-irradiation capabilities at NRI Rez plc are overviewed. Basic information about the research rector LVR-15 and in-pile water loops is given. An experience in the field of Zr-alloy cladding corrosion testing and investigation of cladding corrosion behaviour is demonstrated for two experimental programmes conducted at NRI Rez in the past period. The first example describes results obtained at studying of corrosion behaviour of advanced Zr-alloys under PWR conditions with a special concern to a high lithium content and subcooled surface boiling. The second example informs about completion of the experimental programme supported by the IAEA which is focused on investigation of Zircaloy-4 cladding behaviour under VVER water chemistry, thermal-hydraulic and irradiation conditions with the main to obtain experimental data for an assessment of the Zircaloy-4 cladding compatibility with VVER conditions. (author)

Reactor coolant system hydrostatic test and risk analysis for the first AP1000 unit

International Nuclear Information System (INIS)

Cao Hongjun; Yan Xiuping

2013-01-01

The cold hydrostatic test scheme of the primary coolant circuit, of the first AP1000 unit was described. Based on the up-stream design documents, standard specifications and design technical requirements, the select principle of test boundary was identified. The design requirements for water quality, pressure, temperature and temporary hydro-test pump were proposed. A reasonable argument for heating and pressurization rate, and cooling and depressurization rate was proposed. The possible problems and risks during the hydrostatic test were analyzed. This test scheme can provide guidance for the revisions and implementations of the follow-up test procedures. It is a good reference for hydrostatic tests of AP1000 units in the future in China. (authors)

Follow-up durability measurements and mitigation-performance improvement tests in 38 Eastern Pennsylvania houses having indoor radon-reduction systems. Final report, Oct 89-Feb 90

International Nuclear Information System (INIS)

Findlay, W.O.; Robertson, A.; Scott, A.G.

1991-03-01

The report gives results of follow-up tests in 38 difficult-to-mitigate Pennsylvania houses where indoor radon reduction systems had been installed 2 to 4 years earlier. Objectives were to assess system durability, methods for improving performance, and methods for reducing installation and operating costs. The durability tests indicated that the 38 systems have not experienced any significant degradation in indoor radon levels or in system flows/suctions, except in 6 houses where system fans failed, and in houses where homeowners turned off the systems. Tests to improve performance indicated that nearly all of the elevated residual radon levels are due to re-entrainment back into the house of very-high-radon exhaust gas from the soil depressurization systems, and to radon release from well water. Tests to reduce system costs showed that premitigation sub-slab suction field measurements can help prevent installation of too many suction pipes when communication is good, but suggest a need for too many pipes when communication is poor. Soil depressurization fans could not be turned down to the extent expected in some systems that were over-designed. Between 6 and 42% of the exhausted air was withdrawn from the house

Data report of ROSA/LSTF experiment SB-HL-12. 1% hot leg break LOCA with SG depressurization and gas inflow

International Nuclear Information System (INIS)

Takeda, Takeshi

2016-01-01

An experiment SB-HL-12 was conducted on February 24, 1998 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-HL-12 simulated a 1% hot leg small-break loss-of-coolant accident in a pressurized water reactor under assumptions of total failure of high pressure injection system and non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of emergency core cooling system (ECCS). Steam generator (SG) secondary-side depressurization by fully opening the relief valves in both SGs as an accident management (AM) action was initiated immediately after maximum surface temperature of simulated fuel rod reached 600 K. Auxiliary feedwater injection into the secondary-side of both SGs was started immediately after the initiation of AM action. After the onset of AM action due to first core uncovery by core boil-off, the primary pressure decreased following the SG secondary-side pressure, causing core mixture level swell. The fuel rod surface temperature then increased up to 635 K. Second core uncovery by core boil-off took place before loop seal clearing (LSC) induced by steam condensation on ACC coolant injected into cold legs. The core liquid level recovered rapidly after the LSC. The fuel rod surface temperature then increased up to 696 K. The pressure difference became larger between the primary and SG secondary sides after the ACC tanks started to discharge nitrogen gas, which resulted in no actuation of LPI system of ECCS during the experiment. Third core uncovery by core boil-off occurred during the reflux condensation in the SG U-tubes under nitrogen gas inflow. The core power was automatically decreased by the LSTF core protection system when the maximum fuel rod surface temperature exceeded 908 K. The obtained data would be useful to define the conditions for counterpart testing of other integral test facilities to address scaling problems through thermal

KSC Technical Capabilities Website

Science.gov (United States)

Nufer, Brian; Bursian, Henry; Brown, Laurette L.

2010-01-01

This document is the website pages that review the technical capabilities that the Kennedy Space Center (KSC) has for partnership opportunities. The purpose of this information is to make prospective customers aware of the capabilities and provide an opportunity to form relationships with the experts at KSC. The technical capabilities fall into these areas: (1) Ground Operations and Processing Services, (2) Design and Analysis Solutions, (3) Command and Control Systems / Services, (4) Materials and Processes, (5) Research and Technology Development and (6) Laboratories, Shops and Test Facilities.

Pilot project - demonstration of capabilities and benefits of bridge load rating through physical testing : tech transfer summary.

Science.gov (United States)

2013-08-01

This project demonstrated the capabilities for load testing bridges in Iowa, developed and presented a webinar to local and state engineers, and produced a spreadsheet and benefit evaluation matrix that others can use to preliminarily assess where br...

Evaluation of the gravity-injection capability for core cooling after a loss-of-SDC event

International Nuclear Information System (INIS)

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

1999-01-01

In order to evaluate the gravity-drain capability to maintain core cooling after a loss-of-shutdown-cooling event during shutdown operation, the plant conditions of the Young Gwang Units 3 and 4 were reviewed. The six cases of possible gravity-drain paths using the water of the refueling water storage tank (RWST) were identified and the thermal hydraulic analyses were performed using RELAP5/MOD3.2 code. The core cooling capability was dependent on the gravity-drain paths and the drain rate. In the cases with the injection path and opening on the different leg side, the system was well depressurized after gravity-injection and the core boiling was successfully prevented for a long-term transient. However, in the cases with the injection path and opening on the cold leg side, the core coolant continued boiling although the system pressure remains atmospheric after gravity-injection because the cold water injected from the RWST was bypassed the core region. In the cases with the higher pressurizer opening than the RWST water level, the system was also pressurized by the water-hold in the pressurizer and the core was uncovered because the gravity-injection from the RWST stopped due to the high system pressure. In addition, from the sensitivity study on the gravity-injection flow rates, it was found that about 54 kg/s of RWST drain rate was required to maintain the core cooling. Those analysis results would provide useful information to operators coping with the event

Heat removal tests for pressurized water reactor containment spray by largescale facility

International Nuclear Information System (INIS)

Motoki, Y.; Hashimoto, K.; Kitani, S.; Naritomi, M.; Nishio, G.; Tanaka, M.

1983-01-01

Heat removal tests for pressurized water reactor (PWR) containment spray were carried out to investigate effectiveness of the depressurization by Japan Atomic Energy Research Institute model containment (7-m diameter, 20 m high, and 708-m 3 volume) with PWR spray nozzles. The depressurization rate is influenced by the spray heat transfer efficiency and the containment wall surface heat transfer coefficient. The overall spray heat transfer efficiency was investigated with respect to spray flow rate, weight ratio of steam/air, and spray height. The spray droplet heat transfer efficiency was investigated whether the overlapping of spray patterns gives effect or not. The effect was not detectable in the range of large value of steam/air, however, it was better in the range of small value of it. The experimental results were compared with the calculated results by computer code CONTEMPT-LT/022. The overall spray heat transfer efficiency was almost 100% in the containment pressure, ranging from 2.5 to 0.9 kg/cm 2 X G, so that the code was useful on the prediction of the thermal hydraulic behavior of containment atmosphere in a PWR accident condition

Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de despresurizacion automatica (ADS) para un reactor de agua en ebullicion (BWR) basado en RELAP

Energy Technology Data Exchange (ETDEWEB)

Ramirez G, C.; Chavez M, C., E-mail: ces.raga@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

2012-10-15

The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

Triaxial testing system for pressure core analysis using image processing technique

Science.gov (United States)

Yoneda, J.; Masui, A.; Tenma, N.; Nagao, J.

2013-11-01

In this study, a newly developed innovative triaxial testing system to investigate strength, deformation behavior, and/or permeability of gas hydrate bearing-sediments in deep sea is described. Transport of the pressure core from the storage chamber to the interior of the sealing sleeve of a triaxial cell without depressurization was achieved. An image processing technique was used to capture the motion and local deformation of a specimen in a transparent acrylic triaxial pressure cell and digital photographs were obtained at each strain level during the compression test. The material strength was successfully measured and the failure mode was evaluated under high confining and pore water pressures.

Design, implementation, and testing of a cryogenic loading capability on an engineering neutron diffractometer

Energy Technology Data Exchange (ETDEWEB)

Woodruff, T. R.; Krishnan, V. B.; Vaidyanathan, R. [Department of Mechanical, Materials, and Aerospace Engineering, Advanced Materials Processing and Analysis Center (AMPAC), University of Central Florida, Orlando, Florida 32816 (United States); Clausen, B.; Sisneros, T.; Livescu, V.; Brown, D. W.; Bourke, M. A. M. [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)

2010-06-15

A novel capability was designed, implemented, and tested for in situ neutron diffraction measurements during loading at cryogenic temperatures on the spectrometer for materials research at temperature and stress at Los Alamos National Laboratory. This capability allowed for the application of dynamic compressive forces of up to 250 kN on standard samples controlled at temperatures between 300 and 90 K. The approach comprised of cooling thermally isolated compression platens that in turn conductively cooled the sample in an aluminum vacuum chamber which was nominally transparent to the incident and diffracted neutrons. The cooling/heat rate and final temperature were controlled by regulating the flow of liquid nitrogen in channels inside the platens that were connected through bellows to the mechanical actuator of the load frame and by heaters placed on the platens. Various performance parameters of this system are reported here. The system was used to investigate deformation in Ni-Ti-Fe shape memory alloys at cryogenic temperatures and preliminary results are presented.

Armstrong Flight Research Center Flight Test Capabilities and Opportunities for the Applications of Wireless Data Acquisition Systems

Science.gov (United States)

Hang, Richard

2015-01-01

The presentation will overview NASA Armstrong Flight Research Centers flight test capabilities, which can provide various means for flight testing of passive and active wireless sensor systems, also, it will address the needs of the wireless data acquisition solutions for the centers flight instrumentation issues such as additional weight caused by added instrumentation wire bundles, connectors, wire cables routing, moving components, etc., that the Passive Wireless Sensor Technology Workshop may help. The presentation shows the constraints and requirements that the wireless sensor systems will face in the flight test applications.

Liquid entrainment and off-take at the top of the pressurizer in the case of the actuation of safety depressurization system of APR1400

International Nuclear Information System (INIS)

Kim, Chang Hyun; No, Hee Cheon

2003-01-01

In order to determine the bleed capacity of Safety Depressurization System (SDS) of Advanced Power Reactor 1400 (APR1400) in case of Total Loss of Feed Water (TLOFW), we performed an experimental study of liquid entrainment and liquid off-take from the swelled two-phase mixture surface in a vessel. A total of 208 experimental data on the entrainment and off-take are obtained using a test vessel with the height of 2.0m and the inner diameter of 0.3m having a top break with diameter of 0.05m. Two-phase mixture levels are measured by the ultrasonic sensor within ? .77% with respect to the visual level data. Droplet entrainments are measured and compared with the existing pool entrainment data. The empirical correlation for the onset of off-take is developed in terms of the Froude number (Fr g ) at the break and non-dimensional inception height (h b /d). This correlation shows agreement with the present experimental data within ? 5%. The present off-take quality data is in agreement with Schrock's off-take quality correlation with the r.m.s. error of 15.8%. In the present experiment, droplet entrainment E fg strongly depends upon jg * /h * and is proportional to the 7 th power of jg * /h * in the same way as the off-take data

Judgmental Forecasting of Operational Capabilities

DEFF Research Database (Denmark)

Hallin, Carina Antonia; Tveterås, Sigbjørn; Andersen, Torben Juul

This paper explores a new judgmental forecasting indicator, the Employee Sensed Operational Capabilities (ESOC). The purpose of the ESOC is to establish a practical prediction tool that can provide early signals about changes in financial performance by gauging frontline employees’ sensing...... of changes in the firm’s operational capabilities. We present the first stage of the development of ESOC by applying a formative measurement approach to test the index in relation to financial performance and against an organizational commitment scale. We use distributed lag models to test whether the ESOC...

Operation of the hot test loop facilities

International Nuclear Information System (INIS)

Cheong, Moon Ki; Park, Choon Kyeong; Won, Soon Yeon; Yang, Sun Kyu; Cheong, Jang Whan; Cheon, Se Young; Song, Chul Hwa; Jeon, Hyeong Kil; Chang, Suk Kyu; Jeong, Heung Jun; Cho, Young Ro; Kim, Bok Duk; Min, Kyeong Ho

1994-12-01

The objective of this project is to obtain the available experimental data and to develop the measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics department have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within fuel bundle and to understand the characteristic of pressure drop required for improving the nuclear fuel and to develop the advanced measuring techniques. RCS Loop, which is used to measure the CHF, is presently under design and construction. B and C Loop is designed and constructed to assess the automatic depressurization safety system behavior. 4 tabs., 79 figs., 7 refs. (Author) .new

The multipurpose thermalhydraulic test facility TOPFLOW: an overview on experimental capabilities, instrumentation and results

International Nuclear Information System (INIS)

Prasser, H.M.; Beyer, M.; Carl, H.; Manera, A.; Pietruske, H.; Schuetz, P.; Weiss, F.P.

2006-01-01

A new multipurpose thermalhydraulic test facility TOPFLOW (TwO Phase FLOW) was built and put into operation at Forschungszentrum Rossendorf in 2002 and 2003. Since then, it has been mainly used for the investigation of generic and applied steady state and transient two phase flow phenomena and the development and validation of models of computational fluid dynamic (CFD) codes in the frame of the German CFD initiative. The advantage of TOPFLOW consists in the combination of a large scale of the test channels with a wide operational range both of the flow velocities as well as of the system pressures and temperatures plus finally the availability of a special instrumentation that is capable in high spatial and temporal resolving two phase flow phenomena, for example the wire-mesh sensors. (orig.)

Fuels and materials testing capabilities in Fast Flux Test Facility

International Nuclear Information System (INIS)

Baker, R.B.; Chastain, S.A.; Culley, G.E.; Ethridge, J.L.; Lovell, A.J.; Newland, D.J.; Pember, L.A.; Puigh, R.J.; Waltar, A.E.

1989-01-01

The Fast Flux Test Facility (FFTF) reactor, which started operating in 1982, is a 400 MWt sodium-cooled fast neutron reactor located in Hanford, Washington State, and operated by Westinghouse Hanford Co. under contract with U.S. Department of Energy. The reactor has a wide variety of functions for irradiation tests and special tests, and its major purpose is the irradiation of fuel and material for liquid metal reactor, nuclear reactor and space reactor projects. The review first describes major technical specifications and current conditions of the FFTF reactor. Then the plan for irradiation testing is outlined focusing on general features, fuel pin/assembly irradiation tests, and absorber irradiation tests. Assemblies for special tests include the material open test assembly (MOTA), fuel open test assembly (FOTA), closed loop in-reactor assembly (CLIRA), and other special fuel assemblies. An interim examination and maintenance cell (FFTF/IEM cell) and other hot cells are used for nondestructive/destructive tests and physical/mechanical properties test of material after irradiation. (N.K.)

Building Airport Surface HITL Simulation Capability

Science.gov (United States)

Chinn, Fay Cherie

2016-01-01

FutureFlight Central is a high fidelity, real-time simulator designed to study surface operations and automation. As an air traffic control tower simulator, FFC allows stakeholders such as the FAA, controllers, pilots, airports, and airlines to develop and test advanced surface and terminal area concepts and automation including NextGen and beyond automation concepts and tools. These technologies will improve the safety, capacity and environmental issues facing the National Airspace system. FFC also has extensive video streaming capabilities, which combined with the 3-D database capability makes the facility ideal for any research needing an immersive virtual and or video environment. FutureFlight Central allows human in the loop testing which accommodates human interactions and errors giving a more complete picture than fast time simulations. This presentation describes FFCs capabilities and the components necessary to build an airport surface human in the loop simulation capability.

LOFT Augmented Operator Capability Program

International Nuclear Information System (INIS)

Hollenbeck, D.A.; Krantz, E.A.; Hunt, G.L.; Meyer, O.R.

1980-01-01

The outline of the LOFT Augmented Operator Capability Program is presented. This program utilizes the LOFT (Loss-of-Fluid Test) reactor facility which is located at the Idaho National Engineering Laboratory and the LOFT operational transient experiment series as a test bed for methods of enhancing the reactor operator's capability for safer operation. The design of an Operational Diagnotics and Display System is presented which was backfit to the existing data acquisition computers. Basic color-graphic displays of the process schematic and trend type are presented. In addition, displays were developed and are presented which represent safety state vector information. A task analysis method was applied to LOFT reactor operating procedures to test its usefulness in defining the operator's information needs and workload

Critical Directed Energy Test and Evaluation Infrastructure Shortfalls: Results of the Directed Energy Test and Evaluation Capability Tri-Service Study Update

Science.gov (United States)

2009-06-01

Sensor H11 HPM Chamber Test Capability—Explosive Equivalent Substitute H12 HEL Irradiance & Temperature H13 HEL Near/In-Beam Path Quality H14 HPM Sensor...such things as artillery shells or UAVs and may impact the earth. Possible targets include missiles in flight or a relatively close command, control...capability is a synergy of four high priority shortfalls identified by the T-SS Update. H13 —HEL near/in-beam path quality H13 is the need for a

Scaling and design analyses of a scaled-down, high-temperature test facility for experimental investigation of the initial stages of a VHTR air-ingress accident

International Nuclear Information System (INIS)

Arcilesi, David J.; Ham, Tae Kyu; Kim, In Hun; Sun, Xiaodong; Christensen, Richard N.; Oh, Chang H.

2015-01-01

Highlights: • A 1/8th geometric-scale test facility that models the VHTR hot plenum is proposed. • Geometric scaling analysis is introduced for VHTR to analyze air-ingress accident. • Design calculations are performed to show that accident phenomenology is preserved. • Some analyses include time scale, hydraulic similarity and power scaling analysis. • Test facility has been constructed and shake-down tests are currently being carried out. - Abstract: A critical event in the safety analysis of the very high-temperature gas-cooled reactor (VHTR) is an air-ingress accident. This accident is initiated, in its worst case scenario, by a double-ended guillotine break of the coaxial cross vessel, which leads to a rapid reactor vessel depressurization. In a VHTR, the reactor vessel is located within a reactor cavity that is filled with air during normal operating conditions. Following the vessel depressurization, the dominant mode of ingress of an air–helium mixture into the reactor vessel will either be molecular diffusion or density-driven stratified flow. The mode of ingress is hypothesized to depend largely on the break conditions of the cross vessel. Since the time scales of these two ingress phenomena differ by orders of magnitude, it is imperative to understand under which conditions each of these mechanisms will dominate in the air ingress process. Computer models have been developed to analyze this type of accident scenario. There are, however, limited experimental data available to understand the phenomenology of the air-ingress accident and to validate these models. Therefore, there is a need to design and construct a scaled-down experimental test facility to simulate the air-ingress accident scenarios and to collect experimental data. The current paper focuses on the analyses performed for the design and operation of a 1/8th geometric scale (by height and diameter), high-temperature test facility. A geometric scaling analysis for the VHTR, a time

Staging Options for the Air Force’s Electronic Combat Test Capability: a Cost Analysis

Science.gov (United States)

1990-09-01

strategic in nature and completely different than daily operating decisions (20:6). Horngren , in his book Cost Accounting : A Managerial Emphasis...AFIT/GCA/LSY/90S-3 DTTC S E-191 J) C, STAGING OPTIONS FOR THE AIR FORCE’S ELECTRONIC COMBAT TEST CAPABILITY: A COST ANALYSIS THESIS Joseph J. Landino...Alternative Costs ......... 56 v AFIT/GCA/LSY/90S-3 Abstract This study’s purpose was to identify the lowest cost aircraft staging base( s ) for the Air

Testing the Financial Capability Framework: Findings from YouthSave-Impact Study Kenya.

Science.gov (United States)

Kagotho, Njeri; Ssewamala, Fred M; Patak-Pietrafesa, Michele; Byansi, William

2018-01-01

In sub-Saharan Africa (SSA), youths (23 years or younger)-who account for almost half the population-are particularly vulnerable to poverty and exclusion from financial markets and intermediaries. In addition, a significant factor in the financial instability of the region appears to be the economic functioning of its youths. In recent years, social work interventions throughout the region have focused on investing in the economic functioning of youths. This study looked at baseline data from one such intervention in Kenya (N = 3,965), using the financial capabilities framework to evaluate the factors related to youths' saving behaviors. Authors investigated the association between youths' financial literacy (that is, knowledge, socialization), financial access, and financial capabilities and savings behaviors. Results indicate that adolescents who rate themselves as financially literate and those living in close proximity to a bank are more likely to report higher capabilities. Furthermore, financial capabilities in turn partially mediate the relationship between financial literacy, access, and savings. Overall, the study's findings point to the positive effect of enhanced financial capabilities among youths and offer support for asset-based interventions targeting youths in SSA. © 2017 National Association of Social Workers.

Thermal-Hydraulic Analysis of SWAMUP Facility Using ATHLET-SC Code

Energy Technology Data Exchange (ETDEWEB)

Wang, Zidi; Cao, Zhen; Liu, Xiaojing, E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai (China)

2015-03-16

During the loss of coolant accident (LOCA) of supercritical water-cooled reactor (SCWR), the pressure in the reactor system will undergo a rapid decrease from the supercritical pressure to the subcritical condition. This process is called trans-critical transients, which is of crucial importance for the LOCA analysis of SCWR. In order to simulate the trans-critical transient, a number of system codes for SCWR have been developed up to date. However, the validation work for the trans-critical models in these codes is still missing. The test facility Supercritical WAter MUltiPurpose loop (SWAMUP) with 2 × 2 rod bundle in Shanghai Jiao Tong University (SJTU) will be applied to provide test data for code validation. Some pre-test calculations are important and necessary to show the feasibility of the experiment. In this study, trans-critical transient analysis is performed for the SWAMUP facility with the system code ATHLET-SC, which is modified in SJTU, for supercritical water system. This paper presents the system behavior, e.g., system pressure, coolant mass flow, cladding temperature during the depressurization. The effects of some important parameters such as heating power, depressurization rate on the system characteristics are also investigated in this paper. Additionally, some sensitivities study of the code models, e.g., heat transfer coefficient, critical heat flux correlation are analyzed and discussed. The results indicate that the revised system code ATHLET-SC is capable of simulating thermal-hydraulic behavior during the trans-critical transient. According to the results, the cladding temperature during the transient is kept at a low value. However, the pressure difference of the heat exchanger after depressurization could reach 6 MPa, which should be considered in the experiment.

Sensor for measurement of fuel rod gas pressure during loss-of-fluid-tests

International Nuclear Information System (INIS)

Billeter, T.R.

1979-05-01

Qualification tests have been conducted of a measurement system for determining the pressure of certain fuel rods in the loss-of-fluid-test (LOFT) reactor. Because of physical size (0.35-in. OD by 5.5-in length) and operational characteristics, an eddy current device was selected as the most promising measurement transducer for the application. The sensor must operate at pressure up to 17.2 MPa (2500 psig) and at temperatures up to 800 0 F. During the reactor transient caused by loss of coolant flow, sensor temperature and applied pressure will vary rapidly and significantly. Consequently, qualification tests included subjection of the sensor to rapid depressurization, temperature transients, and blowdowns in an autoclave, as well as to calibrations and various slow temperature cycles

Space Nuclear Facility test capability at the Baikal-1 and IGR sites Semipalatinsk-21, Kazakhstan

Science.gov (United States)

Hill, T. J.; Stanley, M. L.; Martinell, J. S.

1993-01-01

The International Space Technology Assessment Program was established 1/19/92 to take advantage of the availability of Russian space technology and hardware. DOE had two delegations visit CIS and assess its space nuclear power and propulsion technologies. The visit coincided with the Conference on Nuclear Power Engineering in Space Nuclear Rocket Engines at Semipalatinsk-21 (Kurchatov, Kazakhstan) on Sept. 22-25, 1992. Reactor facilities assessed in Semipalatinski-21 included the IVG-1 reactor (a nuclear furnace, which has been modified and now called IVG-1M), the RA reactor, and the Impulse Graphite Reactor (IGR), the CIS version of TREAT. Although the reactor facilities are being maintained satisfactorily, the support infrastructure appears to be degrading. The group assessment is based on two half-day tours of the Baikals-1 test facility and a brief (2 hr) tour of IGR; because of limited time and the large size of the tour group, it was impossible to obtain answers to all prepared questions. Potential benefit is that CIS fuels and facilities may permit USA to conduct a lower priced space nuclear propulsion program while achieving higher performance capability faster, and immediate access to test facilities that cannot be available in this country for 5 years. Information needs to be obtained about available data acquisition capability, accuracy, frequency response, and number of channels. Potential areas of interest with broad application in the U.S. nuclear industry are listed.

Pipe stress intensity factors and coupled depressurization and dynamic crack propagation. 1976 Annual report

International Nuclear Information System (INIS)

Emery, A.F.; Kobayashi, A.S.; Love, W.J.

1978-04-01

This report contains the description of predictive models for the initiation and propagation of cracks in pipes and the numerical results obtained. The initiation of the crack was studied by evaluating stress intensity factors under static conditions for a series of representative flaws. Three-dimensional static stress intensity factors were determined for quarter-elliptical cracks at the corner of a hole in an infinite plate and at the corner of a bore in a rotating disk. Semi-elliptical cracks for plates in bending and in pressurized and thermally stressed hollow cylinders were also evaluated. The stress fields, in the absence of a crack, were used in the ''alternating technique'' to compute the stress intensity factors along the crack front. Parametric studies were made to assess the effects of crack thickness, the ratio of the major and minor axes of the ellipse and the thickness of the cylinders or plates. These parametric results may be used to predict critical flaw sizes for the initiation of the running crack. The initiation and propagation of axial through cracks in pressurized pipes was studied by using an elastic-plastic finite different shell code coupled with a one-dimensional thermal-hydraulic code which computed the leakage through the crack opening and the depressurization of the fluid in the pipe. The effects of large deflections and different fluid pressure profiles were investigated. The results showed that the crack opening shape is dependent upon the fracture criterion used and upon the average pressure on the crack flaps, but not upon the specific pressure profile. The consideration of large deflections changed the opening size of the crack and through the coupling with the pipe pressures, strongly affected the crack tip speed. However, for equal crack lengths, there was little difference between calculations made for large and small deflection

The Global Modeling Test Bed - Building a New National Capability for Advancing Operational Global Modeling in the United States.

Science.gov (United States)

Toepfer, F.; Cortinas, J. V., Jr.; Kuo, W.; Tallapragada, V.; Stajner, I.; Nance, L. B.; Kelleher, K. E.; Firl, G.; Bernardet, L.

2017-12-01

NOAA develops, operates, and maintains an operational global modeling capability for weather, sub seasonal and seasonal prediction for the protection of life and property and fostering the US economy. In order to substantially improve the overall performance and accelerate advancements of the operational modeling suite, NOAA is partnering with NCAR to design and build the Global Modeling Test Bed (GMTB). The GMTB has been established to provide a platform and a capability for researchers to contribute to the advancement primarily through the development of physical parameterizations needed to improve operational NWP. The strategy to achieve this goal relies on effectively leveraging global expertise through a modern collaborative software development framework. This framework consists of a repository of vetted and supported physical parameterizations known as the Common Community Physics Package (CCPP), a common well-documented interface known as the Interoperable Physics Driver (IPD) for combining schemes into suites and for their configuration and connection to dynamic cores, and an open evidence-based governance process for managing the development and evolution of CCPP. In addition, a physics test harness designed to work within this framework has been established in order to facilitate easier like-to-like comparison of physics advancements. This paper will present an overview of the design of the CCPP and test platform. Additionally, an overview of potential new opportunities of how physics developers can engage in the process, from implementing code for CCPP/IPD compliance to testing their development within an operational-like software environment, will be presented. In addition, insight will be given as to how development gets elevated to CPPP-supported status, the pre-cursor to broad availability and use within operational NWP. An overview of how the GMTB can be expanded to support other global or regional modeling capabilities will also be presented.

Testing capabilities of Los Alamos National Laboratory for irradiated materials

International Nuclear Information System (INIS)

Maloy, S.A.; James, M.R.; Sommer, W.F.

1999-01-01

Spallation neutron sources expose materials to high energy (>100 MeV) proton and neutron spectra. Although numerous studies have investigated the effects of radiation damage in a lower energy neutron flux from fission or fusion reactors on the mechanical properties of materials, very little work has been performed on the effects that exposure to a spallation neutron spectrum has on the mechanical properties of materials. These effects can be significantly different than those observed in a fission or fusion reactor spectrum because exposure to high energy protons and neutrons produces more He and H along with the atomic displacement damage. Los Alamos National Laboratory has unique facilities to study the effects of spallation radiation damage on the mechanical properties of materials. The Los Alamos Neutron Science Center (LANSCE) has a pulsed linear accelerator which operates at 800 MeV and 1 mA. The Los Alamos Spallation Radiation Effect Facility (LASREF) located at the end of this accelerator is designed to allow the irradiation of components in a proton beam while water cooling these components and measuring their temperature. After irradiation, specimens can be investigated at hot cells located at the Chemical Metallurgy Research Building. Wing 9 of this facility contains 16 hot cells set up in two groups of eight, each having a corridor in the center to allow easy transfer of radioactive shipments into and out of the hot cells. These corridors have been used to prepare specimens for shipment to collaborating laboratories such as PNNL, ORNL, BNL, and the Paul Scherrer Institute to perform specialized testing at their hot cells. The LANL hot cells contain capabilities for opening radioactive components and testing their mechanical properties as well as preparing specimens from irradiated components

Experiment data of 200% recirculation pump discharge line break integral test run 961 with HPCS failure at ROSA-III and comparison with results of suction line break tests

International Nuclear Information System (INIS)

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

1984-03-01

This report presents the experimental data of RUN 961, a 200% double-ended break test at the recirculation pump discharge line in the ROSA-III test facility. The ROSA-III test facility is a volumetrically scaled (1/424) system of the BWR/6. The facility has the electrically heated core, the break simulator and the scaled ECCS (Emergency Core Cooling System). The MSIV (Main Steam Isolation Valve) closure and the ECCS actuation were tripped by the liquid level in the upper downcomer. The PCT (Peak Cladding Temperature) was 894 K, which was 107 K higher than a 200% pump suction line break test (RUN 926) due to the smaller depressurization rate. The effect of break location on transient LOCA phenomena was clarified by comparing the discharge and suction break tests. The whole core was quenched 71 s after LPCI actuation and the effectiveness of ECCS has been confirmed. (author)

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

Planning calculations of spray tests for the ERCOSAM-SAMARA project

Energy Technology Data Exchange (ETDEWEB)

Liang, Z. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Andreani, M. [Paul Scherrer Institut, Laboratory for Thermal-Hydraulics, Villigen (Switzerland)

2012-07-01

Within the framework of the ERCOSAM-SAMARA project, co-funded by the European Union and the Russian State Atomic Energy Corporation, planning and pre-test calculations are performed to examine sensitivity parameters that can affect the break-up (erosion) of a helium (substitute for hydrogen) layer by mitigation devices (i.e., cooler, spray, or Passive Autocatalytic Recombiner - PAR). This paper reports the GOTHIC analysis results for the spray tests to be performed in the PANDA facility. The effects of spray flow rate, temperature and injection height on depressurization, erosion of helium cloud and gas transport behavior are studied. This analysis is valuable because only a limited number of conditions will be examined in the planned experiments. The study provides a useful understanding of the interaction of spray with a stratified atmosphere. (author)

Measurement of supercritical CO2 critical flow: Effects of L/D and surface roughness

International Nuclear Information System (INIS)

Mignot, Guillaume P.; Anderson, Mark H.; Corradini, Michael L.

2009-01-01

The use of supercritical fluids (SCF) has been proposed for advanced power systems including advanced sodium reactors, since these fluids can provide higher thermal efficiency and reduced system component size. Data characterizing the behavior of SCF during a blowdown or rapid depressurization are essential to validate certain aspects of safety analyses. This paper describes the results of an experiment to measure the critical mass flux for numerous stagnation thermodynamic conditions, geometry and outlet tube roughness. It was found that a 1D homogeneous equilibrium model (HEM) was capable of relatively good (less than 10% error) prediction of the test data.

Leak detection capability in CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Azer, N.; Barber, D.H.; Boucher, P.J. [and others

1997-04-01

This paper addresses the moisture leak detection capability of Ontario Hydro CANDU reactors which has been demonstrated by performing tests on the reactor. The tests confirmed the response of the annulus gas system (AGS) to the presence of moisture injected to simulate a pressure tube leak and also confirmed the dew point response assumed in leak before break assessments. The tests were performed on Bruce A Unit 4 by injecting known and controlled rates of heavy water vapor. To avoid condensation during test conditions, the amount of moisture which could be injected was small (2-3.5 g/hr). The test response demonstrated that the AGS is capable of detecting and annunciating small leaks. Thus confidence is provided that it would alarm for a growing pressure tube leak where the leak rate is expected to increase to kg/hr rapidly. The measured dew point response was close to that predicted by analysis.

Leak detection capability in CANDU reactors

International Nuclear Information System (INIS)

Azer, N.; Barber, D.H.; Boucher, P.J.

1997-01-01

This paper addresses the moisture leak detection capability of Ontario Hydro CANDU reactors which has been demonstrated by performing tests on the reactor. The tests confirmed the response of the annulus gas system (AGS) to the presence of moisture injected to simulate a pressure tube leak and also confirmed the dew point response assumed in leak before break assessments. The tests were performed on Bruce A Unit 4 by injecting known and controlled rates of heavy water vapor. To avoid condensation during test conditions, the amount of moisture which could be injected was small (2-3.5 g/hr). The test response demonstrated that the AGS is capable of detecting and annunciating small leaks. Thus confidence is provided that it would alarm for a growing pressure tube leak where the leak rate is expected to increase to kg/hr rapidly. The measured dew point response was close to that predicted by analysis

Thermal instability observations during ramp tests in the Studsvik R2 reactor

International Nuclear Information System (INIS)

Roennberg, G.; Kjaer-Pedersen, N.

1984-01-01

A series of ramp tests on ENC-built BWR fuel from the Big Rock Point reactor was performed in September 1982 in the Studsvik R2 Reactor. The tests involved segmented rods with a burnup of 18 MWd/KgU, and constituted part of the Fuel Performance Improvement Program sponsored by the United States Department of Energy. Rods of different designs were tested. The reference design had solid, dished pellets and was unpressurized. The alternative designs were annular pellets and sphere-pac. Some of the rods with annular pellets were prepressurized, and some were not. During the ramp tests the rod power is controlled by a helium depressurization loop which causes a strictly linear power ramp versus time. The thermal output of the test rig is measured calorimetrically, the data immediately being recorded on a strip chart and later processed by a computer. Furthermore, elongation detectors permit the immediate recording of the rod length variation versus time. For some of the rods the thermal output went constant for a fraction of a minute after reaching a certain value, then continued to rise, while the helium depressurization continued to proceed linearly with time. For the duration of this plateau of the thermal output curve the slope of the elongation detector signal was significantly higher than before, but fell back to its original value after the plateau. This observation was made only for the reference rods. None of the annular rods, with or without prepressurization, nor the sphere-pac rods, showed the effect. When observed, the effect occurred at about 40 kw/m. The effect is attributed to fission gas release rapidly being enhanced by thermal feedback. The increase in stored energy associated with the temperature rise in the fuel causes the delay in thermal output. The larger available internal volume and/or the prepressurization of the annular rods, and the lack of a distinct fuel-clad gap for the sphere-pac rods prevented the effect from occurring in those other

Defect Localization Capabilities of a Global Detection Scheme: Spatial Pattern Recognition Using Full-field Vibration Test Data in Plates

Science.gov (United States)

Saleeb, A. F.; Prabhu, M.; Arnold, S. M. (Technical Monitor)

2002-01-01

Recently, a conceptually simple approach, based on the notion of defect energy in material space has been developed and extensively studied (from the theoretical and computational standpoints). The present study focuses on its evaluation from the viewpoint of damage localization capabilities in case of two-dimensional plates; i.e., spatial pattern recognition on surfaces. To this end, two different experimental modal test results are utilized; i.e., (1) conventional modal testing using (white noise) excitation and accelerometer-type sensors and (2) pattern recognition using Electronic speckle pattern interferometry (ESPI), a full field method capable of analyzing the mechanical vibration of complex structures. Unlike the conventional modal testing technique (using contacting accelerometers), these emerging ESPI technologies operate in a non-contacting mode, can be used even under hazardous conditions with minimal or no presence of noise and can simultaneously provide measurements for both translations and rotations. Results obtained have clearly demonstrated the robustness and versatility of the global NDE scheme developed. The vectorial character of the indices used, which enabled the extraction of distinct patterns for localizing damages proved very useful. In the context of the targeted pattern recognition paradigm, two algorithms were developed for the interrogation of test measurements; i.e., intensity contour maps for the damaged index, and the associated defect energy vector field plots.

Verification of the computer code ATHLET in the framework of the external verification group ATHLET BETHSY test 9.3 - steam generator U-tube rupture with failure of the high pressure injection. Final report; Verifikation des ATHLET-Rechenprogramms im Rahmen der externen Verifikationsgruppe ATHLET BETHSY Test 9.3 - Heizrohrbruch mit Versagen der Hochdruck-Noteinspeisung. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Krepper, E.; Schaefer, F. [Forschungszentrum Rossendorf e.V. (FZR) (Germany). Inst. fuer Sicherheitsforschung

1998-08-01

In the framework of the external validation of the thermalhydraulic code ATHLET MOD 1.1 CYCLE D, which is being developed by the GRS, post test analyses of two experiments were done, which were performed at the French integral test facility BETHSY. During test 9.3 the consequences of a steam generator U-tube rupture with failure of the high pressure injection and of the auxiliary feedwater supply were investigated. As accident management measures, the depressurization of the secondary sides, first of the two intact steam generators, then of the damaged steam generator and finally the primary depressurization by opening of the pressurizer valve were performed. The results show, that the code ATHLET is able to describe the complex scenario in good accordance with the experiment. The safety relevant statement could be reproduced. Deviations, which did not impose the general results, occurred concerning the break mass flow during the depressurization of the damaged steam generator and the description of the failure of the heat transfer to the damaged steam generator. Reasons are hardly to find, because these processes are highly complex. (orig.) [Deutsch] Im Rahmen der externen Validierung des von der Gesellschaft fuer Anlagen- und Reaktorsicherheit entwickelten Stoerfallcodes ATHLET, der in der Version Mod 1.1 Cycle D vorlag, wurden zwei Experimente nachgerechnet und analysiert, die an der franzoesischen Versuchsanlage BETHSY durchgefuehrt wurden. Im Test 9.3 werden die Konsequenzen untersucht, wenn bei einem Heizrohrbruch die Hochdruckeinspeisung sowie die Not-Speisewasserversorgung der Dampferzeuger versagen und nur die Druckspeicher sowie die Niederdruckeinspeisung zur Verfuegung stehen. Als Accident Management Massnahmen wurde die sekundaere Druckentlastung und schliesslich die primaere Entlastung ueber den Druckhalter untersucht. Die Analyse kommt zu dem Ergebnis, dass der Code ATHLET in der Lage ist, dieses komplexe Szenario recht gut zu beschreiben. Die

The physical capabilities underlying timed "Up and Go" test are time-dependent in community-dwelling older women.

Science.gov (United States)

Coelho-Junior, Hélio José; Rodrigues, Bruno; Gonçalves, Ivan de Oliveira; Asano, Ricardo Yukio; Uchida, Marco Carlos; Marzetti, Emanuele

2018-04-01

Timed 'Up and Go' (TUG) has been widely used in research and clinical practice to evaluate physical function and mobility in older adults. However, the physical capabilities underlying TUG performance are not well elucidated. Therefore, the present study aimed at investigating a selection of physical capacities underlying TUG performance in community-dwelling older women. Four hundred and sixty-eight apparently healthy older women independent to perform the activities of daily living (mean age: 65.8 ± 6.0 years) were recruited from two specialized healthcare centers for older adults to participate in the study. Volunteers had their medical books reviewed and underwent evaluations of anthropometric data as well as physical and functional capacities. Pearson's correlation results indicate that TUG performance was significantly associated with upper (i.e., handgrip strength) and lower (i.e., sit-to-stand) limb muscle strength, balance (i.e., one-leg stand), lower limb muscle power (i.e., countermovement jump), aerobic capacity (i.e., 6-minute walk test), and mobility (i.e., usual and maximal walking speeds). When the analyses were performed based on TUG quartiles, a larger number of physical capabilities were associated with TUG >75% in comparison with TUG <25%. Multiple linear regression results indicate that the variability in TUG (~20%) was explained by lower limb muscle strength (13%) and power (1%), balance (4%), mobility (2%), and aerobic capacity (<1%), even after adjusted by age and age plus body mass index (BMI). However, when TUG results were added as quartiles, a decrease in the impact of physical capacities on TUG performance was determined. As a whole, our findings indicate that the contribution of physical capabilities to TUG performance is altered according to the time taken to perform the test, so that older women in the lower quartiles - indicating a higher performance - have an important contribution of lower limb muscle strength, while

Establishing a Ballistic Test Methodology for Documenting the Containment Capability of Small Gas Turbine Engine Compressors

Science.gov (United States)

Heady, Joel; Pereira, J. Michael; Ruggeri, Charles R.; Bobula, George A.

2009-01-01

A test methodology currently employed for large engines was extended to quantify the ballistic containment capability of a small turboshaft engine compressor case. The approach involved impacting the inside of a compressor case with a compressor blade. A gas gun propelled the blade into the case at energy levels representative of failed compressor blades. The test target was a full compressor case. The aft flange was rigidly attached to a test stand and the forward flange was attached to a main frame to provide accurate boundary conditions. A window machined in the case allowed the projectile to pass through and impact the case wall from the inside with the orientation, direction and speed that would occur in a blade-out event. High-peed, digital-video cameras provided accurate velocity and orientation data. Calibrated cameras and digital image correlation software generated full field displacement and strain information at the back side of the impact point.

Safety analyses for an in-pile SCWR fuel qualification test loop

Energy Technology Data Exchange (ETDEWEB)

Schulenberg, T.; Raque, M. [Karlsruhe Inst. of Tech., Karlsruhe (Germany)

2014-07-01

As a nuclear facility cooled with supercritical water has never been built nor operated in the past, the planned SCWR fuel qualification test will give the first experience with supercritical water-cooled nuclear systems in general. With a fuel inventory of almost 1 kg of UO{sub 2} with almost 20% enrichment, the supercritical pressure test section inside a low pressure, pool type research reactor needs to be cooled properly even in case of a number of postulated design basis accidents. Depressurization systems and emergency cooling systems will need to be designed with similar reliability as for a prototype reactor to ensure the integrity of barriers retaining the radioactive material. The paper reports about the safety concept and summarizes the safety analyses which have been performed in this context. (author)

Experiment data report for LOFT nonnuclear Test L1-4

International Nuclear Information System (INIS)

Batt, D.L.

1977-07-01

Test L1-4 was the fourth in a series of five nonnuclear isothermal blowdown tests conducted by the Loss of Fluid Test (LOFT) Program. Test L1-4 was the first Nuclear Regulatory Commission standard problem (International Problem No. 5 and U.S. Problem No. 7) experiment conducted at LOFT. Data from this test will be compared with predictions generated by the standard problem participants. For this test the LOFT Facility was configured to simulate a loss-of-coolant accident in a large pressurized water reactor resulting from a 200% double-ended offset shear break in a cold leg of the primary coolant system. A hydraulic core simulator assembly was installed in place of the nuclear core. The initial conditions in the primary coolant system intact loop were temperature at 279 0 C, gauge pressure at 15.65 MPa, and intact loop flow at 268.4 kg/s. During system depressurization into a simulated containment, emergency core cooling water was injected into the primary coolant system cold leg to provide data on the effects of emergency core cooling on system thermalhydraulic response

NASA Langley's AirSTAR Testbed: A Subscale Flight Test Capability for Flight Dynamics and Control System Experiments

Science.gov (United States)

Jordan, Thomas L.; Bailey, Roger M.

2008-01-01

As part of the Airborne Subscale Transport Aircraft Research (AirSTAR) project, NASA Langley Research Center (LaRC) has developed a subscaled flying testbed in order to conduct research experiments in support of the goals of NASA s Aviation Safety Program. This research capability consists of three distinct components. The first of these is the research aircraft, of which there are several in the AirSTAR stable. These aircraft range from a dynamically-scaled, twin turbine vehicle to a propeller driven, off-the-shelf airframe. Each of these airframes carves out its own niche in the research test program. All of the airplanes have sophisticated on-board data acquisition and actuation systems, recording, telemetering, processing, and/or receiving data from research control systems. The second piece of the testbed is the ground facilities, which encompass the hardware and software infrastructure necessary to provide comprehensive support services for conducting flight research using the subscale aircraft, including: subsystem development, integrated testing, remote piloting of the subscale aircraft, telemetry processing, experimental flight control law implementation and evaluation, flight simulation, data recording/archiving, and communications. The ground facilities are comprised of two major components: (1) The Base Research Station (BRS), a LaRC laboratory facility for system development, testing and data analysis, and (2) The Mobile Operations Station (MOS), a self-contained, motorized vehicle serving as a mobile research command/operations center, functionally equivalent to the BRS, capable of deployment to remote sites for supporting flight tests. The third piece of the testbed is the test facility itself. Research flights carried out by the AirSTAR team are conducted at NASA Wallops Flight Facility (WFF) on the Eastern Shore of Virginia. The UAV Island runway is a 50 x 1500 paved runway that lies within restricted airspace at Wallops Flight Facility. The

Ventilation systems analysis during tornado conditions. Progress report, January--June 1975

International Nuclear Information System (INIS)

Bennett, G.A.; Gregory, W.S.; Smith, P.R.

1975-11-01

The principal concern of this investigation is to develop the capability to simulate the dynamic effects of a tornado depressurization on a ventilation system. The basic formulation and solution of the two-zone series model ventilation subsystem is based on lumped parameter component response equations, the isothermal compression of air, and the conservation of mass. Solutions based on these assumptions are also presented for the two-zone series model with natural bypass, the two-zone series model with recirculation, and the natural branching model. A parameter study is presented comparing the effects of changes in system resistance, system capacitance, and variable tornado depressurization rates. The adaptability of the basic formulation to adiabatic compression of air and the addition of duct resistance is examined. A quasi-steady formulation is introduced and preliminary considerations of the importance of inertia are presented. Preliminary conclusions in this area indicate that inertial effects can be neglected. For relatively long ducts slow shock development appears possible. Work on the effect of tornado depressurization rates as related to shock development and on the importance of inertia effects is continuing

Structural Capability of an Organization toward Innovation Capability

DEFF Research Database (Denmark)

Nielsen, Susanne Balslev; Momeni, Mostafa

2016-01-01

The scholars in the field of strategic management have developed two major approaches for attainment of competitive advantage: an approach based on environmental opportunities, and another one based on internal capabilities of an organization. Some investigations in the last two decades have...... indicated that the advantages relying on the internal capabilities of organizations may determine the competitive position of organizations better than environmental opportunities do. Characteristics of firms shows that one of the most internal capabilities that lead the organizations to the strongest...... competitive advantage in the organizations is the innovation capability. The innovation capability is associated with other organizational capabilities, and many organizations have focused on the need to identify innovation capabilities.This research focuses on recognition of the structural aspect...

Apparatus and method for depressurizing, degassing, and affording decay of the radioactivity of weakly radioactive condensates in nuclear power plants

International Nuclear Information System (INIS)

Gross, R.; Plotz, J.

1976-01-01

Described is an apparatus for depressurizing, degassing and affording decay of weakly radioactive condensates in nuclear power plants having a turbine and a main condenser turbine wherein exhaust steam of the turbine is condensed and forms a main condensate, and includes a collecting tank for the condensate situated below the condenser. A plurality of horizontal degassing channels, each having a lateral overflow, are disposed in the upper part of the condensate collecting tank and are filled with the main condensate up to the level of the overflow. At least one feedwater preheater which is heated by bleeder steam from the turbine provides a secondary condensate. Below the overflow height of the degassing channels extend horizontal feed pipes for the secondary condensate. The feed pipes are connected to the output of pressure relieving expanding devices and are provided on their underside with discharge openings for the bubbling of the secondary condensate into the main condensate to thereby degass the main condensate. The condensate collecting tank has mutually offset partitions therein providing an adequately long path for the decay of the main and secondary condensates. The condensate which is discharged from the condensate collecting tank is returned into the cycle as feedwater. Also disclosed is a method of operating the foregoing apparatus

Performance assessment of mass flow rate measurement capability in a large scale transient two-phase flow test system

International Nuclear Information System (INIS)

Nalezny, C.L.; Chapman, R.L.; Martinell, J.S.; Riordon, R.P.; Solbrig, C.W.

1979-01-01

Mass flow is an important measured variable in the Loss-of-Fluid Test (LOFT) Program. Large uncertainties in mass flow measurements in the LOFT piping during LOFT coolant experiments requires instrument testing in a transient two-phase flow loop that simulates the geometry of the LOFT piping. To satisfy this need, a transient two-phase flow loop has been designed and built. The load cell weighing system, which provides reference mass flow measurements, has been analyzed to assess its capability to provide the measurements. The analysis consisted of first performing a thermal-hydraulic analysis using RELAP4 to compute mass inventory and pressure fluctuations in the system and mass flow rate at the instrument location. RELAP4 output was used as input to a structural analysis code SAPIV which is used to determine load cell response. The computed load cell response was then smoothed and differentiated to compute mass flow rate from the system. Comparison between computed mass flow rate at the instrument location and mass flow rate from the system computed from the load cell output was used to evaluate mass flow measurement capability of the load cell weighing system. Results of the analysis indicate that the load cell weighing system will provide reference mass flows more accurately than the instruments now in LOFT

Test of the hypothesis; a lymphoma stem cells exist which is capable of self-renewal

DEFF Research Database (Denmark)

Kjeldsen, Malene Krag

  Test of the hypothesis; a lymphoma stem cell exist which is capable of self-renewal   Malene Krag Pedersen, Karen Dybkaer, Hans E. Johnsen   The Research Laboratory, Department of Haematology, Aalborg Hospital, Århus University   Failure of current therapeutics in the treatment of diffuse large B...... and sustaining cells(1-3). My project is based on studies of stem and early progenitor cells in lymphoid cell lines from patients with advanced DLBCL. The cell lines are world wide recognised and generously provided by Dr. Hans Messner and colleagues.   Hypothesis and aims: A lymphoma stem and progenitor cell...

Graphical Visualization of Human Exploration Capabilities

Science.gov (United States)

Rodgers, Erica M.; Williams-Byrd, Julie; Arney, Dale C.; Simon, Matthew A.; Williams, Phillip A.; Barsoum, Christopher; Cowan, Tyler; Larman, Kevin T.; Hay, Jason; Burg, Alex

2016-01-01

NASA's pioneering space strategy will require advanced capabilities to expand the boundaries of human exploration on the Journey to Mars (J2M). The Evolvable Mars Campaign (EMC) architecture serves as a framework to identify critical capabilities that need to be developed and tested in order to enable a range of human exploration destinations and missions. Agency-wide System Maturation Teams (SMT) are responsible for the maturation of these critical exploration capabilities and help formulate, guide and resolve performance gaps associated with the EMC-identified capabilities. Systems Capability Organization Reporting Engine boards (SCOREboards) were developed to integrate the SMT data sets into cohesive human exploration capability stories that can be used to promote dialog and communicate NASA's exploration investments. Each SCOREboard provides a graphical visualization of SMT capability development needs that enable exploration missions, and presents a comprehensive overview of data that outlines a roadmap of system maturation needs critical for the J2M. SCOREboards are generated by a computer program that extracts data from a main repository, sorts the data based on a tiered data reduction structure, and then plots the data according to specified user inputs. The ability to sort and plot varying data categories provides the flexibility to present specific SCOREboard capability roadmaps based on customer requests. This paper presents the development of the SCOREboard computer program and shows multiple complementary, yet different datasets through a unified format designed to facilitate comparison between datasets. Example SCOREboard capability roadmaps are presented followed by a discussion of how the roadmaps are used to: 1) communicate capability developments and readiness of systems for future missions, and 2) influence the definition of NASA's human exploration investment portfolio through capability-driven processes. The paper concludes with a description

Survey of industrial coal conversion equipment capabilities: valves

Energy Technology Data Exchange (ETDEWEB)

Bush, W. A.; Slade, E. C.

1978-06-01

A survey of the industrial capabilities of the valve and valve-actuator industry to supply large, high-pressure stop valves for the future coal conversion industry is presented in this report. Also discussed are development and testing capabilities of valve and valve-actuator manufacturers and anticipated lead times required to manufacture advanced design valves for the most stringent service applications. Results indicate that the valve and valve-actuator industry is capable of manufacturing in quantity equipment of the size and for the pressure and temperature ranges which would be required in the coal conversion industry. Valve manufacturers do not, however, have sufficient product application experience to predict the continuing functional ability of valves used for lock-hopper feeders, slurry feeders, and slag-char letdown service. Developmental and testing efforts to modify existing valve designs or to develop new valve concepts for these applications were estimated to range from 1 to 6 years. A testing facility to simulate actuation of critical valves under service conditions would be beneficial.

Conceptual Model of IT Infrastructure Capability and Its Empirical Justification

Institute of Scientific and Technical Information of China (English)

QI Xianfeng; LAN Boxiong; GUO Zhenwei

2008-01-01

Increasing importance has been attached to the value of information technology (IT) infrastructure in today's organizations. The development of efficacious IT infrastructure capability enhances business performance and brings sustainable competitive advantage. This study analyzed the IT infrastructure capability in a holistic way and then presented a concept model of IT capability. IT infrastructure capability was categorized into sharing capability, service capability, and flexibility. This study then empirically tested the model using a set of survey data collected from 145 firms. Three factors emerge from the factor analysis as IT flexibility, IT service capability, and IT sharing capability, which agree with those in the conceptual model built in this study.

RELAP5 analysis of PKL, main steam line break test

Energy Technology Data Exchange (ETDEWEB)

Jonnet, J.R.; Stempniewicz, M.M., E-mail: stempniewicz@nrg.eu; With, A. de; Wakker, P.H.

2013-12-15

Highlights: • RELAP5/MOD 3.2 code validation is performed by analyzing a main steam line break test in the PKL large scale test facility. • The RELAP5 model reproduces well the important events of the PKL test. • RELAP5 transient results show noticeable sensitivity to small differences in the initial conditions. • Accurate prediction of the coolant temperature is essential for the assessment of potential core re-criticality. - Abstract: PKL is a large scale test facility of the primary system owned by AREVA NP GmbH. It is used for extensive experimental investigations to study the integral behavior of Pressurized Water Reactor (PWR) plants under accident conditions. Since 2001, the test program is a part of an international cooperation project (SETH, followed by PKL1 and PKL2) set up by the OECD. The aim of the present work was to perform a short validation study of the thermo-hydraulics code RELAP5. A model of the PKL test facility has been developed, tested and applied to one of the experiments performed at the PKL. The chosen experiment was the test G3.1. In that experiment, a main steam line break occurs, causing a rapid depressurization of the affected steam generator. This leads to an increase of the heat transfer from the primary to the secondary side and thereby to a fast cool-down transient on the primary side. The main objective of this analysis was the qualification of the RELAP5 code results against heat transfer from the primary to the secondary side in both affected and intact loops, and temperatures in the primary system. The calculation results have been compared to the experimental results. It was concluded that the most important events during the test are reproduced relatively well by the model. The calculated coolant temperature in the core is higher than in the experiment. The minimum temperature is about 5% higher than measured. The secondary pressures in SG-1, 3, and 4 is in very good agreement with the experimental value, but in the

Upgrading of TREAT experimental capabilities

International Nuclear Information System (INIS)

Dickerman, C.E.; Rose, D.; Bhattacharyya, S.K.

1982-01-01

The TREAT facility at the Argonne National Laboratory site in the Idaho National Engineering Laboratory is being upgraded to provide capabilities for fast-reactor-safety transient experiments not possible at any other experimental facility. Principal TREAT Upgrade (TU) goal is provision for 37-pin size experiments on energetics of core-disruptive accidents (CDA) in fast breeder reactor cores with moderate sodium void coefficients. this goal requires a significant enhancement of the capabilities of the TREAT facility, specifically including reactor control, hardened neutron spectrum incident on the test sample, and enlarged building. The upgraded facility will retain the capability for small-size experiments of the types currently being performed in TREAT. Reactor building and crane upgrading have been completed. TU schedules call for the components of the upgraded reactor system to be finished in 1984, including upgraded TREAT fuel and control system, and expanded coverage by the hodoscope fuel-motion diagnostics system

Recent Investments by NASA's National Force Measurement Technology Capability

Science.gov (United States)

Commo, Sean A.; Ponder, Jonathan D.

2016-01-01

The National Force Measurement Technology Capability (NFMTC) is a nationwide partnership established in 2008 and sponsored by NASA's Aeronautics Evaluation and Test Capabilities (AETC) project to maintain and further develop force measurement capabilities. The NFMTC focuses on force measurement in wind tunnels and provides operational support in addition to conducting balance research. Based on force measurement capability challenges, strategic investments into research tasks are designed to meet the experimental requirements of current and future aerospace research programs and projects. This paper highlights recent and force measurement investments into several areas including recapitalizing the strain-gage balance inventory, developing balance best practices, improving calibration and facility capabilities, and researching potential technologies to advance balance capabilities.

Evaluating Self Healing Capability of Bituminous Mastics

NARCIS (Netherlands)

Qiu, J.; Van de Ven, M.; Wu, S.; Yu, J.; Molenaar, A.

2012-01-01

The self-healing capability of bituminous materials has been known for many years. Researches were mostly focused on the self healing behaviour during load repetitions. The tests are either time consuming and/or complex. In this paper, a simple self healing test procedure is presented combining the

Determination of the replacement cooling tower capability at the ETRR-2 research reactor

International Nuclear Information System (INIS)

El-Din El-Morshdy, S.

2004-01-01

The ETRR-2 replacement cooling tower capability has been evaluated by the thermal acceptance test performed in June 2003. All instruments used were calibrated prior to the test. The measured data are collected at regular intervals in accordance with the acceptance test code for water cooling towers of the cooling tower institute recommendations. Both the characteristic curve and the performance curve methods were used to evaluate the tower capability. The test results yield a tower capability of about 105% and so the tower is thermally accepted. (orig.)

Current limiting capability of diffused resistors

International Nuclear Information System (INIS)

Shedd, W.; Cappelli, J.

1979-01-01

An experimental evaluation of the current limiting capability of dielectrically isolated diffused resistors at transient ionizing dose rates up to 6*10 12 rads(Si)/sec is presented. Existing theoretical predictions of the transient response of diffused resistors are summarized and compared to the experimentally measured values. The test resistors used allow the effects of sheet resistance and geometry on the transient response to be determined. The experimental results show that typical dielectrically isolated diffused resistors maintain adequate current limiting capability for use in radiation hardened integrated circuits

Impact of Personnel Capabilities on Organizational Innovation Capability

DEFF Research Database (Denmark)

Nielsen, Susanne Balslev; Momeni, Mostafa

2016-01-01

in this rapidly changing world. This research focuses on definition of the personnel aspect of innovation capability, and proposes a conceptual model based on the scientific articles of academic literature on organisations innovation capability. This paper includes an expert based validation in three rounds...... of the Delphi method. And for the purpose of a better appreciation of the relationship dominating the factors of the model, it has distributed the questionnaire to Iranian companies in the Food industry. This research proposed a direct relationship between Innovation Capability and the Personnel Capability...

Recirculation pump suction line 2.8% break integral test at ROSA-III with HPCS failure, RUN 984

International Nuclear Information System (INIS)

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

1984-06-01

This report presents the experimental data of 2.8% suction line break test RUN 984 at ROSA-III, which was conducted as one of counterpart tests to FIST program sponsored by GE, EPRI and USNRC. The similarity study between the ROSA-III and FIST tests is on the way. The report also presents the information on the ROSA-III test facility, experiment results and the effects of the ADS flow rate and the MSIV trip level comparing with the previously conducted ROSA-III small break tests, RUNs 920 and 922. Major conclusions obtained are as follows. (1) Change of the MSIV trip level from L2 to L1 gives delay of MSIV closure and longer actuation of pressure control system in a small break LOCA. (2) Larger ADS flow gives faster depressurization rate and earlier ECCS actuation, which results in shorter fuel rod dryout period and lower PCT. (author)

Heart Rhythm Monitoring in the Constellation Lunar and Launch/Landing EVA Suit: Recommendations from an Expert Panel

Science.gov (United States)

Scheuring, Richard A.; Hamilton, Doug; Jones, Jeffrey A.; Alexander, David

2009-01-01

There are currently several physiological monitoring requirements for EVA in the Human-Systems Interface Requirements (HSIR) document. There are questions as to whether the capability to monitor heart rhythm in the lunar surface space suit is a necessary capability for lunar surface operations. Similarly, there are questions as to whether the capability to monitor heart rhythm during a cabin depressurization scenario in the launch/landing space suit is necessary. This presentation seeks to inform space medicine personnel of recommendations made by an expert panel of cardiovascular medicine specialists regarding in-suit ECG heart rhythm monitoring requirements during lunar surface operations. After a review of demographic information and clinical cases and panel discussion, the panel recommended that ECG monitoring capability as a clinical tool was not essential in the lunar space suit; ECG monitoring was not essential in the launch/landing space suit for contingency scenarios; the current hear rate monitoring capability requirement for both launch/landing and lunar space suits should be maintained; lunar vehicles should be required to have ECG monitoring capability with a minimum of 5-lead ECG for IVA medical assessments; and, exercise stress testing for astronaut selection and retention should be changed from the current 85% maximum heart rate limit to maximal, exhaustive 'symptom-limited' testing to maximize diagnostic utility as a screening tool for evaluating the functional capacity of astronauts and their cardiovascular health.

Capabilities, innovation, and overall performance in Brazilian export firms.

Directory of Open Access Journals (Sweden)

José Ednilson de Oliveira Cabral

2015-06-01

Full Text Available This article extends the current research on innovation by investigating the relationship between innovative capabilities and export firms’ overall performance. From the perspectives of the resource-based view (RBV and dynamic capability, we examine the differential and interactive effects of exploration and exploitation capabilities in product innovation for external markets and overall performance (direct and mediated by a new product. In addition, we test the moderating effect of market dynamism and the controlling effect of firm size on these relationships. Hence, the main contribution of this article is developing and empirically testing an original model, by combining these constructs that address new relationships, in an emerging country. This model was tested with data from 498 Brazilian export firms, distributed throughout all Brazilian manufacturing sectors, by firm size, and in states. The analysis was made with application of the structural equation modeling (SEM. As a result, we found support for the assumptions that exploitation capabilities influence product innovation and overall performance, whereas exploration capabilities and their interaction to exploitation capabilities influence overall performance, but not product innovation. Additionally, the relationship between exploitation capabilities and overall performance is mediated by product innovation. Unlike hypothesized, market dynamism does not moderate the relationship between product innovation and overall performance. Furthermore, firm size works as a controlling variable in the relationships analyzed. Regarding the implications for theory, this study contributes to grasp that exploitation capabilities influences a firm’s overall performance, both directly and indirectly (via product innovation, and highlights the various direct and mediatory effects of innovation on overall performance. These insights show the importance of considering the role of mediating and

Energy Systems Test Area (ESTA). Power Systems Test Facilities

Science.gov (United States)

Situ, Cindy H.

2010-01-01

This viewgraph presentation provides a detailed description of the Johnson Space Center's Power Systems Facility located in the Energy Systems Test Area (ESTA). Facilities and the resources used to support power and battery systems testing are also shown. The contents include: 1) Power Testing; 2) Power Test Equipment Capabilities Summary; 3) Source/Load; 4) Battery Facilities; 5) Battery Test Equipment Capabilities Summary; 6) Battery Testing; 7) Performance Test Equipment; 8) Battery Test Environments; 9) Battery Abuse Chambers; 10) Battery Abuse Capabilities; and 11) Battery Test Area Resources.

Culturable prokaryotic diversity of deep, gas hydrate sediments: first use of a continuous high-pressure, anaerobic, enrichment and isolation system for subseafloor sediments (DeepIsoBUG)

OpenAIRE

Parkes, R John; Sellek, Gerard; Webster, Gordon; Martin, Derek; Anders, Erik; Weightman, Andrew J; Sass, Henrik

2009-01-01

Deep subseafloor sediments may contain depressurization-sensitive, anaerobic, piezophilic prokaryotes. To test this we developed the DeepIsoBUG system, which when coupled with the HYACINTH pressure-retaining drilling and core storage system and the PRESS core cutting and processing system, enables deep sediments to be handled without depressurization (up to 25 MPa) and anaerobic prokaryotic enrichments and isolation to be conducted up to 100 MPa. Here, we describe the system and its first use...

Reactor core cooling device for nuclear power plant

International Nuclear Information System (INIS)

Tsuda, Masahiko.

1992-01-01

The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

Partnership for the Revitalization of National Wind Tunnel Force Measurement Capability

Science.gov (United States)

Rhew, Ray D.; Skelley, Marcus L.; Woike, Mark R.; Bader, Jon B.; Marshall, Timothy J.

2009-01-01

Lack of funding and lack of focus on research over the past several years, coupled with force measurement capabilities being decentralized and distributed across the National Aeronautics and Space Administration (NASA) research centers, has resulted in a significant erosion of (1) capability and infrastructure to produce and calibrate force measurement systems; (2) NASA s working knowledge of those systems; and (3) the quantity of high-quality, full-capability force measurement systems available for use in aeronautics testing. Simultaneously, and at proportional rates, the capability of industry to design, manufacture, and calibrate these test instruments has been eroding primarily because of a lack of investment by the aeronautics community. Technical expertise in this technology area is a core competency in aeronautics testing; it is highly specialized and experience-based, and it represents a niche market for only a few small precision instrument shops in the United States. With this backdrop, NASA s Aeronautics Test Program (ATP) chartered a team to examine the issues and risks associated with the problem, focusing specifically on strain- gage balances. The team partnered with the U.S. Air Force s Arnold Engineering Development Center (AEDC) to exploit their combined capabilities and take a national level government view of the problem. This paper describes the team s approach, its findings, and its recommendations, and the current status for revitalizing the government s balance capability with respect to designing, fabricating, calibrating, and using the instruments.

Capability Paternalism

NARCIS (Netherlands)

Claassen, R.J.G.|info:eu-repo/dai/nl/269266224

A capability approach prescribes paternalist government actions to the extent that it requires the promotion of specific functionings, instead of the corresponding capabilities. Capability theorists have argued that their theories do not have much of these paternalist implications, since promoting

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Assessment of Literature Related to Combustion Appliance Venting Systems

Energy Technology Data Exchange (ETDEWEB)

Rapp, Vi H. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Singer, Brett C. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Stratton, Chris [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Wray, Craig P. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

2012-06-01

In many residential building retrofit programs, air tightening to increase energy efficiency is constrained by concerns about related impacts on the safety of naturally vented combustion appliances. Tighter housing units more readily depressurize when exhaust equipment is operated, making combustion appliances more prone to backdraft or spillage. Several test methods purportedly assess the potential for depressurization-induced backdrafting and spillage, but these tests are not robustly reliable and repeatable predictors of venting performance, in part because they do not fully capture weather effects on venting performance. The purpose of this literature review is to investigate combustion safety diagnostics in existing codes, standards, and guidelines related to combustion appliances. This review summarizes existing combustion safety test methods, evaluations of these test methods, and also discusses research related to wind effects and the simulation of vent system performance. Current codes and standards related to combustion appliance installation provide little information on assessing backdrafting or spillage potential. A substantial amount of research has been conducted to assess combustion appliance backdrafting and spillage test methods, but primarily focuses on comparing short-term (stress) induced tests and monitoring results. Monitoring, typically performed over one week, indicated that combinations of environmental and house operation characteristics most conducive to combustion spillage were rare. Research, to an extent, has assessed existing combustion safety diagnostics for house depressurization, but the objectives of the diagnostics, both stress and monitoring, are not clearly defined. More research is also needed to quantify the frequency of test “failure” occurrence throughout the building stock and assess the statistical effects of weather (especially wind) on house depressurization and in turn on combustion appliance venting

Preliminary Analysis on Decay Heat Removal Capability of Helium Cooled Solid Breeder Test Blanket Module

International Nuclear Information System (INIS)

Ahn, Mu Young; Cho, Seung Yon; Kim, Duck Hoi; Lee, Eun Seok; Kim, Hyung Seok; Suh, Jae Seung; Yun, Sung Hwan; Cho, Nam Zin

2007-01-01

One of the main ITER goals is to test and validate design concepts of tritium breeding blankets relevant to DEMO or fusion power plants. Korea Helium-Cooled Solid Breeder (HCSB) Test Blanket Module (TBM) has been developed with overall objectives of achieving this goal. The TBM employs high pressure helium to cool down the First Wall (FW), Side Wall (SW) and Breeding Zone (BZ). Therefore, safety consideration is a part of the design process. Each ITER Party performing the TBM program is requested to reach a similar level of confidence in the TBM safety analysis. To meet ITER's request, Failure Mode and Effects Analysis (FMEA) studies have been performed on the TBM to identify the Postulated Initial Event (PIE). Although FMEA on the KO TBM has not been completed, in-vessel, in-box and ex-vessel Loss Of Coolant Accident (LOCA) are considered as enveloping cases of PIE in general. In this paper, accidental analyses for the three selected LOCA were performed to investigate the decay heat removal capability of the TBM. To simulate transient thermo-hydraulic behavior of the TBM for the selected scenarios, RELAP5/MOD3.2 code was used

Transition from depressurization to long term cooling in AP600 scaled integral test facilities

International Nuclear Information System (INIS)

Bessette, D.E.; Marzo, M. di

1999-01-01

A novel light water reactor design called the AP600 has been proposed by the Westinghouse Electric Corporation. In the evaluation of this plant's behavior during a small break loss of coolant accident (LOCA), the crucial transition to low pressure, long-term cooling is marked by the injection of the gravitationally driven flow from the in-containment refueling water storage tank (IRWST). The onset of this injection is characterized by intermittency in the IRWST flow. This happens at a time when the reactor vessel reaches its minimum inventory. Therefore, it is important to understand and scale the behavior of the integral experimental test facilities during this portion of the transient. The explanation is that the periodic liquid drains and refills of the pressurizer are the reason for the intermittent behavior. The momentum balance for the surge line yields the nondimensional parameter controlling this process. Data from one of the three experimental facilities represent the phenomena well at the prototypical scale. The impact of the intermittent IRWST injection on the safe plant operation is assessed and its implications are successfully resolved. The oscillation is found to result from, in effect, excess water in the primary system and it is not of safety significance. (orig.)

Grid sensitivity capability for large scale structures

Science.gov (United States)

Nagendra, Gopal K.; Wallerstein, David V.

1989-01-01

The considerations and the resultant approach used to implement design sensitivity capability for grids into a large scale, general purpose finite element system (MSC/NASTRAN) are presented. The design variables are grid perturbations with a rather general linking capability. Moreover, shape and sizing variables may be linked together. The design is general enough to facilitate geometric modeling techniques for generating design variable linking schemes in an easy and straightforward manner. Test cases have been run and validated by comparison with the overall finite difference method. The linking of a design sensitivity capability for shape variables in MSC/NASTRAN with an optimizer would give designers a powerful, automated tool to carry out practical optimization design of real life, complicated structures.

WFPC2 Science Capability Report

Science.gov (United States)

Brown, David I.

2001-01-01

In the following pages, a brief outline of the salient science features of Wide Field/Planetary Camera 2 (WFPC2) that impact the proposal writing process and conceptual planning of observations is presented. At the time of writing, WFPC2, while having been better defined than in the past, is far from being at the stage where science and engineering details are well enough known that concrete observational/operational sequences can be plannned with assurance. Conceptual issues are another matter. The thrust of the Science Capability Report at this time is to outline the known performance parameters and capabilities of WFPC2, filling in with specifications when necessary to hold a place for these items as they become known. Also, primary scientific and operational differences between WFPC 1 and 2 are discussed section-by-section, along with issues that remain to be determined and idiosyncrasies when known. Clearly the determination of the latter awaits some form of testing, most likely thermal/vacuum testing. All data in this report should be viewed with a jaundiced eye at this time.

Power source evaluation capabilities at Sandia National Laboratories

Energy Technology Data Exchange (ETDEWEB)

Doughty, D.H.; Butler, P.C.

1996-04-01

Sandia National Laboratories maintains one of the most comprehensive power source characterization facilities in the U.S. National Laboratory system. This paper describes the capabilities for evaluation of fuel cell technologies. The facility has a rechargeable battery test laboratory and a test area for performing nondestructive and functional computer-controlled testing of cells and batteries.

NASA DOE POD NDE Capabilities Data Book

Science.gov (United States)

Generazio, Edward R.

2015-01-01

This data book contains the Directed Design of Experiments for Validating Probability of Detection (POD) Capability of NDE Systems (DOEPOD) analyses of the nondestructive inspection data presented in the NTIAC, Nondestructive Evaluation (NDE) Capabilities Data Book, 3rd ed., NTIAC DB-97-02. DOEPOD is designed as a decision support system to validate inspection system, personnel, and protocol demonstrating 0.90 POD with 95% confidence at critical flaw sizes, a90/95. The test methodology used in DOEPOD is based on the field of statistical sequential analysis founded by Abraham Wald. Sequential analysis is a method of statistical inference whose characteristic feature is that the number of observations required by the procedure is not determined in advance of the experiment. The decision to terminate the experiment depends, at each stage, on the results of the observations previously made. A merit of the sequential method, as applied to testing statistical hypotheses, is that test procedures can be constructed which require, on average, a substantially smaller number of observations than equally reliable test procedures based on a predetermined number of observations.

Dynamic capabilities, Marketing Capability and Organizational Performance

Directory of Open Access Journals (Sweden)

Adriana Roseli Wünsch Takahashi

2017-01-01

Full Text Available The goal of the study is to investigate the influence of dynamic capabilities on organizational performance and the role of marketing capabilities as a mediator in this relationship in the context of private HEIs in Brazil. As a research method we carried out a survey with 316 IES and data analysis was operationalized with the technique of structural equation modeling. The results indicate that the dynamic capabilities have influence on organizational performance only when mediated by marketing ability. The marketing capability has an important role in the survival, growth and renewal on educational services offerings for HEIs in private sector, and consequently in organizational performance. It is also demonstrated that mediated relationship is more intense for HEI with up to 3,000 students and other organizational profile variables such as amount of courses, the constitution, the type of institution and type of education do not significantly alter the results.

Design and Testing of an Active Heat Rejection Radiator with Digital Turn-Down Capability

Science.gov (United States)

Sunada, Eric; Birur, Gajanana C.; Ganapathi, Gani B.; Miller, Jennifer; Berisford, Daniel; Stephan, Ryan

2010-01-01

NASA's proposed lunar lander, Altair, will be exposed to vastly different external environment temperatures. The challenges to the active thermal control system (ATCS) are compounded by unfavorable transients in the internal waste heat dissipation profile: the lowest heat load occurs in the coldest environment while peak loads coincide with the warmest environment. The current baseline for this fluid is a 50/50 inhibited propylene glycol/water mixture with a freeze temperature around -35 C. While the overall size of the radiator's heat rejection area is dictated by the worst case hot scenario, a turn-down feature is necessary to tolerate the worst case cold scenario. A radiator with digital turn-down capability is being designed as a robust means to maintain cabin environment and equipment temperatures while minimizing mass and power consumption. It utilizes active valving to isolate and render ineffective any number of parallel flow tubes which span across the ATCS radiator. Several options were assessed in a trade-study to accommodate flow tube isolation and how to deal with the stagnant fluid that would otherwise remain in the tube. Bread-board environmental tests were conducted for options to drain the fluid from a turned-down leg as well an option to allow a leg to freeze/thaw. Each drain option involved a positive displacement gear pump with different methods of providing a pressure head to feed it. Test results showed that a start-up heater used to generate vapor at the tube inlet held the most promise for tube evacuation. Based on these test results and conclusions drawn from the trade-study, a full-scale radiator design is being worked for the Altair mission profile.

PIE on Safety-Tested AGR-1 Compact 5-1-1

Energy Technology Data Exchange (ETDEWEB)

Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Morris, Robert Noel [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Baldwin, Charles A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Fred C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gerczak, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-08-01

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High-Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program; this work continues to be funded by the Department of Energy's Office of Nuclear Energy as part of the Advanced Reactor Technologies (ART) initiative. AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for three years in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to examine fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing of irradiated AGR-1 Compact 5-1-1 in the ORNL Core Conduction Cooldown Test Facility (CCCTF) and post-safety testing PIE.

Design Management Capability and Product Innovation in SMEs

OpenAIRE

Fernández-Mesa, Ana Isabel; ALEGRE VIDAL, JOAQUIN; CHIVA GOMEZ, RICARDO; Gutiérrez Gracia, Antonio

2013-01-01

[EN] Purpose The aim of this paper is to present design management as a dynamic capability and to analyze its mediating role between organizational learning capability and product innovation performance in small and medium enterprises (SMEs). Design/methodology/approach Structural equation modeling is used to test the research hypotheses based on data from the Italian and Spanish ceramic tile industries. The data are derived from the responses of 182 companies (50 percent of the targ...

PWR cold-leg small break loca with faulty HPI

International Nuclear Information System (INIS)

Kumamaru, H.; Kukita, Y.

1991-01-01

The ROSA-IV Large Scale Test Facility (LSTF) is a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). At the LSTF are performed cold-leg small-break loss-of-coolant accident (LOCA) tests with faulty high pressure injection (HPI) system for break areas from 0.5% to 10% and an intentional primary system depressurization test following a small-break LOCA test. A simple prediction model is proposed for prediction of times of major events. Test data and calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of 5% or more, and is insufficient for intermediate break areas to maintain adequate core cooling. (author)

Innovation and dynamic capabilities of the firm: Defining an assessment model

Directory of Open Access Journals (Sweden)

André Cherubini Alves

2017-05-01

Full Text Available Innovation and dynamic capabilities have gained considerable attention in both academia and practice. While one of the oldest inquiries in economic and strategy literature involves understanding the features that drive business success and a firm’s perpetuity, the literature still lacks a comprehensive model of innovation and dynamic capabilities. This study presents a model that assesses firms’ innovation and dynamic capabilities perspectives based on four essential capabilities: development, operations, management, and transaction capabilities. Data from a survey of 1,107 Brazilian manufacturing firms were used for empirical testing and discussion of the dynamic capabilities framework. Regression and factor analyses validated the model; we discuss the results, contrasting with the dynamic capabilities’ framework. Operations Capability is the least dynamic of all capabilities, with the least influence on innovation. This reinforces the notion that operations capabilities as “ordinary capabilities,” whereas management, development, and transaction capabilities better explain firms’ dynamics and innovation.

A model-free approach to eliminate autocorrelation when testing for process capability

DEFF Research Database (Denmark)

Vanmann, Kerstin; Kulahci, Murat

2008-01-01

There is an increasing use of on-line data acquisition systems in industry. This usually leads to autocorrelated data and implies that the assumption of independent observations has to be re-examined. Most decision procedures for capability analysis assume independent data. In this article we pre...

RELAP5 assessment using semiscale SBLOCA test S-NH-1

International Nuclear Information System (INIS)

Lee, E.J.; Chung, B.D.; Kim, H.J.

1993-06-01

2-inch cold leg break test S-NH-1, conducted at the 1/1705 volume scaled facility Semiscale was analyzed using RELAP5/MOD2 Cycle 36.04 and MOD3 Version 5m5. Loss of HPIS was assumed, and reactor trip occurred on a low PZR pressure signal (13.1 MPa), and pumps began an unpowered coastdown on SI signal (12.5 MPa). The system was recovered by opening ADV's when the PCT became higher than 811 K. Accumulator was finally injected into the system when the primary system pressure was less than 4.0 MPa. The experiment was terminated when the pressure reached the LPIS actuation set point RELAP5/MOD2 analysis demonstrated its capability to predict, with a sufficient accuracy, the main phenomena occurring in the depressurization transient, both from a qualitative and quantitative points of view. Nevertheless, several differences were noted regarding the break flow rate and inventory distribution due to deficiencies in two-phase choked flow model, horizontal stratification interfacial drag, and a CCFL model. The main reason for the core to remain nearly fully covered with the liquid was the under-prediction of the break flow by the code. Several sensitivity calculations were tried using the MOD2 to improve the results by using the different options of break flow modeling (downward, homogeneous, and area increase). The break area compensating concept based on ''the integrated break flow matching'' gave the best results than downward junction and homogeneous options. And the MOD3 showed improvement in predicting a CCFL in SG and a heatup in the core

IE Information Notice No. 85-75: Improperly installed instrumentation, inadequate quality control and inadequate postmodification testing

International Nuclear Information System (INIS)

Jordan, E.L.

1992-01-01

On June 10, 1985, the licensee informed the NRC Resident Inspector that for approximately 5 days LaSalle Unit 2 had been without the capability of automatic actuation of emergency core cooling (ECCS) and that for approximately 3 days during this period the plant had been without secondary containment integrity. The major cause of this condition was improper installation (the variable and reference legs were reversed) of the two reactor vessel level actuation switches which control Division 1 automatic depressurization system (ADS), low pressure core spray (LPCS), and reactor core isolation cooling (RCIC). On July 20, 1985, the Trojan Nuclear Power Plant tripped from 100% power because of a turbine trip that was caused by the loss of the unit auxiliary transformer. All systems functioned normally except that low suction pressure caused one auxiliary feedwater pump to trip and then the other auxiliary feedwater pump to trip after restart of the first auxiliary feedwater pump. The cause of the trips of the auxiliary feedwater pumps can be traced back to improper postmodification adjustment and inadequate postmodification testing following retrofit of environmentally qualified controllers for the auxiliary feedwater system. The auxiliary feedwater pump trips on low suction pressure were caused by excessive combined flow from the two auxiliary feedwater pumps that draw from a single header from the condensate storage tank. The flow control valves were open farther than required after new environmentally qualified controllers had been installed during a recent refueling outage

Counterpart experimental study of ISP-42 PANDA tests on PUMA facility

International Nuclear Information System (INIS)

Yang, Jun; Choi, Sung-Won; Lim, Jaehyok; Lee, Doo-Yong; Rassame, Somboon; Hibiki, Takashi; Ishii, Mamoru

2013-01-01

Highlights: ► Counterpart tests were performed on two large-scale BWR integral facilities. ► Similarity of post-LOCA system behaviors observed between two tests. ► Passive core and containment cooling systems work as design in both tests. -- Abstract: A counterpart test to the Passive Nachwärmeabfuhr und Druckabbau Test Anlage (Passive Decay Heat Removal and Depressurization Test Facility, PANDA) International Standard Problem (ISP)-42 test was conducted at the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. Aimed to support code validation on a range of light water reactor (LWR) containment issues, the ISP-42 test consists of six sequential phases (Phases A–F) with separately defined initial and boundary conditions, addressing different stages of anticipated accident scenario and system responses. The counterpart test was performed from Phases A to D, which are within the scope of the normal integral tests performed on the PUMA facility. A scaling methodology was developed by using the PANDA facility as prototype and PUMA facility as test model, and an engineering scaling has been applied to the PUMA facility. The counterpart test results indicated that functions of passive safety systems, such as passive containment cooling system (PCCS) start-up, gravity-driven cooling system (GDCS) discharge, PCCS normal operation and overload function were confirmed in both the PANDA and PUMA facilities with qualitative similarities

Antecedents of network capability and their effects on innovation performance: an empirical test of hi-tech firms in China

NARCIS (Netherlands)

Fang, Gang; Ma, Xiang Yuan; Brouwers-Ren, Liqin; Zhou, Qing

2014-01-01

A firm’s competitive advantage can come not only from internal resources but also from inter-firm innovation networks. This paper shows that network capabilities (i.e., network visioning capability, network constructing capability, network operating capability and network centring capability) are

Dynamic capabilities and innovation capabilities: The case of the ‘Innovation Clinic’

Directory of Open Access Journals (Sweden)

Fred Strønen

2017-01-01

Full Text Available In this explorative study, we investigate the relationship between dynamic capabilities and innovation capabilities. Dynamic capabilities are at the core of strategic management in terms of how firms can ensure adaptation to changing environments over time. Our paper follows two paths of argumentation. First, we review and discuss some major contributions to the theories on ordinary capabilities, dynamic capabilities, and innovation capabilities. We seek to identify different understandings of the concepts in question, in order to clarify the distinctions and relationships between dynamic capabilities and innovation capabilities. Second, we present a case study of the ’Innovation Clinic’ at a major university hospital, including four innovation projects. We use this case study to explore and discuss how dynamic capabilities can be extended, as well as to what extent innovation capabilities can be said to be dynamic. In our conclusion, we discuss the conditions for nurturing ‘dynamic innovation capabilities’ in organizations.

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

International Nuclear Information System (INIS)

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

1975-11-01

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

PIE on Safety-Tested Loose Particles from Irradiated Compact 4-4-2

Energy Technology Data Exchange (ETDEWEB)

Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gerczak, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Morris, Robert Noel [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Baldwin, Charles A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Fred C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-04-01

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High Temperature Gas-cooled Reactors (HTGRs). This work is sponsored by the Department of Energy Office of Nuclear Energy (DOE-NE) through the Advanced Reactor Technologies (ART) Office under the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program. The AGR-1 experiment was the first in a series of TRISO fuel irradiation tests initiated in 2006. The AGR-1 TRISO particles and fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 using laboratory-scale equipment and irradiated for 3 years in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. Post-irradiation examination was performed at INL and ORNL to study how the fuel behaved during irradiation, and to test fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing and post-safety testing PIE conducted at ORNL on loose particles extracted from irradiated AGR-1 Compact 4-4-2.

Capability ethics

OpenAIRE

Robeyns, Ingrid

2012-01-01

textabstractThe capability approach is one of the most recent additions to the landscape of normative theories in ethics and political philosophy. Yet in its present stage of development, the capability approach is not a full-blown normative theory, in contrast to utilitarianism, deontological theories, virtue ethics, or pragmatism. As I will argue in this chapter, at present the core of the capability approach is an account of value, which together with some other (more minor) normative comm...

Stored energy analysis in scale-down test facility

International Nuclear Information System (INIS)

Deng Chengcheng; Qin Benke; Fang Fangfang; Chang Huajian; Ye Zishen

2013-01-01

In the integral test facilities that simulate the accident transient process of the prototype nuclear power plant, the stored energy in the metal components has a direct influence on the simulation range and the test results of the facilities. Based on the heat transfer theory, three methods analyzing the stored energy were developed, and a thorough study on the stored energy problem in the scale-down test facilities was further carried out. The lumped parameter method and power integration method were applied to analyze the transient process of energy releasing and to evaluate the average total energy stored in the reactor pressure vessel of the ACME (advanced core-cooling mechanism experiment) facility, which is now being built in China. The results show that the similarity requirements for such three methods to analyze the stored energy in the test facilities are reduced gradually. Under the condition of satisfying the integral similarity of natural circulation, the stored energy releasing process in the scale-down test facilities can't maintain exact similarity. The stored energy in the reactor pressure vessel wall of ACME, which is released quickly during the early stage of rapid depressurization of system, will not make a major impact on the long-term behavior of system. And the scaling distortion of integral average total energy of the stored heat is acceptable. (authors)

Evaluating the habitat capability model for Merriam's turkeys

Science.gov (United States)

Mark A. Rumble; Stanley H. Anderson

1995-01-01

Habitat capability (HABCAP) models for wildlife assist land managers in predicting the consequences of their management decisions. Models must be tested and refined prior to using them in management planning. We tested the predicted patterns of habitat selection of the R2 HABCAP model using observed patterns of habitats selected by radio-marked Merriam’s turkey (

Gossiping Capabilities

DEFF Research Database (Denmark)

Mogensen, Martin; Frey, Davide; Guerraoui, Rachid

Gossip-based protocols are now acknowledged as a sound basis to implement collaborative high-bandwidth content dissemination: content location is disseminated through gossip, the actual contents being subsequently pulled. In this paper, we present HEAP, HEterogeneity Aware gossip Protocol, where...... nodes dynamically adjust their contribution to gossip dissemination according to their capabilities. Using a continuous, itself gossip-based, approximation of relative capabilities, HEAP dynamically leverages the most capable nodes by (a) increasing their fanouts (while decreasing by the same proportion...... declare a high capability in order to augment their perceived quality without contributing accordingly. We evaluate HEAP in the context of a video streaming application on a 236 PlanetLab nodes testbed. Our results shows that HEAP improves the quality of the streaming by 25% over a standard gossip...

In-situ Creep Testing Capability Development for Advanced Test Reactor

Energy Technology Data Exchange (ETDEWEB)

B. G. Kim; J. L. Rempe; D. L. Knudson; K. G. Condie; B. H. Sencer

2010-08-01

Creep is the slow, time-dependent strain that occurs in a material under a constant strees (or load) at high temperature. High temperature is a relative term, dependent on the materials being evaluated. A typical creep curve is shown in Figure 1-1. In a creep test, a constant load is applied to a tensile specimen maintained at a constant temperature. Strain is then measured over a period of time. The slope of the curve, identified in the figure below, is the strain rate of the test during Stage II or the creep rate of the material. Primary creep, Stage I, is a period of decreasing creep rate due to work hardening of the material. Primary creep is a period of primarily transient creep. During this period, deformation takes place and the resistance to creep increases until Stage II, Secondary creep. Stage II creep is a period with a roughly constant creep rate. Stage II is referred to as steady-state creep because a balance is achieved between the work hardening and annealing (thermal softening) processes. Tertiary creep, Stage III, occurs when there is a reduction in cross sectional area due to necking or effective reduction in area due to internal void formation; that is, the creep rate increases due to necking of the specimen and the associated increase in local stress.

Rights, goals, and capabilities

NARCIS (Netherlands)

van Hees, M.V.B.P.M

This article analyses the relationship between rights and capabilities in order to get a better grasp of the kind of consequentialism that the capability theory represents. Capability rights have been defined as rights that have a capability as their object (rights to capabilities). Such a

Capabilities Report 2012, West Desert Test Center

Science.gov (United States)

2012-03-12

categories for BSAT: long-term storage that includes BSAT not in active use that are stored in the bioholdings facility, located in Building 2029...under varying environmental conditions  Analysis of common battlefield contaminants (e.g., diesel fuel, gasoline, brake fluid, paint) Laboratory tests...and regenerative ( REGEN ) filters. Vapor dissemination is introduced upstream of the air filtration/purification device with challenge

Enhanced H- ion source testing capabilities at LANSCE

International Nuclear Information System (INIS)

Ingalls, W.B.; Hardy, M.W.; Prichard, B.A.; Sander, O.R.; Stelzer, J.E.; Stevens, R.R.; Leung, K.N.; Williams, M.D.

1998-01-01

As part of the on-going beam-current upgrade in the Proton Storage Ring (PSR) at the Los Alamos Neutron Science Center (LANSCE), the current available from the H - injector will be increased from the present 16 to 18 mA to as much as 40 mA. A collaboration between the Ion Beam Technology Group at Lawrence Berkeley National Laboratory (LBNL) and the Ion Sources and Injectors section of LANSCE-2 at Los Alamos National Laboratory (LANL) has been formed to develop and evaluate a new ion source. A new Ion Source Test Stand (ISTS) has been constructed at LANSCE to evaluate candidate ion sources. The ISTS has been constructed to duplicate as closely as possible the beam transport and ancillary systems presently in use in the LANSCE H - injector, while incorporating additional beam diagnostics for source testing. The construction and commissioning of the ISTS will be described, preliminary results for the proof-of-principle ion source developed by the Berkeley group will be presented, and future plans for the extension of the test stand will be presented

Capability ethics

NARCIS (Netherlands)

I.A.M. Robeyns (Ingrid)

2012-01-01

textabstractThe capability approach is one of the most recent additions to the landscape of normative theories in ethics and political philosophy. Yet in its present stage of development, the capability approach is not a full-blown normative theory, in contrast to utilitarianism, deontological

Fifth in situ vitrification engineering-scale test of simulated INEL buried waste sites

International Nuclear Information System (INIS)

Bergsman, T.M.; Shade, J.W.; Farnsworth, R.K.

1992-06-01

In September 1990, an engineering-scale in situ vitrification (ISV) test was conducted on sealed canisters containing a combined mixture of buried waste materials expected to be present at the Idaho National Engineering Laboratory (INEL) Subsurface Disposal Area (SDA). The test was part of a Pacific Northwest Laboratory (PNL) program to assist INEL in treatability studies of the potential application of ISV to mixed transuranic wastes at the INEL SDA. The purpose of this test was to determine the effect of a close-packed layer of sealed containers on ISV processing performance. Specific objectives included determining (1) the effect of releases from sealed containers on hood plenum pressure and temperature, (2) the release pressure ad temperatures of the sealed canisters, (3) the relationships between canister depressurization and melt encapsulation, (4) the resulting glass and soil quality, (5) the potential effects of thermal transport due to a canister layer, (6) the effects on particle entrainment of differing angles of approach for the ISV melt front, and (7) the effects of these canisters on the volatilization of voltatile and semivolatile contaminants into the hood plenum

The establishment and analysis of TRACE model for ultimate response guideline of Chinshan nuclear power plant - 15448

International Nuclear Information System (INIS)

Huang, J.J.; Wang, J.R.; Shih, C.; Chen, S.W.; Liao, L.Y.; Lin, H.T.

2015-01-01

The purpose of this research is to use TRACE code to perform a simulation that executes the procedures of URG (Ultimate Response Guidelines) to deal with Fukushima-like accidents. TRACE is an advanced thermal hydraulic code that has been developed by the United States Nuclear Regulatory Commission for NPP safety analysis. In this work TRACE has been used to analyze the thermal hydraulic model for the URG of the Chinshan nuclear power plant that is composed of 2 BWR-type reactors. URG includes 2-stage depressurization, alternative water injection and removing decay heat through the ejection from containment. The 2-stage depressurization strategy includes controlled depressurization and emergency depressurization to replace traditional one-stage depressurization. Results show that by comparing with one-stage depressurization strategy, 2-stage depressurization strategy is able to reduce peak cladding temperature (PCT) effectively and needs much less minimum flow rate of alternative water injection in the accident

Emergency core cooling system

International Nuclear Information System (INIS)

Abe, Nobuaki.

1993-01-01

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

Methane Lunar Surface Thermal Control Test

Science.gov (United States)

Plachta, David W.; Sutherlin, Steven G.; Johnson, Wesley L.; Feller, Jeffrey R.; Jurns, John M.

2012-01-01

NASA is considering propulsion system concepts for future missions including human return to the lunar surface. Studies have identified cryogenic methane (LCH4) and oxygen (LO2) as a desirable propellant combination for the lunar surface ascent propulsion system, and they point to a surface stay requirement of 180 days. To meet this requirement, a test article was prepared with state-of-the-art insulation and tested in simulated lunar mission environments at NASA GRC. The primary goals were to validate design and models of the key thermal control technologies to store unvented methane for long durations, with a low-density high-performing Multi-layer Insulation (MLI) system to protect the propellant tanks from the environmental heat of low Earth orbit (LEO), Earth to Moon transit, lunar surface, and with the LCH4 initially densified. The data and accompanying analysis shows this storage design would have fallen well short of the unvented 180 day storage requirement, due to the MLI density being much higher than intended, its substructure collapse, and blanket separation during depressurization. Despite the performance issue, insight into analytical models and MLI construction was gained. Such modeling is important for the effective design of flight vehicle concepts, such as in-space cryogenic depots or in-space cryogenic propulsion stages.

Dynamic Capabilities

DEFF Research Database (Denmark)

Grünbaum, Niels Nolsøe; Stenger, Marianne

2013-01-01

The findings reveal a positive relationship between dynamic capabilities and innovation performance in the case enterprises, as we would expect. It was, however, not possible to establish a positive relationship between innovation performance and profitability. Nor was there any positive...... relationship between dynamic capabilities and profitability....

Advanced Simulation Capability for Environmental Management (ASCEM) Phase II Demonstration

Energy Technology Data Exchange (ETDEWEB)

Freshley, M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hubbard, S. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Flach, G. [Savannah River National Lab. (SRNL), Aiken, SC (United States); Freedman, V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Agarwal, D. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Andre, B. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Bott, Y. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chen, X. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Davis, J. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Faybishenko, B. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Gorton, I. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Murray, C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Moulton, D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Meyer, J. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Rockhold, M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Shoshani, A. [LBNL; Steefel, C. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Wainwright, H. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Waichler, S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2012-09-28

In 2009, the National Academies of Science (NAS) reviewed and validated the U.S. Department of Energy Office of Environmental Management (EM) Technology Program in its publication, Advice on the Department of Energy’s Cleanup Technology Roadmap: Gaps and Bridges. The NAS report outlined prioritization needs for the Groundwater and Soil Remediation Roadmap, concluded that contaminant behavior in the subsurface is poorly understood, and recommended further research in this area as a high priority. To address this NAS concern, the EM Office of Site Restoration began supporting the development of the Advanced Simulation Capability for Environmental Management (ASCEM). ASCEM is a state-of-the-art scientific approach that uses an integration of toolsets for understanding and predicting contaminant fate and transport in natural and engineered systems. The ASCEM modeling toolset is modular and open source. It is divided into three thrust areas: Multi-Process High Performance Computing (HPC), Platform and Integrated Toolsets, and Site Applications. The ASCEM toolsets will facilitate integrated approaches to modeling and site characterization that enable robust and standardized assessments of performance and risk for EM cleanup and closure activities. During fiscal year 2012, the ASCEM project continued to make significant progress in capabilities development. Capability development occurred in both the Platform and Integrated Toolsets and Multi-Process HPC Simulator areas. The new Platform and Integrated Toolsets capabilities provide the user an interface and the tools necessary for end-to-end model development that includes conceptual model definition, data management for model input, model calibration and uncertainty analysis, and model output processing including visualization. The new HPC Simulator capabilities target increased functionality of process model representations, toolsets for interaction with the Platform, and model confidence testing and verification for

Relating children's attentional capabilities to intelligence, memory, and academic achievement: a test of construct specificity in children with asthma.

Science.gov (United States)

Annett, Robert D; Bender, Bruce G; Gordon, Michael

2007-01-01

The relationship between attention, intelligence, memory, achievement, and behavior in a large population (N = 939) of children without neuropsychologic problems was investigated in children with mild and moderate asthma. It was hypothesized that different levels of children's attentional capabilities would be associated with different levels of intellectual, memory, and academic abilities. Children ages 6-12 at the eight clinical centers of the Childhood Asthma Management Program (CAMP) were enrolled in this study. Standardized measures of child neuropsychological and behavioral performance were administered to all participants, with analyses examining both the developmental trajectory of child attentional capabilities and the associations between Continuous Performance Test (CPT) scores and intellectual functioning, and measures of memory, academic achievement, and behavioral functioning. Findings demonstrated that correct responses on the CPT increase significantly with age, while commission errors decrease significantly with age. Performance levels on the CPT were associated with differences in child intellectual function, memory, and academic achievement. Overall these findings reveal how impairments in child attention skills were associated with normal levels of performance on measures of children's intelligence, memory, academic achievement, and behavioral functioning, suggesting that CPT performance is a salient marker of brain function.

Capabilities and Incapabilities of the Capabilities Approach to Health Justice.

Science.gov (United States)

Selgelid, Michael J

2016-01-01

This first part of this article critiques Sridhar Venkatapuram's conception of health as a capability. It argues that Venkatapuram relies on the problematic concept of dignity, implies that those who are unhealthy lack lives worthy of dignity (which seems politically incorrect), sets a low bar for health, appeals to metaphysically problematic thresholds, fails to draw clear connections between appealed-to capabilities and health, and downplays the importance/relevance of health functioning. It concludes by questioning whether justice entitlements should pertain to the capability for health versus health achievements, challenging Venkatapuram's claims about the strength of health entitlements, and demonstrating that the capabilities approach is unnecessary to address social determinants of health. © 2016 John Wiley & Sons Ltd.

Proficiency Testing for Bacterial Whole Genome Sequencing: An End-User Survey of Current Capabilities, Requirements and Priorities

DEFF Research Database (Denmark)

Moran-Gilad, Jacob; Sintchenko, Vitali; Karlsmose Pedersen, Susanne

2015-01-01

The advent of next-generation sequencing (NGS) has revolutionised public health microbiology. Given the potential impact of NGS, it is paramount to ensure standardisation of ‘wet’ laboratory and bioinformatic protocols and promote comparability of methods employed by different laboratories...... and their outputs. Therefore, one of the ambitious goals of the Global Microbial Identifier (GMI) initiative (http://​www.​globalmicrobiali​dentifier.​org/​) has been to establish a mechanism for inter-laboratory NGS proficiency testing (PT). This report presents findings from the survey recently conducted...... by Working Group 4 among GMI members in order to ascertain NGS end-use requirements and attitudes towards NGS PT. The survey identified the high professional diversity of laboratories engaged in NGS-based public health projects and the wide range of capabilities within institutions, at a notable range...

Proficiency testing for bacterial whole genome sequencing: an end-user survey of current capabilities, requirements and priorities

DEFF Research Database (Denmark)

Moran-Gilad, Jacob; Sintchenko, Vitali; Karlsmose Pedersen, Susanne

2015-01-01

The advent of next-generation sequencing (NGS) has revolutionised public health microbiology. Given the potential impact of NGS, it is paramount to ensure standardisation of 'wet' laboratory and bioinformatic protocols and promote comparability of methods employed by different laboratories...... and their outputs. Therefore, one of the ambitious goals of the Global Microbial Identifier (GMI) initiative (http://www.globalmicrobialidentifier.org/) has been to establish a mechanism for inter-laboratory NGS proficiency testing (PT). This report presents findings from the survey recently conducted by Working...... Group 4 among GMI members in order to ascertain NGS end-use requirements and attitudes towards NGS PT. The survey identified the high professional diversity of laboratories engaged in NGS-based public health projects and the wide range of capabilities within institutions, at a notable range of costs...

Evaluation of physical conditional capabilities in athletes with physical motors limitations

Directory of Open Access Journals (Sweden)

Indira de las Mercedes Saínz-Reyes

2015-06-01

Full Text Available This research was carried out in the province of Holguin with the disabled athletes competing in T53 mode event 100 m levels. It was determined as a scientific problem: How to evaluate the conditional physical capabilities in athletes with physical motor limitations in athletics? and as objective: To adapt functional tests to evaluate the conditional physical capabilities in athletes with physical motor limitations in athletics. A selection of functional test was carried out applying then some methodological modifications in order to make them easier for disabled athletes in wheelchairs. Theoretical and empirical levels methods as well as consensus techniques were used. The research has an adequate scientific rigor since a methodologies theoretical foundation on conditional physical capabilities and their evaluation by functional testing. Through the nominal group it could be determined the feasibility of the proposal, since the consulted experts came to the consensus that the chosen and modified functional tests can be applied to disabled athletes in wheelchairs, and recognized that through them it can be checked the training level of these athletes and the influence of the physical exercise in their organism.

Capabilities for Strategic Adaptation

DEFF Research Database (Denmark)

Distel, Andreas Philipp

This dissertation explores capabilities that enable firms to strategically adapt to environmental changes and preserve competitiveness over time – often referred to as dynamic capabilities. While dynamic capabilities being a popular research domain, too little is known about what these capabiliti...

Legitimacy, capability, effectiveness and the future of the NPT

International Nuclear Information System (INIS)

Keeley, J.F.

1987-01-01

This chapter looks at the relationship between legitimacy and capability in conceptually and politically contestable regions. This issue was highlighted by India's nuclear test of May 1974 and the Osiraq raid of 1981. These illustrated the general problem of the threat to the coherence and legitimacy of the non-proliferation regime. This threat arose from the spread of nuclear technological capabilities. Two developments in the non-proliferation regime that have helped produce the more specific problems of that regime are discussed. These are the spread of nuclear technological capabilities and the development of complex co-operation networks. The prospects for the modification of the NPT in response to these challenges are considered finally. (U.K.)

Characterization and process technology capabilities for Hanford tank waste disposal

International Nuclear Information System (INIS)

Buelt, J.L.; Weimer, W.C.; Schrempf, R.E.

1996-03-01

The purpose of this document is to describe the Paciflc Northwest National Laboratory's (the Laboratory) capabilities in characterization and unit process and system testing that are available to support Hanford tank waste processing. This document is organized into two parts. The first section discusses the Laboratory's extensive experience in solving the difficult problems associated with the characterization of Hanford tank wastes, vitrified radioactive wastes, and other very highly radioactive and/or heterogeneous materials. The second section of this document discusses the Laboratory's radioactive capabilities and facilities for separations and waste form preparation/testing that can be used to Support Hanford tank waste processing design and operations

Dynamics of vendor innovation capability: Evidence from the Electronics Manufacturing Services industry

DEFF Research Database (Denmark)

Perunovic, Zoran; Mefford, Robert; Christoffersen, Mads

2012-01-01

and innovation. The first is the realization that vendor capabilities have been recognized as one of the most important factors for the success of outsourcing. The second refers to the fact that, even though innovation capability is required, vendors are still being selected, and their performance evaluated......, by traditional manufacturing capabilities, such as cost, quality, delivery, and flexibility. Taking a vendor’s perspective in outsourcing, we develop and present a conceptual framework for studying vendor innovation capability. We propose to test this framework in the Electronic Manufacturing Services Industry....

Verification and validation of COBRA-SFS transient analysis capability

International Nuclear Information System (INIS)

Rector, D.R.; Michener, T.E.; Cuta, J.M.

1998-05-01

This report provides documentation of the verification and validation testing of the transient capability in the COBRA-SFS code, and is organized into three main sections. The primary documentation of the code was published in September 1995, with the release of COBRA-SFS, Cycle 2. The validation and verification supporting the release and licensing of COBRA-SFS was based solely on steady-state applications, even though the appropriate transient terms have been included in the conservation equations from the first cycle. Section 2.0, COBRA-SFS Code Description, presents a capsule description of the code, and a summary of the conservation equations solved to obtain the flow and temperature fields within a cask or assembly model. This section repeats in abbreviated form the code description presented in the primary documentation (Michener et al. 1995), and is meant to serve as a quick reference, rather than independent documentation of all code features and capabilities. Section 3.0, Transient Capability Verification, presents a set of comparisons between code calculations and analytical solutions for selected heat transfer and fluid flow problems. Section 4.0, Transient Capability Validation, presents comparisons between code calculations and experimental data obtained in spent fuel storage cask tests. Based on the comparisons presented in Sections 2.0 and 3.0, conclusions and recommendations for application of COBRA-SFS to transient analysis are presented in Section 5.0

Leak rate test of containment personnel lock

International Nuclear Information System (INIS)

Julien, J.T.; Peters, S.W.

1988-01-01

As part of the US NRC Containment Integrity Program, a leak rate test was performed on a full size personnel airlock for a nuclear containment building. The airlock was subjected to conditions simulating severe accident conditions. The objective of the test was to characterize the performance of airlock door seals when subjected to conditions that exceeded design. The seals tested were a double dog-ear configuration and made from EPDM E603. The data obtained from this test will be used by SNL as a benchmark for development of analytical methods. In addition to leak rate information, strain, temperature, displacements, and pressure data were measured and recorded from over 330 transducers. The test lasted approximately 60 hours. Data were recorded at regular intervals and during heating, pressurization and depressurization. The inner airlock door and bulkhead were exposed to a maximum air temperature of 850 F and a maximum air pressure of 300 psig. The airlock was originally designed for 340 F and 60 psig. Two heating and pressurization cycles were planned; one to heat to 400 F and pressurize to 300 psig, and the second to heat to 800 F and pressurize to 300 psig. No significant leakage was recorded during these two cycles. A third cycle was added to the test program. The air temperature was increased to 850 F and held at this temperature for approximately 10 hours. The inner door seal failed quickly at a pressure of 150.5 psig. The maximum leak rate was 706 SCFM

Organizational Strategic Learning Capability: Exploring the Dimensions

Science.gov (United States)

Moon, Hanna; Sejong, Wendy; Valentine, Tom

2017-01-01

Purpose: How to build and enhance the strategic learning capability (SLC) of an organization becomes crucial to both research and practice. This study was designed with the purpose to conceptualize SLC by translating and interpreting the related literature to develop empirical dimensions that could be tested and used in a survey instrument.…

A new laser reflectance system capable of measuring changing cross-sectional area of soft tissues during tensile testing.

Science.gov (United States)

Pokhai, Gabriel G; Oliver, Michele L; Gordon, Karen D

2009-09-01

Determination of the biomechanical properties of soft tissues such as tendons and ligaments is dependent on the accurate measurement of their cross-sectional area (CSA). Measurement methods, which involve contact with the specimen, are problematic because soft tissues are easily deformed. Noncontact measurement methods are preferable in this regard, but may experience difficulty in dealing with the complex cross-sectional shapes and glistening surfaces seen in soft tissues. Additionally, existing CSA measurement systems are separated from the materials testing machine, resulting in the inability to measure CSA during testing. Furthermore, CSA measurements are usually made in a different orientation, and with a different preload, prior to testing. To overcome these problems, a noncontact laser reflectance system (LRS) was developed. Designed to fit in an Instron 8872 servohydraulic test machine, the system measures CSA by orbiting a laser transducer in a circular path around a soft tissue specimen held by tissue clamps. CSA measurements can be conducted before and during tensile testing. The system was validated using machined metallic specimens of various shapes and sizes, as well as different sizes of bovine tendons. The metallic specimens could be measured to within 4% accuracy, and the tendons to within an average error of 4.3%. Statistical analyses showed no significant differences between the measurements of the LRS and those of the casting method, an established measurement technique. The LRS was successfully used to measure the changing CSA of bovine tendons during uniaxial tensile testing. The LRS developed in this work represents a simple, quick, and accurate way of reconstructing complex cross-sectional profiles and calculating cross-sectional areas. In addition, the LRS represents the first system capable of automatically measuring changing CSA of soft tissues during tensile testing, facilitating the calculation of more accurate biomechanical properties.

INEL design studies in support of the Westinghouse EPRI small plant study

International Nuclear Information System (INIS)

Burtt, J.D.; Kullberg, C.M.

1986-03-01

In support of the design effort of a Westinghouse EPRI small plant study, several analyses were performed at the Idaho National Engineering Laboratory. An analysis was performed to study fuel behavior under conditions of a limiting flow coastdown transient. Depressurization capabilities for the reactor coolant system were studied. The post-accident heat removal for the current containment design was studied. The results of all three studies are reported. 31 figs

Computational fluid dynamics analysis of the initial stages of a VHTR air-ingress accident using a scaled-down model

Energy Technology Data Exchange (ETDEWEB)

Ham, Tae K., E-mail: taekyu8@gmail.com [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Arcilesi, David J., E-mail: arcilesi.1@osu.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Kim, In H., E-mail: ihkim0730@gmail.com [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Sun, Xiaodong, E-mail: sun.200@osu.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Christensen, Richard N., E-mail: rchristensen@uidaho.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Oh, Chang H. [Idaho National Laboratory, Idaho Falls, ID 83402 (United States); Kim, Eung S., E-mail: kes7741@snu.ac.kr [Idaho National Laboratory, Idaho Falls, ID 83402 (United States)

2016-04-15

Highlights: • Uncertainty quantification and benchmark study are performed to validate an ANSYS FLUENT computer model for a depressurization process in a high-temperature gas-cooled reactor. • An ANSYS FLUENT computer model of a 1/8th scaled-down geometry of a VHTR hot exit plenum is presented, which is similar to the experimental test facility that has been constructed at The Ohio State University. • Using the computer model of the scaled-down geometry, the effects of the depressurization process and flow oscillations on the subsequent density-driven stratified flow phenomenology are examined computationally. • The effects of the scaled-down hot exit plenum internal structure temperature on the density-driven stratified flow phenomenology are investigated numerically. - Abstract: An air-ingress accident is considered to be one of the design basis accidents of a very high-temperature gas-cooled reactor (VHTR). The air-ingress accident is initiated, in its worst-case scenario, by a complete break of the hot duct in what is referred to as a double-ended guillotine break. This leads to an initial loss of the primary helium coolant via depressurization. Following the depressurization process, the air–helium mixture in the reactor cavity could enter the reactor core via the hot duct and hot exit plenum. In the event that air ingresses into the reactor vessel, the high-temperature graphite structures in the reactor core and hot plenum will chemically react with the air, which could lead to damage of in-core graphite structures and fuel, release of carbon monoxide and carbon dioxide, core heat up, failure of the structural integrity of the system, and eventually the release of radionuclides to the environment. Studies in the available literature focus on the phenomena of the air ingress accident that occur after the termination of the depressurization, such as density-driven stratified flow, molecular diffusion, and natural circulation. However, a recent study

Capability-based computer systems

CERN Document Server

Levy, Henry M

2014-01-01

Capability-Based Computer Systems focuses on computer programs and their capabilities. The text first elaborates capability- and object-based system concepts, including capability-based systems, object-based approach, and summary. The book then describes early descriptor architectures and explains the Burroughs B5000, Rice University Computer, and Basic Language Machine. The text also focuses on early capability architectures. Dennis and Van Horn's Supervisor; CAL-TSS System; MIT PDP-1 Timesharing System; and Chicago Magic Number Machine are discussed. The book then describes Plessey System 25

Absorptive capacity and mass customization capability

OpenAIRE

Zhang, Min; Zhao, Xiande; Lyles, Marjorie A.; Guo, Hangfei

2015-01-01

Purpose The purpose of this paper is to investigate the effects of a manufacturer’s absorptive capacity (AC) on its mass customization capability (MCC). Design/methodology/approach The authors conceptualize AC within the supply chain context as four processes: knowledge acquisition from customers, knowledge acquisition from suppliers, knowledge assimilation, and knowledge application. The authors then propose and empirically test a model on the relationships among AC processes and MCC using s...

Binomial Test Method for Determining Probability of Detection Capability for Fracture Critical Applications

Science.gov (United States)

Generazio, Edward R.

2011-01-01

The capability of an inspection system is established by applications of various methodologies to determine the probability of detection (POD). One accepted metric of an adequate inspection system is that for a minimum flaw size and all greater flaw sizes, there is 0.90 probability of detection with 95% confidence (90/95 POD). Directed design of experiments for probability of detection (DOEPOD) has been developed to provide an efficient and accurate methodology that yields estimates of POD and confidence bounds for both Hit-Miss or signal amplitude testing, where signal amplitudes are reduced to Hit-Miss by using a signal threshold Directed DOEPOD uses a nonparametric approach for the analysis or inspection data that does require any assumptions about the particular functional form of a POD function. The DOEPOD procedure identifies, for a given sample set whether or not the minimum requirement of 0.90 probability of detection with 95% confidence is demonstrated for a minimum flaw size and for all greater flaw sizes (90/95 POD). The DOEPOD procedures are sequentially executed in order to minimize the number of samples needed to demonstrate that there is a 90/95 POD lower confidence bound at a given flaw size and that the POD is monotonic for flaw sizes exceeding that 90/95 POD flaw size. The conservativeness of the DOEPOD methodology results is discussed. Validated guidelines for binomial estimation of POD for fracture critical inspection are established.

Molecular Contamination Investigation Facility (MCIF) Capabilities

Science.gov (United States)

Soules, David M.

2013-01-01

This facility was used to guide the development of ASTM E 1559 center dot Multiple Quartz Crystal Microbalances (QCMs), large sample and spectral effects capability center dot Several instrumented, high vacuum chamber systems are used to evaluate the molecular outgassing characteristics of materials, flight components and other sensitive surfaces. Test materials for spacecraft/instrument selection center.Test flight components for acceptable molecular outgas levels center dot Determine time/temperature vacuum bake-out requirements center. Data used to set limits for use of materials and specific components center. Provide Input Data to Contamination Transport Models -Applied to numerous flight projects over the past 20 years.

Building Service Provider Capabilities

DEFF Research Database (Denmark)

Brandl, Kristin; Jaura, Manya; Ørberg Jensen, Peter D.

2015-01-01

In this paper we study whether and how the interaction between clients and the service providers contributes to the development of capabilities in service provider firms. In situations where such a contribution occurs, we analyze how different types of activities in the production process...... process. We find that clients influence the development of human capital capabilities and management capabilities in reciprocally produced services. While in sequential produced services clients influence the development of organizational capital capabilities and management capital capabilities....... of the services, such as sequential or reciprocal task activities, influence the development of different types of capabilities. We study five cases of offshore-outsourced knowledge-intensive business services that are distinguished according to their reciprocal or sequential task activities in their production...

Pressure test at the reactor building of the Embalse Nuclear Power Plant (CNE)

International Nuclear Information System (INIS)

Coutsiers, E.E.; Perrino, J.; Moreno, C.; Batistic, J.A.; Lolis, R.R.; Aviles, A.

1991-01-01

Upon request by the Licensing Authority, the reactor building (RB) in a nuclear power plant must be submitted to pressure tests. One of these tests is to be performed before startup and, then, a test must be carried out every 5 years in operation. The pre-operational tests took place in August 1981, under two values of relative pressure: 1.266 kg/cm 2 and 0.422 kg/cm 2 . Operational tests must only be made at the lower pressure and their objective is to verify that the loss speed remains within the range indicated in the corresponding technical specification. The first operational test was performed in August 1989. The personnel of the CNE took care of the preparation of the Work Plan, of aligning the various systems contained in the RB, of pressurization, of monitoring localized tightedness, of depressurization and of the general and quality control of the test. The measurements were carried out by the CISME (Center of Metrology Research and Service) of the National Institute of Industrial Technology (INTI) , which did also supply the necesary instruments and the data collection system. There is also a description of the work performed before the test, of the calculation method used for assessing the loss rate, of the test sequencies and of the results obtained. (Author) [es

NGNP Data Management and Analysis System Analysis and Web Delivery Capabilities

Energy Technology Data Exchange (ETDEWEB)

Cynthia D. Gentillon

2011-09-01

Projects for the Very High Temperature Reactor (VHTR) Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the very high temperature reactor. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high-temperature and high-fluence environments. The NGNP Data Management and Analysis System (NDMAS) at the Idaho National Laboratory has been established to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the third NDMAS objective. It describes capabilities for displaying the data in meaningful ways and for data analysis to identify useful relationships among the measured quantities. The capabilities are described from the perspective of NDMAS users, starting with those who just view experimental data and analytical results on the INL NDMAS web portal. Web display and delivery capabilities are described in detail. Also the current web pages that show Advanced Gas Reactor, Advanced Graphite Capsule, and High Temperature Materials test results are itemized. Capabilities available to NDMAS developers are more extensive, and are described using a second series of examples. Much of the data analysis efforts focus on understanding how thermocouple measurements relate to simulated temperatures and other experimental parameters. Statistical control charts and correlation monitoring provide an ongoing assessment of instrument accuracy. Data analysis capabilities are virtually unlimited for those who use the NDMAS web data download capabilities and the analysis software of their choice. Overall, the NDMAS provides convenient data analysis and web delivery capabilities for studying a very large and rapidly increasing database of well-documented, pedigreed data.

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

Global Monitoring of the CTBT: Progress, Capabilities and Plans (Invited)

Science.gov (United States)

Zerbo, L.

2013-12-01

The Preparatory Commission for the Comprehensive Nuclear-Test-Ban Treaty Organization (CTBTO), established in 1996, is tasked with building up the verification regime of the CTBT. The regime includes a global system for monitoring the earth, the oceans and the atmosphere for nuclear tests, and an on-site inspection (OSI) capability. More than 80% of the 337 facilities of the International Monitoring System (IMS) have been installed and are sending data to the International Data Centre (IDC) in Vienna, Austria for processing. These IMS data along with IDC processed and reviewed products are available to all States that have signed the Treaty. Concurrent with the build-up of the global monitoring networks, near-field geophysical methods are being developed and tested for OSIs. The monitoring system is currently operating in a provisional mode, as the Treaty has not yet entered into force. Progress in installing and operating the IMS and the IDC and in building up an OSI capability will be described. The capabilities of the monitoring networks have progressively improved as stations are added to the IMS and IDC processing techniques refined. Detection thresholds for seismic, hydroacoustic, infrasound and radionuclide events have been measured and in general are equal to or lower than the predictions used during the Treaty negotiations. The measurements have led to improved models and tools that allow more accurate predictions of future capabilities and network performance under any configuration. Unplanned tests of the monitoring network occurred when the DPRK announced nuclear tests in 2006, 2009, and 2013. All three tests were well above the detection threshold and easily detected and located by the seismic monitoring network. In addition, noble gas consistent with the nuclear tests in 2006 and 2013 (according to atmospheric transport models) was detected by stations in the network. On-site inspections of these tests were not conducted as the Treaty has not entered

Liquid Rocket Engine Testing

Science.gov (United States)

Rahman, Shamim

2005-01-01

Comprehensive Liquid Rocket Engine testing is essential to risk reduction for Space Flight. Test capability represents significant national investments in expertise and infrastructure. Historical experience underpins current test capabilities. Test facilities continually seek proactive alignment with national space development goals and objectives including government and commercial sectors.

Design of capability measurement instruments pedagogic content knowledge (PCK) for prospective mathematics teachers

Science.gov (United States)

Aminah, N.; Wahyuni, I.

2018-05-01

The purpose of this study is to find out how the process of designing a tool of measurement Pedagogical Content Knowledge (PCK) capabilities, especially for prospective mathematics teachers are valid and practical. The design study of this measurement appliance uses modified Plomp development step, which consists of (1) initial assessment stage, (2) design stage at this stage, the researcher designs the measuring grille of PCK capability, (3) realization stage that is making measurement tool ability of PCK, (4) test phase, evaluation, and revision that is testing validation of measurement tools conducted by experts. Based on the results obtained that the design of PCK capability measurement tool is valid as indicated by the assessment of expert validator, and the design of PCK capability measurement tool, shown based on the assessment of teachers and lecturers as users of states strongly agree the design of PCK measurement tools can be used.

Capabilities required to conduct the LLNL plutonium mission

International Nuclear Information System (INIS)

Kass, J.; Bish, W.; Copeland, A.; West, J.; Sack, S.; Myers, B.

1991-01-01

This report outlines the LLNL plutonium related mission anticipated over the next decade and defines the capabilities required to meet that mission wherever the Plutonium Facility is located. If plutonium work is relocated to a place where the facility is shared, then some capabilities can be commonly used by the sharing parties. However, it is essential that LLNL independently control about 20000 sq ft of net lab space, filled with LLNL controlled equipment, and staffed by LLNL employees. It is estimated that the cost to construct this facility should range from $140M to $200M. Purchase and installation of equipment to replace that already in Bldg 332 along with additional equipment identified as being needed to meet the mission for the next ten to fifteen years, is estimated to cost $118M. About $29M of the equipment could be shared. The Hardened Engineering Test Building (HETB) with its additional 8000 sq ft of unique test capability must also be replaced. The fully equipped replacement cost is estimated to be about $10M. About 40000 sq ft of setup and support space are needed along with office and related facilities for a 130 person resident staff. The setup space is estimated to cost $8M. The annual cost of a 130 person resident staff (100 programmatic and 30 facility operation) is estimated to be $20M

PTC test bed upgrades to provide ACSES testing support capabilities at transportation technology center.

Science.gov (United States)

2015-06-01

FRA Task Order 314 upgraded the Positive Train Control (PTC) Test Bed at the Transportation Technology Center to support : testing of PTC systems, components, and related equipment associated with the Advanced Civil Speed Enforcement System : (ACSES)...

Helicopter precision approach capability using the Global Positioning System

Science.gov (United States)

Kaufmann, David N.

1992-01-01

The period between 1 July and 31 December, 1992, was spent developing a research plan as well as a navigation system document and flight test plan to investigate helicopter precision approach capability using the Global Positioning System (GPS). In addition, all hardware and software required for the research was acquired, developed, installed, and verified on both the test aircraft and the ground-based reference station.

Measurement of the two track separation capability of hybrid pixel sensors

Energy Technology Data Exchange (ETDEWEB)

Muñoz, F.J., E-mail: Francisca.MunozSanchez@manchester.ac.uk [University of Manchester (United Kingdom); Battaglia, M. [University of California, Santa Cruz, United States of America (United States); CERN, The European Organization for Nuclear Research (Switzerland); Da Vià, C. [University of Manchester (United Kingdom); La Rosa, A. [University of California, Santa Cruz, United States of America (United States); Dann, N. [University of Manchester (United Kingdom)

2017-02-11

Large Hadron Collider experiments face new challenges in Run-2 conditions due to the increased beam energy, the interest for searches of new physics signals with higher jet pT and the consequent longer decay length of heavy hadrons. In this new scenario, the capability of the innermost pixel sensors to distinguish tracks in very dense environment becomes crucial for efficient tracking and flavour tagging performance. In this work, we discuss the measurement in a test beam of the two track separation capability of hybrid pixel sensors using the interaction particles out of the collision of high energy pions on a thin copper target. With this method we are able to evaluate the effect of merged hits in the sensors under test due to tracks closer than the sensor spatial granularity in terms of collected charge, multiplicity and reconstruction efficiency. - Highlights: • Measurement of the two-track separation capability of hybrid pixel sensors. • Emulating track dense environment with a cooper target in a test beam. • Cooper target in between telescope arms to create vertices. • Validation of simulation and reconstruction algorithm for future vertex detectors. • New qualification method for pixel modules in track dense environments.

Gamma-Ray Emission Tomography: Modeling and Evaluation of Partial-Defect Testing Capabilities

International Nuclear Information System (INIS)

Jacobsson Svard, S.; Jansson, P.; Davour, A.; Grape, S.; White, T.A.; Smith, L.E.; Deshmukh, N.; Wittman, R.S.; Mozin, V.; Trellue, H.

2015-01-01

Gamma emission tomography (GET) for spent nuclear fuel verification is the subject for IAEA MSP project JNT1955. In line with IAEA Safeguards R&D plan 2012-2023, the aim of this effort is to ''develop more sensitive and less intrusive alternatives to existing NDA instruments to perform partial defect test on spent fuel assembly prior to transfer to difficult to access storage''. The current viability study constitutes the first phase of three, with evaluation and decision points between each phase. Two verification objectives have been identified; (1) counting of fuel pins in tomographic images without any a priori knowledge of the fuel assembly under study, and (2) quantitative measurements of pinby- pin properties, e.g., burnup, for the detection of anomalies and/or verification of operator-declared data. Previous measurements performed in Sweden and Finland have proven GET highly promising for detecting removed or substituted fuel rods in BWR and VVER-440 fuel assemblies even down to the individual fuel rod level. The current project adds to previous experiences by pursuing a quantitative assessment of the capabilities of GET for partial defect detection, across a broad range of potential IAEA applications, fuel types and fuel parameters. A modelling and performance-evaluation framework has been developed to provide quantitative GET performance predictions, incorporating burn-up and cooling-time calculations, Monte Carlo radiation-transport and detector-response modelling, GET instrument definitions (existing and notional) and tomographic reconstruction algorithms, which use recorded gamma-ray intensities to produce images of the fuel's internal source distribution or conclusive rod-by-rod data. The framework also comprises image-processing algorithms and performance metrics that recognize the inherent tradeoff between the probability of detecting missing pins and the false-alarm rate. Here, the modelling and analysis framework is

Does organizational agility affect organizational learning capability? Evidence from commercial banking

OpenAIRE

Zaina Mustafa Mahmoud Hamad; Uğur Yozgat

2017-01-01

Both organizational agility and learning capability are prerequisites for organizational survival and success. This study explores the contribution of agility practices to organizational learning capabilities at the commercial banks in Jordan. To examine the proposed model, a sample of 158 employees within top and middle managements was used. Structural Equation Modeling was conducted for assessing validity and reliability of measurement instrument, evaluating model fit, and testing hypothese...

Development and design of a high pressure carbon dioxide system for the separation of hazardous contaminants from non-hazardous debris

International Nuclear Information System (INIS)

Adkins, C.L.J.; Russick, E.M.; Smith, H.M.; Olson, R.B.

1995-01-01

Under the Department of Energy (DOE)/United States Air Force (USAF) Memorandum of Understanding, a system is being designed that will use high pressure carbon dioxide for the separation of oils, greases, and solvents from non-hazardous solid waste. The contaminants are dissolved into the high pressure carbon dioxide and precipitated out upon depressurization. The carbon dioxide solvent can then be recycled for continued use. Excellent extraction capability for common manufacturing oils, greases, and solvents has been measured. It has been observed that extraction performance follows the dilution model if a constant flow system is used. The solvents tested are extremely soluble and have been extracted to 100% under both liquid and mild supercritical carbon dioxide conditions. These data are being used to design a 200 liter extraction system

ORGANISATIONAL CAPABILITIES, COMPETITIVE ADVANTAGE AND PERFORMANCE IN SUPPORTING INDUSTRIES IN VIETNAM

Directory of Open Access Journals (Sweden)

Nham Phong Tuan

2010-01-01

Full Text Available This paper focuses on applying the resource-based view (RBV of firms to explain performance in supporting industries in Vietnam. Specifically, we based our research on the comprehensive framework of RBV and reviewed previous empirical researches before deciding on adopting a dynamic capabilities approach to test relationships among organisational capabilities, competitive advantage and performance. A multivariate analysis of survey responses of 102 firms belonging to supporting industries in Vietnam indicates that the organisational capabilities are related to the competitive advantage, that the competitive advantage is related to performance, and that the competitive advantage mediates the relationship between organizational capabilities and performance. These findings have considerable implications for academics as well as practitioners. Finally, this study also provides directions for future research.

Engineering evaluation of the General Motors (GM) diesel rating and capabilities

International Nuclear Information System (INIS)

Gross, R.E.

1992-04-01

K-Reactor's number one GM diesel (GM-lK) suffered recurrent, premature piston pin bushing failures between July 1990 and January 1991. These failures raised a concern that the engine's original design capabilities were being exceeded. Were we asking old engines to do too much by powering 1200 kw (continuous) rated electrical generators? Was excessive wear of the piston pin bushings a result of having exceeded the engine's capabilities (overload), or were the recent failures a direct result of poor quality, poor design, or defective replacement parts? Considering the engine's overall performance for the past 30 years, during which an engine failure of this nature had never occurred, and the fact that 1200 kw was approximately 50% of the engine's original tested capability, Reactor Engineering did not consider it likely that an overloaded engine caused bushing failures. What seemed more plausible was that the engine's failure to perform was caused by deficiencies in, or poor quality of, replacement parts.The following report documents: (1) the results of K-Reactor EDG failure analysis; (2) correlation of P- and C-Reactor GM diesel teardowns; (3) the engine rebuild to blueprint specification; (4) how the engine was determined ready for test; (5) testing parameters that were developed; (6) a summary of test results and test insights; (7) how WSRC determined engine operation was acceptable; (8) independent review of 1200 kw operational data; (9) approval of the engines' 12OOkw continuous rating

Engineering evaluation of the General Motors (GM) diesel rating and capabilities

Energy Technology Data Exchange (ETDEWEB)

Gross, R.E.

1992-04-01

K-Reactor`s number one GM diesel (GM-lK) suffered recurrent, premature piston pin bushing failures between July 1990 and January 1991. These failures raised a concern that the engine`s original design capabilities were being exceeded. Were we asking old engines to do too much by powering 1200 kw (continuous) rated electrical generators? Was excessive wear of the piston pin bushings a result of having exceeded the engine`s capabilities (overload), or were the recent failures a direct result of poor quality, poor design, or defective replacement parts? Considering the engine`s overall performance for the past 30 years, during which an engine failure of this nature had never occurred, and the fact that 1200 kw was approximately 50% of the engine`s original tested capability, Reactor Engineering did not consider it likely that an overloaded engine caused bushing failures. What seemed more plausible was that the engine`s failure to perform was caused by deficiencies in, or poor quality of, replacement parts.The following report documents: (1) the results of K-Reactor EDG failure analysis; (2) correlation of P- and C-Reactor GM diesel teardowns; (3) the engine rebuild to blueprint specification; (4) how the engine was determined ready for test; (5) testing parameters that were developed; (6) a summary of test results and test insights; (7) how WSRC determined engine operation was acceptable; (8) independent review of 1200 kw operational data; (9) approval of the engines` 12OOkw continuous rating.

Study On The Uranium Adsorption Capability Of Bone Black In Radioactive Waste Water Treatment

International Nuclear Information System (INIS)

Phan Dinh Tuan

2008-01-01

It has been found that bone black can adsorb uranium and radium from radioactive wastewater. Nevertheless, bone black is not so competitive for the low adsorption capability and the slow adsorption rate. The article describes the research results in increasing the uranium adsorption capability of bone black by treating it with hydrochloric acid. The influences of pH on adsorption capability and the results of batch- and column tests have been investigated. Column tests for elution process have pointed out that HCl is quite good eluent for uranium. It is recommended to apply the treated bone black for radioactive wastewater treatment and uranium recovery. (author)

Developing Alliance Capabilities

DEFF Research Database (Denmark)

Heimeriks, Koen H.; Duysters, Geert; Vanhaverbeke, Wim

This paper assesses the differential performance effects of learning mechanisms on the development of alliance capabilities. Prior research has suggested that different capability levels could be identified in which specific intra-firm learning mechanisms are used to enhance a firm's alliance...

Post Irradiation Capabilities at the Idaho National Laboratory

International Nuclear Information System (INIS)

Schulthess, J.L.; Rosenberg, K.E.

2011-01-01

The U.S. Department of Energy (DOE), Office of Nuclear Energy (NE) oversees the efforts to ensure nuclear energy remains a viable option for the United States. A significant portion of these efforts are related to post-irradiation examinations (PIE) of highly activated fuel and materials that are subject to the extreme environment inside a nuclear reactor. As the lead national laboratory, Idaho National Laboratory (INL) has a rich history, experience, workforce and capabilities for performing PIE. However, new advances in tools and techniques for performing PIE now enable understanding the performance of fuels and materials at the nano-scale and smaller level. Examination at this level is critical since this is the scale at which irradiation damage occurs. The INL is on course to adopt these advanced tools and techniques to develop a comprehensive nuclear fuels and materials characterization capability that is unique in the world. Because INL has extensive PIE capabilities currently in place, a strong foundation exist to build upon as new capabilities are implemented and work load increases. In the recent past, INL has adopted significant capability to perform advanced PIE characterization. Looking forward, INL is planning for the addition of two facilities that will be built to meet the stringent demands of advanced tools and techniques for highly activated fuels and materials characterization. Dubbed the Irradiated Materials Characterization Laboratory (IMCL) and Advanced Post Irradiation Examination Capability, these facilities are next generation PIE laboratories designed to perform the work of PIE that cannot be performed in current DOE facilities. In addition to physical capabilities, INL has recently added two significant contributors to the Advanced Test Reactor-National Scientific User Facility (ATR-NSUF), Oak Ridge National Laboratory and University of California, Berkeley.

Heat removal capability of core-catcher with inclined cooling channels

International Nuclear Information System (INIS)

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

2009-01-01

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

Molecular Tagging Velocimetry Development for In-situ Measurement in High-Temperature Test Facility

Science.gov (United States)

Andre, Matthieu A.; Bardet, Philippe M.; Burns, Ross A.; Danehy, Paul M.

2015-01-01

The High Temperature Test Facility, HTTF, at Oregon State University (OSU) is an integral-effect test facility designed to model the behavior of a Very High Temperature Gas Reactor (VHTR) during a Depressurized Conduction Cooldown (DCC) event. It also has the ability to conduct limited investigations into the progression of a Pressurized Conduction Cooldown (PCC) event in addition to phenomena occurring during normal operations. Both of these phenomena will be studied with in-situ velocity field measurements. Experimental measurements of velocity are critical to provide proper boundary conditions to validate CFD codes, as well as developing correlations for system level codes, such as RELAP5 (http://www4vip.inl.gov/relap5/). Such data will be the first acquired in the HTTF and will introduce a diagnostic with numerous other applications to the field of nuclear thermal hydraulics. A laser-based optical diagnostic under development at The George Washington University (GWU) is presented; the technique is demonstrated with velocity data obtained in ambient temperature air, and adaptation to high-pressure, high-temperature flow is discussed.

The role of empowering organization capabilities on efficiency of new product development

Directory of Open Access Journals (Sweden)

Hoda Nikakhtar

2014-03-01

Full Text Available This paper presents an empirical investigation on the effects of empowering organizational capabilities on new product-development efficiency improvement and the proposed study is applied in one of Iranian food producers in city of Tehran, Iran. The study considers seven components including technological capabilities, marketing mix capabilities, capabilities for communication with customers, quality of new products, fast entry to market capabilities, customer satisfaction and economic success. The study designs a questionnaire in Likert scale and distributes it among 384 randomly selected people who regularly use different food products. Cronbach alphas for all components of the survey are within acceptable limits and it confirms the overall questionnaire in terms of various questions. The study has used t-student test as well as structural equation modeling to examine different hypotheses of the survey.

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

Modeling and analysis of selected organization for economic cooperation and development PKL-3 station blackout experiments using TRACE

Directory of Open Access Journals (Sweden)

Roman Mukin

2018-04-01

Full Text Available A series of tests dedicated to station blackout (SBO accident scenarios have been recently performed at the Primärkreislauf-Versuchsanlage (primary coolant loop test facility; PKL facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal–hydraulic code TRACE (v5.0 Patch4 of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for

Acquisition Modernization: Transitioning Technology Into Warfighter Capability

Science.gov (United States)

2011-08-01

to test and evaluate the technology and integrate the new capability into operational weapon systems (Figure 4). This funding model creates stove...misalignment between missions, TRLs, and the RDT&E funding model is a major 11 contributor to the valley of death. Technologies become obsolete on... funding model of the acquisition system. Create an individual budget account to fund the development of promising technologies. The Acquisition

Development of an autonomous setup for evaluating self healing capability of asphalt mixtures

NARCIS (Netherlands)

Qiu, J.; Molenaar, A.A.A.; Van de Ven, M.F.C.; Wu, S.

2012-01-01

It is a well known fact that asphalt mixtures have self healing capabilities. Yet most of the self healing investigations are carried out using complex and time consuming fatigue tests. In order to investigate the self healing capability in a simple and efficient manner, a beam on elastic foundation

A Versatile Internet-Accessible Electronics Workbench with Troubleshooting Capabilities

Directory of Open Access Journals (Sweden)

Hamidou Soumare

2009-08-01

Full Text Available The MIT iLab Project was established to expand the range of laboratory experiences available to students in science and engineering education. iLabs are online laboratories that enable students to conduct real experiments remotely. Recently, the iLab Project has focused on building remote laboratories around the NI-ELVIS platform, an all-in-one electronics workbench. This paper will detail our recent efforts in expanding the capabilities of ELVIS-based iLabs by enabling students to test and debug digital and analog circuits. This work will enable students to perform remote experiments characterizing digital logic elements. By merging switching capabilities with the Digital Multimeter available on the ELVIS, students will have the ability to examine and troubleshoot circuits. These added capabilities will provide educators and students with unparalleled flexibility and significantly enrich the remote laboratory experience.

Toward Vision Oriented Organization through Foresight Capability Development

Directory of Open Access Journals (Sweden)

Gianita BLEOJU

2014-11-01

Full Text Available Dealing with complexity is becoming increasingly difficult for organizations, due to limited replicable abilities, once management performance was remunerated by successful decisions on the marketplace. The competitive advantage, based upon current documented organizational management expertise, deployed into patterns of competitive behavior, prove to be unsustainable. Therefore, we assist to a relative emergency of strategic intelligence adjustment framework, to channel the managerial capability mechanisms, from current detective orientation capabilities toward anticipatory ones. Based upon exploitation of an organizational profiling database, we try our contribution to this challenging debate, by formulating recommendations for strategic adjustment and prototype testing of the potential solutions, through a designed transition matrix from market oriented to vision oriented organizations.

Experiment data report for LOFT nonnuclear test L1-3

International Nuclear Information System (INIS)

Millar, G.M.

1977-04-01

Test L1-3 was the third in a series of five nonnuclear isothermal blowdown tests conducted by the Loss of Fluid Test (LOFT) Program. For this test the LOFT Facility was configured to simulate a loss-of-coolant accident in a large pressurized water reactor resulting from a 200 percent double-ended shear break in a cold leg of the primary coolant system. A hydraulic core simulator assembly was installed in place of the nuclear core. The initial conditions in the primary coolant system intact loop were: temperature at 540 0 F, pressure at 2256 psig, and loop flow at 2.34 x 10 6 lbm/hr. During system depressurization, emergency core cooling water was specified to be injected into the lower plenum of the reactor vessel using an accumulator, a low-pressure injection system pump, and a high-pressure injection system pump to provide data on the effects of emergency core cooling on the system thermal-hydraulic response. Injection into the lower plenum was initiated from the high- and low-pressure injection systems. Injection from the accumulator, however, was not initiated because a valve was inadvertently left closed. The experiment, therefore, was not completely successful in that one of the objectives outlined in the experiment operating specification for this test was not accomplished. Test L1-3 was repeated at Test L1-3A to meet the experimental requirements. Despite these difficulties, Test L1-3 did provide very valuable data to verify experiment repeatability

Test Capability of Comparative NAA Method in Analysis of Long Lived Element in SRM 1648

International Nuclear Information System (INIS)

Sri-Wardani

2005-01-01

The comparative NAA method had been examine on the analysis of long-lived elements content in air particulate sample of NIST.SRM 1648 for evaluation of a capability of comparative NAA method that used at P2TRR. From the result of analysis it could be determined analysis elements contained in the sample, namely: Sc, Co, Zn, Br, Rb, Sb, Hf and Th with optimum results in bias of 10%. The optimum result of long-lived elements obtained on a good accuracy and precision. From the analysis data obtained showed that the comparative NAA method with Gamma Trac and APTEC software capable to analyze several kinds of elements in environmental samples. Therefore, this method could be implement in biological and healthy samples. (author)

Engineering evaluation of the General Motors (GM) diesel rating and capabilities

Energy Technology Data Exchange (ETDEWEB)

Gross, R.E.

1992-04-01

K-Reactor's number one GM diesel (GM-lK) suffered recurrent, premature piston pin bushing failures between July 1990 and January 1991. These failures raised a concern that the engine's original design capabilities were being exceeded. Were we asking old engines to do too much by powering 1200 kw (continuous) rated electrical generators Was excessive wear of the piston pin bushings a result of having exceeded the engine's capabilities (overload), or were the recent failures a direct result of poor quality, poor design, or defective replacement parts Considering the engine's overall performance for the past 30 years, during which an engine failure of this nature had never occurred, and the fact that 1200 kw was approximately 50% of the engine's original tested capability, Reactor Engineering did not consider it likely that an overloaded engine caused bushing failures. What seemed more plausible was that the engine's failure to perform was caused by deficiencies in, or poor quality of, replacement parts.The following report documents: (1) the results of K-Reactor EDG failure analysis; (2) correlation of P- and C-Reactor GM diesel teardowns; (3) the engine rebuild to blueprint specification; (4) how the engine was determined ready for test; (5) testing parameters that were developed; (6) a summary of test results and test insights; (7) how WSRC determined engine operation was acceptable; (8) independent review of 1200 kw operational data; (9) approval of the engines' 12OOkw continuous rating.

Sacubitril/Valsartan: Effect on Walking Test and Physical Capability.

Science.gov (United States)

Sgorbini, Luca; Rossetti, Antonella; Galati, Alfonso

The 6-min walk test (6MWT) is a simple and inexpensive exercise test to evaluate physical functional capacity that is widely used in heart failure (HF) patients. With the 6MWT, a distance 50 m is considered clinically relevant. To our knowledge, information on improvement in physical functional capacity with sacubitril/valsartan, as assessed by the 6MWT, is still scant. In our daily practice, we apply this test to all patients whenever possible; therefore, we report here the findings observed in a small series of 5 patients with HF with reduced ejection fraction after a 1-month treatment with sacubitril/valsartan at full dose. The mean distance walked on the 6MWT at baseline was 129 m (±64 SD), and this value increased to 436 m (±156) after 1 month of therapy with sacubitril/valsartan 97/103 mg b.i.d. The mean difference from baseline was 305 m (±110). According to these preliminary findings, in clinical practice, a 1-month therapy of sacubitril/valsartan optimized at a 97/103-mg b.i.d. dose appears to be associated with a relevant improvement in the 6MWT. © 2017 S. Karger AG, Basel.

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

International Nuclear Information System (INIS)

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

1975-05-01

Recorded test data are presented for Test S-01-1B of the semiscale Mod-1 isothermal blowdown test series. System hardware is representative of the LOFT design, selected using volumetric scaling methods, and initial conditions duplicate those identified for the LOFT nonnuclear tests. Test S-01-1B is a repeat of Test S-01-1 with the exception that simulated ECC was injected into the cold leg of the intact loop rather than into the inlet annulus of the downcomer. The principal objective of Test S-01-1B was to determine whether a different ECC injection would significantly alter the system response during the period of ECC injection. Test S-01-1B was conducted from an initial temperature of 541 0 F and an initial pressure of 1630 psig. A simulated intermediate size double-ended hot leg break (0.00145 ft 2 break area on each end) was used to investigate the system response to a slow de-pressurization transient. An orificed structure was used in the pressure vessel to simulate the LOFT core simulator. Following the blowdown portion of Test S-01-1B, coolant spray was introduced into the pressure suppression tank to determine the response of the pressure suppression system. (U.S.)

Refractive Thinking Profile In Solving Mathematical Problem Reviewed from Students Math Capability

Science.gov (United States)

Maslukha, M.; Lukito, A.; Ekawati, R.

2018-01-01

Refraction is a mental activity experienced by a person to make a decision through reflective thinking and critical thinking. Differences in mathematical capability have an influence on the difference of student’s refractive thinking processes in solving math problems. This descriptive research aims to generate a picture of refractive thinking of students in solving mathematical problems in terms of students’ math skill. Subjects in this study consisted of three students, namely students with high, medium, and low math skills based on mathematics capability test. Data collection methods used are test-based methods and interviews. After collected data is analyzed through three stages that are, condensing and displaying data, data display, and drawing and verifying conclusion. Results showed refractive thinking profiles of three subjects is different. This difference occurs at the planning and execution stage of the problem. This difference is influenced by mathematical capability and experience of each subject.

Disentangling the effects of organizational capabilities, innovation and firm size on SME sales growth

NARCIS (Netherlands)

Uhlaner, Lorraine M.; van Stel, Andre; Duplat, Valerie; Zhou, Haibo

2013-01-01

This paper focuses on certain drivers of SME sales growth related to knowledge and innovation. Building on the dynamic capabilities literature, we test whether two organizational capabilities (external sourcing and employee involvement in renewal activities) predict sales growth, and if so, whether

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)

Transforming organizational capabilities in strategizing

DEFF Research Database (Denmark)

Jørgensen, Claus; Friis, Ole Uhrskov; Koch, Christian

2014-01-01

Offshored and networked enterprises are becoming an important if not leading organizational form and this development seriously challenges their organizational capabilities. More specifically, over the last years, SMEs have commenced entering these kinds of arrangements. As the organizational...... capabilities of SMEs are limited at the outset, even more emphasis is needed regarding the issues of developing relevant organizational capabilities. This paper aims at investigating how capabilities evolve during an offshoring process of more than 5 years in two Danish SMEs, i.e. not only short- but long......-term evolvements within the companies. We develop our framework of understanding organizational capabilities drawing on dynamic capability, relational capability and strategy as practice concepts, appreciating the performative aspects of developing new routines. Our two cases are taken from one author’s Ph...

ADVANCED SIMULATION CAPABILITY FOR ENVIRONMENTAL MANAGEMENT- CURRENT STATUS AND PHASE II DEMONSTRATION RESULTS

Energy Technology Data Exchange (ETDEWEB)

Seitz, R.

2013-02-26

The U.S. Department of Energy (USDOE) Office of Environmental Management (EM), Office of Soil and Groundwater, is supporting development of the Advanced Simulation Capability for Environmental Management (ASCEM). ASCEM is a state-of-the-art scientific tool and approach for understanding and predicting contaminant fate and transport in natural and engineered systems. The modular and open source high-performance computing tool facilitates integrated approaches to modeling and site characterization that enable robust and standardized assessments of performance and risk for EM cleanup and closure activities. The ASCEM project continues to make significant progress in development of computer software capabilities with an emphasis on integration of capabilities in FY12. Capability development is occurring for both the Platform and Integrated Toolsets and High-Performance Computing (HPC) Multiprocess Simulator. The Platform capabilities provide the user interface and tools for end-to-end model development, starting with definition of the conceptual model, management of data for model input, model calibration and uncertainty analysis, and processing of model output, including visualization. The HPC capabilities target increased functionality of process model representations, toolsets for interaction with Platform, and verification and model confidence testing. The Platform and HPC capabilities are being tested and evaluated for EM applications in a set of demonstrations as part of Site Applications Thrust Area activities. The Phase I demonstration focusing on individual capabilities of the initial toolsets was completed in 2010. The Phase II demonstration completed in 2012 focused on showcasing integrated ASCEM capabilities. For Phase II, the Hanford Site deep vadose zone (BC Cribs) served as an application site for an end-to-end demonstration of capabilities, with emphasis on integration and linkages between the Platform and HPC components. Other demonstrations

Desert Rats 2011 Mission Simulation: Effects of Microgravity Operational Modes on Fields Geology Capabilities

Science.gov (United States)

Bleacher, Jacob E.; Hurtado, J. M., Jr.; Meyer, J. A.

2012-01-01

Desert Research and Technology Studies (DRATS) is a multi-year series of NASA tests that deploy planetary surface hardware and exercise mission and science operations in difficult conditions to advance human and robotic exploration capabilities. DRATS 2011 (Aug. 30-Sept. 9, 2011) tested strategies for human exploration of microgravity targets such as near-Earth asteroids (NEAs). Here we report the crew perspective on the impact of simulated microgravity operations on our capability to conduct field geology.

Laboratory Evaluation of Energy Recovery Ventilators

Energy Technology Data Exchange (ETDEWEB)

Kosar, D. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

2016-11-01

Over the years, building scientists have characterized the relationship between building airtightness, exhaust-only appliances airflows, and building depressurization. Now, as the use of deep retrofit measures and new construction practices is growing to realize lower infiltration levels in increasingly tighter envelopes, performance issues can arise with the operation of exhaust-only appliances in a depressurized home. As the depressurization levels climb in tighter homes, many of these exhaust-only appliances see their rated airflows reduced and other related performance issues arise as a result. If sufficiently depressurized, atmospherically vented combustion appliances that may be present in the home can backdraft as well. Furthermore, when exhaust-only appliances operate and the tight home becomes depressurized, water vapor intrusion from outdoors can raise additional issues of mold in the building envelope in more humid climates.

Directed Design of Experiments for Validating Probability of Detection Capability of NDE Systems (DOEPOD)

Science.gov (United States)

Generazio, Edward R.

2015-01-01

Directed Design of Experiments for Validating Probability of Detection Capability of NDE Systems (DOEPOD) Manual v.1.2 The capability of an inspection system is established by applications of various methodologies to determine the probability of detection (POD). One accepted metric of an adequate inspection system is that there is 95% confidence that the POD is greater than 90% (90/95 POD). Design of experiments for validating probability of detection capability of nondestructive evaluation (NDE) systems (DOEPOD) is a methodology that is implemented via software to serve as a diagnostic tool providing detailed analysis of POD test data, guidance on establishing data distribution requirements, and resolving test issues. DOEPOD demands utilization of observance of occurrences. The DOEPOD capability has been developed to provide an efficient and accurate methodology that yields observed POD and confidence bounds for both Hit-Miss or signal amplitude testing. DOEPOD does not assume prescribed POD logarithmic or similar functions with assumed adequacy over a wide range of flaw sizes and inspection system technologies, so that multi-parameter curve fitting or model optimization approaches to generate a POD curve are not required. DOEPOD applications for supporting inspector qualifications is included.

Assessment of leak detection capability of Candu 6 annulus gas system using moisture injection tests

International Nuclear Information System (INIS)

Nho, Ki Man; Kim, Wang Bae

1998-01-01

The Candu 6 reactor assembly consists of an array of 380 pressure tubes, which are installed horizontally in a large cylindrical vessel, the Calandria, containing the low pressure heavy water moderator. The pressure tube is located inside calandria tube and the annulus between these tubes, which forms a closed loop with CO 2 gas recirculating, is called the Annulus Gas System (AGS). It is designed to give an alarm to the operator even for a small pressure tube leak by a very sensitive dew point meter so that he can take a preventive action for the pressure tbe rupture incident. To judge whether the operator action time is enough or not in the design of Wolsung 2, 3, and 4, the Leak Before Break (LBB) assessment is required for the analysis of the pressure tube failure accident. In order to provide the required data for the LBB assessment of Wolsung Units 2, 3, 4, a series of leak detection capability tests was performed by injecting controlled rates of heavy water vapour. The data of increased dew point and rates of rise were measured to determine the alarm set point for dew point rate of rise of Wolsung Unit 2. It was found that the response of the dew point depends on the moisture injection rate, CO 2 gas flow rate and the leak location. The test showed that Candu 6 AGS can detect the very small leaks less than few g/hr and dew point rate of rise alarm can be the most reliable alarm signal to warn the operator. Considering the present results, the first response time of dew point to the AGS CO 2 flow rate is approximated. (author)

Development and testing of a zero stitch MLI blanket using plastic pins for space use

Science.gov (United States)

Hatakenaka, Ryuta; Miyakita, Takeshi; Sugita, Hiroyuki; Saitoh, Masanori; Hirai, Tomoyuki

2014-11-01

New types of MLI blanket have been developed to achieve high thermal performance while maintaining production and assembly workability equivalent to the conventional type. Tag-pins, which are widely used in commercial applications to hook price tags to products, are used to fix the films in place and the pin material is changed to polyetheretherketone (PEEK) for use in space. Thermal performance is measured by using a boil-off calorimeter, in which a rectangular liquid nitrogen tank is used to evaluate the degradation at the bending corner and joint of the blanket. Zero-stitch- and multi-blanket-type MLIs show significantly improved thermal performance (ɛeff is smaller than 0.0050 at room temperature) despite having the same fastener interface as traditional blankets, while the venting design and number of tag-pins are confirmed as appropriate in a depressurization test.

A Study Effects Architectural Marketing Capabilities on Performance Marketing unit Based on: Morgan et al case: Past Industry in Tehran

OpenAIRE

Mohammad Reza Dalvi; Robabe Seifi

2014-01-01

Over a period of time architectural marketing capabilities combination of knowledge and skills develop in to capabilities. These architectural marketing capabilities have been identified as one of the important ways firms can achieve a competitive advantage The following research tests effects architectural marketing capabilities on performance marketing unit Based on a survey .a structural equation model was developed to test our hypotheses. the study develops a structural model linking arch...

A business analytics capability framework

Directory of Open Access Journals (Sweden)

Ranko Cosic

2015-09-01

Full Text Available Business analytics (BA capabilities can potentially provide value and lead to better organisational performance. This paper develops a holistic, theoretically-grounded and practically relevant business analytics capability framework (BACF that specifies, defines and ranks the capabilities that constitute an organisational BA initiative. The BACF was developed in two phases. First, an a priori conceptual framework was developed based on the Resource-Based View theory of the firm and a thematic content analysis of the BA literature. Second, the conceptual framework was further developed and refined using a three round Delphi study involving 16 BA experts. Changes from the Delphi study resulted in a refined and confirmed framework including detailed capability definitions, together with a ranking of the capabilities based on importance. The BACF will help academic researchers and industry practitioners to better understand the capabilities that constitute an organisational BA initiative and their relative importance. In future work, the capabilities in the BACF will be operationalised to measure their as-is status, thus enabling organisations to identify key areas of strength and weakness and prioritise future capability improvement efforts.

Sensor Alerting Capability

Science.gov (United States)

Henriksson, Jakob; Bermudez, Luis; Satapathy, Goutam

2013-04-01

There is a large amount of sensor data generated today by various sensors, from in-situ buoys to mobile underwater gliders. Providing sensor data to the users through standardized services, language and data model is the promise of OGC's Sensor Web Enablement (SWE) initiative. As the amount of data grows it is becoming difficult for data providers, planners and managers to ensure reliability of data and services and to monitor critical data changes. Intelligent Automation Inc. (IAI) is developing a net-centric alerting capability to address these issues. The capability is built on Sensor Observation Services (SOSs), which is used to collect and monitor sensor data. The alerts can be configured at the service level and at the sensor data level. For example it can alert for irregular data delivery events or a geo-temporal statistic of sensor data crossing a preset threshold. The capability provides multiple delivery mechanisms and protocols, including traditional techniques such as email and RSS. With this capability decision makers can monitor their assets and data streams, correct failures or be alerted about a coming phenomena.

Fused Reality for Enhanced Flight Test Capabilities, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — While modern ground-based flight simulators continue to improve in fidelity and effectiveness, there remains no substitute for flight test evaluations. In addition...

Variable reluctance displacement transducer temperature compensated to 6500F

International Nuclear Information System (INIS)

1975-01-01

In pressurized water reactor tests, compact instruments for accurate measurement of small displacements in a 650 0 F environment are often required. In the case of blowdown tests such as the Loss of Fluid Test (LOFT) or Semiscale computer code development tests, not only is the initial environment water at 650 0 F and 2200 psi but it undergoes a severe transient due to depressurization. Since the LOFT and Semiscale tests are run just for the purpose of obtaining data during the depressurization, instruments used to obtain the data must not give false outputs induced by the change in environment. A LOFT rho v 2 probe and a Semiscale drag disk are described. Each utilizes a variable reluctance transducer (VRT) for indication of the drag-disk location and a torsion bar for drag-disk restoring force. The VRT, in addition to being thermally gain and null offset stable, is fabricated from materials known to be resistant to large nuclear radiation levels and has successfully passed a fast neutron radiation test of 2.7 x 10 17 nvt without failure

Fused Reality for Enhanced Flight Test Capabilities, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — In terms of relevancy to piloted evaluation, there remains no substitute for actual flight tests even when considering the fidelity and effectiveness of modern...

Inward Internationalization of African-Ethnic Restaurants: The Role of Managerial Dynamic Capabilities

Directory of Open Access Journals (Sweden)

Samson Omuudu OTENGEI

2016-12-01

Full Text Available This paper empirically tests the link between cultural orientation, market orientation, service innovation capabilities of owner-managers and the level of inward internationalisation of African-ethnic restaurants. Data were obtained from 201 African-ethnic restaurants. A twostep approach was used, i.e. (i a confirmatory factor analysis (CFA to generate a measurement model; and (ii a structural equation model (SEM to test the hypotheses. All the predictor variables are positively and significantly related to the degree of inward internationalisation and the relationship is strong enough to cause a 41 per cent variance. Since it is cross-sectional in design, the study cannot be fully relied upon to draw inferences regarding causality among the variables. The results suggest that owner-managers who develop the three capabilities succeed in enhancing inward internationalisation. The current study could be among the very few to propose a framework that brings together major managerial (dynamic capabilities that are presumed to be vital in enhancing inward internationalisation.

Experimental test of the background rejection, through imaging capability, of a highly segmented AGATA germanium detector

International Nuclear Information System (INIS)

Doncel, M.; Recchia, F.; Quintana, B.; Gadea, A.; Farnea, E.

2010-01-01

The development of highly segmented germanium detectors as well as the algorithms to identify the position of the interaction within the crystal opens the possibility to locate the γ-ray source using Compton imaging algorithms. While the Compton-suppression shield, coupled to the germanium detector in conventional arrays, works also as an active filter against the γ rays originated outside the target, the new generation of position sensitive γ-ray detector arrays has to fully rely on tracking capabilities for this purpose. In specific experimental conditions, as the ones foreseen at radioactive beam facilities, the ability to discriminate background radiation improves the sensitivity of the gamma spectrometer. In this work we present the results of a measurement performed at the Laboratori Nazionali di Legnaro (LNL) aiming the evaluation of the AGATA detector capabilities to discriminate the origin of the γ rays on an event-by-event basis. It will be shown that, exploiting the Compton scattering formula, it is possible to track back γ rays coming from different positions, assigning them to specific emitting locations. These imaging capabilities are quantified for a single crystal AGATA detector.

Dynamic capabilities for cooperation in Brazilian multinational and factors determining its management

Directory of Open Access Journals (Sweden)

Priscila Rezende da Costa

2015-05-01

Full Text Available In the context of emerging companies’ growth, current challenges depend on the local generation of product and process innovations, as well as dynamic capability to generate innovative solutions cooperatively and new globe business models. The objective of this study is to analyse the determining managerial factors for the dynamic capability of cooperation in Brazilian multinationals (BMNs. A survey was conducted with a sample of 60 BMNs, and a structured questionnaire and statistical tests with factorial analysis and Cronbach's alpha were used. The aggregate analysis of the results indicates that BMNs are going through a transitional process between the operational capability of cooperation and the capability for dynamic cooperation, which are relevant to the continuous adaptation of already-established cooperative management routines and the evaluation and incorporation of the relational capability of management practices that consider systemic flows, open innovation and integrate strategic intentionality into cooperative processes.

Development of a Fissile Materials Irradiation Capability for Advanced Fuel Testing at the MIT Research Reactor

International Nuclear Information System (INIS)

Hu Linwen; Bernard, John A.; Hejzlar, Pavel; Kohse, Gordon

2005-01-01

A fissile materials irradiation capability has been developed at the Massachusetts Institute of Technology (MIT) Research Reactor (MITR) to support nuclear engineering studies in the area of advanced fuels. The focus of the expected research is to investigate the basic properties of advanced nuclear fuels using small aggregates of fissile material. As such, this program is intended to complement the ongoing fuel evaluation programs at test reactors. Candidates for study at the MITR include vibration-packed annular fuel for light water reactors and microparticle fuels for high-temperature gas reactors. Technical considerations that pertain to the design of the MITR facility are enumerated including those specified by 10 CFR 50 concerning the definition of a research reactor and those contained in a separate license amendment that was issued by the U.S. Nuclear Regulatory Commission to MIT for these types of experiments. The former includes limits on the cross-sectional area of the experiment, the physical form of the irradiated material, and the removal of heat. The latter addresses experiment reactivity worth, thermal-hydraulic considerations, avoidance of fission product release, and experiment specific temperature scrams

Experiment data report for semiscale MOD-1 test S-01-3 (isothermal blowdown with core resistance simulator)

International Nuclear Information System (INIS)

Zender, S.N.

1975-03-01

Recorded test data are presented for Test S-01-3 of the semiscale Mod-1 isothermal blowdown test series. Test S-01-3 is one of several semiscale Mod-1 experiments which are counterparts of the planned Loss-of-Fluid Test (LOFT) nonnuclear experiments. System hardware is of the LOFT design, selected using volumetric scaling methods, and initial conditions duplicate those identified for the LOFT nonnuclear tests. Test S-01-3 employed an intact loop resistance that was low relative to that of the first test in the series (Test S-01-2) to establish the importance of intact loop resistance on system response during blowdown. An orificed structure was used in the pressure vessel to simulate the LOFT core simulator. The test was initiated at isothermal conditions of 2245 psig and 538 0 F by a simulated offset shear of the cold-leg broken loop piping. During system depressurization, coolant was injected into the lower plenum of the pressure vessel to provide data on the effects of emergency core cooling on system response. Additionally, to aid in determination of the effects of accumulator gas on pressure suppression system response, the nitrogen used to charge the accumulator systems for Test S-01-3 was allowed to vent into the lower plenum following depletion of the coolant. (U.S.)

Resource-Based Capability on Development Knowledge Management Capabilities of Coastal Community

Science.gov (United States)

Teniwut, Roberto M. K.; Hasyim, Cawalinya L.; Teniwut, Wellem A.

2017-10-01

Building sustainable knowledge management capabilities in the coastal area might face a whole new challenge since there are many intangible factors involved from openness on new knowledge, access and ability to use the latest technology to the various local wisdom that still in place. The aimed of this study was to identify and analyze the resource-based condition of coastal community in this area to have an empirical condition of tangible and intangible infrastructure on developing knowledge management capability coastal community in Southeast Maluku, Indonesia. We used qualitative and quantitative analysis by depth interview and questionnaire for collecting the data with multiple linear regression as our analysis method. The result provided the information on current state of resource-based capability of a coastal community in this Southeast Maluku to build a sustainability model of knowledge management capabilities especially on utilization marine and fisheries resources. The implication of this study can provide an empirical information for government, NGO and research institution to dictate on how they conducted their policy and program on developing coastal community region.

Advanced Simulation Capability for Environmental Management: Development and Demonstrations - 12532

Energy Technology Data Exchange (ETDEWEB)

Freshley, Mark D.; Freedman, Vicky; Gorton, Ian [Pacific Northwest National Laboratory, MSIN K9-33, P.O. Box 999, Richland, WA 99352 (United States); Hubbard, Susan S. [Lawrence Berkeley National Laboratory, 1 Cyclotron Road, MS 50B-4230, Berkeley, CA 94720 (United States); Moulton, J. David; Dixon, Paul [Los Alamos National Laboratory, MS B284, P.O. Box 1663, Los Alamos, NM 87544 (United States)

2012-07-01

The U.S. Department of Energy Office of Environmental Management (EM), Technology Innovation and Development is supporting development of the Advanced Simulation Capability for Environmental Management (ASCEM). ASCEM is a state-of-the-art scientific tool and approach for understanding and predicting contaminant fate and transport in natural and engineered systems. The modular and open source high-performance computing tool facilitates integrated approaches to modeling and site characterization that enable robust and standardized assessments of performance and risk for EM cleanup and closure activities. The ASCEM project continues to make significant progress in development of capabilities, which are organized into Platform and Integrated Tool-sets and a High-Performance Computing Multi-process Simulator. The Platform capabilities target a level of functionality to allow end-to-end model development, starting with definition of the conceptual model and management of data for model input. The High-Performance Computing capabilities target increased functionality of process model representations, tool-sets for interaction with Platform, and verification and model confidence testing. The new capabilities are demonstrated through working groups, including one focused on the Hanford Site Deep Vadose Zone. The ASCEM program focused on planning during the first year and executing a prototype tool-set for an early demonstration of individual components. Subsequently, ASCEM has focused on developing and demonstrating an integrated set of capabilities, making progress toward a version of the capabilities that can be used to engage end users. Demonstration of capabilities continues to be implemented through working groups. Three different working groups, one focused on EM problems in the deep vadose zone, another investigating attenuation mechanisms for metals and radionuclides, and a third focusing on waste tank performance assessment, continue to make progress. The project

Radon mitigation experience in houses with basements and adjoining crawl spaces

International Nuclear Information System (INIS)

Messing, M.; Henschel, D.B.

1990-01-01

Active soil depressurization systems were installed in four basement houses with adjoining crawl spaces in Maryland. In addition, existing soil depressurization systems were modified in two additional basement-plus-crawl-space houses. These six houses were selected to include both good and poor communication beneath the basement slab, and different degrees of importance of the crawl space as a source of the indoor radon. The radon reduction effectiveness was compared for: depressurization only under the basement slab; depressurization only under a polyethylene liner over the unpaved crawl-space floor; and simultaneous depressurization under both the basement slab and the crawl-space liner. The objective of this paper is to identify under what conditions treatment of the basement alone might provide sufficient radon reductions in houses of this substructure, and what incremental benefits might be achieved by also treating the crawl space

Campus Capability Plan

Energy Technology Data Exchange (ETDEWEB)

Adams, C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Arsenlis, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bailey, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bergman, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Brase, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Brenner, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Camara, L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Carlton, H. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Cheng, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chrzanowski, P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Colson, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); East, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Farrell, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ferranti, L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Gursahani, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hansen, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Helms, L. L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hernandez, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jeffries, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Larson, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Lu, K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); McNabb, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mercer, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Skeate, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sueksdorf, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Zucca, B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Le, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ancria, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Scott, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Leininger, L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Gagliardi, F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Gash, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bronson, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chung, B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hobson, B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Meeker, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sanchez, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Zagar, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Quivey, B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sommer, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Atherton, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2017-06-06

Lawrence Livermore National Laboratory Campus Capability Plan for 2018-2028. Lawrence Livermore National Laboratory (LLNL) is one of three national laboratories that are part of the National Nuclear Security Administration. LLNL provides critical expertise to strengthen U.S. security through development and application of world-class science and technology that: Ensures the safety, reliability, and performance of the U.S. nuclear weapons stockpile; Promotes international nuclear safety and nonproliferation; Reduces global danger from weapons of mass destruction; Supports U.S. leadership in science and technology. Essential to the execution and continued advancement of these mission areas are responsive infrastructure capabilities. This report showcases each LLNL capability area and describes the mission, science, and technology efforts enabled by LLNL infrastructure, as well as future infrastructure plans.

NASA Capabilities That Could Impact Terrestrial Smart Grids of the Future

Science.gov (United States)

Beach, Raymond F.

2015-01-01

Incremental steps to steadily build, test, refine, and qualify capabilities that lead to affordable flight elements and a deep space capability. Potential Deep Space Vehicle Power system characteristics: power 10 kilowatts average; two independent power channels with multi-level cross-strapping; solar array power 24 plus kilowatts; multi-junction arrays; lithium Ion battery storage 200 plus ampere-hours; sized for deep space or low lunar orbit operation; distribution120 volts secondary (SAE AS 5698); 2 kilowatt power transfer between vehicles.

Development of students learning capabilities and professional capabilities

DEFF Research Database (Denmark)

Ringtved, Ulla Lunde; Wahl, Christian; Belle, Gianna

This paper describes the work-in-progress on a project that aims todevelop a tool that via learning analytic methods enable studentsto enhance, document and assess the development of their learningcapabilities and professional capabilities in consequence of theirself-initiated study activities...... during their bachelor educations. Thetool aims at enhancing the development of students’ capabilities toself-initiate, self-regulate and self-assess their study activities.The tool uses the concept of collective intelligence as source formotivation and inspiration in self-initiating study activities...... as wellas self-assessing them. The tool is based on a heutagogical approachto support reflection on learning potential in these activities. Thisenhances the educational use of students self-initiated learningactivities by bringing visibility and evidence to them, and therebybringing value to the assessment...

Organizational Context and Capabilities for Integrating Care: A Framework for Improvement

Directory of Open Access Journals (Sweden)

Jenna M. Evans

2016-08-01

Full Text Available Background: Interventions aimed at integrating care have become widespread in healthcare; however, there is significant variability in their success. Differences in organizational contexts and associated capabilities may be responsible for some of this variability. Purpose: This study develops and validates a conceptual framework of organizational capabilities for integrating care, identifies which of these capabilities may be most important, and explores the mechanisms by which they influence integrated care efforts.  Methods: The Context and Capabilities for Integrating Care (CCIC Framework was developed through a literature review, and revised and validated through interviews with leaders and care providers engaged in integrated care networks in Ontario, Canada. Interviews involved open-ended questions and graphic elicitation. Quantitative content analysis was used to summarize the data.  Results: The CCIC Framework consists of eighteen organizational factors in three categories: Basic Structures, People and Values, and Key Processes. The three most important capabilities shaping the capacity of organizations to implement integrated care interventions include Leadership Approach, Clinician Engagement and Leadership, and Readiness for Change. The majority of hypothesized relationships among organizational capabilities involved Readiness for Change and Partnering, emphasizing the complexity, interrelatedness and importance of these two factors to integrated care efforts.  Conclusions: Organizational leaders can use the framework to determine readiness to integrate care, develop targeted change management strategies, and select appropriate partners with overlapping or complementary profiles on key capabilities. Researchers may use the results to test and refine the proposed framework, with a focus on the hypothesized relationships among organizational capabilities and between organizational capabilities and performance outcomes.

Organizational Context and Capabilities for Integrating Care: A Framework for Improvement

Science.gov (United States)

Grudniewicz, Agnes; Baker, G. Ross; Wodchis, Walter P.

2016-01-01

Background: Interventions aimed at integrating care have become widespread in healthcare; however, there is significant variability in their success. Differences in organizational contexts and associated capabilities may be responsible for some of this variability. Purpose: This study develops and validates a conceptual framework of organizational capabilities for integrating care, identifies which of these capabilities may be most important, and explores the mechanisms by which they influence integrated care efforts. Methods: The Context and Capabilities for Integrating Care (CCIC) Framework was developed through a literature review, and revised and validated through interviews with leaders and care providers engaged in integrated care networks in Ontario, Canada. Interviews involved open-ended questions and graphic elicitation. Quantitative content analysis was used to summarize the data. Results: The CCIC Framework consists of eighteen organizational factors in three categories: Basic Structures, People and Values, and Key Processes. The three most important capabilities shaping the capacity of organizations to implement integrated care interventions include Leadership Approach, Clinician Engagement and Leadership, and Readiness for Change. The majority of hypothesized relationships among organizational capabilities involved Readiness for Change and Partnering, emphasizing the complexity, interrelatedness and importance of these two factors to integrated care efforts. Conclusions: Organizational leaders can use the framework to determine readiness to integrate care, develop targeted change management strategies, and select appropriate partners with overlapping or complementary profiles on key capabilities. Researchers may use the results to test and refine the proposed framework, with a focus on the hypothesized relationships among organizational capabilities and between organizational capabilities and performance outcomes. PMID:28413366

Irradiation capability of Japanese materials test reactor for water chemistry experiments

International Nuclear Information System (INIS)

Hanawa, Satoshi; Hata, Kuniki; Chimi, Yasuhiro; Nishiyama, Yutaka; Nakamura, Takehiko

2012-09-01

Appropriate understanding of water chemistry in the core of LWRs is essential as chemical species generated due to water radiolysis by neutron and gamma-ray irradiation govern corrosive environment of structural materials in the core and its periphery, causing material degradation such as stress corrosion cracking. Theoretical model calculation such as water radiolysis calculation gives comprehensive understanding of water chemistry at irradiation field where we cannot directly monitor. For enhancement of the technology, accuracy verification of theoretical models under wide range of irradiation conditions, i.e. dose rate, temperature etc., with well quantified in-pile measurement data is essential. Japan Atomic Energy Agency (JAEA) has decided to launch water chemistry experiments for obtaining data that applicable to model verification as well as model benchmarking, by using an in-pile loop which will be installed in the Japan Materials Testing Reactor (JMTR). In order to clarify the irradiation capability of the JMTR for water chemistry experiments, preliminary investigations by water radiolysis / ECP model calculations were performed. One of the important irradiation conditions for the experiments, i.e. dose rate by neutron and gamma-ray, can be controlled by selecting irradiation position in the core. In this preliminary study, several representative irradiation positions that cover from highest to low absorption dose rate were chosen and absorption dose rate at the irradiation positions were evaluated by MCNP calculations. As a result of the calculations, it became clear that the JMTR could provide the irradiation conditions close to the BWR. The calculated absorption dose rate at each irradiation position was provided to water radiolysis calculations. The radiolysis calculations were performed under various conditions by changing absorption dose rate, water chemistry of feeding water etc. parametrically. Qualitatively, the concentration of H 2 O 2 , O 2 and

Capitalizing on capabilities.

Science.gov (United States)

Ulrich, Dave; Smallwood, Norm

2004-06-01

By making the most of organizational capabilities--employees' collective skills and fields of expertise--you can dramatically improve your company's market value. Although there is no magic list of proficiencies that every organization needs in order to succeed, the authors identify 11 intangible assets that well-managed companies tend to have: talent, speed, shared mind-set and coherent brand identity, accountability, collaboration, learning, leadership, customer connectivity, strategic unity, innovation, and efficiency. Such companies typically excel in only three of these capabilities while maintaining industry parity in the other areas. Organizations that fall below the norm in any of the 11 are likely candidates for dysfunction and competitive disadvantage. So you can determine how your company fares in these categories (or others, if the generic list doesn't suit your needs), the authors explain how to conduct a "capabilities audit," describing in particular the experiences and findings of two companies that recently performed such audits. In addition to highlighting which intangible assets are most important given the organization's history and strategy, this exercise will gauge how well your company delivers on its capabilities and will guide you in developing an action plan for improvement. A capabilities audit can work for an entire organization, a business unit, or a region--indeed, for any part of a company that has a strategy to generate financial or customer-related results. It enables executives to assess overall company strengths and weaknesses, senior leaders to define strategy, midlevel managers to execute strategy, and frontline leaders to achieve tactical results. In short, it helps turn intangible assets into concrete strengths.

Test and evaluation capabilities at NAVELEXCEN Charleston

Energy Technology Data Exchange (ETDEWEB)

Stalvey, T.W.; Anderson, G.B.; Hinson, T.L. [Naval Electronic Systems Engineering Center, Charleston, SC (United States)

1993-12-31

The Environmental Systems and Instrumentation Engineering Department is located within the Special Programs Directorate of the Naval Electronic Systems Engineering Center (NAVELEXCEN Charleston). This Center is an echelon 4 Command under the Naval Command Control and Ocean Surveillance Center, San Diego (NCCOSC). NCCOSC is an echelon 3 Command under the Space and Warfare Systems Command (SPAWAR) which is located in Washington DC. Radiation Detection, Indication and Computation (RDIAC) equipment life-cycle management for the entire Navy falls under the auspices of the Naval Sea Systems Command (SEA 04R). The RADIAC Program provides centralized management for the execution of research, development, test, evaluation, maintenance, procurement, allowance, and equipment support for all Navy RADIAC instrumentation and assigned special monitoring equipments. RADIAC equipment is used throughout the Navy to support various functions associated with radioactivity, potential contamination, and personnel exposure to sources of ionizing radiation. Common sources in today`s Navy include nuclear reactors, nuclear weapons, industrial radiography, and nuclear medicine. Types of radiation includes gamma, x-ray, alpha, and beta.

The evolution of alliance capabilities

NARCIS (Netherlands)

Heimeriks, K.H.; Duysters, G.M.; Vanhaverbeke, W.P.M.

2004-01-01

This paper assesses the effectiveness and differential performance effects of learning mechanisms on the evolution of alliance capabilities. Relying on the concept of capability lifecycles, prior research has suggested that different capability levels could be identified in which different

Temperature dependent power capability estimation of lithium-ion batteries for hybrid electric vehicles

International Nuclear Information System (INIS)

Zheng, Fangdan; Jiang, Jiuchun; Sun, Bingxiang; Zhang, Weige; Pecht, Michael

2016-01-01

The power capability of lithium-ion batteries affects the safety and reliability of hybrid electric vehicles and the estimate of power by battery management systems provides operating information for drivers. In this paper, lithium ion manganese oxide batteries are studied to illustrate the temperature dependency of power capability and an operating map of power capability is presented. Both parametric and non-parametric models are established in conditions of temperature, state of charge, and cell resistance to estimate the power capability. Six cells were tested and used for model development, training, and validation. Three samples underwent hybrid pulse power characterization tests at varied temperatures and were used for model parameter identification and model training. The other three were used for model validation. By comparison, the mean absolute error of the parametric model is about 29 W, and that of the non-parametric model is around 20 W. The mean relative errors of two models are 0.076 and 0.397, respectively. The parametric model has a higher accuracy in low temperature and state of charge conditions, while the non-parametric model has better estimation result in high temperature and state of charge conditions. Thus, two models can be utilized together to achieve a higher accuracy of power capability estimation. - Highlights: • The temperature dependency of power capability of lithium-ion battery is investigated. • The parametric and non-parametric power capability estimation models are proposed. • An exponential function is put forward to compensate the effects of temperature. • A comparative study on the accuracy of two models using statistical metrics is presented.

The use of stability indices in predicting asphaltene problems in upstream and downstream oil operations

Energy Technology Data Exchange (ETDEWEB)

Asomaning, S. [Baker Petrolite, Sugar Land, TX (United States)

2003-07-01

A series of test methods have been developed to determine the stability of asphaltenes in crude oils. They were developed due to the high cost of remediating asphaltene deposition in offshore operations. This study described the characteristics of the Oliensis Spot Test, two saturates, aromatics, resins and asphaltenes (SARA)-based screens (the Colloidal Instability Index and Asphaltene-Resin ratio), a solvent titration method with near infrared radiation (NIR) solids detection, and live oil depressurization. Each method is used to predict the stability of asphaltenes in crude oils with different API gravity. A complete description of the test methods was provided along with experimental data. The effectiveness of the different tests in predicting the stability of asphaltenes in crude oils was also assessed. Results indicate that the prediction of a crude oil's tendency towards asphaltene precipitation was more accurate with the Colloidal Instability Index and the solvent titration method. Live oil depressurization proved to be very effective in predicting the stability of asphaltenes for light oils, where most stability tests fail. tabs., figs.

Advanced Simulation Capability for Environmental Management - Current Status and Phase II Demonstration Results - 13161

Energy Technology Data Exchange (ETDEWEB)

Seitz, Roger R.; Flach, Greg [Savannah River National Laboratory, Savannah River Site, Bldg 773-43A, Aiken, SC 29808 (United States); Freshley, Mark D.; Freedman, Vicky; Gorton, Ian [Pacific Northwest National Laboratory, MSIN K9-33, P.O. Box 999, Richland, WA 99352 (United States); Dixon, Paul; Moulton, J. David [Los Alamos National Laboratory, MS B284, P.O. Box 1663, Los Alamos, NM 87544 (United States); Hubbard, Susan S.; Faybishenko, Boris; Steefel, Carl I.; Finsterle, Stefan [Lawrence Berkeley National Laboratory, 1 Cyclotron Road, MS 50B-4230, Berkeley, CA 94720 (United States); Marble, Justin [Department of Energy, 19901 Germantown Road, Germantown, MD 20874-1290 (United States)

2013-07-01

The U.S. Department of Energy (US DOE) Office of Environmental Management (EM), Office of Soil and Groundwater, is supporting development of the Advanced Simulation Capability for Environmental Management (ASCEM). ASCEM is a state-of-the-art scientific tool and approach for understanding and predicting contaminant fate and transport in natural and engineered systems. The modular and open source high-performance computing tool facilitates integrated approaches to modeling and site characterization that enable robust and standardized assessments of performance and risk for EM cleanup and closure activities. The ASCEM project continues to make significant progress in development of computer software capabilities with an emphasis on integration of capabilities in FY12. Capability development is occurring for both the Platform and Integrated Tool-sets and High-Performance Computing (HPC) Multi-process Simulator. The Platform capabilities provide the user interface and tools for end-to-end model development, starting with definition of the conceptual model, management of data for model input, model calibration and uncertainty analysis, and processing of model output, including visualization. The HPC capabilities target increased functionality of process model representations, tool-sets for interaction with Platform, and verification and model confidence testing. The Platform and HPC capabilities are being tested and evaluated for EM applications in a set of demonstrations as part of Site Applications Thrust Area activities. The Phase I demonstration focusing on individual capabilities of the initial tool-sets was completed in 2010. The Phase II demonstration completed in 2012 focused on showcasing integrated ASCEM capabilities. For Phase II, the Hanford Site deep vadose zone (BC Cribs) served as an application site for an end-to-end demonstration of capabilities, with emphasis on integration and linkages between the Platform and HPC components. Other demonstrations

Advanced Simulation Capability for Environmental Management - Current Status and Phase II Demonstration Results - 13161

International Nuclear Information System (INIS)

Seitz, Roger R.; Flach, Greg; Freshley, Mark D.; Freedman, Vicky; Gorton, Ian; Dixon, Paul; Moulton, J. David; Hubbard, Susan S.; Faybishenko, Boris; Steefel, Carl I.; Finsterle, Stefan; Marble, Justin

2013-01-01

The U.S. Department of Energy (US DOE) Office of Environmental Management (EM), Office of Soil and Groundwater, is supporting development of the Advanced Simulation Capability for Environmental Management (ASCEM). ASCEM is a state-of-the-art scientific tool and approach for understanding and predicting contaminant fate and transport in natural and engineered systems. The modular and open source high-performance computing tool facilitates integrated approaches to modeling and site characterization that enable robust and standardized assessments of performance and risk for EM cleanup and closure activities. The ASCEM project continues to make significant progress in development of computer software capabilities with an emphasis on integration of capabilities in FY12. Capability development is occurring for both the Platform and Integrated Tool-sets and High-Performance Computing (HPC) Multi-process Simulator. The Platform capabilities provide the user interface and tools for end-to-end model development, starting with definition of the conceptual model, management of data for model input, model calibration and uncertainty analysis, and processing of model output, including visualization. The HPC capabilities target increased functionality of process model representations, tool-sets for interaction with Platform, and verification and model confidence testing. The Platform and HPC capabilities are being tested and evaluated for EM applications in a set of demonstrations as part of Site Applications Thrust Area activities. The Phase I demonstration focusing on individual capabilities of the initial tool-sets was completed in 2010. The Phase II demonstration completed in 2012 focused on showcasing integrated ASCEM capabilities. For Phase II, the Hanford Site deep vadose zone (BC Cribs) served as an application site for an end-to-end demonstration of capabilities, with emphasis on integration and linkages between the Platform and HPC components. Other demonstrations

A Comparison of Hazard Prediction and Assessment Capability (HPAC) Software Dose-Rate Contour Plots to a Sample of Local Fallout Data From Test Detonations in the Continental United States, 1945 - 1962

National Research Council Canada - National Science Library

Chancellor, Richard W

2005-01-01

A comparison of Hazard Prediction and Assessment Capability (HPAC) software dose-rate contour plots to a sample of local nuclear fallout data from test detonations in the continental United States, 1945 - 1962, is performed...

Brandishing Cyberattack Capabilities

Science.gov (United States)

2013-01-01

Advertising cyberwar capabilities may be helpful. It may back up a deterrence strategy. It might dissuade other states from conventional mischief or...to enable the attack.5 Many of the instruments of the attack remain with the target system, nestled in its log files, or even in the malware itself...debat- able. Even if demonstrated, what worked yesterday may not work today. But difficult does not mean impossible. Advertising cyberwar capabilities

Study of magnetorheology and sensing capabilities of MR elastomers

International Nuclear Information System (INIS)

Tian, T F; Li, W H; Alici, G

2013-01-01

This study focuses on the magnetorheology and sensing capability of graphite based Magnetorheological Elastomers (Gr MREs). By introducing graphite (Gr) to conventional MREs, the Gr MREs are derived. The anisotropic sample with 20% graphite weight fraction was selected to be compared with anisotropic conventional MREs. The microstructures of anisotropic Gr MREs and conventional MREs were observed. Both steady state tests and dynamic tests were conducted to study rheological properties of the samples. For dynamic tests, the effects of strain amplitude, and frequency on both storage modulus and loss modulus were measured. For sensing capability, the resistance of selected Gr MREs under different magnetic fields and external loadings is measured with a multi-meter. Either higher magnetic field or more external loading results in the resistance increment. Based on an ideal assumption of perfect chain structure, a mathematical model was proposed to investigate the relationship between the MRE resistance with the external loadings. In this model, the current flowing through the chain structure consists of both tunnel current and conductivity current, both of which depends on external loadings. The modelling parameters were identified and reconstructed from comparison with experimental results. The comparison indicates that both experimental results and modelling prediction agree favourably well.

Experiments on pollutant transport from soil into residential basements by pressure-driven airflow

International Nuclear Information System (INIS)

Nazaroff, W.W.; Lewis, S.R.; Doyle, S.M.; Moed, B.A.; Nero, A.V.

1987-01-01

At two residences in Portland, OR, they have investigated (1) the coupling between residential basements and the air in nearby soil and (2) the influence of basement depressurization on the migration of air in soil. With the basements depressurized 25-50 Pa relative to outdoor air, underpressures as great as 20-40% of those in the basement were observed at sampling points in the soil. Sulfur hexafluoride was injected into the soil near the houses and its concentration monitored in soil air and in the house over time, both with and without basement depressurization. Depressurization was seen to have a substantial effect on the migration of the tracer within the soil. For basement depressurizations of 25-50 Pa, effective transport velocities through the soil and into the houses were observed to exceed 1 m h -1 . Airborne 222 Rn concentration was monitored in the basement of one house during the 6-day investigation and was seen to increase substantially on each of the seven occasions that the house was depressurized. The techniques employed are applicable to the study of problems of excessive radon entry into buildings and the migration of toxic vapors from waste dumps and landfills

Capability of DFIG WTS to ride through recurring asymmetrical grid faults

DEFF Research Database (Denmark)

Chen, Wenjie; Blaabjerg, Frede; Chen, Min

2014-01-01

The Wind Turbine Systems (WTS) are required to ride through recurring grid faults in some countries. In this paper, the capability of Doubly Fed Induction Generator (DFIG) WTS to ride through recurring asymmetrical grid faults is evaluated and compared with the ride through capability under single...... asymmetrical grid fault. A mathematical model of the DFIG under recurring asymmetrical grid faults is represented. The analysis are verified by simulations on a 1.5MW DFIG model and by experiments on a reduced-scale DFIG test system....

Underground mining of the lower 163 zone through groundwater drainage at the Eagle Point Mine

International Nuclear Information System (INIS)

Robson, D.M.; Bashir, R.; Thomson, J.; Klemmer, S.; Rigden, A.

2010-01-01

The Eagle Point Mine is part of the Cameco Rabbit Lake Operation. The mine produces uranium ore using the long-hole, vertical and horizontal retreat mining method. The majority of the mine workings are under Wollaston Lake and cementitious grouting is used as one of the water control measures. Historical groundwater table in the mining area was close to ground surface. The Lower 163 Zone encompasses an estimated 4.2 million pounds U_3O_8 geological resource that was not considered feasible to mine due to the expected groundwater flows in the area. Cross-hole testing was conducted to better understand the groundwater flow through various geologic units. A local depressurization test was conducted to assess the potential for lowering the water table. Following testing an active depressurization was conducted to lower the groundwater table below the planned mining areas. This resulted in safe and drier mining conditions and allowed for the successful extraction of the ore body. (author)

Organizational Economics of Capability and Heterogeneity

DEFF Research Database (Denmark)

Argyres, Nicholas S.; Felin, Teppo; Foss, Nicolai Juul

2012-01-01

For decades, the literatures on firm capabilities and organizational economics have been at odds with each other, specifically relative to explaining organizational boundaries and heterogeneity. We briefly trace the history of the relationship between the capabilities literature and organizational...... economics, and we point to the dominance of a “capabilities first” logic in this relationship. We argue that capabilities considerations are inherently intertwined with questions about organizational boundaries and internal organization, and we use this point to respond to the prevalent capabilities first...... logic. We offer an integrative research agenda that focuses first on the governance of capabilities and then on the capability of governance....

Competing on capabilities: the new rules of corporate strategy.

Science.gov (United States)

Stalk, G; Evans, P; Shulman, L E

1992-01-01

In the 1980s, companies discovered time as a new source of competitive advantage. In the 1990s, they will discover that time is only one piece of a more far-reaching transformation in the logic of competition. Using examples from Wal-Mart and other highly successful companies, Stalk, Evans, and Shulman of the Boston Consulting Group provide managers with a guide to the new world of "capabilities-based competition." In today's dynamic business environment, strategy too must become dynamic. Competition is a "war of movement" in which success depends on anticipation of market trends and quick response to changing customer needs. In such an environment, the essence of strategy is not the structure of a company's products and markets but the dynamics of its behavior. To succeed, a company must weave its key business processes into hard-to-imitate strategic capabilities that distinguish it from its competitors in the eyes of customers. A capability is a set of business processes strategically understood--for example, Wal-Mart's expertise in inventory replenishment, Honda's skill at dealer management, or Banc One's ability to "out-local the national banks and out-national the local banks." Such capabilities are collective and cross-functional--a small part of many people's jobs, not a large part of a few. Finally, competing on capabilities requires strategic investments in support systems that span traditional SBUs and functions and go far beyond what traditional cost-benefit metrics can justify. A CEO's success in building and managing a company's capabilities will be the chief test of management skill in the 1990s. The prize: companies that combine scale and flexibility to outperform the competition.

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.

Benchmarking LWR codes capability to model radionuclide deposition within SFR containments: An analysis of the Na ABCOVE tests

International Nuclear Information System (INIS)

Herranz, Luis E.; Garcia, Monica; Morandi, Sonia

2013-01-01

Highlights: • Assessment of LWR codes capability to model aerosol deposition within SFR containments. • Original hypotheses proposed to partially accommodate drawbacks from Na oxidation reactions. • A defined methodology to derive a more accurate characterization of Na-based particles. • Key missing models in LWR codes for SFR applications are identified. - Abstract: Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide transport, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide deposition, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. The present paper is aimed at assessing the current capability of LWR codes to model aerosol deposition within a SFR containment under BDBA conditions. Through a systematic application of the ASTEC, ECART and MELCOR codes to relevant ABCOVE tests, insights have been gained into drawbacks and capabilities of these computation tools. Hypotheses and approximations have been adopted so that

Metrology Measurement Capabilities

Energy Technology Data Exchange (ETDEWEB)

Dr. Glen E. Gronniger

2007-10-02

This document contains descriptions of Federal Manufacturing & Technologies (FM&T) Metrology capabilities, traceability flow charts, and the measurement uncertainty of each measurement capability. Metrology provides NIST traceable precision measurements or equipment calibration for a wide variety of parameters, ranges, and state-of-the-art uncertainties. Metrology laboratories conform to the requirements of the Department of Energy Development and Production Manual Chapter 13.2, ANSI/ISO/IEC ANSI/ISO/IEC 17025:2005, and ANSI/NCSL Z540-1. FM&T Metrology laboratories are accredited by NVLAP for the parameters, ranges, and uncertainties listed in the specific scope of accreditation under NVLAP Lab code 200108-0. See the Internet at http://ts.nist.gov/Standards/scopes/2001080.pdf. These parameters are summarized. The Honeywell Federal Manufacturing & Technologies (FM&T) Metrology Department has developed measurement technology and calibration capability in four major fields of measurement: (1) Mechanical; (2) Environmental, Gas, Liquid; (3) Electrical (DC, AC, RF/Microwave); and (4) Optical and Radiation. Metrology Engineering provides the expertise to develop measurement capabilities for virtually any type of measurement in the fields listed above. A strong audit function has been developed to provide a means to evaluate the calibration programs of our suppliers and internal calibration organizations. Evaluation includes measurement audits and technical surveys.

Capable design or designing capabilities? An exploration of service design as an emerging organizational capability in Telenor – Martinkenaite

Directory of Open Access Journals (Sweden)

Ieva Martinkenaite

2017-01-01

Full Text Available This empirical paper examines a process, starting with the managerial decision to make service design an organizational capability, and follows it as it unfolds over time within one organization. Service design has become an established business practice of how firms create new products and services to promote differentiation in an increasingly uncertain business landscape. Implicit in the literature on service design are assumptions about strategic implications of adopting the prescribed innovation methods and tools. However, little is known about how service design evolves into an organizational capability enabling firms to transform their existing businesses and sustain competitiveness. Through a longitudinal, exploratory case study of service design practices in one of the world’s largest telecommunications companies, we explicate mechanisms through which service design evolves into an organizational capability by exploring the research question: what are the mechanisms through which service design develops into an organizational capability? Our study reveals the effect of an initial introduction of service design tools, identification of boundaryspanning actors and co-alignment of dedicated resources between internal functions, as well as through co-creation with customers. Over time, these activities lead to the adoption of service design practices, and subsequently these practices spark incremental learning throughout the organization, alter managerial decisions and influence multiple paths for the development of new capabilities. Reporting on this process, we are able to describe how service design practices were disseminated and institutionalized within the organization we observed. This study thus contributes by informing how service design can evolve into an organizational capability, as well as by bridging the emerging literature on service design and design thinking with established strategy theory. Further research will have to

Space Logistics: Launch Capabilities

Science.gov (United States)

Furnas, Randall B.

1989-01-01

The current maximum launch capability for the United States are shown. The predicted Earth-to-orbit requirements for the United States are presented. Contrasting the two indicates the strong National need for a major increase in Earth-to-orbit lift capability. Approximate weights for planned payloads are shown. NASA is studying the following options to meet the need for a new heavy-lift capability by mid to late 1990's: (1) Shuttle-C for near term (include growth versions); and (2) the Advanced Lauching System (ALS) for the long term. The current baseline two-engine Shuttle-C has a 15 x 82 ft payload bay and an expected lift capability of 82,000 lb to Low Earth Orbit. Several options are being considered which have expanded diameter payload bays. A three-engine Shuttle-C with an expected lift of 145,000 lb to LEO is being evaluated as well. The Advanced Launch System (ALS) is a potential joint development between the Air Force and NASA. This program is focused toward long-term launch requirements, specifically beyond the year 2000. The basic approach is to develop a family of vehicles with the same high reliability as the Shuttle system, yet offering a much greater lift capability at a greatly reduced cost (per pound of payload). The ALS unmanned family of vehicles will provide a low end lift capability equivalent to Titan IV, and a high end lift capability greater than the Soviet Energia if requirements for such a high-end vehicle are defined.In conclusion, the planning of the next generation space telescope should not be constrained to the current launch vehicles. New vehicle designs will be driven by the needs of anticipated heavy users.

The Capability to Hold Property

NARCIS (Netherlands)

Claassen, Rutger

2015-01-01

This paper discusses the question of whether a capability theory of justice (such as that of Martha Nussbaum) should accept a basic “capability to hold property.” Answering this question is vital for bridging the gap between abstract capability theories of justice and their institutional

KüFA safety testing of HTR fuel pebbles irradiated in the High Flux Reactor in Petten

Energy Technology Data Exchange (ETDEWEB)

Seeger, O., E-mail: oliver.seeger@rwth-aachen.de [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Laurie, M., E-mail: mathias.laurie@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Abjani, A. El; Ejton, J.; Boudaud, D.; Freis, D.; Carbol, P.; Rondinella, V.V. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Fütterer, M. [European Commission, Joint Research Centre (JRC), Institute for Energy and Transport (IET), Nuclear Reactor Integrity Assessment and Knowledge Management Unit, PO Box 2, 1755 ZG Petten (Netherlands); Allelein, H.-J. [Lehrstuhl für Reaktorsicherheit und -technik an der RWTH Aachen, Kackertstraße 9, 52072 Aachen (Germany)

2016-09-15

The Cold Finger Apparatus (KühlFinger-Apparatur—KüFA) in operation at JRC-ITU is designed to experimentally scrutinize the effects of Depressurization LOss of Forced Circulation (D-LOFC) accident scenarios on irradiated High Temperature Reactor (HTR) fuel pebbles. Up to 1600 °C, the reference maximum temperature for these accidents, high-quality German HTR fuel pebbles have already demonstrated a small fission product release. This paper discusses and compares the releases obtained from KüFA-testing the pebbles HFR-K5/3 and HFR-EU1/3, which were both irradiated in the High Flux Reactor (HFR) in Petten. We present the time-dependent fractional release of the volatile fission product {sup 137}Cs as well as the fission gas {sup 85}Kr for both pebbles. For HFR-EU1/3 the isotopes {sup 134}Cs and {sup 154}Eu as well as the shorter-lived {sup 110m}Ag have also been measured. A detailed description of the experimental setup and its accuracy is given. The data for the recently tested pebbles is discussed in the context of previous results.

Conceptualizing innovation capabilities: A contingency perspective

Directory of Open Access Journals (Sweden)

Tor Helge Aas

2017-01-01

Full Text Available Empirical research has confirmed that a positive relationship exists between the implementation of innovation activities and the future performance of organizations. Firms utilize resources and capabilities to develop innovations in the form of new products, services or processes. Some firms prove to be better at reproducing innovation success than others, and the capacity to do so is referred to as innovation capability. However, the term innovation capability is ambiguously treated in extant literature. There are several different definitions of the concept and the distinction between innovation capabilities and other types of capabilities, such as dynamic capabilities, is neither explicitly stated, nor is the relationship between the concept and other resource- and capability-based concepts within strategy theory established. Although innovation is increasingly identified as crucial for a firm’s sustainable competitiveness in contemporary volatile and complex markets, the strategy-innovation link is underdeveloped in extant research. To overcome this challenge this paper raises the following research question: What type of innovation capabilities are required to innovate successfully? Due to the status of the extant research, we chose a conceptual research design to answer our research question and the paper contributes with a conceptual framework to discuss what innovation capabilities firms need to reproduce innovation success. Based on careful examination of current literature on innovation capability specifically, and the strategy-innovation link in general, we suggest that innovation capability must be viewed along two dimensions – innovation novelty and market characteristics. This framework enables the identification of four different contexts for innovation capabilities in a two-bytwo matrix. We discuss the types of innovation capabilities necessary within the four different contexts. This novel framework contributes to the

Simulating the gas hydrate production test at Mallik using the pilot scale pressure reservoir LARS

Science.gov (United States)

Heeschen, Katja; Spangenberg, Erik; Schicks, Judith M.; Priegnitz, Mike; Giese, Ronny; Luzi-Helbing, Manja

2014-05-01

LARS, the LArge Reservoir Simulator, allows for one of the few pilot scale simulations of gas hydrate formation and dissociation under controlled conditions with a high resolution sensor network to enable the detection of spatial variations. It was designed and built within the German project SUGAR (submarine gas hydrate reservoirs) for sediment samples with a diameter of 0.45 m and a length of 1.3 m. During the project, LARS already served for a number of experiments simulating the production of gas from hydrate-bearing sediments using thermal stimulation and/or depressurization. The latest test simulated the methane production test from gas hydrate-bearing sediments at the Mallik test site, Canada, in 2008 (Uddin et al., 2011). Thus, the starting conditions of 11.5 MPa and 11°C and environmental parameters were set to fit the Mallik test site. The experimental gas hydrate saturation of 90% of the total pore volume (70 l) was slightly higher than volumes found in gas hydrate-bearing formations in the field (70 - 80%). However, the resulting permeability of a few millidarcy was comparable. The depressurization driven gas production at Mallik was conducted in three steps at 7.0 MPa - 5.0 MPa - 4.2 MPa all of which were used in the laboratory experiments. In the lab the pressure was controlled using a back pressure regulator while the confining pressure was stable. All but one of the 12 temperature sensors showed a rapid decrease in temperature throughout the sediment sample, which accompanied the pressure changes as a result of gas hydrate dissociation. During step 1 and 2 they continued up to the point where gas hydrate stability was regained. The pressure decreases and gas hydrate dissociation led to highly variable two phase fluid flow throughout the duration of the simulated production test. The flow rates were measured continuously (gas) and discontinuously (liquid), respectively. Next to being discussed here, both rates were used to verify a model of gas

Developing Collaborative Product Development Capabilities

DEFF Research Database (Denmark)

Mahnke, Volker; Tran, Yen

2012-01-01

innovation strategies’. Our analyses suggest that developing such collaboration capabilities benefits from the search for complementary practices, the combination of learning styles, and the development of weak and strong ties. Results also underscore the crucial importance of co-evolution of multi......Collaborative product development capabilities support a company’s product innovation activities. In the context of the fast fashion sector, this paper examines the development of the product development capabilities (PDC) that align product development capabilities in a dual innovation context......, one, slow paced, where the firm is well established and the other, fast paced, which represents a new competitive arena in which the company competes. To understand the process associated with collaborative capability development, we studied three Scandinavian fashion companies pursuing ‘dual...

Experimental studies on the thermal stratification and its influence on BLEVEs

Energy Technology Data Exchange (ETDEWEB)

Lin, Wensheng; Gong, Yanwu; Gao, Ting; Gu, Anzhong; Lu, Xuesheng [Institute of Refrigeration and Cryogenics, Shanghai Jiao Tong University, Shanghai 200240 (China)

2010-10-15

The thermal stratification of Liquefied Petroleum Gas (LPG) and its effect on the occurrence of the boiling liquid expanding vapor explosion (BLEVE) have been investigated experimentally. Stratifications in liquid and vapor occur when the LPG tank is heated. The degree of the liquid stratification {beta} increases with an increasing heat flux and decreasing filling ratio. The effect of stratification on the BLEVE has been examined with depressurization tests of LPG. The results show that the pressure recovery for the stratified LPG ({beta} = 1.4) upon sudden depressurization is much lower than that for the isothermal LPG ({beta} = 1). It can be concluded that the liquid stratification decreases the liquid energy and the occurrence of the BLEVE. (author)

Update: Partnership for the Revitalization of National Wind Tunnel Force Measurement Technology Capability

Science.gov (United States)

Rhew, Ray D.

2010-01-01

NASA's Aeronautics Test Program (ATP) chartered a team to examine the issues and risks associated with the lack of funding and focus on force measurement over the past several years, focusing specifically on strain-gage balances. NASA partnered with the U.S. Air Force's Arnold Engineering Development Center (AEDC) to exploit their combined capabilities and take a national level government view of the problem and established the National Force Measurement Technology Capability (NFMTC) project. This paper provides an update on the team's status for revitalizing the government's balance capability with respect to designing, fabricating, calibrating, and using the these critical measurement devices.

Dynamic Capabilities and Performance

DEFF Research Database (Denmark)

Wilden, Ralf; Gudergan, Siegfried P.; Nielsen, Bo Bernhard

2013-01-01

are contingent on the competitive intensity faced by firms. Our findings demonstrate the performance effects of internal alignment between organizational structure and dynamic capabilities, as well as the external fit of dynamic capabilities with competitive intensity. We outline the advantages of PLS...

Synthetic aperture radar capabilities in development

Energy Technology Data Exchange (ETDEWEB)

Miller, M. [Lawrence Livermore National Lab., CA (United States)

1994-11-15

The Imaging and Detection Program (IDP) within the Laser Program is currently developing an X-band Synthetic Aperture Radar (SAR) to support the Joint US/UK Radar Ocean Imaging Program. The radar system will be mounted in the program`s Airborne Experimental Test-Bed (AETB), where the initial mission is to image ocean surfaces and better understand the physics of low grazing angle backscatter. The Synthetic Aperture Radar presentation will discuss its overall functionality and a brief discussion on the AETB`s capabilities. Vital subsystems including radar, computer, navigation, antenna stabilization, and SAR focusing algorithms will be examined in more detail.

Radix Achyranthis Bidentatae improves learning and memory capabilities in ovariectomized rats

Institute of Scientific and Technical Information of China (English)

Yuefen Wang; Ya Xu; Yanshu Pan; Weihong Li; Wei Zhang; Yang Liu; Jing Jia; Pengtao Li

2013-01-01

Kidney-tonifying recipe can reduce the accumulation of advanced glycation end products, prevent neuronal degeneration and improve cognitive functions in ovariectomized rats. Radix Achyranthis Bidentatae alcohol extracts may dose-dependently inhibit non-enzymatic saccharification in vitro. This study aimed to examine the effect of Radix Achyranthis Bidentatae on advanced glycation end products and on learning and memory capabilities in ovariectomized rats. Ovariectomized rats were treated with Radix Achyranthis Bidentatae alcohol extracts (containing 1.5 g/kg crude drug) or 0.1% aminoguanidine for 12 weeks and behavioral testing was performed with the Y-electrical maze. This test revealed that Radix Achyranthis Bidentatae and aminoguanidine could improve the learning and memory capabilities of ovariectomized rats. Results of competitive enzyme-linked immunosorbent assay showed that treatment with Radix Achyranthis Bidentatae or aminoguanidine reduced the accumulation of advanced glycation end products in the frontal cortex of ovariectomized rats, while increasing content in the blood and urine. Biochemical tests showed that treatment with Radix Achyranthis Bidentatae or aminoguanidine decreased superoxide dismutase activity in the serum and frontal cortex, and increased serum levels of glutathione peroxidase in ovariectomized rats. In addition, there was no apparent effect on malondialdehyde levels. These experimental findings indicate that Radix Achyranthis Bidentatae inhibits production of advanced glycation end products and its accumulation in brain tissue, and improves learning and memory capabilities in ovariectomized rats. These effects may be associated with an anti-oxidative action of the extract.

Marketing Capability in Strategy Research

DEFF Research Database (Denmark)

Ritter, Thomas; Distel, Andreas Philipp

Following the call for a demand-side perspective of strategic management (e.g., Priem et al., 2012), a firm’s marketing capability, i.e. its ability to interact with down-stream stakeholders, becomes a pivotal element in explaining a firm’s competitiveness. While marketing capability is recognized...... in the strategic management literature as an important driver of firm performance, our review of 86 articles reveals a lack of a generally accepted definition of marketing capability, a lack of a common conceptualization as well as differences in the measurement of marketing capability. In order to build a common...... ground for advancing marketing capability research and thus supporting the demand-side perspective in strategic management, we develop an integrative framework to explain the differences and propose a research agenda for developing the field....

Evaluation of the NASA Arc Jet Capabilities to Support Mission Requirements

Science.gov (United States)

Calomino, Anthony; Bruce, Walt; Gage, Peter; Horn, Dennis; Mastaler, Mike; Rigali, Don; Robey, Judee; Voss, Linda; Wahlberg, Jerry; Williams, Calvin

2010-01-01

NASA accomplishes its strategic goals through human and robotic exploration missions. Many of these missions require launching and landing or returning spacecraft with human or return samples through Earth's and other planetary atmospheres. Spacecraft entering an atmosphere are subjected to extreme aerothermal loads. Protecting against these extreme loads is a critical element of spacecraft design. The safety and success of the planned mission is a prime concern for the Agency, and risk mitigation requires the knowledgeable use of thermal protection systems to successfully withstand the high-energy states imposed on the vehicle. Arc jets provide ground-based testing for development and flight validation of re-entry vehicle thermal protection materials and are a critical capability and core competency of NASA. The Agency's primary hypersonic thermal testing capability resides at the Ames Research Center and the Johnson Space Center and was developed and built in the 1960s and 1970s. This capability was critical to the success of Apollo, Shuttle, Pioneer, Galileo, Mars Pathfinder, and Orion. But the capability and the infrastructure are beyond their design lives. The complexes urgently need strategic attention and investment to meet the future needs of the Agency. The Office of Chief Engineer (OCE) chartered the Arc Jet Evaluation Working Group (AJEWG), a team of experienced individuals from across the Nation, to capture perspectives and requirements from the arc jet user community and from the community that operates and maintains this capability and capacity. This report offers the AJEWG's findings and conclusions that are intended to inform the discussion surrounding potential strategic technical and investment strategies. The AJEWG was directed to employ a 30-year Agency-level view so that near-term issues did not cloud the findings and conclusions and did not dominate or limit any of the strategic options.

The Impact of IT Capability on Employee Capability, Customer Value, Customer Satisfaction, and Business Performance

Science.gov (United States)

Chae, Ho-Chang

2009-01-01

This study empirically examines the impact of IT capability on firms' performance and evaluates whether firms' IT capabilities play a role in improving employee capability, customer value, customer satisfaction, and ultimately business performance. The results were based on comparing the business performance of the IT leader companies with that of…

An Overview of Facilities and Capabilities to Support the Development of Nuclear Thermal Propulsion

Energy Technology Data Exchange (ETDEWEB)

James Werner; Sam Bhattacharyya; Mike Houts

2011-02-01

Abstract. The future of American space exploration depends on the ability to rapidly and economically access locations of interest throughout the solar system. There is a large body of work (both in the US and the Former Soviet Union) that show that Nuclear Thermal Propulsion (NTP) is the most technically mature, advanced propulsion system that can enable this rapid and economical access by its ability to provide a step increase above what is a feasible using a traditional chemical rocket system. For an NTP system to be deployed, the earlier measurements and recent predictions of the performance of the fuel and the reactor system need to be confirmed experimentally prior to launch. Major fuel and reactor system issues to be addressed include fuel performance at temperature, hydrogen compatibility, fission product retention, and restart capability. The prime issue to be addressed for reactor system performance testing involves finding an affordable and environmentally acceptable method to test a range of engine sizes using a combination of nuclear and non-nuclear test facilities. This paper provides an assessment of some of the capabilities and facilities that are available or will be needed to develop and test the nuclear fuel, and reactor components. It will also address briefly options to take advantage of the greatly improvement in computation/simulation and materials processing capabilities that would contribute to making the development of an NTP system more affordable. Keywords: Nuclear Thermal Propulsion (NTP), Fuel fabrication, nuclear testing, test facilities.

Structural integrity test of prestressed concrete containment vessel for Tsuruga Unit No. 2 Nuclear Power Station

International Nuclear Information System (INIS)

Tamura, S.; Nagata, K.; Takeda, T.; Yamaguchi, T.; Nakayama, T.

1987-01-01

In introducing the PCCV to Japan, various verification tests were carried out to understand the structural performance of the PCCV and confirm the reliability of its design. In addition to those tests, a Structural Integrity Test (SIT) was conducted in Feb. 1986 as a final acceptance test. This report discusses the results of the SIT on the PCCV. The test was carried out simultaneously with an Integrated Leak Rate Test (ILRT) under the same pressure sequence. 1) Pressure-displacement relationships and pressure-strain relationships were more or less linear. 2) The measured displacement values at the maximum pressure (4.5 kgf/cm 2 G) corresponded well with calculated values. Correspondence with converted displacement obtained from strain and measured displacement was also good. 3) The residual displacement when 24 hours had elapsed after completion of depressurization was not more than 10% of the displacement at the maximum pressure. 4) The variation in tendon force at the maximum pressure is smaller than the calculated value in proportion to the elongation of the PCCV. 5) Although fine surface cracks due to shrinkage of concrete were seen, new structural cracks due to pressure were not observed. The leakage rate was evaluated at 0.016% of volume per day. It is much smaller than the design value of 0.1% of volume per day. (orig./HP)

Amartya Sen's Capability Approach and Education

Science.gov (United States)

Walker, Melanie

2005-01-01

The human capabilities approach developed by the economist Amartya Sen links development, quality of life and freedom. This article explores the key ideas in the capability approach of: capability, functioning, agency, human diversity and public participation in generating valued capabilities. It then considers how these ideas relate specifically…

Capabilities for innovation

DEFF Research Database (Denmark)

Nielsen, Peter; Nielsen, Rene Nesgaard; Bamberger, Simon Grandjean

2012-01-01

is a survey that collected information from 601 firms belonging to the private urban sector in Denmark. The survey was carried out in late 2010. Keywords: dynamic capabilities/innovation/globalization/employee/employer cooperation/Nordic model Acknowledgment: The GOPA study was financed by grant 20080053113......Technological developments combined with increasing levels of competition related to the ongoing globalization imply that firms find themselves in dynamic, changing environments that call for dynamic capabilities. This challenges the internal human and organizational resources of firms in general...

Non-Destructive Investigation on Short Circuit Capability of Wind-Turbine-Scale IGBT Power Modules

DEFF Research Database (Denmark)

Wu, Rui; Iannuzzo, Francesco; Wang, Huai

2014-01-01

This paper presents a comprehensive investigation on the short circuit capability of wind-turbine-scale IGBT power modules by means of a 6 kA/1.1 kV non-destructive testing system. A Field Programmable Gate Array (FPGA) supervising unit is adpoted to achieve an accurate time control for short...... circuit test, which enables to define the driving signals with an accuracy of 10 ns. Thanks to the capability and the effectiveness of the constructed setup, oscillations appearing during short circuits of the new-generation 1.7 kV/1 kA IGBT power modules have been evidenced and characterized under...

The Greater Sekhukhune-CAPABILITY outreach project.

Science.gov (United States)

Gregersen, Nerine; Lampret, Julie; Lane, Tony; Christianson, Arnold

2013-07-01

The Greater Sekhukhune-CAPABILITY Outreach Project was undertaken in a rural district in Limpopo, South Africa, as part of the European Union-funded CAPABILITY programme to investigate approaches for capacity building for the translation of genetic knowledge into care and prevention of congenital disorders. Based on previous experience of a clinical genetic outreach programme in Limpopo, it aimed to initiate a district clinical genetic service in Greater Sekhukhune to gain knowledge and experience to assist in the implementation and development of medical genetic services in South Africa. Implementing the service in Greater Sekhukhune was impeded by a developing staff shortage in the province and pressure on the health service from the existing HIV/AIDS and TB epidemics. This situation underscores the need for health needs assessment for developing services for the care and prevention of congenital disorders in middle- and low-income countries. However, these impediments stimulated the pioneering of innovate ways to offer medical genetic services in these circumstances, including tele-teaching of nurses and doctors, using cellular phones to enhance clinical care and adapting and assessing the clinical utility of a laboratory test, QF-PCR, for use in the local circumstances.

Development and industrial tests of the first LNG hydraulic turbine system in China

OpenAIRE

Jie Chen; Yihuai Hua; Qingbo Su; Xueli Wan; Zhenlin Li

2016-01-01

The cryogenic hydraulic turbine can be used to replace the conventional J–T valve for LNG or mixed refrigerant throttling and depressurization in a natural gas liquefaction plant. This advanced technology is not only to enhance the efficiency of the liquefaction plant, but to usher a new trend in the development of global liquefaction technologies. China has over 136 liquefaction plants, but the cryogenic hydraulic turbines have not been deployed in industrial utilization. In addition, these ...

Telematics Options and Capabilities

Energy Technology Data Exchange (ETDEWEB)

Hodge, Cabell [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

2017-09-05

This presentation describes the data tracking and analytical capabilities of telematics devices. Federal fleet managers can use the systems to keep their drivers safe, maintain a fuel efficient fleet, ease their reporting burden, and save money. The presentation includes an example of how much these capabilities can save fleets.

Capability Handbook- offline metrology

DEFF Research Database (Denmark)

Islam, Aminul; Marhöfer, David Maximilian; Tosello, Guido

This offline metrological capability handbook has been made in relation to HiMicro Task 3.3. The purpose of this document is to assess the metrological capability of the HiMicro partners and to gather the information of all available metrological instruments in the one single document. It provides...

The Capability Approach

OpenAIRE

Robeyns, Ingrid

2011-01-01

textabstract In its most general description, the capability approach is a flexible and multi-purpose normative framework, rather than a precise theory of well-being, freedom or justice. At its core are two normative claims: first, the claim that the freedom to achieve well-being is of primary moral importance, and second, that freedom to achieve well-being is to be understood in terms of people’s capabilities, that is, their real opportunities to do and be what they have reason to value. Thi...

Sandia QIS Capabilities.

Energy Technology Data Exchange (ETDEWEB)

Muller, Richard P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-07-01

Sandia National Laboratories has developed a broad set of capabilities in quantum information science (QIS), including elements of quantum computing, quantum communications, and quantum sensing. The Sandia QIS program is built atop unique DOE investments at the laboratories, including the MESA microelectronics fabrication facility, the Center for Integrated Nanotechnologies (CINT) facilities (joint with LANL), the Ion Beam Laboratory, and ASC High Performance Computing (HPC) facilities. Sandia has invested $75 M of LDRD funding over 12 years to develop unique, differentiating capabilities that leverage these DOE infrastructure investments.

Nuclear reactor with makeup water assist from residual heat removal system

Science.gov (United States)

Corletti, Michael M.; Schulz, Terry L.

1993-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

Nuclear reactor with makeup water assist from residual heat removal system

International Nuclear Information System (INIS)

Corletti, M.M.; Schulz, T.L.

1993-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures

Nuclear reactor with makeup water assist from residual heat removal system

Science.gov (United States)

Corletti, M.M.; Schulz, T.L.

1993-12-07

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

A Study on the Impact of Marketing Capability, Operations Capability, Environmental Capability and Diversification Strategy on the Performance of Hotel Industry in the UK

OpenAIRE

ZHANG, YUBO

2009-01-01

Hotel industry has been an increasingly significant service industry across the whole world. The performance measurement method is crucial for hotels’ operations. This study examines the impacts of marketing capability, operations capability, environmental capability and diversification strategy on the business performance of hotel industry in the UK. Based on the financial archival data in the database and information obtained in the firm websites from 2004 to 2007, the investigation is carr...

Loneliness and objectively measured physical capability in middle-aged adults

DEFF Research Database (Denmark)

Lund, Rikke; Laban, J; Petersen, GL

2018-01-01

and women compared with the ‘not lonely’ were 1.2 kg (95% CI − 0.5;2.9)/1.0 kg (−0.7;2.6). Low occupational social class was associated with poorer physical capability, and living alone was associated with poorer handgrip strength in men [−2.4 kg (95% CI − 3.2;−1.5)] and poorer chair rise test in women [−0......Background: Loneliness is associated with poor functional ability in older people. Little is known about this association in the middle-aged. The aim is to investigate if perceived loneliness is associated with lower physical capability among middle-aged men and women and if the associations...... of loneliness with physical capability interact with socioeconomic position and cohabitation status. Methods: 5224 participants from Copenhagen Aging and Midlife Biobank (CAMB) aged 49–62 years (mean age 54) were included. Handgrip strength (measured by a dynamometer) and maximal number of chair rises in 30 s...

PHEBUS/test-218, Behaviour of a Fuel Rod Bundle during a Large Break LOCA Transient with a two Peaks Temperature History

International Nuclear Information System (INIS)

1987-01-01

1 - Description of test facility: PHEBUS test facility operated at CEA Research Center Cadarache consists of a pressurized circuit involving pumps, heat exchangers and a blowdown tank - 25 nuclear fuel rod bundle, coupled to a separate driver core; - active length 0.8 m, cosine axial power profile; - pressurized and un-pressurized fuel rods; - controlled cooling conditions at the bundle inlet (blowdown, refill and reflood period); - de-pressurized test rig volume 0.22 m 3 . The following 'as measured' boundary conditions (B.C.) were offered to participants as options with decreasing challenge to their analytical approach: Boundary conditions B.C.0: - full thermal-hydraulic analysis of PHEBUS test rig (was not recommended). Boundary conditions B.C.1: - thermal power level of fuel bundle; - fluid inlet conditions to bundle section. Boundary conditions B.C.2: - local cladding temperatures of rods; - heat transfer coefficients. Boundary conditions B.C.3: - cladding temperatures of rods; - internal pressure of rods. 2 - Description of test: Post-test investigation into the response of a nuclear fuel bundle to a large break loss of coolant accident with respect to - local fuel temperatures, - cladding strain at the time of burst, - time to burst and under given thermal-hydraulic boundary conditions of PHEBUS-test 218

The capability concept – On how to define and describe capability in relation to risk, vulnerability and resilience

International Nuclear Information System (INIS)

Lindbom, Hanna; Tehler, Henrik; Eriksson, Kerstin; Aven, Terje

2015-01-01

Capabilities-based planning and capability assessment are high on the agendas of several countries and organisations as part of their risk management and emergency preparedness. Despite this, few definitions of capability exist, and they are not easily related to concepts such as risk, vulnerability and resilience. The aim of the present study was thus to broaden the scientific basis of the risk field to also include the concept of capability. The proposed definition is based on a recently developed risk framework, and we define capability as the uncertainty about and the severity of the consequences of an activity given the occurrence of the initiating event and the performed task. We provide examples of how the response capability for a fictive scenario can be described using this definition, and illustrate how our definition can be used to analyse capability assessments prepared according to the Swedish crisis management system. We have analysed the content of 25 capability assessments produced in 2011 by stakeholders on local, regional and national level. It was concluded that none addressed uncertainty to any appreciable extent, and only a third described capability in terms of consequences and task, making it difficult to relate these capability assessments to risk assessments. - Highlights: • Few definitions of capability relate to definitions of risk, vulnerability and resilience. • We relate capability to risk, vulnerability and resilience. • We define capability using the components uncertainty, consequences, event and task

MTR fuel plate qualification capabilities at SCK-CEN

International Nuclear Information System (INIS)

Koonen, E.; Jacquet, P.

2002-01-01

In order to enhance the capabilities of BR2 in the field of MTR fuel plate testing, a dedicated irradiation device has been designed. In its basic version this device allows the irradiation of 3 fuel plates. The central fuel plate may be replaced by a dummy plate or a plate carrying dosimeters. A first FUTURE device has been built. A benchmark irradiation has been executed with standard BR2 fuel plates in order to qualify this device. Detailed neutronic calculations were performed and the results compared to the results of the post-irradiation examinations of the plates. These comparisons demonstrate the capability to conduct a fuel plate irradiation program under requested and well-known irradiation conditions. Further improvements are presently being designed in order to extend the ranges of heat flux and surface temperature of the fuel plates that can be handled with the FUTURE device. (author)

Human push capability.

Science.gov (United States)

Barnett, Ralph L; Liber, Theodore

2006-02-22

Use of unassisted human push capability arises from time to time in the areas of crowd and animal control, the security of locked doors, the integrity of railings, the removal of tree stumps and entrenched vehicles, the manoeuvering of furniture, and athletic pursuits such as US football or wrestling. Depending on the scenario, human push capability involves strength, weight, weight distribution, push angle, footwear/floor friction, and the friction between the upper body and the pushed object. Simple models are used to establish the relationships among these factors.

Benchmarking LWR codes capability to model radionuclide deposition within SFR containments: An analysis of the Na ABCOVE tests

Energy Technology Data Exchange (ETDEWEB)

Herranz, Luis E., E-mail: luisen.herranz@ciemat.es [CIEMAT, Unit of Nuclear Safety Research, Av. Complutense, 40, 28040 Madrid (Spain); Garcia, Monica, E-mail: monica.gmartin@ciemat.es [CIEMAT, Unit of Nuclear Safety Research, Av. Complutense, 40, 28040 Madrid (Spain); Morandi, Sonia, E-mail: sonia.morandi@rse-web.it [Nuclear and Industrial Plant Safety Team, Power Generation System Department, RSE, via Rubattino 54, 20134 Milano (Italy)

2013-12-15

Highlights: • Assessment of LWR codes capability to model aerosol deposition within SFR containments. • Original hypotheses proposed to partially accommodate drawbacks from Na oxidation reactions. • A defined methodology to derive a more accurate characterization of Na-based particles. • Key missing models in LWR codes for SFR applications are identified. - Abstract: Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide transport, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide deposition, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. The present paper is aimed at assessing the current capability of LWR codes to model aerosol deposition within a SFR containment under BDBA conditions. Through a systematic application of the ASTEC, ECART and MELCOR codes to relevant ABCOVE tests, insights have been gained into drawbacks and capabilities of these computation tools. Hypotheses and approximations have

White Sands Missile Range Overview & Introduction: Test Capabilities Briefing

Science.gov (United States)

2011-11-07

MOA BRONCO MOAS TALON MOAS UAV COA Regional Air Space Joint Military Operating Area RESERVE MOA MORENCI MOA TOMBSTONE MOA El Paso Alamogordo Clovis...MANPADS target at Aerial Cable Range QF-4 Full-scale drone Sub-scale drone launch Army Proven Battle Ready Ft. Wingate - 250 miles El Paso...Major Nuclear Effects Characterization Test Facilities and Army Proven Battle Ready  Gamma Radiation: El Dorado Gamma Facility • EGF is an

Advanced Simulation Capability for Turbopump Cavitation Dynamics Guided by Experimental Validation, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — Numerical cavitation modeling capability is critical in the design of liquid rocket engine turbopumps, feed lines, injector manifolds and engine test facilities....

A variable capacitance based modeling and power capability predicting method for ultracapacitor

Science.gov (United States)

Liu, Chang; Wang, Yujie; Chen, Zonghai; Ling, Qiang

2018-01-01

Methods of accurate modeling and power capability predicting for ultracapacitors are of great significance in management and application of lithium-ion battery/ultracapacitor hybrid energy storage system. To overcome the simulation error coming from constant capacitance model, an improved ultracapacitor model based on variable capacitance is proposed, where the main capacitance varies with voltage according to a piecewise linear function. A novel state-of-charge calculation approach is developed accordingly. After that, a multi-constraint power capability prediction is developed for ultracapacitor, in which a Kalman-filter-based state observer is designed for tracking ultracapacitor's real-time behavior. Finally, experimental results verify the proposed methods. The accuracy of the proposed model is verified by terminal voltage simulating results under different temperatures, and the effectiveness of the designed observer is proved by various test conditions. Additionally, the power capability prediction results of different time scales and temperatures are compared, to study their effects on ultracapacitor's power capability.

People Capability Maturity Model. SM.

Science.gov (United States)

1995-09-01

tailored so it consumes less time and resources than a traditional software process assessment or CMU/SEI-95-MM-02 People Capability Maturity Model...improved reputation or customer loyalty. CMU/SEI-95-MM-02 People Capability Maturity Model ■ L5-17 Coaching Level 5: Optimizing Activity 1...Maturity Model CMU/SEI-95-MM-62 Carnegie-Mellon University Software Engineering Institute DTIC ELECTE OCT 2 7 1995 People Capability Maturity

Depressurization of a spread of Brazil-Bolivia gas pipeline and the emergency repair of a weld crack in a instrument derivation at Campo Grande compression station; Despressurizacao de trecho do gasoduto Bolivia-Brazil para reparo emergencial de trinca em uma derivacao de instrumento de temperatura na Estacao de Compressao de Campo Grande - MS

Energy Technology Data Exchange (ETDEWEB)

Dietrich, Carlos Ribeiro; Leite Junior, Ismael Casano [TBG - Transportadora Brasileira Gasoduto Bolivia Brasil S.A., Rio de Janeiro, RJ (Brazil)

2005-07-01

The purpose of this paper is to report the actions taken to repair a gas leak, at an original pressure of 100 kgf/cm{sup 2}, occurred due to a 1 1/2'' branch pipe weld crack, located on the 24'' Campo Grande - Mato Grosso do Sul Compression Station discharge pipe. This branch pipe was used to a thermo well installation and was submitted to an additional strength caused by thermo well vibration. The weld repair actions required an urgent depressurization of a 33 km spread of Bolivia-Brazil Pipeline in a timely manner, to avoid any negative impact in the operational schedule. (author)