ADDING ECOLOGICAL REALISM TO PLANT TESTING

Science.gov (United States)

Current test protocols for the protection of nontarget plants used when registering pesticides in the United States and many other countries depend on two tests using greenhouse grown, agricultural seedling plants. The seedling emergence and vegetative vigor tests are used to as...

The AP600 advanced simplified nuclear power plant. Results of the test program and progress made toward final design approval

International Nuclear Information System (INIS)

Bruschi, H.J.

1996-01-01

At the 1994 Pacific Basin Conference, Mr. Bruschi presented a paper describing the AP600, Westinghouse's advanced light water reactor design with passive safety features. Since then, a rigorous test program was completed and AP600 became the most thoroughly tested advanced reactor system design in history. Westinghouse is now well on its way toward receiving Final Design Approval from the U.S. Nuclear Regulatory Commission for AP600. In this paper, the results of the test program will be discussed and an update on prospects for building the plant will be covered. (author)

The AP600 advanced simplified nuclear power plant. Results of the test program and progress made toward final design approval

Energy Technology Data Exchange (ETDEWEB)

Bruschi, H.J. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1996-10-01

At the 1994 Pacific Basin Conference, Mr. Bruschi presented a paper describing the AP600, Westinghouse`s advanced light water reactor design with passive safety features. Since then, a rigorous test program was completed and AP600 became the most thoroughly tested advanced reactor system design in history. Westinghouse is now well on its way toward receiving Final Design Approval from the U.S. Nuclear Regulatory Commission for AP600. In this paper, the results of the test program will be discussed and an update on prospects for building the plant will be covered. (author)

Novel weapons testing: are invasive plants more chemically defended than native plants?

Directory of Open Access Journals (Sweden)

Eric M Lind

2010-05-01

Full Text Available Exotic species have been hypothesized to successfully invade new habitats by virtue of possessing novel biochemistry that repels native enemies. Despite the pivotal long-term consequences of invasion for native food-webs, to date there are no experimental studies examining directly whether exotic plants are any more or less biochemically deterrent than native plants to native herbivores.In a direct test of this hypothesis using herbivore feeding assays with chemical extracts from 19 invasive plants and 21 co-occurring native plants, we show that invasive plant biochemistry is no more deterrent (on average to a native generalist herbivore than extracts from native plants. There was no relationship between extract deterrence and length of time since introduction, suggesting that time has not mitigated putative biochemical novelty. Moreover, the least deterrent plant extracts were from the most abundant species in the field, a pattern that held for both native and exotic plants. Analysis of chemical deterrence in context with morphological defenses and growth-related traits showed that native and exotic plants had similar trade-offs among traits.Overall, our results suggest that particular invasive species may possess deterrent secondary chemistry, but it does not appear to be a general pattern resulting from evolutionary mismatches between exotic plants and native herbivores. Thus, fundamentally similar processes may promote the ecological success of both native and exotic species.

Operation result of 40kW class MCFC pilot plant

Energy Technology Data Exchange (ETDEWEB)

Saitoh, H.; Hatori, S.; Hosaka, M.; Uematsu, H. [Ishikawajima-Harima Heavy Industries Co., Ltd., Tokyo (Japan)

1996-12-31

Ishikawajima-Harima Heavy Industries Co., Ltd. developed unique Molten Carbonate Fuel Cell (MCFC) system based on our original concept. To demonstrate the possibility of this system, based on MCFC technology of consigned research from New Energy and Industrial Technology Development Organization (NEDO) in Japan, we designed 40kW class MCFC pilot plant which had all equipments required as a power plant and constructed in our TO-2 Technical Center. This paper presents the test results of the plant.

Test requirements for the integral effect test to simulate Korean PWR plants

Energy Technology Data Exchange (ETDEWEB)

Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K

2001-02-01

In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time.

Test requirements for the integral effect test to simulate Korean PWR plants

International Nuclear Information System (INIS)

Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K.

2001-02-01

In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time

Process Testing Results and Scaling for the Hanford Waste Treatment and Immobilization Plant (WTP) Pretreatment Engineering Platform - 10173

International Nuclear Information System (INIS)

Kurath, Dean E.; Daniel, Richard C.; Baldwin, David L.; Rapko, Brian M.; Barnes, Steven M.; Gilbert, Robert A.; Mahoney, Lenna A.; Huckaby, James L.

2010-01-01

The U.S. Department of Energy-Office of River Protections Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being designed and built to pretreat and then vitrify a large portion of the wastes in Hanfords 177 underground waste storage tanks at Richland, Washington. In support of this effort, engineering-scale tests at the Pretreatment Engineering Platform (PEP) have been completed to confirm the process design and provide improved projections of system capacity. The PEP is a 1/4.5-scale facility designed, constructed, and operated to test the integrated leaching and ultrafiltration processes being deployed at the WTP. The PEP replicates the WTP leaching processes with prototypic equipment and control strategies and non-prototypic ancillary equipment to support the core processing. The testing approach used a nonradioactive aqueous slurry simulant to demonstrate the unit operations of caustic and oxidative leaching, cross-flow ultrafiltration solids concentration, and solids washing. Parallel tests conducted at the laboratory scale with identical simulants provided results that allow scale-up factors to be developed between the laboratory and PEP performance. This paper presents the scale-up factors determined between the laboratory and engineering-scale results and presents arguments that extend these results to the full-scale process.

Dynamic testing of nuclear power plant structures: an evaluation

International Nuclear Information System (INIS)

Weaver, H.J.

1980-02-01

Lawrence Livermore Laboratory (LLL) evaluated the applications of system identification techniques to the dynamic testing of nuclear power plant structures and subsystems. These experimental techniques involve exciting a structure and measuring, digitizing, and processing the time-history motions that result. The data can be compared to parameters calculated using finite element or other models of the test systems to validate the model and to verify the seismic analysis. This report summarizes work in three main areas: (1) analytical qualification of a set of computer programs developed at LLL to extract model parameters from the time histories; (2) examination of the feasibility of safely exciting nuclear power plant structures and accurately recording the resulting time-history motions; (3) study of how the model parameters that are extracted from the data be used best to evaluate structural integrity and analyze nuclear power plants

Reports on 1979 result of Sunshine Project. Testing research for detailed design of solar thermal power generation plant (tower converging method); 1979 nendo taiyonetsu hatsuden plant (tower shuko hoshiki) no shosai sekkei no tame no shiken kenkyu seika hokokusho

Energy Technology Data Exchange (ETDEWEB)

NONE

1980-03-31

This report briefly describes first, in the summary of the results, the contents of the research results of the year with the outcome as the central point. Then, it explains in detail, in the contents of the research, the substance of each research item, results, their examination contents, and future research subjects. The objective of the research is, for the purpose of technically seeking the cost performance of a solar thermal power generation plant, (1) to develop equipment constituting the plant and (2) to develop a pilot plant having an electrical output of about 1,000kW at the peak by the tower converging method. The research results were as follows. (1) In the confirming test of a heliostat test unit, the final conclusion was obtained of the wind resistance calculated value by the wind tunnel test of the heliostat; the design materials of the assembly jigs were obtained; the data of the operation forecast was obtained in a tracking test; and a sensor was developed for the tracking instrumentation. (2) In the confirming test of the improved absorbing surface/mirror, the improvement/trial production including the manufacturing method was carried out as the absorbing surface for the actual unit. (3) In the heat collecting test, steam generation and a loop control test were performed. (4) The plant system was analyzed, with data obtained for the operating method. (NEDO)

Experience with RTD response time testing in nuclear power plants

International Nuclear Information System (INIS)

Hashemian, H.M.; Kerlin, T.W.

1985-01-01

The reactor coolant temperatures in pressurized water reactors are measured with platinum resistance temperature detectors (RTDs). The information furnished by these RTDs is used for plant protection as well as control. As a part of the plant protection system, the RTDs must respond to temperature changes in a timely fashion. The RTD response time requirements are different for the various plant types. These requirements are specified in the plant technical specifications in terms of an RTd time constant. The current time constant requirements for nuclear plant RTDs varies from 0.5 seconds to 13.0 seconds depending on the type of the plant. Therefore, different types of RTDs are used in different plants to achieve the required time constants. In addition, in-situ response time tests are periodically performed on protective system RTDs to ensure that the in-service time constants are within acceptable limits as the plant is operating. The periodic testing is important because response time degradation may occur while the RTD ages in the process. Recent response time tests in operating plants revealed unacceptable time constants for several protection system RTDs. As a result, these plants had to be shut down to resolve the problem which in one case was due to improper installation and in another case was because of degradation of a thermal compound used in the thermowell

Guidelines for confirmatory inplant tests of safety-relief valve discharges for BWR plants

International Nuclear Information System (INIS)

Su, T.M.

1981-05-01

Inplant tests of safety/relief valve (SRV) discharges may be required to confirm generically established specifications for SRV loads and the maximum suppression pool temperature, and to evaluate possible effects of plant-unique parameters. These tests are required in those plants which have features that differ substantially from those previously tested. Guidelines for formulating appropriate test matrices, establishing test procedures, selecting necessary instrumentation, and reporting the test results are provided in this report. Guidelines to determine if inplant tests are required on the basis of the plant unique parameters are also included in the report

Results of an EMI/RFI plant survey

International Nuclear Information System (INIS)

Shankar, R.; Mollerus, F.J.

1993-01-01

This paper summarizes the results of a telephone survey to collect information concerning electromagnetic interference/radio frequency interference (EMI/RFI) problems primarily at nuclear power plants. The survey found that problem sources such as two-way radios and welding have been largely resolved by procedural control and use of sound-power phone systems. Additional investigation and testing appear appropriate for noise interference related to grounding of electrical equipment and instrumentation

Summary of results from sodium-heated steam generator test program

Energy Technology Data Exchange (ETDEWEB)

McDonald, J S

1975-07-01

A 28 MWt sodium-heated steam generator test unit developed and fabricated by Atomics International was operated in the Sodium Component Test Installation. The SCTI is located at the Liquid Metal Engineering Center which is operated for the Atomic Energy Commission by Atomics International, Reviewed in this paper are the results of the test operations and the findings of the post-test examination of the module. Testing was performed to assure the mechanical integrity of the unit over a wide range of simulated plant operating conditions and to develop a variety of performance data. Specific tests conducted included preheat, vibration, startup-shutdown, pressurization, steady state and parametric performance mapping, endurance, simulated leak injection, low- flow stability and simulated plant transients. (author)

Test and evaluation results of the 252Cf shuffler at the Savannah River Plant

International Nuclear Information System (INIS)

Crane, T.W.

1981-03-01

The 252 Cf Shuffler, a nondestructive assay instrument employing californium neutron source irradiation and delayed-neutron counting, was developed for measuring 235 U content of scrap and waste items generated at the Savannah River Plant (SRP) reactor fuel fabrication facility. The scrap and waste items include high-purity uranium-aluminum alloy ingots as well as pieces of castings, saw and lathe chips from machining operations, low-purity items such as oxides of uranium or uranium intermixed with flux materials found in recovery operations, and materials not recoverable at SRP such as floor sweepings or residues from the uranium scrap recovery operation. The uranium contains about 60% 235 U with the remaining isotopes being 236 U, 238 U, and 234 U in descending order. The test and evaluation at SRP concluded that the accuracy, safety, reliability, and ease of use made the 252 Cf Shuffler a suitable instrument for routine use in an industrial, production-oriented plant

FY 1980 Report on results of Sunshine Project. Research and development of coal liquefaction techniques (Development of direct hydrogenation type liquefaction plant and 2.4 T/D test plant); 1980 nendo sekitan ekika gijutsu no kenkyu kaihatsu, chokusetsu suiten ekika plant no kaihatsu seika hokokusho. 2.4T/nichi jikken plant no kaihatsu

Energy Technology Data Exchange (ETDEWEB)

NONE

1981-03-01

This program is aimed at construction and operation of a 2.4 T/D test plant for eventual commercialization of the direct hydrogenation type liquefaction process. The FY 1980 program includes designs of the test plant, procurement and manufacture of some equipment, and works for construction of the bases, buildings and scaffolds. The construction site for the 2.4 T/D plant was changed in July 1980 from Mitsui Engineering and Shipbuilding's Chiba works to NKK's Keihin Steelworks, which was accompanied by some changes in the basic and detailed designs. The detailed designs were reviewed for construction of the test plant, to reflect the results of the individual elementary researches. The works for the FY 1980 program also include preparations for obtaining approvals of plant construction, based on the revised designs, from the related government offices, and equipment procurement. This paper presents the major drawings for the plant construction, including those for PID designs, overall plant layouts, piping systems, buildings, scaffolds, and pipe racks. The loading data are also included. (NEDO)

Pilot plant UF6 to UF4 test operations report

International Nuclear Information System (INIS)

Bicha, W.J.; Fallings, M.; Gilbert, D.D.; Koch, G.E.; Levine, P.J.; McLaughlin, D.F.; Nuhfer, K.R.; Reese, J.C.

1987-02-01

The FMPC site includes a plant designed for the reduction of uranium hexafluoride (UF 6 ) to uranium tetrafluoride (UF 4 ). Limited operation of the upgraded reduction facility began in August 1984 and continued through January 19, 1986. A reaction vessel ruptured on that date causing the plant operation to be shut down. The DOE conducted a Class B investigation with the findings of the investigation board issued in preliminary form in May 1986 and as a final recommendation in July 1986. A two-phase restart of the plant was planned and implemented. Phase I included implementing safety system modifications, changing reaction vessel temperature control strategy, and operating the reduction plant under an 8-week controlled test. The results of the test period are the subject of this report. 41 figs., 11 tabs

On-site tests on the nuclear power plants

International Nuclear Information System (INIS)

Morilhat, P.; Favennec, J.M.; Neau, P.; Preudhomme, E.

1996-01-01

On-site tests and experiments are performed by EDF Research and Development Division on the nuclear power plants to assess the behaviour of major components submitted to thermal and vibratory solicitations. On-going studies deal with the qualification of new nuclear power plant standard and with the feedback of plants under operation. The tests, particularly the investigation tests, correspond to large investments and entail an important data volume which must ensure the continuity over a long period of the order of magnitude of the in-service plant life (around 40 years). This paper addresses the on-site experimental activities, describes the means to be used, and gives an example: the qualification of SG of new 1450 MW nuclear power plants. (author)

French nuclear plant safeguard pump qualification testing: EPEC test loop

International Nuclear Information System (INIS)

Guesnon, H.

1985-01-01

This paper reviews the specifications to which nuclear power plant safeguard pumps must be qualified, and surveys the qualification methods and program used in France to verify operability of the pump assembly and major pump components. The EPEC test loop is described along with loop capabilities and acheivements up to now. This paper shows, through an example, the Medium Pressure Safety Injection Pump designed for service in 1300 MW nuclear power plants, and the interesting possibilities offered by qualification testing

Coupling solar photo-Fenton and biotreatment at industrial scale: Main results of a demonstration plant

International Nuclear Information System (INIS)

Malato, Sixto; Blanco, Julian; Maldonado, Manuel I.; Oller, Isabel; Gernjak, Wolfgang; Perez-Estrada, Leonidas

2007-01-01

This paper reports on the combined solar photo-Fenton/biological treatment of an industrial effluent (initial total organic carbon, TOC, around 500 mg L -1 ) containing a non-biodegradable organic substance (α-methylphenylglycine at 500 mg L -1 ), focusing on pilot plant tests performed for design of an industrial plant, the design itself and the plant layout. Pilot plant tests have demonstrated that biodegradability enhancement is closely related to disappearance of the parent compound, for which a certain illumination time and hydrogen peroxide consumption are required, working at pH 2.8 and adding Fe 2+ = 20 mg L -1 . Based on pilot plant results, an industrial plant with 100 m 2 of CPC collectors for a 250 L/h treatment capacity has been designed. The solar system discharges the wastewater (WW) pre-treated by photo-Fenton into a biotreatment based on an immobilized biomass reactor. First, results of the industrial plant are also presented, demonstrating that it is able to treat up to 500 L h -1 at an average solar ultraviolet radiation of 22.9 W m -2 , under the same conditions (pH, hydrogen peroxide consumption) tested in the pilot plant

Testing of mobile surveillance robot at a nuclear power plant

International Nuclear Information System (INIS)

White, J.R.; Harvey, H.W.; Farnstrom, K.A.

1987-01-01

In-plant testing of a mobile surveillance robot (SURBOT) was performed at the Browns Ferry Nuclear Plant by TVA personnel. The results verified that SURBOT can be used for remote surveillance in 54 separate controlled radiation rooms at the plant. High-quality color video, audio, and other data are collected, digitized by an on-board computer, and transmitted through a cable to the control console for real-time display and videotaping. TVA projects that the use of SURBOT for surveillance during plant operation will produce annual savings of about 100 person-rem radiation exposure and $200,000 in operating costs. Based on the successful results of this program, REMOTEC is now commercializing the SURBOT technology on both wheeled and tracked mobile robots for use in nuclear power plants and other hazardous environments

Medicinal Plants Based Products Tested on Pathogens Isolated from Mastitis Milk

Directory of Open Access Journals (Sweden)

Claudia Pașca

2017-09-01

Full Text Available Bovine mastitis a major disease that is commonly associated with bacterial infection. The common treatment is with antibiotics administered intramammary into infected quarters of the udder. The excessive use of antibiotics leads to multidrug resistance and associated risks for human health. In this context, the search for alternative drugs based on plants has become a priority in livestock medicine. These products have a low manufacturing cost and no reports of antimicrobial resistance to these have been documented. In this context, the main objective of this study was to determine the antimicrobial effect of extracts and products of several indigenous, or acclimatized plants on pathogens isolated from bovine mastitis. A total of eleven plant alcoholic extracts and eight plant-derived products were tested against 32 microorganisms isolated from milk. The obtained results have shown an inhibition of bacterial growth for all tested plants, with better results for Evernia prunastri, Artemisia absinthium, and Lavandula angustifolia. Moreover, E. prunastri, Populus nigra, and L. angustifolia presented small averages of minimum inhibitory and bactericidal concentrations. Among the plant-derived products, three out of eight have shown a strong anti-microbial effect comparable with the effect of florfenicol and enrofloxacin, and better than individual plant extracts possibly due to synergism. These results suggest an important anti-microbial effect of these products on pathogens isolated from bovine mastitis with a possible applicability in this disease.

Operating results obtained in a nuclear power plant with a sensor surveillance prototype

International Nuclear Information System (INIS)

Jacquot, J.P.; Poujol, A.; Beaubatie, J.; Ciaramitaro, W.

1983-03-01

Surveillance methods have been validated and specific equipment have been built to measure the response time of sensors from a nuclear power plant protection channel. The reason of the choice of this parameter is twofold: the sensor response time is representative of the sensor physical status and is also part of the overall channel response time. Two surveillance methods are used: noise analysis (by AR or PSD modeling), and loop current step response (for resistance thermometer detectors only). The methods were validated on test facilities and on nuclear power plants. Two test equipments were built and tested on plants. Results are represented and conclusions are drawn on the feasibility of such methods for sensor surveillance [fr

Test results for fuel cell operation on anaerobic digester gas

Science.gov (United States)

Spiegel, R. J.; Preston, J. L.

EPA, in conjunction with ONSI, embarked on a project to define, design, test, and assess a fuel cell energy recovery system for application at anaerobic digester waste water (sewage) treatment plants. Anaerobic digester gas (ADG) is produced at these plants during the process of treating sewage anaerobically to reduce solids. ADG is primarily comprised of methane (57-66%), carbon dioxide (33-39%), nitrogen (1-10%), and a small amount of oxygen (sulfur-bearing compounds (principally hydrogen sulfide) and halogen compounds (chlorides). The project has addressed two major issues: development of a cleanup system to remove fuel cell contaminants from the gas and testing/assessing of a modified ONSI PC25 C fuel cell power plant operating on the cleaned, but dilute, ADG. Results to date demonstrate that the ADG fuel cell power plant can, depending on the energy content of the gas, produce electrical output levels close to full power (200 kW) with measured air emissions comparable to those obtained by a natural gas fuel cell. The cleanup system results show that the hydrogen sulfide levels are reduced to below 10 ppbv and halides to approximately 30 ppbv.

Tele-maintenance 'intelligent' system for technical plants result management

International Nuclear Information System (INIS)

Concetti, Massimo; Cuccioletta, Roberto; Fedele, Lorenzo; Mercuri, Giampiero

2009-01-01

The management of technical plant for productivity and safety is generally a complex activity, particularly when many plants distributed in the territory are considered (i.e. the more and more frequent case of outsourced plants maintenance by specialized companies), granted quality and cost results are required (i.e. the case of some rather innovative contract solutions) and the technology involved is heterogeneous and innovative (i.e. electro-mechanical plants). In order to efficiently achieve the above aims an 'intelligent' maintenance-management system for the distant monitoring and controlling by a remote control center has been developed. The so-called GrAMS (granted availability management system) system is conceived to give to organizations involved in technical-industrial plants management the possibility to tend to a 'well-known availability' and 'zero-failures' management. In particular, this study deals with the diagnostic aspects and safety level of technical plants (such as elevators, thermo-technical plants, etc.), and with the involvement of ad hoc designed software analysis tools based on neural networks and reliability indicators. Part of the research dealing with the tele-maintenance intelligent system has been financed by the Italian High Institute for Safety (ISPESL) and led to the development of a pre-industrial prototype whose realization and testing is here described

Acceptance test report for project C-157 ''T-Plant electrical upgrade''

International Nuclear Information System (INIS)

Jeppson, L.A.

1997-01-01

This Acceptance Test Report (ATR) documents for record purposes the field results, acceptance, and approvals of the completed acceptance test per WHC-SD-Cl57-ATP-001, Rev. 0, ''Acceptance Test Proceedure for Project C-157 'T Plant Electrical Upgrade''' The test was completed and approved without any problems or exceptions

Acceptance test report for project C-157 ``T-Plant electrical upgrade``

Energy Technology Data Exchange (ETDEWEB)

Jeppson, L.A.

1997-08-05

This Acceptance Test Report (ATR) documents for record purposes the field results, acceptance, and approvals of the completed acceptance test per WHC-SD-Cl57-ATP-001, Rev. 0, ``Acceptance Test Proceedure for Project C-157 `T Plant Electrical Upgrade``` The test was completed and approved without any problems or exceptions.

The comparative analysis of model and prototype test results of Bulb turbine

International Nuclear Information System (INIS)

Benisek, M; Bozic, I; Ignjatovic, B

2010-01-01

This paper presents the problem of the hydropower plant oblique water inflow and its influence on the turbines operation. Oblique water inflow on the low head hydropower plant with bulb turbines influences turbine characteristics. The characteristics change occurs due to swirl incidence in the turbine inlet which spreads to the guide vanes inlet. Downstream, the flow conditions change is caused in the turbine runner in relation to the flow conditions without swirl inflow. Special attention is paid to the phenomenon of swirl flow incidence in the turbine conduit. With the aim of presenting and analyzing the oblique water inflow consequences on the hydropower plant operation, the existing turbine model tests results, performed in the laboratories, and the in situ prototype testing results have been used.

Results of extended plant tests using more realistic exposure scenarios for improving environmental risk assessment of veterinary pharmaceuticals

OpenAIRE

Richter, Elisabeth; Berkner, Silvia; Ebert, Ina; Förster, Bernhard; Graf, Nadin; Herrchen, Monika; Kuehnen, Ute; Römbke, Jörg; Simon, Markus

2016-01-01

Background Residues of veterinary medicinal products (VMPs) enter the environment via application of manure onto agricultural areas where in particular antibiotics can cause phytotoxicity. Terrestrial plant tests according to OECD guideline 208 are part of the environmental risk assessment of VMPs. However, this standard approach might not be appropriate for VMPs which form non-extractable residues or transformation products in manure and manure-amended soil. Therefore, a new test design with...

10-MWe pilot-plant-receiver panel test requirements document solar thermal test facility

Energy Technology Data Exchange (ETDEWEB)

1978-08-25

Testing plans for a full-scale test receiver panel and supporting hardware which essentially duplicate both physically and functionally, the design planned for the Barstow Solar Pilot Plant are presented. Testing is to include operation during normal start and shutdown, intermittent cloud conditions, and emergencies to determine the panel's transient and steady state operating characteristics and performance under conditions equal to or exceeding those expected in the pilot plant. The effects of variations of input and output conditions on receiver operation are also to be investigated. Test hardware are described, including the pilot plant receiver, the test receiver assembly, receiver panel, flow control, electrical control and instrumentation, and structural assembly. Requirements for the Solar Thermal Test Facility for the tests are given. The safety of the system is briefly discussed, and procedures are described for assembly, installation, checkout, normal and abnormal operations, maintenance, removal and disposition. Also briefly discussed are quality assurance, contract responsibilities, and test documentation. (LEW)

Summary revaluation of cold testing of the first block of nuclear power plant Mochovce

International Nuclear Information System (INIS)

Miskolci, M.; Sarvaic, I.

1998-01-01

The document contents summary revaluation of the stage of cold testing of the first unit of nuclear power plant Mochovce. The valuation is processed in individual systems with safety significance. The process and results of system testing and their conclusions for the block readiness for active testing are summarized in the document. The valuation has been elaborated by a scientific management for start-up of nuclear power plant Mochovce as an independent conductor assistance for activation check from the nuclear safety point of view. The valuation of the activation results of systems in the first unit of nuclear power plant was processed as of 15.3.1998

Integrated leak rate test results of JOYO reactor containment vessel

International Nuclear Information System (INIS)

Tamura, M.; Endo, J.

1982-02-01

Integrated leak rate tests of JOYO after the reactor coolant system had been filled with sodium have been performed two times since 1978 (February 1978 and December 1979). The tests were conducted with the in-containment sodium systems, primary argon cover gas system and air conditioning systems operating. Both the absolute pressure method and the reference chamber method were employed during the test. The results of both tests confirmed the functioning of the containment vessel, and leak rate limits were satisfied. In Addition, the adequancy of the test instrumentation system and the test method was demonstrated. Finally the plant conditions required to maintain reasonable accuracy for the leak rate testing of LMFBR were established. In this paper, the test conditions and the test results are described. (author)

Preventive maintenance instrumentation results in Spanish nuclear power plants

International Nuclear Information System (INIS)

Curiel, M.; Palomo, M. J.; Verdu, G.; Arnaldos, A.

2010-10-01

This paper is a recompilation of the most significance results in relation to the researching in preventive and predictive maintenance in critical nuclear instrumentation for power plant operation, which it is being developed by Logistica y Acondicionamientos Industriales and the Isirym Institute of the Polytechnic University of Valencia. Instrumentation verification and test, it is a priority of the power plants control and instrumentation department's technicians. These procedures are necessary information for the daily power plant work. It is performed according to different procedures and in different moments of the fuel cycle depending on the instrumentation critical state and the monitoring process. Normally, this study is developed taking into account the instantaneous values of the instrumentation measures and, after their conversion to physical magnitude, they are analyzed according to the power plant operation point. Moreover, redundant sensors measurements are taken into consideration to the equipment and/or power plant monitoring. This work goes forward and it is in advanced to the instrument analysis as it is, independently of the operation point, using specific signal analysis techniques for preventive and predictive maintenance, with the object to obtain not only information about possible malfunctions, but the degradation scale presented in the instrument or in the system measured. We present seven real case studies of Spanish nuclear power plants each of them shall give a significant contribution to problem resolution and power plant performance. (Author)

Production Facility Prototype Blower 1000 Hour Test Results II

Energy Technology Data Exchange (ETDEWEB)

Wass, Alexander Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Woloshun, Keith Albert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dale, Gregory E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dalmas, Dale Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Romero, Frank Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2018-01-08

Long duration tests of the Aerzen GM 12.4 roots style blower in a closed loop configuration provides valuable data and lessons learned for long-term operation at the Mo-99 production facility. The blower was operated in a closed loop configuration with the flow conditions anticipated in plant operation with a Mo-100 target inline. The additional thermal energy generated from beam heating of the Mo-100 disks were not included in these tests. Five 1000 hour tests have been completed since the first test was performed in January of 2016. All five 1000 hour tests have proven successful in exposing preventable issues related to oil and helium leaks. All blower tests to this date have resulted in stable blower performance and consistency. A summary of the results for each test, including a review of the first and second tests, are included in this report.

Remote maintenance demonstration tests at a pilot plant for high level waste vitrification

International Nuclear Information System (INIS)

Selig, M.

1984-01-01

The remote maintenance and replacement technique designed for a radioactive vitrification plant have been developed and tested in a full scale handling mockup and in an inactive pilot plants by the Central Engineering Department of the Karlsruhe Nuclear Research Center. As a result of the development work and the tests it has been proved that the remote maintenance technique and remote handling equipment can be used without any technical problems and are suited for application in a radioactive waste vitrification plant

FFTF thermal-hydraulic testing results affecting piping and vessel component design in LMFBR's

International Nuclear Information System (INIS)

Stover, R.L.; Beaver, T.R.; Chang, S.C.

1983-01-01

The Fast Flux Test Facility completed four years of pre-operational testing in April 1982. This paper describes thermal-hydraulic testing results from this period which impact piping and vessel component design in LMFBRs. Data discussed are piping flow oscillations, piping thermal stratification and vessel upper plenum stratification. Results from testing verified that plant design limits were met

Interim results: fines recycle testing using the 4-inch diameter primary graphite burner

International Nuclear Information System (INIS)

Palmer, W.B.

1975-05-01

The results of twenty-two HTGR primary burner runs in which graphite fines were recycled pneumatically to the 4-inch diameter pilot-plant primary fluidized-bed burner are described. The result of the tests showed that zero fines accumulation can easily be achieved while operating at plant equivalent burn rates. (U.S.)

Methodological approaches to conducting pilot and proof tests on reverse-osmosis systems: Results of comparative studies

Science.gov (United States)

Panteleev, A. A.; Bobinkin, V. V.; Larionov, S. Yu.; Ryabchikov, B. E.; Smirnov, V. B.; Shapovalov, D. A.

2017-10-01

When designing large-scale water-treatment plants based on reverse-osmosis systems, it is proposed to conduct experimental-industrial or pilot tests for validated simulation of the operation of the equipment. It is shown that such tests allow establishing efficient operating conditions and characteristics of the plant under design. It is proposed to conduct pilot tests of the reverse-osmosis systems on pilot membrane plants (PMPs) and test membrane plants (TMPs). The results of a comparative experimental study of pilot and test membrane plants are exemplified by simulating the operating parameters of the membrane elements of an industrial plant. It is concluded that the reliability of the data obtained on the TMP may not be sufficient to design industrial water-treatment plants, while the PMPs are capable of providing reliable data that can be used for full-scale simulation of the operation of industrial reverse-osmosis systems. The test membrane plants allow simulation of the operating conditions of individual industrial plant systems; therefore, potential areas of their application are shown. A method for numerical calculation and experimental determination of the true selectivity and the salt passage are proposed. An expression has been derived that describes the functional dependence between the observed and true salt passage. The results of the experiments conducted on a test membrane plant to determine the true value of the salt passage of a reverse-osmosis membrane are exemplified by magnesium sulfate solution at different initial operating parameters. It is shown that the initial content of a particular solution component has a significant effect on the change in the true salt passage of the membrane.

Numerical and field tests of hydraulic transients at Piva power plant

International Nuclear Information System (INIS)

Giljen, Z

2014-01-01

In 2009, a sophisticated field investigation was undertaken and later, in 2011, numerical tests were completed, on all three turbine units at the Piva hydroelectric power plant. These tests were made in order to assist in making decisions about the necessary scope of the reconstruction and modernisation of the Piva hydroelectric power plant, a plant originally constructed in the mid-1970s. More specifically, the investigation included several hydraulic conditions including both the start-up and stopping of each unit, load rejection under governor control from different initial powers, as well as emergency shut-down. Numerical results were obtained using the method of characteristics in a representation that included the full flow system and the characteristics of each associated Francis turbine. The impact of load rejection and emergency shut-down on the penstock pressure and turbine speed changes are reported and numerical and experimental results are compared, showing close agreement

Estimation of Cadmium uptake by tobacco plants from laboratory leaching tests.

Science.gov (United States)

Marković, Jelena P; Jović, Mihajlo D; Smičiklas, Ivana D; Šljivić-Ivanović, Marija Z; Smiljanić, Slavko N; Onjia, Antonije E; Popović, Aleksandar R

2018-03-21

The objective of the present study was to determine the impact of cadmium (Cd) concentration in the soil on its uptake by tobacco plants, and to compare the ability of diverse extraction procedures for determining Cd bioavailability and predicting soil-to-plant transfer and Cd plant concentrations. The pseudo-total digestion procedure, modified Tessier sequential extraction and six standard single-extraction tests for estimation of metal mobility and bioavailability were used for the leaching of Cd from a native soil, as well as samples artificially contaminated over a wide range of Cd concentrations. The results of various leaching tests were compared between each other, as well as with the amounts of Cd taken up by tobacco plants in pot experiments. In the native soil sample, most of the Cd was found in fractions not readily available under natural conditions, but with increasing pollution level, Cd amounts in readily available forms increased. With increasing concentrations of Cd in the soil, the quantity of pollutant taken up in tobacco also increased, while the transfer factor (TF) decreased. Linear and non-linear empirical models were developed for predicting the uptake of Cd by tobacco plants based on the results of selected leaching tests. The non-linear equations for ISO 14870 (diethylenetriaminepentaacetic acid extraction - DTPA), ISO/TS 21268-2 (CaCl 2 leaching procedure), US EPA 1311 (toxicity characteristic leaching procedure - TCLP) single step extractions, and the sum of the first two fractions of the sequential extraction, exhibited the best correlation with the experimentally determined concentrations of Cd in plants over the entire range of pollutant concentrations. This approach can improve and facilitate the assessment of human exposure to Cd by tobacco smoking, but may also have wider applicability in predicting soil-to-plant transfer.

Preventive maintenance instrumentation results in Spanish nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Curiel, M. [Logistica y Acondicionamientos Industriales SAU, Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain); Palomo, M. J.; Verdu, G. [ISIRYM, Universidad Politecnica de Valencia, Camino de Vera s/n, Valencia (Spain); Arnaldos, A., E-mail: m.curiel@lainsa.co [TITANIA Servicios Tecnologicos SL, Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain)

2010-10-15

This paper is a recompilation of the most significance results in relation to the researching in preventive and predictive maintenance in critical nuclear instrumentation for power plant operation, which it is being developed by Logistica y Acondicionamientos Industriales and the Isirym Institute of the Polytechnic University of Valencia. Instrumentation verification and test, it is a priority of the power plants control and instrumentation department's technicians. These procedures are necessary information for the daily power plant work. It is performed according to different procedures and in different moments of the fuel cycle depending on the instrumentation critical state and the monitoring process. Normally, this study is developed taking into account the instantaneous values of the instrumentation measures and, after their conversion to physical magnitude, they are analyzed according to the power plant operation point. Moreover, redundant sensors measurements are taken into consideration to the equipment and/or power plant monitoring. This work goes forward and it is in advanced to the instrument analysis as it is, independently of the operation point, using specific signal analysis techniques for preventive and predictive maintenance, with the object to obtain not only information about possible malfunctions, but the degradation scale presented in the instrument or in the system measured. We present seven real case studies of Spanish nuclear power plants each of them shall give a significant contribution to problem resolution and power plant performance. (Author)

SPES-2, the full-height, full-pressure, test facility simulating the AP600 plant: Main results from the experimental campaign

International Nuclear Information System (INIS)

Medich, C.; Rigamonti, M.; Martinelli, R.; Tarantini, M.; Conway, L.

1995-01-01

The SPES-2 is a full height, full pressure experimental test facility reproducing the Westinghouse AP600 reactor with a scaling factor of 1/395. The experimental plant, designed and operated by SIET in Piacenza, consists of a full simulation of the AP600 primary core cooling system including all the passive and active safety systems. In 1992, Westinghouse, in cooperation with ENEL, ENEA, SIET and ANSALDO developed an experimental program to test the integrated behavior of the AP600 passive safety systems. The SPES-2 test matrix, concluded in November 1994, has examined the AP600 passive safety system response for a range of small break LOCAs at different locations on the primary system and on the passive system lines; single steam generator tube ruptures with both passive and active non-safety systems, and a main steam line break transient to demonstrate the capability of passive safety systems for rapid cooldown. Each of the tests has provided detailed experimental results for verification of the capability of the analysis methods to predict the integrated passive safety system behavior

Potential safety enhancements to nuclear plant control: proof testing at EBR-II

International Nuclear Information System (INIS)

Lindsay, R.W.; Chisholm, G.H.

1984-01-01

Future changes in nuclear plant control and protective systems will reflect an evolutionary improvement through increased use of computers coupled with a better integration of man and machine. Before improvements can be accepted into the licensed commercial plant environment, significant testing must be accomplished to answer safety questions and to prove the worth of new ideas. The Experimental Breeder Reactor-II (EBR-II) is being used as a test-bed for both in-house development and testing for others in a DOE sponsored Man-Machine Integration program. The ultimate result of the development and testing would be a control system for which safety credit could be taken in the licensing process

High Temperature Calcination - MACT Upgrade Equipment Pilot Plant Test

Energy Technology Data Exchange (ETDEWEB)

Richard D. Boardman; B. H. O& #39; Brien; N. R. Soelberg; S. O. Bates; R. A. Wood; C. St. Michel

2004-02-01

About one million gallons of acidic, hazardous, and radioactive sodium-bearing waste are stored in stainless steel tanks at the Idaho Nuclear Technology and Engineering Center (INTEC), which is a major operating facility of the Idaho National Engineering and Environmental Laboratory. Calcination at high-temperature conditions (600 C, with alumina nitrate and calcium nitrate chemical addition to the feed) is one of four options currently being considered by the Department of Energy for treatment of the remaining tank wastes. If calcination is selected for future processing of the sodium-bearing waste, it will be necessary to install new off-gas control equipment in the New Waste Calcining Facility (NWCF) to comply with the Maximum Achievable Control Technology (MACT) standards for hazardous waste combustors and incinerators. This will require, as a minimum, installing a carbon bed to reduce mercury emissions from their current level of up to 7,500 to <45 {micro}g/dscm, and a staged combustor to reduce unburned kerosene fuel in the off-gas discharge to <100 ppm CO and <10 ppm hydrocarbons. The staged combustor will also reduce NOx concentrations of about 35,000 ppm by 90-95%. A pilot-plant calcination test was completed in a newly constructed 15-cm diameter calciner vessel. The pilot-plant facility was equipped with a prototype MACT off-gas control system, including a highly efficient cyclone separator and off-gas quench/venturi scrubber for particulate removal, a staged combustor for unburned hydrocarbon and NOx destruction, and a packed activated carbon bed for mercury removal and residual chloride capture. Pilot-plant testing was performed during a 50-hour system operability test January 14-16, followed by a 100-hour high-temperature calcination pilot-plant calcination run January 19-23. Two flowsheet blends were tested: a 50-hour test with an aluminum-to-alkali metal molar ratio (AAR) of 2.25, and a 50-hour test with an AAR of 1.75. Results of the testing

Results and implications of the EBR-II inherent safety demonstration tests

International Nuclear Information System (INIS)

Planchon, H.P.; Golden, G.H.; Sackett, J.I.; Mohr, D.; Chang, L.K.; Feldman, E.E.; Betten, P.R.

1987-01-01

On April 3, 1986 two milestone tests were conducted in Experimental Breeder Reactor-2 (EBR-II). The first test was a loss of flow without scram and the second was a loss of heat sink without scram. Both tests were initiated from 100% power and in both tests the reactor was shut down by natural processes, principally thermal expansion, without automatic scram, operator intervention or the help of special in-core devices. The temperature transients during the tests were mild, as predicted, and there was no damage to the core or reactor plant structures. In a general sense, therefore, the tests plus supporting analysis demonstrated the feasibility of inherent passive shutdown for undercooling accidents in metal-fueled LMRs. The results provide a technical basis for future experiments in EBR-II to demonstrate inherent safety for overpower accidents and provide data for validation of computer codes used for design and safety analysis of inherently safe reactor plants

Verification tests for remote controlled inspection system in nuclear power plants

International Nuclear Information System (INIS)

Kohno, Tadaaki

1986-01-01

Following the increase of nuclear power plants, the total radiation exposure dose accompanying inspection and maintenance works tended to increase. Japan Power Engineering and Inspection Corp. carried out the verification test of a practical power reactor automatic inspection system from November, 1981, to March, 1986, and in this report, the state of having carried out this verification test is described. The objects of the verification test were the equipment which is urgently required for reducing radiation exposure dose, the possibility of realization of which is high, and which is important for ensuring the safety and reliability of plants, that is, an automatic ultrasonic flaw detector for the welded parts of bend pipes, an automatic disassembling and inspection system for control rod driving mechanism, a fuel automatic inspection system, and automatic decontaminating equipments for steam generator water chambers, primary system crud and radioactive gas in coolant. The results of the verification test of these equipments were judged as satisfactory, therefore, the application to actual plants is possible. (Kako, I.)

TASK 3: PILOT PLANT GASIFIER TESTING

Energy Technology Data Exchange (ETDEWEB)

Fusselman, Steve

2015-11-01

Aerojet Rocketdyne (AR) has developed an innovative gasifier concept incorporating advanced technologies in ultra-dense phase dry feed system, rapid mix injector, and advanced component cooling to significantly improve gasifier performance, life, and cost compared to commercially available state-of-the-art systems. Design, fabrication and initial testing of the pilot plant compact gasifier was completed in 2011 by a development team led by AR. Findings from this initial test program, as well as subsequent gasifier design and pilot plant testing by AR, identified a number of technical aspects to address prior to advancing into a demonstration-scale gasifier design. Key among these were an evaluation of gasifier ability to handle thermal environments with highly reactive coals; ability to handle high ash content, high ash fusion temperature coals with reliable slag discharge; and to develop an understanding of residual properties pertaining to gasification kinetics as carbon conversion approaches 99%. The gasifier did demonstrate the ability to withstand the thermal environments of highly reactive Powder River Basin coal, while achieving high carbon conversion in < 0.15 seconds residence time. Continuous operation with the high ash fusion temperature Xinyuan coal was demonstrated in long duration testing, validating suitability of outlet design as well as downstream slag discharge systems. Surface area and porosity data were obtained for the Xinyuan and Xinjing coals for carbon conversion ranging from 85% to 97%, and showed a pronounced downward trend in surface area per unit mass carbon as conversion increased. Injector faceplate measurements showed no incremental loss of material over the course of these experiments, validating the commercially traceable design approach and supportive of long injector life goals. Hybrid testing of PRB and natural gas was successfully completed over a wide range of natural gas feed content, providing test data to anchor predictions

Mobile test stand for evaluation of electric power plants for unmanned aircraft

Directory of Open Access Journals (Sweden)

Serbezov Vladimir

2017-01-01

Full Text Available The absence of accurate performance data is a common problem with most civilian unmanned aerial vehicle (UAV power plant producers. The reasons for this are the small size of most of the manufacturers and the high price of precise wind tunnel testing and computer simulations. To overcome this problem at Dronamics Ltd., with support from the Department of Aeronautics of TU-Sofia, a mobile test stand for evaluation of electric power plants for unmanned aircraft was developed. The stand may be used statically, or may be installed on the roof of an automobile. The measurement system of the stand is based on popular hardware that is used in radio controlled models and in general automation. The verification of the measurement system is performed by comparing static test results with data published by the manufacturer of the tested electric motor. Tests were carried out with 2 different types of propellers and the results were compared with published results for common propellers as well as with results of theoretical studies. The results are satisfactory for practical applications. The use of this type of test stands can be a cheap and effective alternative for research and development start-up companies like Dronamics.

Trace coupled with PARCS benchmark against Leibstadt plant data during the turbine trip test

Energy Technology Data Exchange (ETDEWEB)

Sekhri, Abdelkrim; Baumann, Peter, E-mail: abdelkrim.sekhri@kkl.ch, E-mail: peter.Baumann@kkl.ch [KernkraftwerkLeibstadt AG, Leibstadt (Switzerland); Hidalga, Patricio; Morera, Daniel; Miro, Rafael; Barrachina, Teresa; Verdu, Gumersindo, E-mail: pathigar@etsii.upv.es, E-mail: dmorera@isirym.upv.es, E-mail: rmiro@isirym.upv.es, E-mail: tbarrachina@isirym.upv.es, E-mail: gverdu@isirym.upv.es [Universitat Politecnica de Valencia (ISIRYM/UPV), Valencia, (Spain). Institute for Industrial, Radiophysical and Environmental Safety

2013-07-01

In order to enhance the modeling of Nuclear Power Plant Leibstadt (KKL), the coupling of 3D neutron kinetics PARCS code with TRACE has been developed. To test its performance a complex transient of Turbine Trip has been simulated comparing the results with the existing plant data of Turbine Trip test. For this transient also Cross Sections have been generated and used by PARCS. The thermal-hydraulic TRACE model is retrieved from the already existing model. For the benchmarking the Turbine Trip transient has been simulated according to the test resulting in the closure of the turbine control valve (TCV) and the following opening of the bypass valve (TBV). This transient caused a pressure shock wave towards the Reactor Pressure Vessel (RPV) which provoked the decreasing of the void level and the consequent slight power excursion. The power control capacity of the system showed a good response with the procedure of a Selected Rod Insertion (SRI) and the recirculation loops performance which resulted in the proper thermal power reduction comparable to APRM data recorder from the plant. The comparison with plant data shows good agreement in general and assesses the performance of the coupled model. Due to this, it can be concluded that the coupling of PARCS and TRACE codes in addition with the Cross Section used works successfully for simulating the behavior of the reactor core during complex plant transients. Nevertheless the TRACE model shall be improved and the core neutronics corresponding to the test shall be used in the future to allow quantitative comparison between TRACE and plant recorded data. (author)

Inferences from new plant design from fast flux test facility operation

International Nuclear Information System (INIS)

Peterson, R.E.; Peckinpaugh, C.L.; Simpson, D.E.

1985-04-01

Experience gained through operation of the Fast Flux Test Facility (FFTF) is now sufficiently extensive that this experience can be utilized in designing the next generation of liquid metal fast reactors. Experience with FFTF core and plant components is cited which can result in design improvements to achieve inherently safe, economic reactor plants. Of particular interest is the mixed oxide fuel system which has demonstrated large design margins. Other plant components have also demonstrated high reliability and offer capital cost reduction opportunities through design simplifications. The FFTF continues to be a valuable US resource which affords prototypic development and demonstration, contributing to public acceptability of future plants

UF6 test loop for evaluation and implementation of international enrichment plant safeguards

International Nuclear Information System (INIS)

Cooley, J.N.; Fields, L.W.; Swindle, D.W. Jr.

1987-06-01

A functional test loop capable of simulating UF 6 flows, pressures, and pipe deposits characteristic of gas centrifuge enrichment plant piping has been designed and fabricated by the Enrichment Safeguards Program of Martin Marietta Energy Systems, Inc., for use by International Atomic Energy Agency (IAEA) at its Safeguards Analytical Laboratory in Seibersdorf, Austria. Purpose of the test loop is twofold: (1) to enable the IAEA to evaluate and to calibrate enrichment safeguards measurement instrumentation to be used in limited frequency-unannounced access (LFUA) inspection strategy measurements at gas centrifuge enrichment plants and (2) to train IAEA inspectors in the use of such instrumentation. The test loop incorporates actual sections of cascade header pipes from the centrifuge enrichment plants subject to IAEA inspections. The test loop is described, applications for its use by the IAEA are detailed, and results from an initial demonstration session using the test loop are summarized

The use of NPAR [Nuclear Plant Aging Research] results in plant inspection activities

International Nuclear Information System (INIS)

Gunther, W.; Taylor, J.

1989-01-01

The US NRC's Nuclear Plant Aging Research (NPAR) Program is a hardware oriented research program which has produced a large data base of equipment and system operating, maintenance, and testing information. Equipment and systems which have a propensity for age related degradation are identified, and methods for detecting and mitigating aging effects have been evaluated. As plants age, it becomes increasingly important that NRC inspectors be cognizant of plant aging phenomena. This paper describes the NPAR information which can enhance inspection activities, and provides a mechanism for making pertinent research available to the inspectors. 7 refs., 2 figs

Round Robin Test for Performance Demonstration System of Ultrasound Examination Personnel in Nuclear Power Plants

International Nuclear Information System (INIS)

Kim, Young Ho; Yang, Seung Han; Kim, Yong Sik; Yoon, Byung Sik; Lee, Hee Jong

2005-01-01

Ultrasound testing performance during in-service inspection for the main components of NPPs is strongly affected by each examination person. Therefore, ASME established a more strict qualification requirement in Sec. XI Appendix VIII for the ultrasound testing personnel in nuclear power plants. The Korean Performance Demonstration (KPD) System according to the ASME code for the ultrasonic testing personnel, equipments, and procedures to apply to the Class 1 and 2 piping ultrasound examination of nuclear power plants in Korea was established. And a round robin test was conducted in order to verify the effectiveness of PD method by comparing the examination results from the method of Performance Demonstration (PD) and a traditional ASME code dB-drop method. The round robin test shows that the reliability of the PD method is better than that of the dB-drop method. As a result, application of the PD method to the in-service inspection of the nuclear power plants will improve the performance of ultrasound testing

Boeing's STAR-FODB test results

Science.gov (United States)

Fritz, Martin E.; de la Chapelle, Michael; Van Ausdal, Arthur W.

1995-05-01

Boeing has successfully concluded a 2 1/2 year, two phase developmental contract for the STAR-Fiber Optic Data Bus (FODB) that is intended for future space-based applications. The first phase included system analysis, trade studies, behavior modeling, and architecture and protocal selection. During this phase we selected AS4074 Linear Token Passing Bus (LTPB) protocol operating at 200 Mbps, along with the passive, star-coupled fiber media. The second phase involved design, build, integration, and performance and environmental test of brassboard hardware. The resulting brassboard hardware successfully passed performance testing, providing 200 Mbps operation with a 32 X 32 star-coupled medium. This hardware is suitable for a spaceflight experiment to validate ground testing and analysis and to demonstrate performace in the intended environment. The fiber bus interface unit (FBIU) is a multichip module containing transceiver, protocol, and data formatting chips, buffer memory, and a station management controller. The FBIU has been designed for low power, high reliability, and radiation tolerance. Nine FBIUs were built and integrated with the fiber optic physical layer consisting of the fiber cable plant (FCP) and star coupler assembly (SCA). Performance and environmental testing, including radiation exposure, was performed on selected FBIUs and the physical layer. The integrated system was demonstrated with a full motion color video image transfer across the bus while simultaneously performing utility functions with a fiber bus control module (FBCM) over a telemetry and control (T&C) bus, in this case AS1773.

Preventive maintenance instrumentation results in Spanish nuclear power plants

International Nuclear Information System (INIS)

Palomo Anaya, M. Jose; Verdu Martin, Gumersindo; Arnaldos Gonzalvez, Adoracion; Nieva, Marcelino Curiel

2011-01-01

This paper is a recompilation of the most significant results in relation to the researching in Preventive and Predictive Maintenance in critical nuclear instrumentation for power plant operation, which it is being developed by Logistica y Acondicionamientos Industriales and The Isirym Institute of the Polytechnic University of Valencia. Instrumentation verification and test, it is a priority of the Power Plants Control and Instrumentation Department technicians. These procedures are necessary information for the daily power plant work. It is performed according to different procedures and in different moments of the fuel cycle depending on the instrumentation critical state and the monitoring process. Normally, this study is developed taking into account the instantaneous values of the instrumentation measures and, after their conversion to physical magnitude, they are analyzed according to the power plant operation point. Moreover, redundant sensors measurements are taken into consideration to the equipment and/or power plant monitoring. This work goes forward and it is in advanced to the instrument analysis as it is, independently of the operation point, using specific signal analysis techniques for preventive and predictive maintenance, with the aim to obtain not only information about possible malfunctions, but the degradation scale presented in the instrument or in the system measured. We present seven real case studies of Spanish Nuclear Power Plants each of them shall give a significant contribution to problem resolution and power plant performance: Fluctuations in sensor lines (case 1), Air presence in feed water lines (case 2), Root valve partially closed (case 3), Sensor malfunctions (case 4), Electrical source malfunctions (case 5), RTD malfunctions (case 6) and LPRM malfunctions (case 7). (author)

Preventive maintenance instrumentation results in Spanish nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Palomo Anaya, M. Jose; Verdu Martin, Gumersindo, E-mail: mpalomo@iqn.upv.es, E-mail: gverdu@iqn.upv.es [ISIRYM Universidad Politecnica de Valencia, Valencia (Spain); Arnaldos Gonzalvez, Adoracion, E-mail: a.arnaldos@titaniast.com [TITANIA Servicios Tecnologicos SL, Valencia (Spain); Nieva, Marcelino Curiel, E-mail: m.curiel@lainsa.com [Logistica y Acondicionamientos Industriales SAU (LAINSA), Valencia (Spain)

2011-07-01

This paper is a recompilation of the most significant results in relation to the researching in Preventive and Predictive Maintenance in critical nuclear instrumentation for power plant operation, which it is being developed by Logistica y Acondicionamientos Industriales and The Isirym Institute of the Polytechnic University of Valencia. Instrumentation verification and test, it is a priority of the Power Plants Control and Instrumentation Department technicians. These procedures are necessary information for the daily power plant work. It is performed according to different procedures and in different moments of the fuel cycle depending on the instrumentation critical state and the monitoring process. Normally, this study is developed taking into account the instantaneous values of the instrumentation measures and, after their conversion to physical magnitude, they are analyzed according to the power plant operation point. Moreover, redundant sensors measurements are taken into consideration to the equipment and/or power plant monitoring. This work goes forward and it is in advanced to the instrument analysis as it is, independently of the operation point, using specific signal analysis techniques for preventive and predictive maintenance, with the aim to obtain not only information about possible malfunctions, but the degradation scale presented in the instrument or in the system measured. We present seven real case studies of Spanish Nuclear Power Plants each of them shall give a significant contribution to problem resolution and power plant performance: Fluctuations in sensor lines (case 1), Air presence in feed water lines (case 2), Root valve partially closed (case 3), Sensor malfunctions (case 4), Electrical source malfunctions (case 5), RTD malfunctions (case 6) and LPRM malfunctions (case 7). (author)

Safety demonstration test on solvent fire in fuel reprocessing plant

International Nuclear Information System (INIS)

Nishio, Gunji; Hashimoto, Kazuichiro

1989-03-01

This report summarizes a fundamental of results obtained in the Reprocessing Plant Safety Demonstration Test Program which was performed under the contract between the Science and Technology Agency of Japan and the Japan Atomic Energy Research Institute. In this test program, a solvent fire was hypothesized, and such data were obtained as fire behavior, smoke behavior and integrity of exhaust filters in the ventilation system. Through the test results, it was confirmed that under the fire condition in hypothetical accident, the integrity of the cell and the cell ventilation system were maintained, and the safety function of the exhaust filters was maintained against the smoke loading. Analytical results by EVENT code agreed well with the present test data on the thermofluid flow in a cell ventilation system. (author)

Power plant cable condition monitoring and testing at Georgia Power

International Nuclear Information System (INIS)

Champion, T.C.

1988-01-01

Georgia Power's Research Center has been heavily involved in the evaluation of electrical insulating materials and cables since its inception more than 17 years ago. For the past ten years that expertise has been applied to cables used in generation plants. This paper discusses the results of two test programs. The first is a quality control inspection on 169 samples of new power generation cables. The second is a material degradation evaluation on four short cable samples removed from a coal fired plant during an equipment upgrade. The new material evaluation was performed to identify the cause of a high failure rate upon initial hi-pot testing of newly installed cables. The material degradation evaluation was performed to evaluate the need for replacement of existing cables during an equipment upgrade. Results of the evaluations have led to development of a detailed proposal for a program to evaluate cable degradation and remaining life for cables used in power generation facilities

Results of the gas carrier reliquefaction plant trial

Directory of Open Access Journals (Sweden)

Y. Fatyhov

2007-12-01

Full Text Available In the paper results of the gas carrier reliquefaction plant trial are considered. Safe transportation of liquefied gases is explained. The construction of the ship on trial is described. Designed parameters of the reliquefaction plant are presented. Heat gain into cargo tanks is obtained. Volumetric capacity, cooling capacity, volumetric efficiency and power consumption of the compressors are determined. Results of the main engine trial, diesel generator trial, reliquefaction plant trial, and calculations performed after wards are represented in five tables. The results obtained may be used for optimisation calculations of gas carriers’ reliquefaction plants.

A study of statistical tests for near-real-time materials accountancy using field test data of Tokai reprocessing plant

International Nuclear Information System (INIS)

Ihara, Hitoshi; Nishimura, Hideo; Ikawa, Koji; Miura, Nobuyuki; Iwanaga, Masayuki; Kusano, Toshitsugu.

1988-03-01

An Near-Real-Time Materials Accountancy(NRTA) system had been developed as an advanced safeguards measure for PNC Tokai Reprocessing Plant; a minicomputer system for NRTA data processing was designed and constructed. A full scale field test was carried out as a JASPAS(Japan Support Program for Agency Safeguards) project with the Agency's participation and the NRTA data processing system was used. Using this field test data, investigation of the detection power of a statistical test under real circumstances was carried out for five statistical tests, i.e., a significance test of MUF, CUMUF test, average loss test, MUF residual test and Page's test on MUF residuals. The result shows that the CUMUF test, average loss test, MUF residual test and the Page's test on MUF residual test are useful to detect a significant loss or diversion. An unmeasured inventory estimation model for the PNC reprocessing plant was developed in this study. Using this model, the field test data from the C-1 to 85 - 2 campaigns were re-analyzed. (author)

9 CFR 113.6 - Animal and Plant Health Inspection Service testing.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Animal and Plant Health Inspection Service testing. 113.6 Section 113.6 Animals and Animal Products ANIMAL AND PLANT HEALTH INSPECTION... STANDARD REQUIREMENTS Applicability § 113.6 Animal and Plant Health Inspection Service testing. A...

Technology and testing for the extension of plant life

International Nuclear Information System (INIS)

Blumer, U.R.; Edelmann, X.

1988-01-01

This paper describes selected portions of a recommended program for the application of equipment-manufacturing-related technology and testing for the extension of life for operating nuclear power plants. It is appropriate to mention that the Swiss nuclear plants, their staffs, and the supporting Swiss nuclear industry are rightfully proud of their record of performance. Plant staffs have been intimately involved in system and equipment design and engineering from the very beginnings of their plants. Maintenance of the plant systems and equipment is referred to as engineering rather than maintenance, because it is viewed as a technical effort and an extension of the original plant and equipment design and construction effort. Care, competence, cleanliness, and attention to detail have been bywords for the Swiss plants. Success has been demonstrated through enviable availability performance. With operation and availability capability already demonstrated, the Swiss are now turning their attention to the extension of plant life. This summary describes some aspects of this work, which is fundamentally based on the application of technology and testing skills developed for equipment manufacture and the original installation of this equipment in the plants, but has been enhanced by research and development (R and D) and an ongoing effort to serve utilities in their maintenance activities

Assessment of impacts at the advanced test reactor as a result of chemical releases at the Idaho Chemical Processing Plant

International Nuclear Information System (INIS)

Rood, A.S.

1991-02-01

This report provides an assessment of potential impacts at the Advanced Test Reactor Facility (ATR) resulting from accidental chemical spill at the Idaho Chemical Processing Plant (ICPP). Spills postulated to occur at the Lincoln Blvd turnoff to ICPP were also evaluated. Peak and time weighted average concentrations were calculated for receptors at the ATR facility and the Test Reactor Area guard station at a height above ground level of 1.0 m. Calculated concentrations were then compared to the 15 minute averaged Threshold Limit Value - Short Term Exposure Limit (TLV-STEL) and the 30 minute averaged Immediately Dangerous to Life and Health (IDLH) limit. Several different methodologies were used to estimate source strength and dispersion. Fifteen minute time weighted averaged concentrations of hydrofluoric acid and anhydrous ammonia exceeded TLV-STEL values for the cases considered. The IDLH value for these chemicals was not exceeded. Calculated concentrations of ammonium hydroxide, hexone, nitric acid, propane, gasoline, chlorine and liquid nitrogen were all below the TLV-STEL value

RESULTS OF THE FIRST MI-171A2 FLYING LABORATORY TEST PHASE

OpenAIRE

V. A. Ivchin; K. Y. Samsonov

2014-01-01

The present publication describes the results of the first stage of the flying laboratory (Mi-171 helicopter) flight tests performed at Mil Moscow Helicopter Plant, JSC facilities. Main rotor components with blades made of polymer composite materials and X-type tail rotor were tested on the Mi-171 № 14987, flying laboratory, under Mi-171A Helicopter Retrofit Program.

CRBRP design and test results for fuel handling systems, plugs, and seals

International Nuclear Information System (INIS)

Berg, G.E.

1977-01-01

The fuel handling system and reactor rotating plugs for the Clinch River Breeder Reactor Plant (CRBRP) are based primarily on existing technology and, in many respects, follow the concept developed for the Fast Flux Test Facility (FFTF). The equipment and the development programs initiated to verify its performance are described. Test results obtained from the development program, and the extent to which these results verified original design selections, or suggested potential improvements, are discussed

Sequential Design of Experiments to Maximize Learning from Carbon Capture Pilot Plant Testing

Energy Technology Data Exchange (ETDEWEB)

Soepyan, Frits B.; Morgan, Joshua C.; Omell, Benjamin P.; Zamarripa-Perez, Miguel A.; Matuszewski, Michael S.; Miller, David C.

2018-02-06

Pilot plant test campaigns can be expensive and time-consuming. Therefore, it is of interest to maximize the amount of learning and the efficiency of the test campaign given the limited number of experiments that can be conducted. This work investigates the use of sequential design of experiments (SDOE) to overcome these challenges by demonstrating its usefulness for a recent solvent-based CO2 capture plant test campaign. Unlike traditional design of experiments methods, SDOE regularly uses information from ongoing experiments to determine the optimum locations in the design space for subsequent runs within the same experiment. However, there are challenges that need to be addressed, including reducing the high computational burden to efficiently update the model, and the need to incorporate the methodology into a computational tool. We address these challenges by applying SDOE in combination with a software tool, the Framework for Optimization, Quantification of Uncertainty and Surrogates (FOQUS) (Miller et al., 2014a, 2016, 2017). The results of applying SDOE on a pilot plant test campaign for CO2 capture suggests that relative to traditional design of experiments methods, SDOE can more effectively reduce the uncertainty of the model, thus decreasing technical risk. Future work includes integrating SDOE into FOQUS and using SDOE to support additional large-scale pilot plant test campaigns.

The Benchmark Test Results of QNX RTOS

Energy Technology Data Exchange (ETDEWEB)

Kim, Jang Yeol; Lee, Young Jun; Cheon, Se Woo; Lee, Jang Soo; Kwon, Kee Choon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-10-15

A Real-Time Operating System(RTOS) is an Operating System(OS) intended for real-time applications. Benchmarking is a point of reference by which something can be measured. The QNX is a Real Time Operating System(RTOS) developed by QSSL(QNX Software Systems Ltd.) in Canada. The ELMSYS is the brand name of commercially available Personal Computer(PC) for applications such as Cabinet Operator Module(COM) of Digital Plant Protection System(DPPS) and COM of Digital Engineered Safety Features Actuation System(DESFAS). The ELMSYS PC Hardware is being qualified by KTL(Korea Testing Lab.) for use as a Cabinet Operator Module(COM). The QNX RTOS is being dedicated by Korea Atomic Energy Research Institute (KAERI). This paper describes the outline and benchmarking test results on Context Switching, Message Passing, Synchronization and Deadline Violation of QNX RTOS under the ELMSYS PC platform

The Benchmark Test Results of QNX RTOS

International Nuclear Information System (INIS)

Kim, Jang Yeol; Lee, Young Jun; Cheon, Se Woo; Lee, Jang Soo; Kwon, Kee Choon

2010-01-01

A Real-Time Operating System(RTOS) is an Operating System(OS) intended for real-time applications. Benchmarking is a point of reference by which something can be measured. The QNX is a Real Time Operating System(RTOS) developed by QSSL(QNX Software Systems Ltd.) in Canada. The ELMSYS is the brand name of commercially available Personal Computer(PC) for applications such as Cabinet Operator Module(COM) of Digital Plant Protection System(DPPS) and COM of Digital Engineered Safety Features Actuation System(DESFAS). The ELMSYS PC Hardware is being qualified by KTL(Korea Testing Lab.) for use as a Cabinet Operator Module(COM). The QNX RTOS is being dedicated by Korea Atomic Energy Research Institute (KAERI). This paper describes the outline and benchmarking test results on Context Switching, Message Passing, Synchronization and Deadline Violation of QNX RTOS under the ELMSYS PC platform

Simulation of Valve Operation for Flow Interrupt Test in Nuclear Power Plant

International Nuclear Information System (INIS)

Kim, Jae Hyung; Shin, Dae Yong; Shin, Dong Woo; Kim, Charn Jung; Lee, Jung Hee

2012-01-01

The valve used in nuclear power plant must be qualified for the function according to the KEPIC MF. The test valve must be selected by shape and size, which is given by KEPIC MF. In the functional test, the mathematical model for the valve operation is needed. The mathematical model must be verified by the test, whose method and procedure is defined in KEPIC MF. The lack of analytical technique has lead to the poor mathematical model, with which the functional test for the big valve is impossible with analytical method. Especially, the tank and rupture disk in the flow test is not considered and the result of the analysis is so different to the real one. In these days, the 3D model for the flow interrupt test makes more accurate analysis. And no facility about functional test reduces the research will for the nuclear power plant valve. For this problem, the test facility for the functional test of the valve and pump in nuclear power plant has been made until 2012. With the test facility, the research project related the valve were initiated in KIMM( Korea Institute of Machinery and Materials). And the joint project to SNU(Seoul National University) has been going on the numerical analysis for the valve in nuclear power plant. Using the commercial software and user subroutine, UDF, the co-simulation with multi-body dynamic and fluid flow analysis and the addition of tank and rupture disk to the user subroutine make possible to simulate the flow interrupt test numerically. This is not simple and regular analysis, which was introduced in user subroutine. In order to simulate the real situation, the engineering work, related mathematical model, and the programming in the user subroutine are needed. This study is on the making the mathematical model for the functional test of the valve in nuclear power plan. The functional test is the real test procedure and defined in KEPIC MF

Survey results of corroding problems at biological treatment plants, Stage II Protection of concrete - State of the Art

Energy Technology Data Exchange (ETDEWEB)

Edwards, Ylva (CBI, Boraas (Sweden)); Henriksson, Gunilla (SP, Boraas (Sweden))

2011-07-01

A pilot study on the degradation and corrosion of concrete in biological treatment plants was conducted in 2009/2010 in a Waste Refinery Project WR-27 'Survey results of corroding problems at biological treatment plants'. The results showed that the concrete does not have sufficient resistance in the current aggressive plant environment. Furthermore, it is stated that some form of surface protection system is needed to ensure the good performance of concrete constructions, and that the system must withstand the aggressive environment and the traffic that occurs on site. Consequently, a new study was proposed in order to develop specifications for surface protection of concrete in aggressive food waste environments. Results from that study are presented in this report. The report includes various types of waterproofing/protection coating for concrete in biological treatment plants. A number of proposals from the industry are presented in the light of results from project WR-27, i.e., the materials must, among other things, withstand the aggressive leachate from waste food at temperatures up to 70 deg C, and some degree of wear. Some systems are compared in terms of technical material properties as reported by the manufacturer. It turns out that different testing methods were used, and the test results are thus generally not directly comparable. A proposal for a test program has been developed, focusing on chemical resistance and wear resistance. A test solution corresponding to leachate is specified. Laboratory tests for verification of the proposed methodology and future requirements are proposed, as well as test sites and follow-up in the field

Test operation of the uranium ore processing pilot plant and uranium conversion plant

International Nuclear Information System (INIS)

Suh, I.S.; Lee, K.I.; Whang, S.T.; Kang, Y.H.; Lee, C.W.; Chu, J.O.; Lee, I.H.; Park, S.C.

1983-01-01

For the guarantee of acid leaching process of the Uranium Ore Processing Pilot Plnat, the KAERI team performed the test operation in coorperation with the COGEMA engineers. The result of the operation was successful achieving the uranium leaching efficiency of 95%. Completing the guarentee test, a continuous test operation was shifted to reconform the reproducibility of the result and check the functions of every units of the pilot plant feeding the low-grade domestic ore, the consistency of the facility was conformed that the uranium can easily be dissolved out form the ore between the temperature range of 60degC-70degC for two hours of leaching with sulfuric acid and could be obtained the leaching efficiency of 92% to 95%. The uranium recovery efficiencies for the processes of extraction and stripping were reached to 99% and 99.6% respectively. As an alternative process for the separation of solid from the ore pulp, four of the Counter Current Decanters were shifted replacing the Belt Filter and those were connected in a series, which were not been tested during the guarantee operation. It was found out that the washing efficiencies of the ore pulp in each tests for the decanters were proportionally increased according to the quantities of the washing water. As a result of the test, it was obtained that washing efficiencies were 95%, 85%, 83% for the water to ore ratio of 3:1, 2:1, 1.5:1 respectively. (Author)

Meta-analysis in plant pathology: synthesizing research results.

Science.gov (United States)

Rosenberg, M S; Garrett, K A; Su, Z; Bowden, R L

2004-09-01

ABSTRACT Meta-analysis is a set of statistical procedures for synthesizing research results from a number of different studies. An estimate of a statistical effect, such as the difference in disease severity for plants with or without a management treatment, is collected from each study along with a measure of the variance of the estimate of the effect. Combining results from different studies will generally result in increased statistical power so that it is easier to detect small effects. Combining results from different studies may also make it possible to compare the size of the effect as a function of other predictor variables such as geographic region or pathogen species. We present a review of the basic methodology for meta-analysis. We also present an example of meta-analysis of the relationship between disease severity and yield loss for foliar wheat diseases, based on data collected from a decade of fungicide and nematicide test results.

Mercury exposure on potential plant Ludwigia octovalvis L. - Preliminary toxicological testing

Science.gov (United States)

Alrawiq, Huda S. M.; Mushrifah, I.

2013-11-01

The preliminary test in phytoremediation is necessaryto determine the ability of plant to survive in media with different concentrations of contaminant. It was conducted to determine the maximum concentration of the contaminant that isharmful to the plant and suppress the plant growth. This study showed the ability of Ludwigia octovalvisto resist mercury (Hg) contaminant in sand containing different concentrations of Hg (0, 0.5, 1, 2, 4, 6 and 8 mg/L). The experimental work wasperformed under greenhouse conditions for an observation period of 4 weeks. Throughout the 4 weeks duration, the resultsshowed that 66.66% of the plants withered for on exposure to Hg concentration of 4 mg/L and 100% withered at higher concentrations of 6 and 8 mg/L. The results of this study may serve as a basis for research that aims to study uptake and accumulation of Hg using potential phytoremediation plants.

RESULTS OF THE FIRST MI-171A2 FLYING LABORATORY TEST PHASE

Directory of Open Access Journals (Sweden)

V. A. Ivchin

2014-01-01

Full Text Available The present publication describes the results of the first stage of the flying laboratory (Mi-171 helicopter flight tests performed at Mil Moscow Helicopter Plant, JSC facilities. Main rotor components with blades made of polymer composite materials and X-type tail rotor were tested on the Mi-171 № 14987, flying laboratory, under Mi-171A Helicopter Retrofit Program.

Limits to the Recognizability of Flaws in Non-Destructive Testing Steam-Generator Tubes for Nuclear-Power Plants

International Nuclear Information System (INIS)

Kuhlmann, A.; Adamsky, F.-J.

1965-01-01

In the Federal Republic of Germany there are nuclear reactors under construction with steam generators inside the reactor pressure-vessel. As a result design repairs of steam- generator tubes are very difficult and cause large shut-down times of the nuclear-power plant. It is known that numerous troubles in operating conventional power plants are results of steam-generator tube damages. Because of the high total costs of these reactors it. is necessary to construct the steam generators especially in such a manner that the load factor of the power plant is as high as possible. The Technischer Überwachungs-Verein Rheinland was charged to supervise and to test fabrication and construction of the steam generators to see that this part of the plant was as free of defects as possible. The experience gained during this work is of interest for manufacture and construction of steam generators for nuclear-power plants in general. This paper deals with the efficiency limits of non-destructive testing steam-generator tubes. The following tests performed will be discussed in detail: (a) Automatic ultrasonic testing of the straight tubes in the production facility; (b) Combined ultrasonic and radiographic testing of the bent tubes and tube weldings; (c) Other non-destructive tests. (author) [fr

TERRESTRIAL PLANT REPRODUCTIVE TESTING: SHOULD WILDLIFE TOXICOLOGISTS CARE?

Science.gov (United States)

Standard phytotoxicity testing using the seedling emergence and vegetative vigor tests have been shown to be inadequate for the protection of plant reproduction. Both experimental evidence and unintended field exposures have shown vegetation can be minimally or not significantly...

The use of NPAR results in plant inspection activities

International Nuclear Information System (INIS)

Gunther, W.; Taylor, J.

1990-01-01

The US Nuclear Regulatory Commission's (NRC's) Nuclear Plant Aging Research (NPAR) Program is a hardware oriented research program which has produced a large data base of equipment and system operating, maintenance, and testing information. Equipment and systems which have a propensity for age related degradation are identified, and methods for detecting and mitigating aging effects have been evaluated. As plants age, it becomes increasingly important that NRC inspectors be cognizant of plant aging phenomena. This paper describes the NPAR information which can enhance inspection activities, and provides a mechanism for making pertinent research available to the inspectors

ENERGY EFFICIENCY OF DIESEL LOCOMOTIVE HYDRAULIC TRANSMISSION TESTS AT LOCOMOTIVE REPAIR PLANT

Directory of Open Access Journals (Sweden)

B. E. Bodnar

2015-10-01

Full Text Available Purpose. In difficult economic conditions, cost reduction of electricity consumption for the needs of production is an urgent task for the country’s industrial enterprises. Technical specifications of enterprises, which repair diesel locomotive hydraulic transmission, recommend conducting a certain amount of evaluation and regulatory tests to monitor their condition after repair. Experience shows that a significant portion of hydraulic transmission defects is revealed by bench tests. The advantages of bench tests include the ability to detect defects after repair, ease of maintenance of the hydraulic transmission and relatively low labour intensity for eliminating defects. The quality of these tests results in the transmission resource and its efficiency. Improvement of the technology of plant post-repairs hydraulic tests in order to reduce electricity consumption while testing. Methodology. The possible options for hydraulic transmission test bench improvement were analysed. There was proposed an energy efficiency method for diesel locomotive hydraulic transmission testing in locomotive repair plant environment. This is achieved by installing additional drive motor which receives power from the load generator. Findings. Based on the conducted analysis the necessity of improving the plant stand testing of hydraulic transmission was proved. The variants of the stand modernization were examined. The test stand modernization analysis was conducted. Originality. The possibility of using electric power load generator to power the stand electric drive motor or the additional drive motor was theoretically substantiated. Practical value. A variant of hydraulic transmission test stand based on the mutual load method was proposed. Using this method increases the hydraulic transmission load range and power consumption by stand remains unchanged. The additional drive motor will increase the speed of the input shaft that in its turn wil allow testing in

Results from four Pinus patula water planting trials in the summer ...

African Journals Online (AJOL)

Planting with water is used by some forestry companies in South Africa to reduce post-planting water stress. Four trials were implemented to test the response in survival of Pinus patula to water applied at planting. Two trials each were situated in the KwaZulu-Natal Midlands and Mpumalanga escarpment. The first trial at ...

Pipe rupture test results; 6 in. pipe whip test under BWR LOCA conditions

International Nuclear Information System (INIS)

Kurihara, Ryoichi; Yano, Toshikazu; Ueda, Shuzo; Isozaki, Toshikuni; Miyazaki, Noriyuki; Kato, Rokuro; Miyazono, Shohachiro

1983-02-01

A series of pipe rupture tests has been performed in JAERI to demonstrate the safety of the primary coolant circuits in the event of pipe rupture, in nuclear power plants. The present report summarizes the results of 6 in. pipe whip tests (RUN 5605, 5606), under BWR LOCA conditions (285 0 C, 6.8 MPa), which were performed in August, 1981. The test pipe is made of Type 304 stainless steel and its outer diameter is 6 in. and its thickness is 11.1 mm. The restraints are made of Type 304 stainless steel and its diameter is 16.0 mm. Two restraints were set on the restraint support with clearance of 100 mm. Overhang length was varied as the parameter in these tests and was 300 mm or 700 mm. The following results are obtained. (1) The deformations of a pipe and restraints are limited effectively by shorter overhang length of 300. However, they become larger when the overhang length is 700 mm, and the pipe deforms especially at the setting point of restraints. (2) Velocity at the free end of pipe becomes about 30 m/sec just after the break. However, velocity at the setting point of restraint becomes about only 4 m/sec just after the break. (3) It seems from the comparison between the 4 in. tests and 6 in. tests that the maximum restraint force of 6 in. tests is about two times as large as that of 4 in. tests. (author)

Field Lysimeter Test Facility: Second year (FY 1989) test results

International Nuclear Information System (INIS)

Campbell, M.D.; Gee, G.W.; Kanyid, M.J.; Rockhold, M.L.

1990-04-01

The Record of Decision associated with the Hanford Defense Waste Environmental Impact Statement (53 FR 12449-53) commits to an evaluation of the use of protective barriers placed over near-surface wastes. The barrier must protect against wind and water erosion and limit plant and animal intrusion and infiltration of water. Successful conclusion of this program will yield the necessary protective barrier design for near-surface waste isolation. This report presents results from the second year of tests at the FLTF. The primary objective of testing protective barriers at the FLTF was to measure the water budgets within the various barriers and assess the effectiveness of their designs in limiting water intrusion into the zone beneath each barrier. Information obtained from these measurements is intended for use in refining barrier designs. Four elements of water budget were measured during the year: precipitation, evaporation, storage, and drainage. Run-off, which is a fifth element of a complete water budget, was made negligible by a lip on the lysimeters that protrudes 5 cm above the soil surface to prevent run-off. A secondary objective of testing protective barriers at the FLTF was to refine procedures and equipment to support data collection for verification of the computer model needed for long-term projections of barrier performance. 6 refs

Operating results 2015. Nuclear power plants. Pt. 1

Energy Technology Data Exchange (ETDEWEB)

Anon.

2016-05-15

A report is given on the opening results achieved in 2015, events important to plant safety, special and relevant repair, and retrofit measures from nuclear power plants in Germany. Reports about nuclear power plants in Belgium, Finland, the Netherlands, Switzerland, and Spain will be published in further issue.

Full-scale dynamic structural testing of Paks nuclear power plant

International Nuclear Information System (INIS)

Da Rin, E.M.; Muzzi, F.P.

1995-01-01

Within the framework of the IAEA coordinated 'Benchmark Study for the seismic analysis and testing of WWER-type NPPs', in-situ dynamic structural testing activities have been performed at the Paks Nuclear Power Plant in Hungary. The specific objective of the investigation was to obtain experimental data on the actual dynamic structural behaviour of the plant's major constructions and equipment under normal operating conditions, for enabling a valid seismic safety review to be made. This paper gives a synthetic description of the conducted experiments and presents some results, regarding in particular the free-field excitations produced during the earthquake-simulation experiments and an experiment of the dynamic soil-structure interaction global effects at the base of the reactor containment structure. Moreover, a method which can be used for inferring dynamic structural characteristics from the recorded time-histories is briefly described and a simple illustrative example given. (author)

Identification of Radioactive Pilot-Plant test requirements

Energy Technology Data Exchange (ETDEWEB)

Powell, W.J.; Riebling, E.F.

1995-05-09

Radioactive Pilot-Plant testing needs and alternatives are evaluated for enhanced Sludge Washing and High and Low-Level Vitrification efforts. Also investigated was instrument and equipment testing needs associated with the vitrification and retrieval process. The scope of this document is to record the existing March 1994 letter report for future use. A structured Kepner-Trego{trademark} decision analysis process was used to assist analysis of the testing needs. This analysis provided various combinations of laboratory and radioactive (hot) and cold pilot testing options associated with the above need areas. Recommendations for testing requirements were made.

Identification of Radioactive Pilot-Plant test requirements

International Nuclear Information System (INIS)

Powell, W.J.; Riebling, E.F.

1995-01-01

Radioactive Pilot-Plant testing needs and alternatives are evaluated for enhanced Sludge Washing and High and Low-Level Vitrification efforts. Also investigated was instrument and equipment testing needs associated with the vitrification and retrieval process. The scope of this document is to record the existing March 1994 letter report for future use. A structured Kepner-Trego trademark decision analysis process was used to assist analysis of the testing needs. This analysis provided various combinations of laboratory and radioactive (hot) and cold pilot testing options associated with the above need areas. Recommendations for testing requirements were made

Monitoring of biogas test plants

DEFF Research Database (Denmark)

Holm-Nielsen, Jens Bo; Esbensen, Kim H.

2011-01-01

realistic bioreactor scales, it is necessary to obtain a fairly constant level of volatile fatty acid (VFA) concentration, which furthers a stable biogas production. Uncontrolled VFA contents have a significant negative impact on biogas production; VFA concentrations should not exceed 5–6000 mg/L lest......Most studies reported in the literature have investigated near infrared spectroscopy (NIR) in laboratory-scale or minor pilot biogas plants only; practically no other studies have examined the potential for meso-scale/full-scale on-line process monitoring. The focus of this study is on a meso......-scale biogas test plant implementation of process analytical technologies (PAT) to develop multivariate calibration/prediction models for anaerobic digestion (AD) processes. A 150 L bioreactor was fitted with a recurrent loop at which NIR spectroscopy and attendant reference sampling were carried out. In all...

Experimental results from pressure testing a 1:6-scale nuclear power plant containment

International Nuclear Information System (INIS)

Horschel, D.S.

1992-01-01

This report discusses the testing of a 1:6-scale, reinforced-concrete containment building at Sandia National Laboratories, in Albuquerque, New Mexico. The scale-model, Light Water Reactor (LWR) containment building was designed and built to the American Society of Mechanical Engineers (ASME) code by United Engineers and Constructors, Inc., and was instrumented with over 1200 transducers to prepare for the test. The containment model was tested to failure to determine its response to static internal overpressurization. As part of the US Nuclear Regulatory Commission's program on containment integrity, the test results will be used to assess the capability of analytical methods to predict the performance of containments under severe-accident loads. The scaled dimensions of the cylindrical wall and hemispherical dome were typical of a full-size containment. Other typical features included in the heavily reinforced model were equipment hatches, personnel air locks, several small piping penetrations, and a ihin steel liner that was attached to the concrete by headed studs. In addition to the transducers attached to the model, an acoustic detection system and several video and still cameras were used during testing to gather data and to aid in the conduct of the test. The model and its instrumentation are briefly discussed, and is followed by the testing procedures and measured response of the containment model. A summary discussion is included to aid in understanding the significance of the test as it applies to real world reinforced concrete containment structures. The data gathered during SIT and overpressure testing are included as an appendix

Experimental results from pressure testing a 1:6-scale nuclear power plant containment

Energy Technology Data Exchange (ETDEWEB)

Horschel, D.S. [Sandia National Labs., Albuquerque, NM (United States)

1992-01-01

This report discusses the testing of a 1:6-scale, reinforced-concrete containment building at Sandia National Laboratories, in Albuquerque, New Mexico. The scale-model, Light Water Reactor (LWR) containment building was designed and built to the American Society of Mechanical Engineers (ASME) code by United Engineers and Constructors, Inc., and was instrumented with over 1200 transducers to prepare for the test. The containment model was tested to failure to determine its response to static internal overpressurization. As part of the US Nuclear Regulatory Commission`s program on containment integrity, the test results will be used to assess the capability of analytical methods to predict the performance of containments under severe-accident loads. The scaled dimensions of the cylindrical wall and hemispherical dome were typical of a full-size containment. Other typical features included in the heavily reinforced model were equipment hatches, personnel air locks, several small piping penetrations, and a ihin steel liner that was attached to the concrete by headed studs. In addition to the transducers attached to the model, an acoustic detection system and several video and still cameras were used during testing to gather data and to aid in the conduct of the test. The model and its instrumentation are briefly discussed, and is followed by the testing procedures and measured response of the containment model. A summary discussion is included to aid in understanding the significance of the test as it applies to real world reinforced concrete containment structures. The data gathered during SIT and overpressure testing are included as an appendix.

Experiences with drug testing at a nuclear power plant

International Nuclear Information System (INIS)

Ray, H.B.

1987-01-01

After more than 2 yr of operation of a drug testing program at the San Onofre nuclear power plant site, the Southern California Edison Co. has had a number of experiences and lessons considered valuable. The drug testing program at San Onofre, implemented in September of 1984, continues in essentially the same form today. Prior to describing the program, the paper reviews several underlying issues that believed to be simultaneously satisfied by the program: trustworthiness, fitness and safety, public trust, and privacy and search. The overall drug testing program, periodic drug monitoring program, and unannounced drug testing program are described. In addition to the obvious features of a good drug testing program, which are described in the EEI guide, it is essential to consider such issues as the stated program rationale, employee relations, and disciplinary action measures when contemplating or engaging in drug testing at nuclear power plants

Testing of a naturally aged nuclear power plant inverter and battery charger

International Nuclear Information System (INIS)

Gunther, W.E.

1988-09-01

A naturally aged inverter and battery charger were obtained from the Shippingport facility. This equipment was manufactured in 1974, and was installed at Shippingport in 1975 as part of a major plant modification. Testing was performed on this equipment under the auspices of the NRC's Nuclear Plant Aging Research (NPAR) Program to evaluate the type and extent of degradation due to aging, and to determine the effectiveness of condition monitoring techniques which could be used to detect aging effects. Steady state testing was conducted over the equipment's entire operating range. Step load changes were also initiated in order to monitor the electrical response. During this testing, component temperatures were monitored and circuit waveforms analyzed. Results indicated that aging had not substantially affected equipment operation. On the other hand, when compared with original acceptance test data, the monitoring techniques employed were sensitive to changes in measurable component and equipment parameters indicating the viability of detecting degradation prior to catastrophic failure. 7 refs., 34 figs., 12 tabs

Full-scale impact test data for tornado-missile design of nuclear plants

International Nuclear Information System (INIS)

Stephenson, A.E.; Sliter, G.E.

1977-01-01

It is standard practice to consider the effects of low-probability impacts of tornado-borne debris (''tornado missiles'' such as utility poles and steel pipes) in the structural design of nuclear power plants in the United States. To provide data that can be used directly in the design procedure, a series of full-scale tornado-missile impact tests was performed. This paper is a brief summary of the results and conclusions from these tests. The tests consisted of reinforced concrete panels impacted by poles, pipes, and rods propelled by a rocket sled. The panels were constructed to current minimum standards and had thicknesses typical of auxiliary buildings of nuclear power plants. A specific objective was the determination of the impact velocities below which the panels do not experience backface scabbing. Another objective was to assess the adequacy of (1) conventional design formulae for penetration and scabbing and (2) conventional design methods for overall structural response. Test missiles and velocities represented those in current design standards. Missiles included utility poles, steel pipes, and steel bars. It is important to interpret the data in this paper in recognition that the test conditions represent conservative assumptions regarding maximum wind speeds, injection of the missile into the wind stream, aerodynamic trajectory, and orientation of missile at impact. Even with the severe assumptions made, the full-scale tests described demonstrate the ability of prototypical nuclear plant walls and roofs to provide adequate protection against postulated tornado-missile impact

Low-flow operation and testing of pumps in nuclear plants

International Nuclear Information System (INIS)

Greenstreet, W.L.

1989-01-01

Low-flow operation of centrifugal pumps introduces hydraulic instability and other factors that can cause damage to these machines. The resulting degradation has been studied and recorded for pumps in electric power plants. The objectives of this paper are to (1) describe the damage-producing phenomena, including their sources and consequences; (2) relate these observations to expectations for damage caused by low-flow operation of pumps in nuclear power plants; and (3) assess the utility of low-flow testing. Hydraulic behavior during low-flow operation is reviewed for a typical centrifugal pump stage, and the damage-producing mechanisms are described. Pump monitoring practices, in conjunction with pump performance characteristics, are considered; experience data are reviewed; and the effectiveness of low-flow surveillance monitoring is examined. Degradation caused by low-flow operation is shown to be an important factor, and low-flow surveillance testing is shown to be inadequate. 18 refs., 5 figs., 4 tabs

MAAP4 CANDU analysis of a generic CANDU-6 plant: preliminary results

Energy Technology Data Exchange (ETDEWEB)

Petoukhov, S.M.; Mathew, P.M

2001-10-01

To support the generic probabilistic safety analysis (PSA) program at AECL, in particular to conduct Level 2 PSA analysis of a CANDU 6 plant undergoing a postulated severe accident, the capability to conduct severe accident consequence analysis for a CANDU plant is required. For this purpose, AECL selected MAAP4 CANDU from a number of other severe accident codes. The necessary models for a generic CANDU 6 station have been implemented in the code, and the code version 0.2 beta was tested using station data, which were assembled for a generic CANDU 6 station. This paper describes the preliminary results of the consequence analysis using MAAP4 CANDU for a generic CANDU 6 station, when it undergoes a station blackout and a large loss-of-coolant accident scenario. The analysis results show that the plant response is consistent with the physical phenomena modeled and the failure criteria used. The results also confirm that the CANDU design is robust with respect to severe accidents, which is reflected in the calculated long times that are available for administering accident management measures to arrest the accident progression before the calandria vessel or containment become at risk. (author)

Cooperative biogas plants. Economic results and analyses. Status report 1998

International Nuclear Information System (INIS)

Hjort-Gregersen, K.

1998-11-01

The years 1995 - 1998 have been characterised by stabilisation of operation and economy of the Danish co-operative biogas plants. Most of the plants have obtained increasingly better economic results although the increase has been less significant than during earlier periods. There are several reasons for the increase. Most of the plants have been able to increase the sales income because of larger amounts of biomass available resulting in an increased biogas production. Furthermore it has been possible to contain the income level for biomass receipt. Several plants have established gas collection in storage tanks, which has resulted in increased gas yield. The operational stability related to both technique and processes have improved. The operational costs have been stabilised and are under control at most of the plants. The improved economic results have resulted in most of the plants having a satisfactory operation and economy. However, it must be stressed that some of the oldest plants have not been able to settle the investment dept at normal conditions. Also some, even rather new plants, still are in a difficult economic situation. Most of the plants established in the 90'ies have had a good start both operationally and economically. Thus the economic risk of establishing a plant has been reduced compared to earlier years. Generally the prerequisites for establishing a biogas plant are favourable economic conditions and quality assurance of the project. (LN)

Pyrolysis of rubber gloves in integral pyrolysis test plant

International Nuclear Information System (INIS)

Norasalwa Zakaria; Mohd Noor Muhd Yunus; Mohd Annuar Assadat Husain; Farid Nasir Ani

2010-01-01

Previously, pyrolysis of rubber gloves in laboratory study was described. In order to visualize the practical application of rubber gloves pyrolysis in terms of treating rubber gloves in medical waste, a new test plant was designed and constructed. The semi-continuous test plant was designed to accommodate rubber gloves that were not cut or shredded. The test plant has a capacity of 2kg/ hr and employed auxiliary fuel instead of the conventional electrical power for heating. The concept was based on moving bed reactor, but additional feature of sand jacket feature was also introduced in the design. Pyrolysis of the gloves was conducted at three temperatures, namely 350 degree Celsius, 400 degree Celsius and 450 degree Celsius. Oxygen presents inside of the reactor due to the combined effect of imperfect sealing and suction effect. This study addresses the performance of this test plant covering the time temperature profile, gas evolution profile and product yield. Comparison between the yield of the liquid, gas and char pyrolyzate was made against the laboratory study. It was found that the oil yield was less than the one obtained from bench scale study. Water formation was more pronounced. The presence of the oxygen also altered the tail gas composition but eliminate the sticky nature of solid residue, making it easier to handle. The chemical composition of the oil was determined and the main compounds in the oil were esters and phtalic acid. (author)

Field vibration test of principal equipment of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Shiraki, Kazuhiro; Fujita, Katsuhisa; Kajimura, Motohiko; Ikegami, Yasuhiko; Hanzawa, Katsumi; Sakai, Yoshiyuki; Kokubo, Eiji; Igarashi, Shigeru

1984-09-01

Japan is one of the most earthquake-stricken countries in the world, and demands for aseismic design have become severer recently. In a nuclear power plant in particular, consisting of a reactor vessel and other facilities dealing with a radioactive substance in some form or other, it is essential from the standpoint of safety to eliminate any possibility of radioactive hazards for the local public, and the employees at the plant as well, if these facilities are struck by an earthquake. This paper is related to the reactor vessel, reactor primary cooling equipment and piping system and important general piping as examples of important facilities of a nuclear power plant, and discusses vibration tests of an actual plant in the field from the standpoint of enhancing the aseismic safety of the Mitsubishi PWR nuclear power plant. Especially concerning vibration test technology, the effects in the evaluation of aseismic safety and its limits are studied to prove how it contributes to the enhancement of the reliability of aseismic design of nuclear power plants.

78 FR 25488 - Qualification Tests for Safety-Related Actuators in Nuclear Power Plants

Science.gov (United States)

2013-05-01

... Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; request for... regulatory guide (DG), DG-1235, ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants... entitled ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants'' is temporarily...

Large-Scale Spray Releases: Additional Aerosol Test Results

Energy Technology Data Exchange (ETDEWEB)

Daniel, Richard C.; Gauglitz, Phillip A.; Burns, Carolyn A.; Fountain, Matthew S.; Shimskey, Rick W.; Billing, Justin M.; Bontha, Jagannadha R.; Kurath, Dean E.; Jenks, Jeromy WJ; MacFarlan, Paul J.; Mahoney, Lenna A.

2013-08-01

One of the events postulated in the hazard analysis for the Waste Treatment and Immobilization Plant (WTP) and other U.S. Department of Energy (DOE) nuclear facilities is a breach in process piping that produces aerosols with droplet sizes in the respirable range. The current approach for predicting the size and concentration of aerosols produced in a spray leak event involves extrapolating from correlations reported in the literature. These correlations are based on results obtained from small engineered spray nozzles using pure liquids that behave as a Newtonian fluid. The narrow ranges of physical properties on which the correlations are based do not cover the wide range of slurries and viscous materials that will be processed in the WTP and in processing facilities across the DOE complex. To expand the data set upon which the WTP accident and safety analyses were based, an aerosol spray leak testing program was conducted by Pacific Northwest National Laboratory (PNNL). PNNL’s test program addressed two key technical areas to improve the WTP methodology (Larson and Allen 2010). The first technical area was to quantify the role of slurry particles in small breaches where slurry particles may plug the hole and prevent high-pressure sprays. The results from an effort to address this first technical area can be found in Mahoney et al. (2012a). The second technical area was to determine aerosol droplet size distribution and total droplet volume from prototypic breaches and fluids, including sprays from larger breaches and sprays of slurries for which literature data are mostly absent. To address the second technical area, the testing program collected aerosol generation data at two scales, commonly referred to as small-scale and large-scale testing. The small-scale testing and resultant data are described in Mahoney et al. (2012b), and the large-scale testing and resultant data are presented in Schonewill et al. (2012). In tests at both scales, simulants were used

FY 1992 report on the results of the development of an entrained bed coal gasification power plant. Part 3. Operation test of pilot plant (2/2); 1992 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu - Sono 3. Pilot plant unten shiken hen (2/2)

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-02-01

The study of operation test was made of the 200 t/d entrained bed coal gasification pilot plant that was constructed with the aim of establishing technology of the integrated coal gasification combined cycle power generation, and the details of the FY 1992 results were summarized. As to the test of gas turbine facilities, at RUN 10, the scheduled test on coal gas mixed combustion continuous operation/coal gas fired operation was not carried out because of the worsening of the state of gasifier operation. The operation was just made for a short time. At RUN 11, it was confirmed that the motion of equipment in the bleeding system was good at the time of the bleeding cooperation test and there was no anomaly also in the state of gas turbine operation. At RUN 12, it was confirmed that the motion of load/pressure control functions was normal in the state of GT load of about 5MW. At RUN 13, it was confirmed that the control function of GT governor was confirmed in the total pressure control test. As a result of the operation test of these, items of improvement were extracted for actual pressure/actual size combustor facilities, safety environment facilities and electric/control facilities. The measures taken for each item were studied. (NEDO)

Experimental results of wind powered pumping plant with electrical transmission

International Nuclear Information System (INIS)

Falchetta, M.; Prischich, D.; Benedetti, A.; Cara, G.

1992-01-01

A demonstrative application of deep well pumping system employing a wind powered pumping plant with an electric transmission was set-up and tested for two years at the test field of the Casaccia center of ENEA (Italian Agency for Energy, New Technologies and the Environment), near Rome. The tests permitted the evaluation of the practical performance, advantages and drawbacks of a wind pumping plant of this type, in order to permit a design optimization and a proper choice of components and of control strategies for future commercial applications. The main point of investigation was the evaluation of the effectiveness of a control scheme based on a 'permanent link' between electric generator and electric motor, avoiding any electronics and switching components, and leading to a very robust and reliable means of transferring energy to the pump at variable speed, and at low cost

Plant Atrium System for Food Production in NASA's Deep Space Habitat Tests

Science.gov (United States)

Massa, Gioia D.; Simpson, Morgan; Wheeler, Raymond M.; Newsham, Gerald; Stutte, Gary W.

2013-01-01

In preparation for future human exploration missions to space, NASA evaluates habitat concepts to assess integration issues, power requirements, crew operations, technology, and system performance. The concept of a Food Production System utilizes fresh foods, such as vegetables and small fruits, harvested on a continuous basis, to improve the crew's diet and quality of life. The system would need to fit conveniently into the habitat and not interfere with other components or operations. To test this concept, a plant growing "atrium" was designed to surround the lift between the lower and upper modules of the Deep Space Habitat and deployed at NASA Desert Research and Technology Studies (DRATS) test site in 2011 and at NASA Johnson Space Center in 2012. With this approach, no-utilized volume provided an area for vegetable growth. For the 2011 test, mizuna, lettuce, basil, radish and sweetpotato plants were grown in trays using commercially available red I blue LED light fixtures. Seedlings were transplanted into the atrium and cared for by the. crew. Plants were then harvested two weeks later following completion of the test. In 2012, mizuna, lettuce, and radish plants were grown similarly but under flat panel banks of white LEDs. In 2012, the crew went through plant harvesting, including sanitizing tlie leafy greens and radishes, which were then consumed. Each test demonstrated successful production of vegetables within a functional hab module. The round red I blue LEDs for the 2011 test lighting cast a purple light in the hab, and were less uniformly distributed over the plant trays. The white LED panels provided broad spectrum light with more uniform distribution. Post-test questionnaires showed that the crew enjoyed tending and consuming the plants and that the white LED light in 2012 provided welcome extra light for the main HAB AREA.

Possibilities of avoidance and control of bacterial plant diseases when using pathogen-tested (certified) or - treated planting material

NARCIS (Netherlands)

Janse, J.; Wenneker, M.

2002-01-01

Testing of planting material for freedom from phytopathogenic bacteria is an important, although not exclusive, method for control of bacterial diseases of plants. Ideally, pathogen-free or pathogen-/disease-resistant planting material is desirable, but this situation is not always possible on a

UF/sub 6/ test loop for evaluation and implementation of international enrichment plant safeguards

International Nuclear Information System (INIS)

Cooley, J.N.; Fields, L.W.; Swindle, D.W. Jr.

1987-01-01

A functional test loop capable of simulating UF/sub 6/ flows, pressures, and pipe deposits characteristic of gas centrifuge enrichment plant piping has been designed and fabricated by the Enrichment Safeguards Program of Martin Marietta Energy Systems, Inc., for use by the International Atomic Energy Agency (IAEA) at its Safeguards Analytical Laboratory in Seibersdorf, Austria. The purpose of the test loop is twofold: (1) to enable the IAEA to evaluate and to calibrate enrichment safeguards measurement instrumentation to be used in limited frequency-unannounced access (LFUA) inspection strategy measurements at gas centrifuge enrichment plants and (2) to train IAEA inspectors in the use of such instrumentation. The test loop incorporates actual sections of cascade header pipes from the centrifuge enrichment plants subject to IAEA inspections. The test loop is described, applications for its use by the IAEA are detailed, and results from an initial demonstration session using the test loop are summarized. By giving the IAEA the in-house capability to evaluate LFUA inspection strategy approaches, to develop inspection procedures, to calibrate instrumentation, and to train inspectors, the UF/sub 6/ cascade header pipe test loop will contribute to the IAEA's success in implementing LFUA strategy inspections at gas centrifuge enrichment facilities subject to international safeguards inspections

Full scale vibration test on nuclear power plant auxiliary building: Part I

International Nuclear Information System (INIS)

Langer, V.; Tinic, S.; Berger, E.; Zwicky, P.; Prater, E.G.

1987-01-01

In connection with the construction of the reinforced concrete auxiliary building housing the two boric water tanks (so-called BOTA building) of the Beznau Nuclear Power Plant in Switzerland the opportunity was given to carry out full scale vibration tests in November 1985. The overall aim of the tests was to validate computational models and parameters widely used in the seismic analysis of the structures and critical components of nuclear power plants. The scope of the experimental investigation was the determination of the eigenfrequencies and damping values for the fundamental soil-structure interaction (SSI) modes. The excitation level was aimed to be as high as feasibly possible. A working group was formed of representatives of the owner, NOK, the consulting firm Basler and Hofmann and the ETH to supervise the project. The project's main phases were the planning and execution of the tests, the evaluation of recorded data, numerical simulation of the tests using different computer models and finally the comparison and interpretation of measured and computed results

Shaking table test study on seismic performance of dehydrogenation fan for nuclear power plants

International Nuclear Information System (INIS)

Liu Kaiyan; Shi Weixing; Cao Jialiang; Wang Yang

2011-01-01

Seismic performance of the dehydrogenation fan for nuclear power plants was evaluated based on the shaking table test of earthquake simulation. Dynamic characteristics including the orthogonal tri-axial fundamental frequencies and equivalent damping ratios were measured by the white noise scanning method. Artificial seismic waves were generated corresponding to the floor acceleration response spectra for nuclear power plants. Furthermore, five OBE and one SSE shaking table tests for dehydrogenation fan were performed by using the artificial seismic waves as the seismic inputs along the orthogonal axis simultaneity. Operating function of dehydrogenation fan was monitored and observed during all seismic tests, and performance indexes of dehydrogenation fan were compared before and after seismic tests. The results show that the structural integrity and operating function of the dehydrogenation fan are perfect during all seismic tests; and the performance indexes of the dehydrogenation fan can remain consistent before and after seismic tests; the seismic performance of the dehydrogenation fan can satisfy relevant technical requirements. (authors)

Safety demonstration tests of postulated solvent fire accidents in extraction process of a fuel reprocessing plant, (2)

International Nuclear Information System (INIS)

Tukamoto, Michio; Takada, Junichi; Koike, Tadao; Nishio, Gunji; Uno, Seiichiro; Kamoshida, Atsusi; Watanabe, Hironori; Hashimoto, Kazuichiro; Kitani, Susumu.

1992-03-01

Demonstration tests of hypothetical solvent fire in an extraction process of the reprocessing plant were carried out from 1984 to 1985 in JAERI, focusing on the confinement of radioactive materials during the fire by a large-scale fire facility (FFF) to evaluate the safety of air-ventilation system in the plant. Fire data from the demonstration test were obtained by focusing on fire behavior at cells and ducts in the ventilation system, smoke generation during the fire, transport and deposition of smoke containing simulated radioactive species in the ventilation system, confinement of radioactive materials, and integrity of HEPA filters by using the FFF simulating an air-ventilation system of the reference reprocessing plant in Japan. The present report is published in a series of the report Phase I (JAERI-M 91-145) of the demonstration test. Test results in the report will be used for the verification of a computer code FACE to evaluate the safety of postulated fire accidents in the reprocessing plant. (author)

Small-Scale Spray Releases: Additional Aerosol Test Results

Energy Technology Data Exchange (ETDEWEB)

Schonewill, Philip P.; Gauglitz, Phillip A.; Kimura, Marcia L.; Brown, G. N.; Mahoney, Lenna A.; Tran, Diana N.; Burns, Carolyn A.; Kurath, Dean E.

2013-08-01

One of the events postulated in the hazard analysis at the Waste Treatment and Immobilization Plant (WTP) and other U.S. Department of Energy (DOE) nuclear facilities is a breach in process piping that produces aerosols with droplet sizes in the respirable range. The current approach for predicting the size and concentration of aerosols produced in a spray leak involves extrapolating from correlations reported in the literature. These correlations are based on results obtained from small engineered spray nozzles using pure liquids with Newtonian fluid behavior. The narrow ranges of physical properties on which the correlations are based do not cover the wide range of slurries and viscous materials that will be processed in the WTP and across processing facilities in the DOE complex. To expand the data set upon which the WTP accident and safety analyses were based, an aerosol spray leak testing program was conducted by Pacific Northwest National Laboratory (PNNL). PNNL’s test program addressed two key technical areas to improve the WTP methodology (Larson and Allen 2010). The first technical area was to quantify the role of slurry particles in small breaches where slurry particles may plug the hole and prevent high-pressure sprays. The results from an effort to address this first technical area can be found in Mahoney et al. (2012a). The second technical area was to determine aerosol droplet size distribution and total droplet volume from prototypic breaches and fluids, including sprays from larger breaches and sprays of slurries for which literature data are largely absent. To address the second technical area, the testing program collected aerosol generation data at two scales, commonly referred to as small-scale and large-scale. The small-scale testing and resultant data are described in Mahoney et al. (2012b) and the large-scale testing and resultant data are presented in Schonewill et al. (2012). In tests at both scales, simulants were used to mimic the

PIE of test assembly of Qinshan nuclear power plant

International Nuclear Information System (INIS)

Ran, M.; Yan, J.; Wang, S.

2000-01-01

The small dimensional test fuel assembly (3x3-2) for the Qinshan Nuclear Power Plant was irradiated up to 25.7 Gwd/tU in the in-pile loop (15.5 Mpa,320 C) in Heavy Water Research Reactor (HWRR), CIAE, at simulative condition to Qinshan PWR normal and short time overpower operation for verifying the design, technology, and material properties of the fuel assembly. Comprehensive post-irradiation examination (PIE) including dimension measurement, gamma scanning, eddy current test, X ray, radiography, measurement of fission gas release, and quantitative metallography etc. were performed. PIE results show that the diameter of the fuel rods changed, ridges appeared on the cladding, pellets swelled, and the rate of fission gas release was higher than what we expected. The results would be an important basis for further improvement of design, technology and material properties for Qinshan PWR assembly. (author)

The scheme optimization and management innovation for the first containment integrated in-service test of nuclear power plant

International Nuclear Information System (INIS)

Wang Haiwei; Yang Gang

2014-01-01

The containment integrated test is a large-scale, high risk and very difficult test in pressurized water reactor nuclear power plants. By simulating peak pressure inside the containment in DESIGN-BASIS accident conditions, measuring the total leakage rate of the containment with the peak pressure, and implementing the structure inspection test on several pressure levels, the containment's performance can be verified. Containment integrated test is an important witness point supervised by NNSA. The test results crucially decide the reactor to be started or not. The containment integrated test in 301 overhaul is the first in-service test of Unit 3. By the experience of the same 6 former tests in Qinshan Second Nuclear Power Plant and the feedback from other plants, the test scheme get more scientific and the organization management more standardized. This article discusses the containment integrated test in 301 overhaul and summarizes the experience to provide some references for the following containment integrated tests in the future. (authors)

Evaluation of the control system checkout test results for YGN 3

International Nuclear Information System (INIS)

Hong, Eon Young; Shon, Suk Whun; Kim, Shim Whan; Sung, Kang Sik; Seo, Jong Tae.

1996-11-01

During the Yonggwang Nuclear Power Plant Unit 3 (YGN3) Power Ascension Test (PAT) period, the Control System Checkout tests were performed at 10%, 20%, 50%, 80%, and 100% respectively. This test evaluates the performance of the feedwater control system, reactor regulating system, pressurizer level and pressure control system in controlling their respective parameters within specified control bands at different power levels. The first test evaluates the ability of the FWCS to control steam generator 1 and 2 water levels during steady and transient conditions. The SG level setpoint was changed from normal SG level. The FWCS no.1 and no.2 controlled the SG water level to the new setpoint within the acceptable band. The second test evaluates the ability of the reactor regulating system (RRS) to control reactor coolant system (RCS) average temperature with respect to the reference temperature. The final test evaluates the ability of all of the control systems to work in an integrated manner controlling their respective parameters while the plant is at steady state conditions. The FWCS, RRS, SBCS, PLCS control their respective parameters within the control bands. The tests performed at Unit 3 were successful by meeting all of the test acceptable criteria. The measured test data for the major plant parameters were collected and evaluated. (author). 14 tabs., 217 figs., 7 refs

Evaluation of the control system checkout test at 100% power for Yonggwang Nuclear Power Plant Unit 3

International Nuclear Information System (INIS)

Kim, Shin Whan; Lee, Joo Han; Baek, Jong Man; Seo, Jong Tae; Lee, Sang Keun; Kang, In Koo; Ju, Hee Wan; Min, Kyung Soo; Kim, Byung Gon

1995-01-01

Control system checkout tests at various powers for Yonggwang Nuclear Power Plant Unit 3(YGN3) were performed to demonstrate the accuracies and proper performances of the control systems of the plant. Tested control systems included the feedwater control system, steam bypass control system, reactor regulation system, control element drive mechanism control system, pressurizer level control system, and pressurizer pressure control system. The measured test data during the control system checkout test at 100% power are evaluated. The test results showed that the control systems of YGN 3 properly control system was simulated by using the LTC code which is the performance analysis code for YGN 3 and 4 design. Comparisons of the predicted results with the measured data confirmed that the feedwater control system controls the steam generator level as designed

Quality assurance and non-destructive testing for nuclear power plants

International Nuclear Information System (INIS)

Manlucu, F.A.

1991-01-01

This article discussed the quality assurance requirements which have been extensively applied in plant design, fabrication, construction and operation and has played a major role in the excellent safety record of nuclear power plants. The application of non-destructive testing techniques, plays a very important role during the in-service inspection (ISI) in order to prevent dangerous accident and to assure continuous safe operation of nuclear power plants. (IMA). 12 refs

CCTF CORE I test results

International Nuclear Information System (INIS)

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

1982-07-01

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

Qualification tests for shift personnel in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Fechner, J B [Bundesministerium des Innern, Bonn (Germany, F.R.)

1978-02-01

The selection of personnel for training as shift supervisors or reactor operators so far used to be made by a plant operator mainly on the basis of such criteria as examinations, diplomas and other documents verifying the educational background, the type of activity exercised, and professional success. In addition, there are the opininons of trainers and supervisors based on personal observation of future shift personnel on training for specific plants at a training center, at the manufacturer's, the operator's or in activities in the construction and commissioning of the respective nuclear power plant. In the course of this phase, which normally takes several years, supervisors asses not only the professional capabilities of a trainee, but also bis psychic and physical performance and aptitude, e.g., with respect to decision making, leadership qualifications or behavior unter stress. The advisability of introducing psychological aptitude tests was also studied. However, a decision was recently taken to defer such psychological tests for the time being. Yet, nuclear power plant operators are required to submit a statement to their responsible authorities about industrial medical checkups and qualification assessments by supervisors.

Qualification tests for shift personnel in nuclear power plants

International Nuclear Information System (INIS)

Fechner, J.B.

1978-01-01

The selection of personnel for training as shift supervisors or reactor operators so far used to be made by a plant operator mainly on the basis of such criteria as examinations, diplomas and other documents verifying the educational background, the type of activity exercised, and professional success. In addition, there are the opininons of trainers and supervisors based on personal observation of future shift personnel on training for specific plants at a training center, at the manufacturer's, the operator's or in activities in the construction and commissioning of the respective nuclear power plant. In the course of this phase, which normally takes several years, supervisors asses not only the professional capabilities of a trainee, but also bis psychic and physical performance and aptitude, e.g., with respect to decision making, leadership qualifications or behavior unter stress. The advisability of introducing psychological aptitude tests was also studied. However, a decision was recently taken to defer such psychological tests for the time being. Yet, nuclear power plant operators are required to submit a statement to their responsible authorities about industrial medical checkups and qualification assessments by supervisors. (orig.) [de

Helium turbomachinery operating experience from gas turbine power plants and test facilities

International Nuclear Information System (INIS)

McDonald, Colin F.

2012-01-01

The closed-cycle gas turbine, pioneered and deployed in Europe, is not well known in the USA. Since nuclear power plant studies currently being conducted in several countries involve the coupling of a high temperature gas-cooled nuclear reactor with a helium closed-cycle gas turbine power conversion system, the experience gained from operated helium turbomachinery is the focus of this paper. A study done as early as 1945 foresaw the use of a helium closed-cycle gas turbine coupled with a high temperature gas-cooled nuclear reactor, and some two decades later this was investigated but not implemented because of lack of technology readiness. However, the first practical use of helium as a gas turbine working fluid was recognized for cryogenic processes, and the first two small fossil-fired helium gas turbines to operate were in the USA for air liquefaction and nitrogen production facilities. In the 1970's a larger helium gas turbine plant and helium test facilities were built and operated in Germany to establish technology bases for a projected future high efficiency large nuclear gas turbine power plant concept. This review paper covers the experience gained, and the lessons learned from the operation of helium gas turbine plants and related test facilities, and puts these into perspective since over three decades have passed since they were deployed. An understanding of the many unexpected events encountered, and how the problems, some of them serious, were resolved is important to avoid them being replicated in future helium turbomachines. The valuable lessons learned in the past, in many cases the hard way, particularly from the operation in Germany of the Oberhausen II 50 MWe helium gas turbine plant, and the technical know-how gained from the formidable HHV helium turbine test facility, are viewed as being germane in the context of current helium turbomachine design work being done for future high efficiency nuclear gas turbine plant concepts. - Highlights: �

Waste Isolation Pilot Plant Salt Decontamination Testing

Energy Technology Data Exchange (ETDEWEB)

Rick Demmer; Stephen Reese

2014-09-01

On February 14, 2014, americium and plutonium contamination was released in the Waste Isolation Pilot Plant (WIPP) salt caverns. At the request of WIPP’s operations contractor, Idaho National Laboratory (INL) personnel developed several methods of decontaminating WIPP salt, using surrogate contaminants and also americium (241Am). The effectiveness of the methods is evaluated qualitatively, and to the extent possible, quantitatively. One of the requirements of this effort was delivering initial results and recommendations within a few weeks. That requirement, in combination with the limited scope of the project, made in-depth analysis impractical in some instances. Of the methods tested (dry brushing, vacuum cleaning, water washing, strippable coatings, and mechanical grinding), the most practical seems to be water washing. Effectiveness is very high, and it is very easy and rapid to deploy. The amount of wastewater produced (2 L/m2) would be substantial and may not be easy to manage, but the method is the clear winner from a usability perspective. Removable surface contamination levels (smear results) from the strippable coating and water washing coupons found no residual removable contamination. Thus, whatever is left is likely adhered to (or trapped within) the salt. The other option that shows promise is the use of a fixative barrier. Bartlett Nuclear, Inc.’s Polymeric Barrier System (PBS) proved the most durable of the coatings tested. The coatings were not tested for contaminant entrapment, only for coating integrity and durability.

Boraflex test results and evaluation

International Nuclear Information System (INIS)

Lindquist, K.; Kline, D.E.; Haley, T.C.

1993-02-01

New data developed, collected, and evaluated to further assess the in-pool performance of the neutron absorber material, Boraflex. The data are from new EPRI test programs, utility surveillance programs, and blackness testing at a number of plants. This new data provides a basis for quantifying the gap phenomenon in full length panels of Boraflex in spent fuel racks; the maximum anticipated gap size, frequency of gap occurrence, and axial distribution of gaps. Methods have been developed to assess the reactivity effects of gaps and Boraflex shrinkage. The analyses presented demonstrates that the reactivity effect of gaps is very small, not much larger than the statistical variations inherent in the calculational method. The data and analyses presented serve to close the issue of gap formation and shrinkage in panels of Boraflex and the effect of such gaps and shrinkage on the reactivity of the fuel/rack configuration. Ongoing EPRI programs to assess the long term performance of Boraflex in spent fuel storage racks are described

RCGVS design improvement and depressurization capability tests for Ulchin nuclear power plant units 3 and 4

Energy Technology Data Exchange (ETDEWEB)

Sung, Kang Sik; Seong, Ho Je; Jeong, Won Sang; Seo, Jong Tae; Lee, Sang Keun [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of); Lim, Keun Hyo; Choi, Kwon Sik; Oh, Chul Sung [Korea Electric Power Cooperation, Taejon (Korea, Republic of)

1999-12-31

The Reactor Coolant Gas Vent System(RCGVS) design for Ulchin Nuclear Power Plant Units 3 and 4 (UCN 3 and 4) has been improved from the Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3 and 4) based on the evaluation results for depressurization capability tests performed at YGN 3 and 4. There has been a series of plant safety analyses for Natural Circulation Cooldown(NCC) event and thermo-dynamic analyses with RELAP5 code for the steam blowdown phenomena in order to optimize the orifice size of UCN 3 and 4 RCGVS. Based on these analyses results, the RCGVS orifics size for UCN 3 and 4 has been reduced to 9/32 inch from the 11/32 inch for YGN 3 and 4. The depressurization capability tests, which were performed at UCN 3 in order to verify the FSAR NCC analysis results, show that the RCGVS depressurization rates are being within the acceptable ranges. Therefore, it is concluded that the orificed flow path of UCN 3 and 4 RCGVS is adequately designed, and can provide the safety-grade depressurization capability required for a safe plant operation. 6 refs., 5 figs., 1 tab. (Author)

RCGVS design improvement and depressurization capability tests for Ulchin nuclear power plant units 3 and 4

Energy Technology Data Exchange (ETDEWEB)

Sung, Kang Sik; Seong, Ho Je; Jeong, Won Sang; Seo, Jong Tae; Lee, Sang Keun [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of); Lim, Keun Hyo; Choi, Kwon Sik; Oh, Chul Sung [Korea Electric Power Cooperation, Taejon (Korea, Republic of)

1998-12-31

The Reactor Coolant Gas Vent System(RCGVS) design for Ulchin Nuclear Power Plant Units 3 and 4 (UCN 3 and 4) has been improved from the Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3 and 4) based on the evaluation results for depressurization capability tests performed at YGN 3 and 4. There has been a series of plant safety analyses for Natural Circulation Cooldown(NCC) event and thermo-dynamic analyses with RELAP5 code for the steam blowdown phenomena in order to optimize the orifice size of UCN 3 and 4 RCGVS. Based on these analyses results, the RCGVS orifics size for UCN 3 and 4 has been reduced to 9/32 inch from the 11/32 inch for YGN 3 and 4. The depressurization capability tests, which were performed at UCN 3 in order to verify the FSAR NCC analysis results, show that the RCGVS depressurization rates are being within the acceptable ranges. Therefore, it is concluded that the orificed flow path of UCN 3 and 4 RCGVS is adequately designed, and can provide the safety-grade depressurization capability required for a safe plant operation. 6 refs., 5 figs., 1 tab. (Author)

Validation of the Engineering Plant Analyzer methodology with Peach Bottom 2 stability tests

International Nuclear Information System (INIS)

Rohatgi, U.S.; Mallen, A.N.; Cheng, H.S.; Wulff, W.

1994-01-01

The Engineering Plant Analyzer (EPA) had been developed in 1984 at Brookhaven National Laboratory to simulate plant transients in boiling water reactors (BWR). Recently, the EPA with its High-Speed Interactive Plant Analyzer code for BWRs ( ppercase HIPA-BWR ) simulated for the first time oscillatory transients with large, non-linear power and flow amplitudes; transients which are centered around the March 9, 1988 instability at the LaSalle-2 BWR power plant.The EPA's capability to simulate oscillatory transients has been demonstrated first by comparing simulation results with LaSalle-2 plant data (Wulff et al., NUREG/CR-5816, BNL-NUREG-52312, Brookhaven National Laboratory, 1992). This paper presents an EPA assessment on the basis of the Peach Bottom 2 instability tests (Carmichael and Niemi, EPRI NP-564, Electric Power Research Institute, Palo Alto, CA, 1978). This assessment of the EPA appears to constitute the first validation of a time-domain reactor systems code on the basis of frequency-domain criteria, namely power spectral density, gain and phase shift of the pressure-to-power transfer function.The reactor system pressure was disturbed in the Peach Bottom 2 power plant tests, and in their EPA simulation, by a pseudo-random, binary sequence signal. The data comparison revealed that the EPA predicted for Peach Bottom tests PT1, PT2, and PT4 the gain of the power-to-pressure transfer function with the biases and standard deviations of (-10±28)%, (-1±40)% and (+28±52)%, respectively. The respective frequencies at the peak gains were predicted with the errors of +6%, +3%, and -28%. The differences between the predicted and the measured phase shift increased with increasing frequency, but stayed within the margin of experimental uncertainty. ((orig.))

Application of wire sawing method to decommissioning of nuclear power plant. Cutting test with turbine pedestal of thermal power plant

International Nuclear Information System (INIS)

Hasegawa, Hideki; Uchiyama, Noriyuki; Sugiyama, Kazuya; Yamashita, Yoshitaka; Watanabe, Morishige

1995-01-01

It is very important to reduce radioactive waste volume, and to reduce radiation dose to workers and to the public during dismantling of the activated concrete in the decommissioning stage of a nuclear power plant. For the above, we studied a dismantling method which can separate activated concrete from non-activated concrete safely and effectively. Considering the state of legal regulation about radioactive waste disposal, and the state of developing of decommissioning technologies, we come to a conclusion that wire sawing method is feasible as a concrete cutting method. This study was carried out to evaluate the availability of the wire sawing method to dismantling of concrete structures of nuclear power plants. This study consists of concrete cutting rate test and concrete block cutting test. The former is to obtain data about cutting rate with various steel ratios while the latter is to obtain data about working time and man hour of the whole work with wire sawing. Thirty-six year old turbine pedestal of a thermal power plant was selected as a test piece to simulate actual decommissioning work of nuclear power plant, taking its massive concrete volume and age. Taking account of the handling in the building, the wire sawing machine with motor driven was used in this study considering that it did not produce exhaust gas. The concrete cutting rate test was performed with parameter of steel ratio in the concrete, wire tension and cutting direction. In the concrete block cutting test, imaging the actual cutting situation, cubic blocks which side was approximately 1 meter were taken out, and a large block to be cut and to be taken out is a section of 1m x 1.5m x 10m. Test results are shown below. The difference of cutting rate was mainly caused by the difference of reinforcement steel ratio. Working time data of installation, removal of machines and cutting were obtained. Data on secondary waste (dust, drainage and sludge) and environmental effect (noise and

Small-Scale Spray Releases: Initial Aerosol Test Results

Energy Technology Data Exchange (ETDEWEB)

Mahoney, Lenna A.; Gauglitz, Phillip A.; Kimura, Marcia L.; Brown, Garrett N.; Kurath, Dean E.; Buchmiller, William C.; Smith, Dennese M.; Blanchard, Jeremy; Song, Chen; Daniel, Richard C.; Wells, Beric E.; Tran, Diana N.; Burns, Carolyn A.

2013-05-29

One of the events postulated in the hazard analysis at the Waste Treatment and Immobilization Plant (WTP) and other U.S. Department of Energy (DOE) nuclear facilities is a breach in process piping that produces aerosols with droplet sizes in the respirable range. The current approach for predicting the size and concentration of aerosols produced in a spray leak involves extrapolating from correlations reported in the literature. These correlations are based on results obtained from small engineered spray nozzles using pure liquids with Newtonian fluid behavior. The narrow ranges of physical properties on which the correlations are based do not cover the wide range of slurries and viscous materials that will be processed in the WTP and across processing facilities in the DOE complex. Two key technical areas were identified where testing results were needed to improve the technical basis by reducing the uncertainty due to extrapolating existing literature results. The first technical need was to quantify the role of slurry particles in small breaches where the slurry particles may plug and result in substantially reduced, or even negligible, respirable fraction formed by high-pressure sprays. The second technical need was to determine the aerosol droplet size distribution and volume from prototypic breaches and fluids, specifically including sprays from larger breaches with slurries where data from the literature are scarce. To address these technical areas, small- and large-scale test stands were constructed and operated with simulants to determine aerosol release fractions and net generation rates from a range of breach sizes and geometries. The properties of the simulants represented the range of properties expected in the WTP process streams and included water, sodium salt solutions, slurries containing boehmite or gibbsite, and a hazardous chemical simulant. The effect of antifoam agents was assessed with most of the simulants. Orifices included round holes and

Multifrequency tests in the EBR-II reactor plant

International Nuclear Information System (INIS)

Feldman, E.E.; Mohr, D.; Gross, K.C.

1989-01-01

A series of eight multifrequency tests was conducted on the Experimental Breeder Reactor II. In half of the tests a control rod was oscillated and in the other half the controller input voltage to the intermediate-loop-sodium pump was perturbed. In each test the input disturbance consisted of several superimposed single-frequency sinusoidal harmonics of the same fundamental. The tests are described along with the theoretical and practical aspects of their development and design. Samples of measured frequency responses are also provided for both the reactor and the power plant. 22 refs., 5 figs., 2 tabs

Failure cause and failure rate evaluation on pumps of BWR plants in PSA. Hypothesis testing for typical or plant specific failure rate of pumps

International Nuclear Information System (INIS)

Sanada, Takahiro; Nakamura, Makoto

2009-01-01

In support of domestic nuclear industry effort to gather and analyze failure data of components concerning nuclear power plants, Nuclear Information Archives (NUCIA) are published for useful information to help PSA. This report focuses on NUCIA pertaining to pumps in domestic nuclear power plants, and provides the reliable estimation on failure rate of pumps resulting from failure cause analysis and hypothesis testing of classified and plant specific failure rate of pumps for improving quality in PSA. The classified and plant specific failure rate of pumps are estimated by analyzing individual domestic nuclear power plant's data of 26 Boiling Water Reactors (BWRs) concerning functionally structurally classified pump failures reported from beginning of commercial operation to March 31, 2007. (author)

Results of Testing the Relative Oxidizing Hazard of Wipes and KMI Zeolite

Energy Technology Data Exchange (ETDEWEB)

Ams, Bridget Elaine [Los Alamos National Laboratory

2017-05-09

This report includes the results from testing performed on the relative oxidizing hazard of a number of organic sorbing wipe materials, as well as KMI zeolite. These studies were undertaken to address a need by the Los Alamos National Laboratory (LANL) Hazardous Materials Management group, which requires a material that can sorb small spills in a glovebox without creating a disposal hazard due to the potential for oxidation reactions, as requested in Request for Testing of Wipes and Zeolite for Los Alamos National Laboratory Hazardous Materials Group (NPl-7) (NPl-7-17-002) and Request for Testing of Chamois Material for Los Alamos National Laboratory Hazardous Materials Group (NPl-7) (NPl-7-17-005). This set oftests is a continuation of previous testing described in Results from Preparation and Testing of Sorbents Mixed with (DWT-RPT-003), which provided data for the Waste Isolation Pilot Plant's Basis of Knowledge. The Basis of Knowledge establishes criteria for evaluating transuranic (TRU) waste that contains oxidizing chemicals.

Large-Scale Spray Releases: Initial Aerosol Test Results

Energy Technology Data Exchange (ETDEWEB)

Schonewill, Philip P.; Gauglitz, Phillip A.; Bontha, Jagannadha R.; Daniel, Richard C.; Kurath, Dean E.; Adkins, Harold E.; Billing, Justin M.; Burns, Carolyn A.; Davis, James M.; Enderlin, Carl W.; Fischer, Christopher M.; Jenks, Jeromy WJ; Lukins, Craig D.; MacFarlan, Paul J.; Shutthanandan, Janani I.; Smith, Dennese M.

2012-12-01

One of the events postulated in the hazard analysis at the Waste Treatment and Immobilization Plant (WTP) and other U.S. Department of Energy (DOE) nuclear facilities is a breach in process piping that produces aerosols with droplet sizes in the respirable range. The current approach for predicting the size and concentration of aerosols produced in a spray leak involves extrapolating from correlations reported in the literature. These correlations are based on results obtained from small engineered spray nozzles using pure liquids with Newtonian fluid behavior. The narrow ranges of physical properties on which the correlations are based do not cover the wide range of slurries and viscous materials that will be processed in the WTP and across processing facilities in the DOE complex. Two key technical areas were identified where testing results were needed to improve the technical basis by reducing the uncertainty due to extrapolating existing literature results. The first technical need was to quantify the role of slurry particles in small breaches where the slurry particles may plug and result in substantially reduced, or even negligible, respirable fraction formed by high-pressure sprays. The second technical need was to determine the aerosol droplet size distribution and volume from prototypic breaches and fluids, specifically including sprays from larger breaches with slurries where data from the literature are scarce. To address these technical areas, small- and large-scale test stands were constructed and operated with simulants to determine aerosol release fractions and generation rates from a range of breach sizes and geometries. The properties of the simulants represented the range of properties expected in the WTP process streams and included water, sodium salt solutions, slurries containing boehmite or gibbsite, and a hazardous chemical simulant. The effect of anti-foam agents was assessed with most of the simulants. Orifices included round holes and

77 FR 73056 - Initial Test Programs for Water-Cooled Nuclear Power Plants

Science.gov (United States)

2012-12-07

... Plants AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; request for comment... (DG), DG-1259, ``Initial Test Programs for Water-Cooled Nuclear Power Plants.'' This guide describes... (ITPs) for light water cooled nuclear power plants. DATES: Submit comments by January 31, 2013. Comments...

[Prediction of 137Cs behaviour in the soil-plant system in the territory of Semipalatinsk test site].

Science.gov (United States)

Spiridonov, S I; Mukusheva, M K; Gontarenko, I A; Fesenko, S V; Baranov, S A

2005-01-01

A mathematical model of 137Cs behaviour in the soil-plant system is presented. The model has been parameterized for the area adjacent to the testing area Ground Zero of the Semipalatinsk Test Site. The model describes the main processes responsible for the changes in 137Cs content in the soil solution and, thereby, dynamics of the radionuclide uptake by vegetation. The results are taken from predictive and retrospective calculations that reflect the dynamics of 137Cs distribution by species in soil after nuclear explosions. The importance of factors governing 137Cs accumulation in plants within the STS area is assessed. The analysis of sensitivity of the output model variable to changes in its parameters revealed that the key soil properties significantly influence the results of prediction of 137Cs content in plants.

Small-Scale Spray Releases: Initial Aerosol Test Results

Energy Technology Data Exchange (ETDEWEB)

Mahoney, Lenna A.; Gauglitz, Phillip A.; Kimura, Marcia L.; Brown, Garrett N.; Kurath, Dean E.; Buchmiller, William C.; Smith, Dennese M.; Blanchard, Jeremy; Song, Chen; Daniel, Richard C.; Wells, Beric E.; Tran, Diana N.; Burns, Carolyn A.

2012-11-01

One of the events postulated in the hazard analysis at the Waste Treatment and Immobilization Plant (WTP) and other U.S. Department of Energy (DOE) nuclear facilities is a breach in process piping that produces aerosols with droplet sizes in the respirable range. The current approach for predicting the size and concentration of aerosols produced in a spray leak involves extrapolating from correlations reported in the literature. These correlations are based on results obtained from small engineered spray nozzles using pure liquids with Newtonian fluid behavior. The narrow ranges of physical properties on which the correlations are based do not cover the wide range of slurries and viscous materials that will be processed in the WTP and across processing facilities in the DOE complex. Two key technical areas were identified where testing results were needed to improve the technical basis by reducing the uncertainty due to extrapolating existing literature results. The first technical need was to quantify the role of slurry particles in small breaches where the slurry particles may plug and result in substantially reduced, or even negligible, respirable fraction formed by high-pressure sprays. The second technical need was to determine the aerosol droplet size distribution and volume from prototypic breaches and fluids, specifically including sprays from larger breaches with slurries where data from the literature are scarce. To address these technical areas, small- and large-scale test stands were constructed and operated with simulants to determine aerosol release fractions and generation rates from a range of breach sizes and geometries. The properties of the simulants represented the range of properties expected in the WTP process streams and included water, sodium salt solutions, slurries containing boehmite or gibbsite, and a hazardous chemical simulant. The effect of anti-foam agents was assessed with most of the simulants. Orifices included round holes and

Hvdc valve development outstrips testing capacity: new synthetic plant at Marchwood

Energy Technology Data Exchange (ETDEWEB)

1966-01-07

A new test plant for high voltage direct current converters is being built at the CEGB's Marchwood Engineering Laboratories. Valve development has outstripped the 60 kV 600 A bridge rating of the existing zero power factor plant.

Tests of qualification of national components of nuclear power plants under design basis accident

International Nuclear Information System (INIS)

Mesquita, A.Z.

1990-01-01

With the purpose of qualifying national components of nuclear power plants, whose working must be maintained during and after an accident, the Thermohydraulic Division of CDTN have done tests to check the equipment stability, under Design Basis Accident conditions. Until this moment, the following components were tested: electrical junction boxes (connectors); coating systems for wall, inside cover and steel containment; hydraulics components of personnel and equipment airlock. This work describes the test instalation, the tests performed and its results. The components tested, in a general way, fulfil the specified requirements. (author) [pt

The validation of waste assay systems during active test at Rokkasho Reprocessing Plant

International Nuclear Information System (INIS)

Tamura, Takayuki; Miura, Yasushi; Iwamoto, Tomonori

2007-01-01

In order to implement accurate material accountancy at Rokkasho Reprocessing Plant (RRP) as a large scale reprocessing plant, it is necessary to introduce accurate measurement systems not only for mainstream material, but also appropriate measurement systems for solid waste materials. In this sense, the generated wastes by the active test operation have been measured with the Non-Destructive Assay Systems, such as Rokkasho Hulls Measurement System (RHMS) and Waste Crate Assay System (WCAS) for accountancy. This paper describes the experience of the NDA operation and the evaluation results for accountancy. (author)

Pilot plant SERSE: Description and results of the experimental tests under treatment of simulated chemical liquid waste; L'impianto pilota SERSE: Descrizione e risultati delle prove sperimentali del trattamento chimico di un rifiuto liquido simulato

Energy Technology Data Exchange (ETDEWEB)

Calle, C; Gili, M; Luce, A; Marrocchelli, A; Pietrelli, L; Troiani, F [ENEA - Dipartimento Ciclo del Combustibile, Centro Ricerche Energia, Casaccia (Italy)

1989-11-15

The chemical processes for the selective separation of the actinides and long lived fission products from aged liquid wastes is described. The SERSE pilot plant is a cold facility which has been designed, by ENEA, for the engineering scale demonstration of the chemical separation processes. The experimental tests carried out in the plant are described and the results confirm the laboratory data. (author)

Software test and validation of wireless sensor nodes used in nuclear power plant

International Nuclear Information System (INIS)

Deng Changjian; Chen Dongyi; Zhang Heng

2015-01-01

The software test and validation of wireless sensor nodes is one of the key approaches to improve or guarantee the reliability of wireless network application in nuclear power plants (NPPs). At first, to validate the software test, some concepts are defined quantitatively, for example the robustness of software, the reliability of software, and the security of software. Then the development tools and simulators of discrete event drive operating system are compared, in order to present robustness, reliability and security of software test approach based on input-output function. Some simple preliminary test results are given to show that different development software can obtain almost same measurement and communication results although the software of special application may be different than normal application. (author)

ESCRIME: testing bench for advanced operator workstations in future plants

International Nuclear Information System (INIS)

Poujol, A.; Papin, B.

1994-01-01

The problem of optimal task allocation between man and computer for the operation of nuclear power plants is of major concern for the design of future plants. As the increased level of automation induces the modification of the tasks actually devoted to the operator in the control room, it is very important to anticipate these consequences at the plant design stage. The improvement of man machine cooperation is expected to play a major role in minimizing the impact of human errors on plant safety. The CEA has launched a research program concerning the evolution of the plant operation in order to optimize the efficiency of the human/computer systems for a better safety. The objective of this program is to evaluate different modalities of man-machine share of tasks, in a representative context. It relies strongly upon the development of a specific testing facility, the ESCRIME work bench, which is presented in this paper. It consists of an EDF 1300MWe PWR plant simulator connected to an operator workstation. The plant simulator model presents at a significant level of details the instrumentation and control of the plant and the main connected circuits. The operator interface is based on the generalization of the use of interactive graphic displays, and is intended to be consistent to the tasks to be performed by the operator. The functional architecture of the workstation is modular, so that different cooperation mechanisms can be implemented within the same framework. It is based on a thorough analysis and structuration of plant control tasks, in normal as well as in accident situations. The software architecture design follows the distributed artificial intelligence approach. Cognitive agents cooperate in order to operate the process. The paper presents the basic principles and the functional architecture of the test bed and describes the steps and the present status of the program. (author)

Information on the Advanced Plant Experiment (APEX) Test Facility

International Nuclear Information System (INIS)

Smith, Curtis Lee

2015-01-01

The purpose of this report provides information related to the design of the Oregon State University Advanced Plant Experiment (APEX) test facility. Information provided in this report have been pulled from the following information sources: Reference 1: R. Nourgaliev and et.al, 'Summary Report on NGSAC (Next-Generation Safety Analysis Code) Development and Testing,' Idaho National Laboratory, 2011. Note that this is report has not been released as an external report. Reference 2: O. Stevens, Characterization of the Advanced Plant Experiment (APEX) Passive Residual Heat Removal System Heat Exchanger, Master Thesis, June 1996. Reference 3: J. Reyes, Jr., Q. Wu, and J. King, Jr., Scaling Assessment for the Design of the OSU APEX-1000 Test Facility, OSU-APEX-03001 (Rev. 0), May 2003. Reference 4: J. Reyes et al, Final Report of the NRC AP600 Research Conducted at Oregon State University, NUREG/CR-6641, July 1999. Reference 5: K. Welter et al, APEX-1000 Confirmatory Testing to Support AP1000 Design Certification (non-proprietary), NUREG-1826, August 2005.

French nuclear power plants. Results and outlooks

International Nuclear Information System (INIS)

Serres, S.; Carbonnier, D.

1999-01-01

Operating results were good in 1997 for French nuclear power plants: safety levels were perfectly satisfactory; operating expenses continued to decrease (by 2% per annum from 1992 to 1997); there were spectacular results in radiation protection; and they had one of the world's highest availability rates (nearly 83%). (orig.) [de

Pressure test behaviour of embalse nuclear power plant containment structure

International Nuclear Information System (INIS)

Bruschi, S.; Marinelli, C.

1984-01-01

It's described the structural behaviour of the containment structure during the pressure test of the Embalse plant (CANDU type, 600MW), made of prestressed concrete with an epoxi liner. Displacement, strain, temperature, and pressure measurements of the containment structure of the Embalse Nuclear Power Plant are presented. The instrumentation set up and measurement specifications are described for all variables of interest before, during and after the pressure test. The analytical models to simulate the heat transfer due to sun heating and air convenction and to predict the associated thermal strains and displacements are presented. (E.G.) [pt

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Survey of analysis results from preservation tests on condensation water

Energy Technology Data Exchange (ETDEWEB)

Hall, Bjoern (Bjoern Hall, Miljoe och Foerbraenningskemi, Onsala (Sweden))

2010-03-15

Avfall Sverige - Swedish Waste Management has together with seven waste incineration plants made a study with the purpose of examining the necessity of preservatives when analyzing process water from wet flue gas treatment. The analyzed water in this study is not water leaving the plant, but process water that has yet not undergone any water treatment. The analysis after two weeks showed that a statistically proven difference between the preserved and non-preserved samples was obtained only from mercury samples at one plant (Plant 1) and from lead samples at another (Plant 2). The difference between the values of the lead samples was, however, so small that it was easily covered by the margin of error for the analysis, and could be considered as coincidental. The differences between the values for other metals were either very small or had values that fell below the detection limit. The analysis made after six weeks also showed that there was a considerable difference between the preserved and non-preserved samples of mercury from plant 1, which confirms the trend seen in the 2 week analysis. Other samples, which were analyzed after six weeks, show that another plant (Plant 4) stands out, in that the preserved samples for most metals had considerably higher levels compared to the non-preserved samples. Plant 4 is different from other plants also when comparing 2-week samples and 6-week samples. For most metals, levels were higher in the 2-week sample. The levels of the samples were very high, in general, at this plant. Other than this there was no statistically clear difference in levels between the preserved and non-preserved samples. This test series showed that the difference between preserved and non-preserved samples was very small for most metals and at most plants. For mercury, there is a statistical and experimental difference between the preserved and non-preserved samples from plant 1. However, the difference between samples preserved in nitric acid and

Verification Results of Safety-grade Optical Modem for Core Protection Calculator (CPC) in Korea Standard Nuclear Power Plant (KSNP)

Energy Technology Data Exchange (ETDEWEB)

Kim, Jangyeol; Son, Kwangseop; Lee, Youngjun; Cheon, Sewoo; Cha, Kyoungho; Lee, Jangsoo; Kwon, Keechoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

We confirmed that the coverage criteria for a safety-grade optical modem of a Core Protection Calculator is satisfactory using a traceability analysis matrix between high-level requirements and lower-level system test case data set. This paper describes the test environment, test components and items, a traceability analysis, and system tests as a result of system verification and validation based on Software Requirement Specifications (SRS) for a safety-grade optical modem of a Core Protection Calculator (CPC) in a Korea Standard Nuclear Power Plant (KSNP), and Software Design Specifications (SDS) for a safety-grade optical modem of a CPC in a KSNP. All tests were performed according to the test plan and test procedures. Functional testing, performance testing, event testing, and scenario based testing for a safety-grade optical modem of a Core Protection Calculator in a Korea Standard Nuclear Power Plant as a thirty-party verifier were successfully performed.

Nuclear power plant Olkiluoto 3. Containment leakage test under extreme conditions

Energy Technology Data Exchange (ETDEWEB)

Fleckenstein, Tobias [TUEV SUED Industrie Service GmbH, Munich (Germany). Measaruement Technology Dept.

2015-01-15

Modern nuclear power plants place high demands on the design and execution of safety checks. TUEV SUED supported the containment leakage test for the largest- capacity third generation nuclear power plant in the world - Olkiluoto 3 in Finland. The experts successfully met the challenges presented by exceptional parameters of the project. The containment of Olkiluoto 3 is unique in that the vessel's volume is 80,000 m{sup 3} while measurements were carried out over a period of ten days. To execute the test, 75 temperature and 15 humidity sensors had to be installed and correctly interlinked by more than ten kilometres of cable. These instruments also needed to withstand an absolute pressure of 6 bar, ambient temperatures of 30 C and high levels of humidity. These conditions required comprehensive preparation and a high amount of qualification tests. Parts of the qualifications were carried out at the autoclave system of the Technical University in Munich, Germany, where the project test conditions could be simulated. The software required to determine the tests was developed by TUEV SUED and verified by German's national accreditation body DAkkS under ISO 17025. TUEV SUED enabled the test schedule to continue without delay by analysing all recorded data continuously on site, including pressure, temperature, humidity and leakage mass flow curves. With the comprehensive preparation, data acquisition system recording measurements continuously and the on-time result calculation, all components of the leak-tightness assessment were successfully completed in accordance with requirements.

FY 1992 report on the results of the development of an entrained bed coal gasification power plant. Part 3. Operation test of pilot plant (1/2); 1992 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu - Sono 3. Pilot plant unten shiken hen (1/2)

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-02-01

The study of operation test was made of the 200 t/d entrained bed coal gasification pilot plant that was constructed with the aim of establishing technology of the integrated coal gasification combined cycle power generation, and the details of the FY 1992 results were summarized. At RUN 10, conducted were the confirmation of the quantity of state of gasifier in the coal-fired high load operation and coal-fired constant load operation, and slagging survey. At RUN 11, the evaluation made after the work for prevention of slagging, and confirmation of the quantity of state of gasifier at a load of 80% heat input. At RUN 12, the evaluation of the measures taken against slagging, and test on the high load stable operation. At RUN 13, the evaluation of the measures taken against slagging, and large combustor response/total pressure control response tests. At RUN D1, test on the change of coal kind from A coal to D coal, and test on the initial adjusting operation of D coal. In the trial operation of these, the following were generated and the preventive measures were studied: impossibility of circulation of desulfurizing agent in gas refining facilities (dry desulfurizing system), bolt fracture of gas refining facilities (separator of dedusting facilities). (NEDO)

Current developments in mechanized non-destructive testing in nuclear power plants

International Nuclear Information System (INIS)

Zeilinger, R.

2008-01-01

Nuclear power plants require frequent in-service activities to be carried out conscientiously in areas potentially hazardous to human operators (because of the associated radiation exposure), such as non-destructive testing of pressurized components of the steam system. Locations to be inspected in this way include the reactor pressure vessel, core internals, steam generators, pressurizers, and pipes. The codes to be used as a basis of these inspections demand high absolute positioning and repeating accuracy. These requirements can be met by mechanized test procedures. Accordingly, many new applications of, mostly mobile, robots have been developed over the past few years. The innovative control and sensor systems for stationary and mobile robots now on the market offer a potential for economic application in a large number of new areas in inspection, maintenance and service in nuclear power plants. More progress in this area is expected for the near future. Areva NP founded the new NDT Center, NETEC (Non-destructive Examination Technical Center), as a global technical center for non-destructive materials testing. NETEC is to advance research and development of all basic NDT technologies, robotics included. For many years, intelligeNDT has offered solutions and products for a variety of inspection and testing purposes and locations in nuclear power plants and is involved in continuous further development of the experience collected in nuclear power plants on the spot. (orig.)

Results of automatic system implementation for the friction control rods execution in Cofrentes nuclear power plant

International Nuclear Information System (INIS)

Curiel, M.; Palomo, M. J.; Urrea, M.; Arnaldos, A.

2010-10-01

The purpose of this presentation is to show the obtained results in Cofrentes nuclear power plant (Spain) of control rods Pcc/24 friction test procedure. In order to perform this, a control rod friction test system has been developed. Principally, this system consists on software and data acquisition hardware that obtains and analyzes the control rod pressure variation on which the test is being made. The Pcc/24 procedure objective is to detect an excessive friction in the control rod movement that could cause a control rod drive movement slower than usual. This test is necessary every time that an anomalous alteration is produced in the reactor core that could affect to a fuel rod, and it is executed before the time measure of control rods rapid scram test of the affected rods. This test has to be carried out to all the reactor control rods and takes valuable time during plant refuelling. So, by means of an automatic system to perform the test, we obtain an important time saving during refuelling. On the other hand, the on-line monitoring of the control rod insertion and changes in differential pressure, permits a control rod operation fast and safe validation. Moreover, an automatic individual report of every rod is generated by the system and a final global result report of the entire test developed in refuelling is generated. The mentioned reports can be attached directly to the procedure documents obtaining an office data processing important saving time. (Author)

Results of automatic system implementation for the friction control rods execution in Cofrentes nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Curiel, M. [Logistica y Acondicionamientos Industriales SAU, Sorolla Center, local 10, Av. de las Cortes Valencianas, 46015 Valencia (Spain); Palomo, M. J. [ISIRYM, Universidad Politecnica de Valencia, Camino de Vera s/n, Valencia (Spain); Urrea, M. [Iberdrola Generacion S. A., Central Nuclear Cofrentes, Carretera Almansa Requena s/n, 04662 Cofrentes, Valencia (Spain); Arnaldos, A., E-mail: m.curiel@lainsa.co [TITANIA Servicios Tecnologicos SL, Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain)

2010-10-15

The purpose of this presentation is to show the obtained results in Cofrentes nuclear power plant (Spain) of control rods Pcc/24 friction test procedure. In order to perform this, a control rod friction test system has been developed. Principally, this system consists on software and data acquisition hardware that obtains and analyzes the control rod pressure variation on which the test is being made. The Pcc/24 procedure objective is to detect an excessive friction in the control rod movement that could cause a control rod drive movement slower than usual. This test is necessary every time that an anomalous alteration is produced in the reactor core that could affect to a fuel rod, and it is executed before the time measure of control rods rapid scram test of the affected rods. This test has to be carried out to all the reactor control rods and takes valuable time during plant refuelling. So, by means of an automatic system to perform the test, we obtain an important time saving during refuelling. On the other hand, the on-line monitoring of the control rod insertion and changes in differential pressure, permits a control rod operation fast and safe validation. Moreover, an automatic individual report of every rod is generated by the system and a final global result report of the entire test developed in refuelling is generated. The mentioned reports can be attached directly to the procedure documents obtaining an office data processing important saving time. (Author)

FY 1990 report on the results of the development of the entrained bed coal gasification power plant. Part 2. Fabrication/installation of pilot plant; 1990 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu - Sono 2. Pilot plant seisaku suetsuke hen

Energy Technology Data Exchange (ETDEWEB)

NONE

1991-03-01

For the purpose of establishing the technology of the integrated coal gasification combined cycle power generation, fabrication/installation work, etc. were made for a pilot plant of 200t/d entrained bed coal gasification power generation, and the FY 1990 results were summarized. Construction work of a pilot plant of coal gasification power generation was at its peak in April 1990, and installation/piping work for each facility/equipment was carried out. In May, transportation/installation of gas turbine and generator were started. In June, installation of equipment of the 66kV special high voltage switching station was conducted, and the initial power receiving of 6.9kV was conducted. In August, inspection before use was made of the main piping of the gasifier equipment, gas refining equipment and gas turbine equipment. In December, trial unit operation of each equipment and interlock test were carried out. 'The integrated plant protection interlock test' was made from January 21 to February 21, 1991, and the favorable results were obtained. On February 28, a ceremony to celebrate the completion of all facilities of pilot plant was made. In March, drying of gasifier and initial firing by light oil were conducted, and all the work was completed on March 25. (NEDO)

FY 1990 report on the results of the development of the entrained bed coal gasification power plant. Part 2. Fabrication/installation of pilot plant; 1990 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu - Sono 2. Pilot plant seisaku suetsuke hen

Energy Technology Data Exchange (ETDEWEB)

NONE

1991-03-01

For the purpose of establishing the technology of the integrated coal gasification combined cycle power generation, fabrication/installation work, etc. were made for a pilot plant of 200t/d entrained bed coal gasification power generation, and the FY 1990 results were summarized. Construction work of a pilot plant of coal gasification power generation was at its peak in April 1990, and installation/piping work for each facility/equipment was carried out. In May, transportation/installation of gas turbine and generator were started. In June, installation of equipment of the 66kV special high voltage switching station was conducted, and the initial power receiving of 6.9kV was conducted. In August, inspection before use was made of the main piping of the gasifier equipment, gas refining equipment and gas turbine equipment. In December, trial unit operation of each equipment and interlock test were carried out. 'The integrated plant protection interlock test' was made from January 21 to February 21, 1991, and the favorable results were obtained. On February 28, a ceremony to celebrate the completion of all facilities of pilot plant was made. In March, drying of gasifier and initial firing by light oil were conducted, and all the work was completed on March 25. (NEDO)

Applicability study on a ceramic filter with hot-test conducted in a BWR plant

International Nuclear Information System (INIS)

Yamada, K.; Shirai, T.; Wada, M.; Nakamizo, H.

1991-01-01

Radioactive crud removal and filtration performance recovery by backwashing were examined with a BWR plant pool water using a ceramic filter element, 0.1 micron in nominal pore size and 0.2m 2 in filtration area. Totally 1114 hours filter operation were accumulated. Ten backwashings were accomplished during the test period. The following results were obtained. (1) Radioactive crud concentration in the filter effluent remained below 10 5 Bq/m 3 . (2) Both pressure loss through the filter and dose rate at the filter vessel surface were recovered to the initial level by each backwashing. The surface dose rate after backwashing was approximately 0.01mSv/h. According to these test results, it is confirmed that the ceramic filter is appropriate for the treatment of highly crud concentrated radioactive liquid, which is generated in nuclear facilities, such as spent fuel reprocessing plants. (author)

FY 1992 report on the results of the demonstration test on the methanol conversion at oil-fired power plant. Demonstration test on a methanol reformation type power generation total system; 1992 nendo sekiyu karyoku hatsudensho metanoru tenkan tou jissho shiken. Metanoru kaishitsu gata hatsuden total system jissho shiken

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-03-01

For the promotion of introduction of methanol to oil-fired power plant, based on the results of the element study, operational study was conducted of a 1,000kW class total system plant for which each of the elements was combined, and the FY 1992 results were summarized. In the operational study, data on various kinds of operational study were sampled of each of the simple cycle/regeneration cycle of liquid methanol and simple cycle/regeneration cycle of gas methanol. As to the reformed gas/water injection/regeneration cycle, all functions as a total system plant worked normally, and it was confirmed that the reformed gas/water injection/regeneration cycle operation could be made possible. Besides, the following were conducted: confirmation test on the performance of the developmental catalyst used in the operational study by bench-scale test device, trial operation for adjustment of gas turbine and combustion study such as the performance test in each cycle, manufacture/study of catalyst for the total system, study for longevity of catalyst for the total system, etc. (NEDO)

Small punch creep test: A promising methodology for high temperature plant components life evaluation

Energy Technology Data Exchange (ETDEWEB)

Tettamanti, S [CISE SpA, Milan (Italy); Crudeli, R [ENEL SpA, Milan (Italy)

1999-12-31

CISE and ENEL are involved for years in a miniaturization creep methodology project to obtain similar non-destructive test with the same standard creep test reliability. The goal can be reached with `Small punch creep test` that collect all the requested characteristics; quasi nondestructive disk specimens extracted both on external or internal side of components, than accurately machined and tested on little and cheap apparatus. CISE has developed complete creep small punch procedure that involved peculiar test facility and correlation`s law comparable with the more diffused isostress methodology for residual life evaluation on ex-serviced high temperature plant components. The aim of this work is to obtain a simple and immediately applicable relationship useful for plant maintenance managing. More added work is need to validate the Small Punch methodology and for relationship calibration on most diffusion high temperature structural materials. First obtained results on a comparative work on ASTM A355 P12 ex-serviced pipe material are presented joint with a description of the Small Punch apparatus realized in CISE. (orig.) 6 refs.

Small punch creep test: A promising methodology for high temperature plant components life evaluation

Energy Technology Data Exchange (ETDEWEB)

Tettamanti, S. [CISE SpA, Milan (Italy); Crudeli, R. [ENEL SpA, Milan (Italy)

1998-12-31

CISE and ENEL are involved for years in a miniaturization creep methodology project to obtain similar non-destructive test with the same standard creep test reliability. The goal can be reached with `Small punch creep test` that collect all the requested characteristics; quasi nondestructive disk specimens extracted both on external or internal side of components, than accurately machined and tested on little and cheap apparatus. CISE has developed complete creep small punch procedure that involved peculiar test facility and correlation`s law comparable with the more diffused isostress methodology for residual life evaluation on ex-serviced high temperature plant components. The aim of this work is to obtain a simple and immediately applicable relationship useful for plant maintenance managing. More added work is need to validate the Small Punch methodology and for relationship calibration on most diffusion high temperature structural materials. First obtained results on a comparative work on ASTM A355 P12 ex-serviced pipe material are presented joint with a description of the Small Punch apparatus realized in CISE. (orig.) 6 refs.

FY 1991 report on the results of the development of an entrained bed coal gasification power plant. Part 3. Adjustment of the operation test of pilot plant (1/2); 1991 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu - Sono 3. Pilot plant unten shiken chosei hen (1/2)

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-01-01

The adjustment was made of the operation test of the 200 t/d entrained bed coal gasification pilot plant that was constructed with the aim of establishing technology of the integrated coal gasification combined cycle power generation, and the results were reported. As to the adjustment of the operation test of gasifier facilities, the light-oil firing test and tests of RUN 1-9 were conducted, and the paper reported cases of 49-item troubles, the causes, measures against them, improvement of facilities, etc. For slagging, operation conditions, improvement of various facilities, etc. were studied. Relating to the adjustment of the operation test of gas refining facilities (dry desulfurization facilities), the following were carried out: test on empty column gas circulation, test on desulfurizer circulation characteristics, test on SO2 reduction tower, test on warming-up characteristics, tests 1-3 on initial performance, tests on change in gas turbine fuel and loads. And, 21 cases of troubles, the causes and measures against them were reported. Concerning the adjustment of the operation test of gas refining facilities (dry dedusting facilities), the following were conducted: test on non-load sequence, test on confirmation prior to letting gas through, tests 1-3 on initial performance, test on dedusting characteristics. And, 22 cases of troubles, the causes and measures against them were reported. (NEDO)

Degradation of PVC/HC blends. II. Terrestrial plant growth test.

Science.gov (United States)

Pascu, Mihaela; Agafiţei, Gabriela-Elena; Profire, Lenuţa; Vasile, Cornelia

2009-01-01

The behavior at degradation by soil burial of some plasticized polyvinyl chloride (PVC) based blends with a variable content of hydrolyzed collagen (HC) has been followed. The modifications induced in the environment by the polymer systems (pH variation, physiologic state of the plants, assimilatory pigments) were studied. Using the growth test of the terrestrial plants, we followed the development of Triticum (wheat), Helianthus annus minimus (little sunflower), Pisum sativum (pea), and Vicia X hybrida hort, during a vegetation cycle. After the harvest, for each plant, the quantities of chlorophyll and carotenoidic pigments and of trace- and macroelements were determined. It was proved that, in the presence of polymer blends, the plants do not suffer morphological and physiological modifications, the products released in the culture soil being not toxic for the plants growth.

European stress tests for nuclear power plants. The Swedish National Report

International Nuclear Information System (INIS)

2011-01-01

On 11 March 2011, the Tohoku region in north Honshu, Japan, suffered a severe earthquake with an ensuing tsunami and an accident at the Fukushima Dai-ichi nuclear power plant. Due to the accident the Council of the European Union declared in late March that Member States were prepared to begin reviewing safety at nuclear facilities in the European Union by means of a comprehensive assessment of risk and safety ('stress testing'). On 25 May, SSM ordered the licensees of the nuclear power plants to conduct renewed analyses of the facilities' resilience against different kinds of natural phenomena. They were also to analyse how the facilities would be capable of dealing with a prolonged loss of electrical power, regardless of cause. On 31 October, the licensees reported on their stress tests to SSM. After reviewing these reports, SSM produced a summary stress test report, which was submitted to the Government on the 15 December. The present report is the national report on Swedish stress tests of nuclear power plants. The report will be submit to the European Commission no later than 31 December. Based on the review SSM has drawn the conclusion that the stress tests carried out by Swedish licensees are largely performed in accordance with the specification resolved within the European Union. The scope and depth of these analyses and assessments are essentially in accordance with ENSREG's definition of 'a comprehensive assessment of risk and safety'. The stress tests show that Swedish facilities are robust, but the tests also identify a number of opportunities to further strengthen the facilities' robustness. SSM will order the respective licensees to present an action plan for dealing with the results from the stress tests. The Authority will then examine the plans and adopt a standpoint on proposed measures as well as check that the necessary safety improvements are made. In a number of cases, the stress tests indicate deficiencies in relation to, or alternatively

Integrated automatic non-destructive testing in industrial production and in the operation of technical plant

International Nuclear Information System (INIS)

Hoeller, P.

1989-01-01

The article deals with non-destructive testing (NDT) in automated manufacture and in the automated operation of industrial plant. In both areas of application, the tests are coupled to the process (real time operation) and the results are used for the control of manufacture or of the course of the process. The control process can be coupled to the process in open loop or closed loop. The subject is explained by the following examples: 1) Automated testing of sheets in a steelworks. 2) Automatic NDT on machine parts in tempering and machining by the 3MA system (3MA: micro-magnetic, multi-parameter, micro-structure and stress analysis). 3) Automated ultrasonic testing in manufacture and in the operation of plants with the ALOK data collection and processing system (ALOK: amplitude, running time, location curves). 4) Automated wheel running surface test on Intercity experimental train, and 5) automated level measurement on BWR pressure vessels. (orig./MM) [de

Test and maintenance of the emergency power supply in the nuclear power plant Biblis

International Nuclear Information System (INIS)

Kotthoff, K.; Huren, H.

1986-01-01

Besides design and construction test and maintenance play an important role for the availability of the emergency power supply. As an example test and maintenance provided for the emergency power supply in a German 4-loop PWR will be described. In general one has to differentiate between test and maintenance performed during power operation of the plant and those carried out during the refuelling outage. For both periods of operation detailed information will be given including type, extent and frequency of test and maintenance work. The results of test and maintenance up to now will be discussed. (authors)

Test and maintenance of the emergency power supply in the nuclear power plant Biblis

Energy Technology Data Exchange (ETDEWEB)

Kotthoff, K. [Gesellschaft fuer Reaktorsicherheit - GRS mbH, Schwertnergasse 1, D-5000 Koeln 1, Cologne (Germany); Huren, H. [Rheinisch-Westfaelisches Elektrizitaetswerk AG, Betriebsverwaltung Biblis, Biblis (Germany)

1986-02-15

Besides design and construction test and maintenance play an important role for the availability of the emergency power supply. As an example test and maintenance provided for the emergency power supply in a German 4-loop PWR will be described. In general one has to differentiate between test and maintenance performed during power operation of the plant and those carried out during the refuelling outage. For both periods of operation detailed information will be given including type, extent and frequency of test and maintenance work. The results of test and maintenance up to now will be discussed. (authors)

FY 1991 report on the results of the development of an entrained bed coal gasification power plant. Part 3. Adjustment of the operation test of pilot plant (2/2); 1991 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu - Sono 3. Pilot plant unten shiken chosei hen (2/2)

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-01-01

The adjustment was made of the operation test of the 200 t/d entrained bed coal gasification pilot plant that was constructed with the aim of establishing technology of the integrated coal gasification combined cycle power generation, and the results were reported. As to the adjustment of the operation test of gas turbine facilities, the following were conducted: tests 1 and 2 on light-oil firing characteristics, test on coal gas ignition, tests on fuel change/gas firing, test on fuel change. And, 12 cases of troubles, the causes and measures against them were reported. Relating to the adjustment of the operation test of actual pressure/actual size combustor testing facilities, tests on hot air device/air heating device and tests 1-3 on light-oil firing were carried out, and 7 cases of troubles, the causes and measures against them were reported. Concerning the adjustment of the operation test of safety environment facilities, tests were made of RUN 3-6, RUN 7 (1 and 2), RUN 8 (1-4) and RUN 9 (1-3), and 20 cases of troubles, the causes and measures against them were reported. As to the adjustment of the operation test of electric/control facilities, items of improvement were reported of gasifier facilities, gas refining facilities, gas turbine facilities, actual pressure/actual size combustor testing facilities, safety environment facilities and total control facilities. (NEDO)

Pilot plant for flue gas treatment - continuous operation tests

International Nuclear Information System (INIS)

Chmielewski, A.G.; Tyminski, B.; Iller, E.; Zimek, Z.; Licki, J.; Radzio, B.

1995-01-01

Tests of continuous operation have been performed on pilot plant at EPS Kaweczyn in the wide range of SO 2 concentration (500-3000 ppm). The bag filter has been applied for aerosol separation. The high efficiencies of SO 2 and NO x removal, approximately 90% were obtained and influenced by such process parameters as: dose, gas temperature and ammonia stoichiometry. The main apparatus of the pilot plant (e.g. both accelerators) have proved their reliability in hard industrial conditions. (Author)

The in-pile proving test for fuel assembly of Qinshan nuclear power plant

International Nuclear Information System (INIS)

Chen Dianshan; Zhang Shucheng; Kang Rixin; Wang Huarong; Chen Guanghan

1989-10-01

The in-pile proving test for fuel assembly of Qinshan nuclear power plant had been conducted in the experimental loop of HWRR at IAE (Institute of Atomic Energy) in Beijing, China, from January 1985 to December 1986. Average burnup of 27000 MWd/tU and peak burnup of 34000 MWd/tU of fuel rod had already been reached. The basic status of the experiment are described, emphasis is placed on the discussion of proving test parameters and analysis of experiment results

Chinese nuclear heating test reactor and demonstration plant

International Nuclear Information System (INIS)

Wang Dazhong; Ma Changwen; Dong Duo; Lin Jiagui

1992-01-01

In this report the importance of nuclear district heating is discussed. From the viewpoint of environmental protection, uses of energy resources and transport, the development of nuclear heating in China is necessary. The development program of district nuclear heating in China is given in the report. At the time being, commissioning of the 5 MW Test Heating Reactor is going on. A 200 MWt Demonstration Plant will be built. In this report, the main characteristics of these reactors are given. It shows this type of reactor has a high inherent safety. Further the report points out that for this type of reactor the stability is very important. Some experimental results of the driving facility are included in the report. (orig.)

IVO participation in IAEA benchmark for VVER-type nuclear power plants seismic analysis and testing

International Nuclear Information System (INIS)

Varpasuo, P.

1997-12-01

This study is a part of the IAEA coordinated research program 'Benchmark study for the Seismic Analysis and Testing of VVER Type NPPs'. The study reports the numerical simulation of the blast test for Paks and Kozloduy nuclear power plants beginning from the recorded free-field response and computing the structural response at various points inside the reactor building. The full-scale blast tests of the Paks and Kozloduy NPPs took place in December 1994 and in July 1996. During the tests the plants operated normally. The instrumentation for the tests consisted of 52 recording channels with 200 Hz sampling rate. Detonating 100 kg charges in 50-meter deep boreholes at 2.5-km distance from the plant carried out the blast tests. The 3D structural models for both reactor buildings were analyzed in the frequency domain. The number of modes extracted in both cases was about 500 and the cut-off frequency was 25 Hz. In the response history run the responses of the selected points were evaluated. The input values for response history run were the three components of the excitation, which were transformed from time domain to the frequency domain with the aid of Fourier transform. The analysis was carried out in frequency domain and responses were transferred back to time domain with inverse Fourier transform. The Paks and Kozloduy blast tests produced a wealth of information on the behavior of the nuclear power plant structures excited by blast type loads containing also the low frequency wave train if albeit with small energy content. The comparison of measured and calculated results gave information about the suitability of the selected analysis approach for the investigated blast type loading

The use of NPAR [Nuclear Plant Aging Research] results in plant inspection activities

International Nuclear Information System (INIS)

Gunther, W.; Taylor, J.

1989-01-01

The Nuclear Plant Aging Research (NPAR) Program is a hardware oriented research program which has produced a large data base of equipment and system operating, maintenance, and testing information. A review of the NRC Inspection Program and discussions with NRC inspection personnel have revealed several areas where NPAR research results would be valuable to the inspector. This paper describes the NPAR information which can enhance inspection activities, and provides alternatives for making these pertinent research results available to the inspectors. The NRC Inspection Program emphasis is on evaluating the performance of licensees by focusing on requirements and standards associated with administrative, managerial, engineering, and operational aspects of licensee activities. The Program recognizes that licensees may satisfy NRC requirements differently, and therefore expresses inspection guidance in the form of performance objectives and evaluation criteria. for the resident and regional inspectors, procedures have been written covering various subject areas, such as operations, maintenance, and surveillance. Some of these procedures contain guidance related to aging degradation. The types of information generated by NPAR which were found to be relevant to inspection needs include the following: functional indicators; failure modes, causes, effects; stresses which cause degradation; maintenance recommendations; inspection prioritization. 3 refs

Field Lysimeter Test Facility: Protective barrier test results (FY 1990, the third year)

International Nuclear Information System (INIS)

Campbell, M.D.; Gee, G.W.

1990-11-01

The Field Lysimeter Test Facility (FLTF) was constructed to test protective barriers for isolating low-level radioactive and hazardous wastes from the biosphere. Protective barriers are specially configured earth materials placed over near-surface wastes to prevent intrusion of water, plants, and animals. Low-level radioactive waste is stored in near-surface repositories at the Hanford Site and can be transported into the biosphere by water, plants, and animals. The purpose of the FLTF is to measure water balance within barriers as precipitation is partitioned to evaporation (including transpiration), storage, and drainage. Runoff was prevented by raised edges on the lysimeters. Water balance in protective barriers depends on the water-holding capacity of the soil, the gradient of a potential, and the conductivity of the underlying capillary barrier. Current barrier design uses soil with a high water storage capacity and a capillary barrier underlying the soil to increase its water storage capacity. This increased storage capacity is to hold water, which would normally drain, near the the surface where evaporation can cycle it back to the atmosphere. 7 refs., 23 figs., 5 tabs

Results of automatic system implementation for the friction control rods execution in Cofrentes nuclear power plant

International Nuclear Information System (INIS)

Palomo, M.; Urrea, M.; Arnaldos, A.

2011-01-01

The purpose of this presentation is to show the obtained results in Cofrentes Nuclear Power Plant (Spain) of Control Rods PCC/24 Friction Test Procedure. In order to perform this, a Control Rod Friction Test System has been developed. Principally, this system consists on software and data acquisition hardware that obtains and analyzes the control rod pressure variation on which the test is being made. The PCC/24 Procedure objective is to detect an excessive friction in the control rod movement that could cause a CRD (Control Rod Drive) movement slower than usual. This test is necessary every time that an anomalous alteration is produced in the reactor core that could affect to a fuel rod, and it is executed before the time measure of control rods rapid scram test of the affected rods. This test has to be carried out to all the reactor control rods and takes valuable time during plant refuelling. So, by means of an automatic system to perform the test, we obtain an important time saving during refuelling. On the other hand, the on-line monitoring of the control rod insertion and changes in differential pressure, permits a control rod operation fast and safe validation. Moreover, an automatic individual report of every rod is generated by the system and a final global result report of the entire test developed in refuelling is generated. The mentioned reports can be attached directly to the procedure documents obtaining an office data processing important saving time.(author)

Results of automatic system implementation for the friction control rods execution in Cofrentes nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Palomo, M., E-mail: mpalomo@iqn.upv.es [Universidad Politecnica de Valencia (UPV) (Spain); Urrea, M., E-mail: matias.urrea@iberdrola.es [Iberdrola Generacion S.A. Valencia (Spain). C.N. Cofrentes; Curiel, M., E-mail: m.curiel@lainsa.com [Logistica y Acondicionamientos Industriales (LAINSA), Valencia (Spain); Arnaldos, A., E-mail: a.arnaldos@titaniast.com [TITANIA Servicios Teconologicos, Valencia (Spain)

2011-07-01

The purpose of this presentation is to show the obtained results in Cofrentes Nuclear Power Plant (Spain) of Control Rods PCC/24 Friction Test Procedure. In order to perform this, a Control Rod Friction Test System has been developed. Principally, this system consists on software and data acquisition hardware that obtains and analyzes the control rod pressure variation on which the test is being made. The PCC/24 Procedure objective is to detect an excessive friction in the control rod movement that could cause a CRD (Control Rod Drive) movement slower than usual. This test is necessary every time that an anomalous alteration is produced in the reactor core that could affect to a fuel rod, and it is executed before the time measure of control rods rapid scram test of the affected rods. This test has to be carried out to all the reactor control rods and takes valuable time during plant refuelling. So, by means of an automatic system to perform the test, we obtain an important time saving during refuelling. On the other hand, the on-line monitoring of the control rod insertion and changes in differential pressure, permits a control rod operation fast and safe validation. Moreover, an automatic individual report of every rod is generated by the system and a final global result report of the entire test developed in refuelling is generated. The mentioned reports can be attached directly to the procedure documents obtaining an office data processing important saving time.(author)

New large solar photocatalytic plant: set-up and preliminary results.

Science.gov (United States)

Malato, S; Blanco, J; Vidal, A; Fernández, P; Cáceres, J; Trincado, P; Oliveira, J C; Vincent, M

2002-04-01

A European industrial consortium called SOLARDETOX has been created as the result of an EC-DGXII BRITE-EURAM-III-financed project on solar photocatalytic detoxification of water. The project objective was to develop a simple, efficient and commercially competitive water-treatment technology, based on compound parabolic collectors (CPCs) solar collectors and TiO2 photocatalysis, to make possible easy design and installation. The design, set-up and preliminary results of the main project deliverable, the first European industrial solar detoxification treatment plant, is presented. This plant has been designed for the batch treatment of 2 m3 of water with a 100 m2 collector-aperture area and aqueous aerated suspensions of polycrystalline TiO2 irradiated by sunlight. Fully automatic control reduces operation and maintenance manpower. Plant behaviour has been compared (using dichloroacetic acid and cyanide at 50 mg l(-1) initial concentration as model compounds) with the small CPC pilot plants installed at the Plataforma Solar de Almería several years ago. The first results with high-content cyanide (1 g l(-1)) waste water are presented and plant treatment capacity is calculated.

78 FR 67206 - Qualification Tests for Safety-Related Actuators in Nuclear Power Plants

Science.gov (United States)

2013-11-08

... Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Revision to regulatory guide; issuance..., ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants.'' This RG is being revised to provide... Operators Installed Inside the Containment of Nuclear Power Plants,'' dated January 1974. ADDRESSES: Please...

Studies and testing in water and steam of valves and fittings, and nuclear components. The result of 25 years of testing using a comprehensive range of test facilities under service conditions

International Nuclear Information System (INIS)

Berail, J.F.; Bruneau, S.; Crouzet, D.; Haas, J.L.; Zbinden, M.

1998-05-01

Electricite de France operates 58 PWR nuclear power stations, for which the behaviour of valves and fittings is of major importance for safety, for the availability of the plants, and for maintenance costs. Since the early 70's, EDF has developed a comprehensive range of facilities to test valves and fittings in PWR service and accident conditions. It has carried out studies, tests, development work, experimental and numerical research in collaboration with external organisations and manufacturers, to improve the technologies of these equipment as well as maintenance tools and methods. In the present paper, the authors quantify the importance of valves and fittings studies for EDF, which has led to the drawing up of a catalogue of approved equipment. They describe the principle test facilities, and the structure of the EDF 'valves and fittings tests results' data base. They show the importance of twenty-five years of testing experience for both the evolution of equipment and for the increase in French nuclear plants availability. (author)

FY 1980 Report on results of Sunshine Project. Development of coal liquefaction techniques (Development of 1 T/D test plant, and researches on the solvent-extraction type liquefaction process); 1980 nendo sekitan ekika gijutsu no kaihatsu, yozai chushutsu ekika plant no kaihatsu seika hokokusho. 1t/nichi jikken plant no kaihatsu, yozai chushutsu ekika process no kenkyu

Energy Technology Data Exchange (ETDEWEB)

NONE

1981-03-01

This program is aimed at establishing the techniques for solvent-extraction type coal liquefaction plant by constructing and operating a 1 T/D test plant to obtain the technical data for the efficient plant. The test plant is operated to confirm the effects of temperature and coal slurry concentration on liquefaction conversion by the solvent-extraction for a short time in the furnace for the extraction unit. The extraction type coal liquefaction tests can be conducted for a reaction time of around 1 hour by the test plant. The recycled solvent purification unit is installed, to regenerate the hydrogen donor solvent. For researches on the solvent-extraction type coal liquefaction process, the continuous extraction is conducted, to investigate the effects of extraction reaction rate at relatively low pressure. The optimum hydrogenation conditions are studied for the test plant. It is confirmed that a Mo-based catalyst is suitable for the hydrogenation. The batch type reaction system is operated to investigate the liquid yield of Wandoan coal, and recycled solvent balances and compositions. (NEDO)

Testing in power plant construction as well as in the petrochemical and chemical industry

International Nuclear Information System (INIS)

Riess, N.; Schittko, H.

1978-01-01

In general, the upgrading of requirements for the most different fields of engineering is also characterized by a corresponding effort in testing. In this context especially nondestructive tests of materials are of outstanding importance. In the fields of power plant construction (among others, components for nuclear power plants) as well as petrochemical and chemical industry considered here, almost all nondestructive test methods are applied. This paper discusses not so much theoretical testing problems, but rather test objects as well as specifications and testing equipment. (orig./HP) [de

Results of the DIOS pilot plant test and summary of the joint research; DIOS pilot plant no shiken sogyo kekka to kenkyu seika no matome

Energy Technology Data Exchange (ETDEWEB)

Sugiyama, T [Center for Coal Utilization, Japan, Tokyo (Japan); Kawaoka, K [The Japan Iron and Steel Federation, Tokyo (Japan)

1996-09-01

A joint research had been carried out with a subsidy from the Agency of Natural Resources and Energy since fiscal 1988 to fiscal 1995 on the direct iron ore smelting reduction process (DIOS process). The process utilizes coal directly as a process to use the strong points and supplement the weak points of the blast furnace process. During the period, a pilot plant had been operated since 1993. Upon having completed the feasibility study, this paper reports the result thereof. The main facilities consist of a smelting and reducing furnace of iron bath type, a spare reducing furnace of fluidized bed type, and a preheating furnace. The former two furnaces constitute a unit structure with the two furnaces connected vertically. The pilot plant achieved a three-day continuous operation producing 500 tons of iron every day. The production rate reached 21 tons an hour at an upward oxygen blowing velocity of about 13,000 Nm {sup 3} per hour. The coal unit requirement showed a result of <1000 kg/t for high VM coal and <900 kg/t for low VM coal. These results verified a possibility that this process can supplement or replace the blast furnace process even for a production scale of 9000 tons a day. 7 refs., 15 figs., 3 tabs.

Bioactivity test and GRW biogas yield test. Methods for optimizing biogas plants for anaerobic digestion of biowaste; Rostocker Aktivitaets- und GRW-Biogasertragstest. Einsatz zur Optimierung von Abfallvergaerungsanlagen

Energy Technology Data Exchange (ETDEWEB)

Engler, Nils [Rostock Univ. (Germany). Lehrstuhl Abfall- und Soffstromwirtschaft; Schiffner, Maik [Rostock Univ. (Germany). Forschungsvorhaben ' ' Bilanzierung von Stoff- und Energiestroemen' ' ; Nelles, Michael [Rostock Univ. (Germany). Lehrstuhl Abfall- und Soffstromwirtschaft; Rostock Univ. (Germany). Inst. fuer Umweltingenieurwesen; Fritz, Thomas

2010-03-15

Anaerobic digestion to obtain biogas is one option for energetic use of biodegradable waste. Data as e. g. the expected biogas yield, the biogas composition or inhibition effects are essentially to estimate the potentials and risks of the use of biowaste in commercial bio gas plants. To deliver such data, several test methods were developed. The GRW biogas yield test was first applied at the university of applied science in Goettingen and enhanced in cooperation with the University of Rostock. The test is particularly suitable for inhomogeneous samples as e. g. biowaste. The Bioactivity Test is still under development. First results have shown that the test can be applied for the detection of potentially inhibition effects. Combination of both Tests can deliver data for optimizing biogas plants for anaerobic digestion of biowaste (orig.)

Production characteristics of lettuce Lactuca sativa L. in the frame of the first crop tests in the Higher Plant Chamber integrated into the MELiSSA Pilot Plant

Science.gov (United States)

Tikhomirova, Natalia; Lawson, Jamie; Stasiak, Michael; Dixon, Mike; Paille, Christel; Peiro, Enrique; Fossen, Arnaud; Godia, Francesc

Micro-Ecological Life Support System Alternative (MELiSSA) is an artificial closed ecosystem that is considered a tool for the development of a bioregenerative life support system for manned space missions. One of the five compartments of MELiSSA loop -Higher Plant Chamber was recently integrated into the MELiSSA Pilot Plant facility at Universitat Aut`noma deo Barcelona. The main contributions expected by integration of this photosynthetic compartment are oxygen, water, vegetable food production and CO2 consumption. Production characteristics of Lactuca sativa L., as a MELiSSA candidate crop, were investigated in this work in the first crop experiments in the MELiSSA Pilot Plant facility. The plants were grown in batch culture and totaled 100 plants with a growing area 5 m long and 1 m wide in a sealed controlled environment. Several replicates of the experiments were carried out with varying duration. It was shown that after 46 days of lettuce cultivation dry edible biomass averaged 27, 2 g per plant. However accumulation of oxygen in the chamber, which required purging of the chamber, and decrease in the food value of the plants was observed. Reducing the duration of the tests allowed uninterrupted test without opening the system and also allowed estimation of the crop's carbon balance. Results of productivity, tissue composition, nutrient uptake and canopy photosynthesis of lettuce regardless of test duration are discussed in the paper.

USING POTATOES IN PROPAGATION TESTS FOR NONTARGET PLANT EFFECTS

Science.gov (United States)

Current tests required for pesticide registration under the FIFRA only investigate seedling emergence and early growth. Previous research with sulfonylurea (SU) herbicides has shown that significant impacts can occur to plant reproduction with little or no visible effect on vege...

Results from uranium deposition studies for development of a Limited Frequency-Unannounced Access (LFUA) inspection strategy for gas centrifuge enrichment plants

International Nuclear Information System (INIS)

Cooley, J.N.; Fields, L.W.; Swindle, D.W.

1985-06-01

Uranium deposition studies were performed on a test loop system designed to simulate process gas flow through the header piping of a gas centrifuge enrichment plant. The objectives of these studies were to investigate the effectiveness of an in-line gaseous cleaning agent in removing uranium in pipe deposits and to analyze long-term deposition growth and isotopic exchange under simulated centrifuge plant operating conditions. The test loop studies are described, the results are reported, and the implications for analyzing actual plant data are discussed. Results indicate that: 93% of the uranium deposit is removed within 15 min when a pipe is pressurized with gaseous ClF 3 ; the isotopic abundance of a highly enriched uranium deposit remains unchanged when UF 6 of a lower assay is introduced into the pipe; and air inleakage will be the cause of the largest deposits in centrifuge plant process header pipes. 3 refs., 3 figs., 3 tabs

Preparation and testing of plant seed meal-based wood adhesives.

Science.gov (United States)

He, Zhongqi; Chapital, Dorselyn C

2015-03-05

Recently, the interest in plant seed meal-based products as wood adhesives has steadily increased, as these plant raw materials are considered renewable and environment-friendly. These natural products may serve as alternatives to petroleum-based adhesives to ease environmental and sustainability concerns. This work demonstrates the preparation and testing of the plant seed-based wood adhesives using cottonseed and soy meal as raw materials. In addition to untreated meals, water washed meals and protein isolates are prepared and tested. Adhesive slurries are prepared by mixing a freeze-dried meal product with deionized water (3:25 w/w) for 2 hr. Each adhesive preparation is applied to one end of 2 wood veneer strips using a brush. The tacky adhesive coated areas of the wood veneer strips are lapped and glued by hot-pressing. Adhesive strength is reported as the shear strength of the bonded wood specimen at break. Water resistance of the adhesives is measured by the change in shear strength of the bonded wood specimens at break after water soaking. This protocol allows one to assess plant seed-based agricultural products as suitable candidates for substitution of synthetic-based wood adhesives. Adjustments to the adhesive formulation with or without additives and bonding conditions could optimize their adhesive properties for various practical applications.

Non-destructive testing for plant life assessment

International Nuclear Information System (INIS)

2005-01-01

The International Atomic Energy Agency (IAEA) is promoting industrial applications of nondestructive testing (NDT) technology, which includes radiography testing (RT) and related methods, to assure safety and reliability of operation of industrial facilities and processes. NDT technology is essentially needed for improvement of the quality of industrial products, safe performance of equipment and plants, including safety of metallic and concrete structures and constructions. The IAEA is playing an important role in promoting the NDT use and technology support to Member States, in harmonisation for training and certification of NDT personnel, and in establishing national accreditation and certifying bodies. All these efforts have led to a stage of maturity and self sufficiency in numerous countries especially in the field of training and certification of personnel, and in provision of services to industries. This has had a positive impact on the improvement of the quality of industrial goods and services. NDT methods are primarily used for detection, location and sizing of surface and internal defects (in welds, castings, forging, composite materials, concrete and many more). Various NDT methods are applied for preventive maintenance (aircraft, bridge), for the inspection of raw materials, half-finished and finished products, for in-service-inspection and for plant life assessment studies. NDT is essential for quality control of the facilities and products, and for fitness - for purpose assessment (so-called plant life assessment). NDT evaluates remaining operation life of plant components (processing lines, pipes, vessels) providing an accurate diagnosis that allows predicting extended life operation beyond design life. Status and trends on the NDT for plant life assessment have been discussed in many IAEA meetings related with NDT development, training and education. Experts have largely demonstrated that, using NDT methods, a comprehensive assessment of the life

Program outline of seismic fragility capacity tests on nuclear power plant equipment

International Nuclear Information System (INIS)

Lijima, T.; Abe, H.; Fujita, T.

2004-01-01

A seismic probabilistic safety assessment (PSA) is an available method to evaluate residual risk of nuclear plant that is designed with definitive seismic design condition. Seismic fragility capacity data are necessary for seismic PSA, but we don't have sufficient data of active components of nuclear plants in Japan. This paper describes a plan of seismic fragility capacity tests on nuclear power plant equipment. The purpose of those tests is to obtain seismic fragility capacity of important equipment from a safety design point of view. And the equipment for the fragility capacity tests were selected considering effect on core damage frequency (CDF) that was evaluated by our preliminary seismic PSA. Consequently horizontal shaft pump, electric cabinets, Control Rod Drive system (CRD system) of BWR and PWR plant and vertical shaft pump were selected. The seismic fragility capacity tests are conducted from phase-1 to phase-3, and horizontal shaft pump and electric cabinets are tested on phase-1. The fragility capacity test consists of two types of tests. One is actual equipment test and another is element test. On actual equipment test, a real size model is tested with high-level seismic motion, and critical acceleration and failure mode are investigated. Regarding fragility test phase-1, we selected typical type horizontal shaft pump and electric cabinets for the actual equipment test. Those were Reactor Building Closed Cooling Water (RCW) Pump and eight kinds of electric cabinets such as relay cabinet, motor control center. On the test phase-1, maximum input acceleration for the actual equipment test is intended to be 6-G-force. Since the shaking table of TADOTSU facility did not have capability for high acceleration, we made vibration amplifying system. In this system, amplifying device is mounted on original shaking table and it moves in synchronization with original table. The element test is conducted with many samples and critical acceleration, median and

Overview on recent results of the VTT's research programme on assuring nuclear power plant structural safety

International Nuclear Information System (INIS)

Rintamaa, R.; Aaltonen, P.; Kauppinen, P.; Keinaenen, H.; Talja, H.; Valo, M.; Wallin, K.; Toerroenen, K.

1994-01-01

An overview of the Finnish national research programme on the Nuclear Power Plant Structural Safety, being carried out from 1990 to 1994, is presented. The focus of this paper is on recent results in the areas of experimental and computational fracture mechanics, material deterioration due to neutron irradiation, corrosion and water chemistry, nondestructive testing methods and procedures, and verification of structural integrity assessment methods by large scale component tests. (author). 21 refs, 21 figs, 2 tabs

Guidelines for Electromagnetic Interference Testing of Power Plant Equipment: Revision 3 to TR-102323

International Nuclear Information System (INIS)

Cunningham, J.; Shank, J.

2004-01-01

To continue meeting safety and reliability requirements while controlling costs, operators of nuclear power plants must be able to replace and upgrade equipment in a cost-effective manner. One issue that has been problematic for new plant equipment and especially for digital instrumentation and control (I and C) systems in recent years is electromagnetic compatibility (EMC). The EMC issue usually involves testing to show that critical equipment will not be adversely affected by electromagnetic interference (EMI) in the plant environment. This guide will help nuclear plant engineers address EMC issues and qualification testing in a consistent, comprehensive manner

Guidelines for Electromagnetic Interference Testing of Power Plant Equipment: Revision 3 to TR-102323

Energy Technology Data Exchange (ETDEWEB)

J. Cunningham and J. Shank

2004-11-01

To continue meeting safety and reliability requirements while controlling costs, operators of nuclear power plants must be able to replace and upgrade equipment in a cost-effective manner. One issue that has been problematic for new plant equipment and especially for digital instrumentation and control (I&C) systems in recent years is electromagnetic compatibility (EMC). The EMC issue usually involves testing to show that critical equipment will not be adversely affected by electromagnetic interference (EMI) in the plant environment. This guide will help nuclear plant engineers address EMC issues and qualification testing in a consistent, comprehensive manner.

Development of the Chinshan plant analyzer and its assessment with plant data

International Nuclear Information System (INIS)

Shihjen Wang; Chunsheng Chien; Jungyuh Jang; Shawcuang Lee

1993-01-01

To apply fast and accurate simulation techniques to Taiwanese nuclear power plants, plant analyzer technology was transferred to Taiwan from the Brookhaven National Laboratory (BNL) through a cooperative program. The Chinshan plant analyzer is developed on the AD100 peripheral processor systems, based on the BNL boiling water reactor plant analyzer. The BNL plant analyzer was first converted from MPS10 programming for AD10 to ADSIM programming for AD100. It was then modified for the Taiwan Power Company's Chinshan power station. The simulation speed of the Chinshan plant analyzer is eight times faster than real time. A load rejection transient performed at 100% of full power during startup tests was simulated with the Chinshan plant analyzer, and the results were benchmarked against test data. The comparison shows good agreement between calculated results and test data

Test phase plan for the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

1993-03-01

The US Department of Energy (DOE) has prepared this Test Phase Plan for the Waste Isolation Pilot Plant to satisfy the requirements of Public Law 102-579, the Waste Isolation Pilot Plant (WIPP) Land Withdrawal Act (LWA). The Act provides seven months after its enactment for the DOE to submit this Plan to the Environmental Protection Agency (EPA) for review. A potential geologic repository for transuranic wastes, including transuranic mixed wastes, generated in national-defense activities, the WIPP is being constructed in southeastern New Mexico. Because these wastes remain radioactive and chemically hazardous for a very long time, the WIPP must provide safe disposal for thousands of years. The DOE is developing the facility in phases. Surface facilities for receiving waste have been built and considerable underground excavations (2150 feet below the surface) that are appropriate for in-situ testing, have been completed. Additional excavations will be completed when they are required for waste disposal. The next step is to conduct a test phase. The purpose of the test phase is to develop pertinent information and assess whether the disposal of transuranic waste and transuranic mixed waste in the planned WIPP repository can be conducted in compliance with the environmental standards for disposal and with the Solid Waste Disposal Act (SWDA) (as amended by RCRA, 42 USC. 6901 et. seq.). The test phase includes laboratory experiments and underground tests using contact-handled transuranic waste. Waste-related tests at WIPP will be limited to contact-handled transuranic and simulated wastes since the LWA prohibits the transport to or emplacement of remote-handled transuranic waste at WIPP during the test phase

76 FR 52355 - NUREG-1482, Revision 2, “Guidelines for Inservice Testing at Nuclear Power Plants, Draft Report...

Science.gov (United States)

2011-08-22

... Testing at Nuclear Power Plants, Draft Report for Comment'' AGENCY: Nuclear Regulatory Commission. ACTION... Testing at Nuclear Power Plants, Draft Report for Comment,'' and subtitled ``Inservice Testing of Pumps... Examination and Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants'' is available electronically...

Phased array UT (Ultrasonic Testing) used in electricity production plants

International Nuclear Information System (INIS)

Kodaira, Takeshi

2012-01-01

Phased Array-Ultrasonic testing techniques widely used for detection and quantitative determination of the lattice defects which have been formed from fatigues or stress corrosion cracking in the materials used in the electricity production plants are presented with particular focus on the accurate determination of the defects depth (sizing) and defects discrimination applicable to weld metals of austenite stainless steels and Ni base alloys. The principle of this non-destructive analysis is briefly explained, followed by point and matrix focus phased array methods developed by Mitsubishi Heavy Industries, Ltd are explained rather in detail with illustration and the evaluated results. (S. Ohno)

Loss of benefits resulting from mandated nuclear plant shutdowns

International Nuclear Information System (INIS)

Peerenboom, J.P.; Buehring, W.A.

1982-01-01

This paper identifies and discusses some of the important consequences of nuclear power plant unavailability, and quantifies a number of technical measures of loss of benefits that result from regulatory actions such as licensing delays and mandated nuclear plant outages. The loss of benefits that accompany such regulatory actions include increased costs of systems generation, increased demand for nonnuclear and often scarce fuels, and reduced system reliability. This paper is based on a series of case studies, supplemented by sensitivity studies, on hypothetical nuclear plant shutdowns. These studies were developed by Argonne in cooperation with four electric utilities

Results from study of potential early commercial MHD power plants and from recent ETF design work. [Engineering Test Facility

Science.gov (United States)

Hals, F.; Kessler, R.; Swallom, D.; Westra, L.; Zar, J.; Morgan, W.; Bozzuto, C.

1980-01-01

The study deals with different 'moderate technology' entry-level commercial MHD power plants. Two of the reference plants are based on combustion of coal with air preheated in a high-temperature regenerative air heater separately fired with a low-BTU gas produced in a gasifier integrated with the power plant. The third reference plant design is based on the use of oxygen enriched combustion air. Performance calculations show that an overall power plant efficiency of the order of 44% can be reached with the use of oxygen enrichment.

Comparison of EPRI safety valve test data with analytically determined hydraulic results

International Nuclear Information System (INIS)

Smith, L.C.; Howe, K.S.

1983-01-01

NUREG-0737 (November 1980) and all subsequent U.S. NRC generic follow-up letters require that all operating plant licensees and applicants verify the acceptability of plant specific pressurizer safety valve piping systems for valve operation transients by testing. To aid in this verification process, the Electric Power Research Institute (EPRI) conducted an extensive testing program at the Combustion Engineering Test Facility. Pertinent tests simulating dynamic opening of the safety valves for representative upstream environments were carried out. Different models and sizes of safety valves were tested at the simulated operating conditions. Transducers placed at key points in the system monitored a variety of thermal, hydraulic and structural parameters. From this data, a more complete description of the transient can be made. The EPRI test configuration was analytically modeled using a one-dimensional thermal hydraulic computer program that uses the method of characteristics approach to generate key fluid parameters as a function of space and time. The conservative equations are solved by applying both the implicit and explicit characteristic methods. Unbalanced or wave forces were determined for each straight run of pipe bounded on each side by a turn or elbow. Blowdown forces were included, where appropriate. Several parameters were varied to determine the effects on the pressure, hydraulic forces and timings of events. By comparing these quantities with the experimentally obtained data, an approximate picture of the flow dynamics is arrived at. Two cases in particular are presented. These are the hot and cold loop seal discharge tests made with the Crosby 6M6 spring-loaded safety valve. Included in the paper is a description of the hydraulic code, modeling techniques and assumptions, a comparison of the numerical results with experimental data and a qualitative description of the factors which govern pipe support loading. (orig.)

Culinary Spice Plants in Dietary Supplement Products and Tested in Clinical Trials.

Science.gov (United States)

Saldanha, Leila G; Dwyer, Johanna T; Betz, Joseph M

2016-03-01

Dried plant parts used as culinary spices (CSs) in food are permitted as dietary ingredients in dietary supplements (DSs) within certain constraints in the United States. We reviewed the amounts, forms, and nutritional support (structure/function) claims of DSs that contain CS plants listed in the Dietary Supplement Label Database (DSLD) and compared this label information with trial doses and health endpoints for CS plants that were the subject of clinical trials listed in clinicaltrials.gov. According to the DSLD, the CS plants occurring most frequently in DSs were cayenne, cinnamon, garlic, ginger, pepper, rosemary, and turmeric. Identifying the botanical species, categorizing the forms used, and determining the amounts from the information provided on DS labels was challenging. CS plants were typically added as a component of a blend, as the powered biomass, dried extracts, and isolated phytochemicals. The amounts added were declared on about 55% of the labels, rendering it difficult to determine the amount of the CS plant used in many DSs. Clinicaltrials.gov provided little information about the composition of test articles in the intervention studies. When plant names were listed on DS labels and in clinical trials, generally the common name and not the Latin binomial name was given. In order to arrive at exposure estimates and enable researchers to reproduce clinical trials, the Latin binomial name, form, and amount of the CS plant used in DSs and tested in clinical trials must be specified. © 2016 American Society for Nutrition.

Seismic proving tests on the reliability for large components and equipment of nuclear power plants

International Nuclear Information System (INIS)

Ohno, Tokue; Tanaka, Nagatoshi

1988-01-01

Since Japan has destructive earthquakes frequently, the structural reliability for large components and equipment of nuclear power plants are rigorously required. They are designed using sophisticated seismic analyses and have not yet encountered a destructive earthquake. When nuclear power plants are planned, it is very important that the general public understand the structural reliability during and after an earthquake. Seismic Proving Tests have been planned by Ministry of International Trade and Industry (Miti) to comply with public requirement in Japan. A large-scale high-performance vibration table was constructed at Tasted Engineering Laboratory of Nuclear Power Engineering Test Center (NU PEC), in order to prove the structural reliability by vibrating the test model (of full scale or close to the actual size) in the condition of a destructive earthquake. As for the test models, the following four items were selected out of large components and equipment important to the safety: Reactor Containment Vessel; Primary Coolant Loop or Primary Loop Recirculation System; Reactor Pressure Vessel; and Reactor Core Internals. Here is described a brief of the vibration table, the test method and the results of the tests on PWR Reactor Containment Vessel and BWR Primary Loop Recirculation System (author)

Testing the Efficacy of DNA Barcodes for Identifying the Vascular Plants of Canada.

Science.gov (United States)

Braukmann, Thomas W A; Kuzmina, Maria L; Sills, Jesse; Zakharov, Evgeny V; Hebert, Paul D N

2017-01-01

Their relatively slow rates of molecular evolution, as well as frequent exposure to hybridization and introgression, often make it difficult to discriminate species of vascular plants with the standard barcode markers (rbcL, matK, ITS2). Previous studies have examined these constraints in narrow geographic or taxonomic contexts, but the present investigation expands analysis to consider the performance of these gene regions in discriminating the species in local floras at sites across Canada. To test identification success, we employed a DNA barcode reference library with sequence records for 96% of the 5108 vascular plant species known from Canada, but coverage varied from 94% for rbcL to 60% for ITS2 and 39% for matK. Using plant lists from 27 national parks and one scientific reserve, we tested the efficacy of DNA barcodes in identifying the plants in simulated species assemblages from six biogeographic regions of Canada using BLAST and mothur. Mean pairwise distance (MPD) and mean nearest taxon distance (MNTD) were strong predictors of barcode performance for different plant families and genera, and both metrics supported ITS2 as possessing the highest genetic diversity. All three genes performed strongly in assigning the taxa present in local floras to the correct genus with values ranging from 91% for rbcL to 97% for ITS2 and 98% for matK. However, matK delivered the highest species discrimination (~81%) followed by ITS2 (~72%) and rbcL (~44%). Despite the low number of plant taxa in the Canadian Arctic, DNA barcodes had the least success in discriminating species from this biogeographic region with resolution ranging from 36% with rbcL to 69% with matK. Species resolution was higher in the other settings, peaking in the Woodland region at 52% for rbcL and 87% for matK. Our results indicate that DNA barcoding is very effective in identifying Canadian plants to a genus, and that it performs well in discriminating species in regions where floristic diversity is

Innovative test method for the estimation of the foaming tendency of substrates for biogas plants

Energy Technology Data Exchange (ETDEWEB)

Moeller, Lucie, E-mail: lucie.moeller@ufz.de [UFZ – Helmholtz Centre for Environmental Research, Centre for Environmental Biotechnology, Permoserstrasse 15, 04318 Leipzig (Germany); Eismann, Frank, E-mail: info@antoc.de [Eismann & Stöbe GbR, GeoPark, Geb. A12, Bautzner Strasse 67, 04347 Leipzig (Germany); Wißmann, Daniel, E-mail: d.s.wissmann@gmx.de [University of Hohenheim, State Institute of Agricultural Engineering and Bioenergy (LA740), Garbenstrasse 9, 70599 Stuttgart (Germany); Nägele, Hans-Joachim, E-mail: hajo.naegele@uni-hohenheim.de [University of Hohenheim, State Institute of Agricultural Engineering and Bioenergy (LA740), Garbenstrasse 9, 70599 Stuttgart (Germany); Zielonka, Simon, E-mail: simon.zielonka@uni-hohenheim.de [University of Hohenheim, State Institute of Agricultural Engineering and Bioenergy (LA740), Garbenstrasse 9, 70599 Stuttgart (Germany); Müller, Roland A., E-mail: roland.mueller@ufz.de [UFZ – Helmholtz Centre for Environmental Research, Centre for Environmental Biotechnology, Permoserstrasse 15, 04318 Leipzig (Germany); Zehnsdorf, Andreas, E-mail: andreas.zehnsdorf@ufz.de [UFZ – Helmholtz Centre for Environmental Research, Centre for Environmental Biotechnology, Permoserstrasse 15, 04318 Leipzig (Germany)

2015-07-15

Graphical abstract: Display Omitted - Highlights: • Foaming in biogas plants depends on the interactions between substrate and digestate. • Foaming tests enable the evaluation of substrate foaming tendency in biogas plants. • Leipzig foam tester enables foaming tests of substrates prior to use. - Abstract: Excessive foaming in anaerobic digestion occurs at many biogas plants and can cause problems including plugged gas pipes. Unfortunately, the majority of biogas plant operators are unable to identify the causes of foaming in their biogas reactor. The occurrence of foaming is often related to the chemical composition of substrates fed to the reactor. The consistency of the digestate itself is also a crucial part of the foam formation process. Thus, no specific recommendations concerning substrates can be given in order to prevent foam formation in biogas plants. The safest way to avoid foaming is to test the foaming tendency of substrates on-site. A possible solution is offered by an innovative foaming test. With the help of this tool, biogas plant operators can evaluate the foaming disposition of new substrates prior to use in order to adjust the composition of substrate mixes.

On the possibility of using biological toxicity tests to monitor the work of wastewater treatment plants

Directory of Open Access Journals (Sweden)

Zorić Jelena

2008-01-01

Full Text Available The aim of this study was to ascertain the possibility of using biological toxicity tests to monitor influent and effluent wastewaters of wastewater treatment plants. The information obtained through these tests is used to prevent toxic pollutants from entering wastewater treatment plants and discharge of toxic pollutants into the recipient. Samples of wastewaters from the wastewater treatment plants of Kragujevac and Gornji Milanovac, as well as from the Lepenica and Despotovica Rivers immediately before and after the influx of wastewaters from the plants, were collected between October 2004 and June 2005. Used as the test organism in these tests was the zebrafish Brachydanio rerio Hamilton - Buchanon (Cyprinidae. The acute toxicity test of 96/h duration showed that the tested samples had a slight acutely toxic effect on B. rerio, except for the sample of influent wastewater into the Cvetojevac wastewater treatment plant, which had moderately acute toxicity, indicating that such water should be prevented from entering the system in order to eliminate its detrimental effect on the purification process.

Adaptive Management Plan for Sensitive Plant Species on the Nevada Test Site

International Nuclear Information System (INIS)

Wills, C. A.

2001-01-01

The Nevada Test Site supports numerous plant species considered sensitive because of their past or present status under the Endangered Species Act and with federal and state agencies. In 1998, the U.S. Department of Energy, Nevada Operation Office (DOE/NV) prepared a Resource Management Plan which commits to protects and conserve these sensitive plant species and to minimize accumulative impacts to them. This document presents the procedures of a long-term adaptive management plan which is meant to ensure that these goals are met. It identifies the parameters that are measured for all sensitive plant populations during long-term monitoring and the adaptive management actions which may be taken if significant threats to these populations are detected. This plan does not, however, identify the current list of sensitive plant species know to occur on the Nevada Test Site. The current species list and progress on their monitoring is reported annually by DOE/NV in the Resource Management Plan

Benchmark Analysis for Condition Monitoring Test Techniques of Aged Low Voltage Cables in Nuclear Power Plants. Final Results of a Coordinated Research Project

International Nuclear Information System (INIS)

2017-10-01

used for in situ testing of installed cables while a nuclear power plant is operating.The results of these benchmark tests were then compared to identify the best condition monitoring methods and establish recommendations for improvements. The conclusions of the data analysis provided insight into condition monitoring techniques which yield usable or traceable results

Full scale dynamic testing of Paks nuclear power plant structures

International Nuclear Information System (INIS)

Da Rin, E.M.

1995-01-01

This report refers to the full-scale dynamic structural testing activities that have been performed in December 1994 at the Paks (H) Nuclear Power Plant, within the framework of: the IAEA Coordinated research Programme 'Benchmark Study for the Seismic Analysis and Testing of WWER-type Nuclear Power Plants, and the nuclear research activities of ENEL-WR/YDN, the Italian National Electricity Board in Rome. The specific objective of the conducted investigation was to obtain valid data on the dynamic behaviour of the plant's major constructions, under normal operating conditions, for enabling an assessment of their actual seismic safety to be made. As described in more detail hereafter, the Paks NPP site has been subjected to low level earthquake like ground shaking, through appropriately devised underground explosions, and the dynamic response of the plant's 1 st reactor unit important structures was appropriately measured and digitally recorded. In-situ free field response was measured concurrently and, moreover, site-specific geophysical and seismological data were simultaneously acquired too. The above-said experimental data is to provide basic information on the geophysical and seismological characteristics of the Paks NPP site, together with useful reference information on the true dynamic characteristics of its main structures and give some indications on the actual dynamic soil-structure interaction effects for the case of low level excitation

FY 1991 report on the results of the development of an entrained bed coal gasification power plant. Part 2. Support study for the development of an entrained bed coal gasification power plant (Summarization of the operation study); 1991 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu (Funryusho sekitan gaska hatsuden plant kaihatsu no shien kenkyu) - Sono 2. Unten kenkyu sokatsu hen

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-04-01

The paper summarized the results of the following tests on a 40 t/d pilot plant which were conducted as the support study for the development of an entrained bed coal gasification power plant: dry desulfurization test, dry dedusting test, high-temperature NOx gas combustion test (gas turbine element test). In the dry desulfurization test, tested were the operational automation and monitoring technology, performance of desulfurization, characteristics of load response, sulfur recovery, etc. In the dry dedusting test, tested were the filter medium/dust separation, powder level meter, continuous dust densitometer, operational automation, characteristics of partial load/load response, improvement of filtration materials, test on characteristics of air conveyance, etc. In the high-temperature low-NOx gas combustion test (gas turbine element test), the following were conducted: examination of the test method of gas turbine materials, inspection/examination of the corrosion of gas turbine combustion test device, inspection/examination of the internal corrosion of the stationary blade for gas turbine deposit test, inspection/examination of gasification system. The results of these support studies were reflected one after another in the project on the development of an entrained bed coal gasification power plant. (NEDO)

Assessment of engineering plant analyzer with Peach Bottom 2 stability tests

International Nuclear Information System (INIS)

Rohatgi, U.S.; Mallen, A.N.; Cheng, H.S.; Wulff, W.

1992-01-01

Engineering Plant Analyzer (EPA) has been developed to simulate plant transients for Boiling Water Reactor (BWR). Recently, this code has been used to simulate LaSalle-2 instability event which was initiated by a failure in the feed water heater. The simulation was performed for the scram conditions and for the postulated failure in the scram. In order to assess the capability of the EPA to simulate oscillatory flows as observed in the LaSalle event, EPA has been benchmarked with the available data from the Peach Bottom 2 (PB2) Instability tests PT1, PT2, and PT4. This document provides a description of these tests

On-line testing of calibration of process instrumentation channels in nuclear power plants. Phase 2, Final report

International Nuclear Information System (INIS)

Hashemian, H.M.

1995-11-01

The nuclear industry is interested in automating the calibration of process instrumentation channels; this report provides key results of one of the sponsored projects to determine the validity of automated calibrations. Conclusion is that the normal outputs of instrument channels in nuclear plants can be monitored over a fuel cycle while the plant is operating to determine calibration drift in the field sensors and associated signal conversion and signal conditioning equipment. The procedure for on-line calibration tests involving calculating the deviation of each instrument channel from the best estimate of the process parameter that the instrument is measuring. Methods were evaluated for determining the best estimate. Deviation of each signal from the best estimate is updated frequently while the plant is operating and plotted vs time for entire fuel cycle, thereby providing time history plots that can reveal channel drift and other anomalies. Any instrument channel that exceeds allowable drift or channel accuracy band is then scheduled for calibration during a refueling outage or sooner. This provides calibration test results at the process operating point, one of the most critical points of the channel operation. This should suffice for most narrow-range instruments, although the calibration of some instruments can be verified at other points throughout their range. It should be pointed out that the calibration of some process signals such as the high pressure coolant injection flow in BWRs, which are normally off- scale during plant operation, can not be tested on-line

Utilization of aging program results in plant inspections

International Nuclear Information System (INIS)

Gunther, W.; Fullwood, R.

1989-01-01

Research conducted under the auspices of the US Nuclear Regulatory Commission Nuclear Plant Aging Research (NPAR) Program has resulted in a large data base of component and system operating experience. This data base has been used to determine equipment aging susceptibility and the potential for equipment aging to impact plant safety and reliability. Methods of detecting and mitigating component and system aging have also been identified. This paper discusses how the NPAR results could be used to focus inspection activities on age-sensitive components and systems and on the specific modes and mechanisms of age degradation. These activities range from the regular inspections conducted by resident inspectors to extensive special inspections such as the Safety System Functional Inspection typically conducted by a team of inspectors. 5 refs., 3 figs., 1 tab

Utilization of aging program results in plant inspections

International Nuclear Information System (INIS)

Gunther, W.; Fullwood, R.

1988-01-01

Research conducted under the auspices of the US Nuclear Regulatory Commission Nuclear Plant Aging Research (NPAR) Program has resulted in a large data base of component and system operating experience. This data base has been used to determine equipment aging susceptibility and the potential for equipment aging to impact plant safety and reliability. Methods of detecting and mitigating component and system aging have also been identified. This paper discusses how the NPAR results could be used to focus inspection activities on age-sensitive components and systems and on the specific modes and mechanisms of age degradation. These activities range from the regular inspections conducted by resident inspectors to extensive special inspections such as the Safety System Functional Inspection typically conducted by a team of inspectors. 5 refs., 3 figs., 1 tab

Lessons learned from full-scale vibration tests on nuclear power plant auxiliary structure in Switzerland

International Nuclear Information System (INIS)

Berger, E.; Tinic, S.

1988-01-01

The Beznau Nuclear Power Plant is located in northern Switzerland. The plant is owned and operated by the Nordostschweizerische Kraftwerke AG (NOK) in Baden, Switzerland. It is a twin unit plant (2 x 350 MWe) which was designed in the early 1960's and placed into commercial operation between 1969 and 1971. In connection with a major backfit project, which will improve the safety of the plant against external events, the free-standing boric water tanks had to be relocated and were replaced by two boric water tanks in a new building (the so called BOTA-building). It enabled to plan and perform full scale vibration tests.The scope of experimental investigation was to determine the eigenfrequencies and damping values for fundamental soil-structure interaction. The vibration tests allowed identification of the important modes of the soil-structure system in the range 3 to 15 Hz. The excitation was strung enough to generate accelerations in the structure comparable to those of a small earthquake. From the comparisons of computed and measured results it is concluded that the rocking frequency can be reasonably well predicted by either Finite Element or Lumped Parameter models with springs simulating the soil-foundation stiffness, provided in the case of the latter the embedment is taken into account. The prediction of the amplitude of structural response appears to be more difficult, as shown by the differences in the mode shapes. In the frequency range 8 to 10 Hz the agreement between computed and test results was less satisfactory. The actual structural behaviour turned out to be more complex than expected and needs further investigation with the aid of more refined models for the soil-structure system

Lessons learned from full-scale vibration tests on nuclear power plant auxiliary structure in Switzerland

Energy Technology Data Exchange (ETDEWEB)

Berger, E [Basler and Hofmann AG, Consulting Engineers, Zurich (Switzerland); Tinic, S [Nordostschweizerische Kraftwerke AG, Baden (Switzerland)

1988-07-01

The Beznau Nuclear Power Plant is located in northern Switzerland. The plant is owned and operated by the Nordostschweizerische Kraftwerke AG (NOK) in Baden, Switzerland. It is a twin unit plant (2 x 350 MWe) which was designed in the early 1960's and placed into commercial operation between 1969 and 1971. In connection with a major backfit project, which will improve the safety of the plant against external events, the free-standing boric water tanks had to be relocated and were replaced by two boric water tanks in a new building (the so called BOTA-building). It enabled to plan and perform full scale vibration tests.The scope of experimental investigation was to determine the eigenfrequencies and damping values for fundamental soil-structure interaction. The vibration tests allowed identification of the important modes of the soil-structure system in the range 3 to 15 Hz. The excitation was strung enough to generate accelerations in the structure comparable to those of a small earthquake. From the comparisons of computed and measured results it is concluded that the rocking frequency can be reasonably well predicted by either Finite Element or Lumped Parameter models with springs simulating the soil-foundation stiffness, provided in the case of the latter the embedment is taken into account. The prediction of the amplitude of structural response appears to be more difficult, as shown by the differences in the mode shapes. In the frequency range 8 to 10 Hz the agreement between computed and test results was less satisfactory. The actual structural behaviour turned out to be more complex than expected and needs further investigation with the aid of more refined models for the soil-structure system.

Survival results of a biomass planting in the Missouri River floodplain

Science.gov (United States)

W. D. ' Dusty' Walter; John P. Dwyer

2003-01-01

A factor essential to successful tree planting in unprotected floodplain environments is survival. Two-year survival results from tree planting in an unprotected floodplain adjacent to the Missouri River are presented. Species planted included silver maple, locally collected cottonwood, and a superior cottonwood selection from Westvaco Corporation. Two spacings, 4 x 4...

Numerical forecast test on local wind fields at Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Chen Xiaoqiu

2005-01-01

Non-hydrostatic, full compressible atmospheric dynamics model is applied to perform numerical forecast test on local wind fields at Qinshan nuclear power plant, and prognostic data are compared with observed data for wind fields. The results show that the prognostic of wind speeds is better than that of wind directions as compared with observed results. As the whole, the results of prognostic wind field are consistent with meteorological observation data, 54% of wind speeds are within a factor of 1.5, about 61% of the deviation of wind direction within the 1.5 azimuth (≤33.75 degrees) in the first six hours. (authors)

Acute and chronic toxicity testing of bisphenol A with aquatic invertebrates and plants.

Science.gov (United States)

Mihaich, Ellen M; Friederich, Urs; Caspers, Norbert; Hall, A Tilghman; Klecka, Gary M; Dimond, Stephen S; Staples, Charles A; Ortego, Lisa S; Hentges, Steven G

2009-07-01

Bisphenol A (BPA, 4,4'-isopropylidine diphenol) is a commercially important chemical used primarily as an intermediate in the production of polycarbonate plastic and epoxy resins. Extensive effect data are currently available, including long-term studies with BPA on fish, amphibians, crustaceans, and mollusks. The aim of this study was to perform additional tests with a number of aquatic invertebrates and an aquatic plant. These studies include acute tests with the midge (Chironomus tentans) and the snail (Marisa cornuarietis), and chronic studies with rotifers (Brachionus calyciflorus), amphipods (Hyalella azteca), and plants (Lemna gibba). The effect data on different aquatic invertebrate and plant species presented in this paper correspond well with the effect and no-effect concentrations (NOECs) available from invertebrate studies in the published literature and are within the range found for other aquatic species tested with BPA.

Most experiments done so far with limited plants. Large-scale testing ...

Indian Academy of Sciences (India)

First page Back Continue Last page Graphics. Most experiments done so far with limited plants. Large-scale testing needs to be done with objectives such as: Apart from primary transformants, their progenies must be tested. Experiments on segregation, production of homozygous lines, analysis of expression levels in ...

Creep tests on clean and argillaceous salt from the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Mellegard, K.D.; Pfeifle, T.W.

1993-05-01

Fifteen triaxial compression creep tests were performed on clean and argillaceous salt from the Waste Isolation Pilot Plant (WIPP). The temperatures in the tests were either 25 degrees C or 100 degrees C while the stress difference ranged from 3.5 MPa to 21.0 MPa. In all tests, the confining pressure was 15 MPa. Test duration ranged from 23 to 613 days with an average duration of 300 days. The results of the creep tests supplemented earlier testing and were used to estimate two parameters in the Modified Munson-Dawson constitutive law for the creep behavior of salt. The two parameters determined from each test were the steady-state strain rate and the transient strain limit. These estimates were combined with parameter estimates determined from previous testing to study the dependence of both transient and steady-state creep deformation on stress difference. The exponents on stress difference determined in this study were found to be consistent with revised estimates of the exponents reported by other investigators

Testing of multistep soil washing for radiocesium-contaminated soil containing plant matter

International Nuclear Information System (INIS)

Funakawa, Masafumi; Tagawa, Akihiro; Okuda, Nobuyasu

2012-01-01

Decontamination work following radiocesium exposure requires a vast reduction in the amount of contaminated soil generated. The current study subjected 4 types of contaminated soil with different properties to multistep soil washing under the same conditions. This study also determined the effectiveness of radiocesium decontamination and the extent to which the amount of contaminated soil was reduced. In addition, the effectiveness of plant matter separation, adsorbent addition, and grinding as part of multistep soil washing was determined using the same contaminated soil. Results of testing indicated that the rate of radiocesium decontamination ranged from 73.6 to 89.2% and the recovery rate ranged from 51.5 to 84.2% for twice-treated soil, regardless of the soil properties or cesium level. Plant matter in soil had a high radiocesium level. However, there was little plant matter in our soil sample. Therefore, plant matter separation had little effect on the improvement in the percentage of radiocesium decontamination of twice-treated soil. Soil surface grinding improved the rate of radiocesium decontamination of twice-treated soil. However, radiocesium in soil tightly bound with minerals in the soil; thus, the addition of an adsorbent also failed to improve the rate of radiocesium decontamination. (author)

Results of a nuclear power plant Application of a new technique for human error analysis (ATHEANA)

International Nuclear Information System (INIS)

Forester, J.A.; Whitehead, D.W.; Kolaczkowski, A.M.; Thompson, C.M.

1997-01-01

A new method to analyze human errors has been demonstrated at a pressurized water reactor (PWR) nuclear power plant. This was the first application of the new method referred to as A Technique for Human Error Analysis (ATHEANA). The main goals of the demonstration were to test the ATHEANA process as described in the frame-of-reference manual and the implementation guideline, test a training package developed for the method, test the hypothesis that plant operators and trainers have significant insight into the error-forcing-contexts (EFCs) that can make unsafe actions (UAs) more likely, and to identify ways to improve the method and its documentation. A set of criteria to evaluate the open-quotes successclose quotes of the ATHEANA method as used in the demonstration was identified. A human reliability analysis (HRA) team was formed that consisted of an expert in probabilistic risk assessment (PRA) with some background in HRA (not ATHEANA) and four personnel from the nuclear power plant. Personnel from the plant included two individuals from their PRA staff and two individuals from their training staff. Both individuals from training are currently licensed operators and one of them was a senior reactor operator open-quotes on shiftclose quotes until a few months before the demonstration. The demonstration was conducted over a 5 month period and was observed by members of the Nuclear Regulatory Commission's ATHEANA development team, who also served as consultants to the HRA team when necessary. Example results of the demonstration to date, including identified human failure events (HFEs), UAs, and EFCs are discussed. Also addressed is how simulator exercises are used in the ATHEANA demonstration project

The Low Temperature CFB Gasifier - Further Test Results and Possible Applications

DEFF Research Database (Denmark)

Stoholm, P.; Nielsen, Rasmus Glar; Sarbæk, L.

2002-01-01

located in the CFB particle re-circulation path. The 50 kW test plant was built and commissioned during 1999 and since then experiences have been gained from more than 80 hours of operation. Nearly all of the test work has been performed on fuel that is expected to be worst-case conditions, i.e. the fuel...

FY 1992 report on the results of the development of an entrained bed coal gasification power plant. Part 1. Element study/Technical survey; 1992 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu - Sono 1. Youso kenkyu hen, gijutsu chosa hen

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-02-01

For the purpose of establishing the technology of the integrated coal gasification combined cycle power generation, element study of a 200 t/d entrained bed coal gasification pilot plant was made, and the FY 1992 results were summarized. In the gasification test using a 2 t/d furnace equipment, study was made of the extraction of subjects on the operation, performance of gasification, slagging characteristics and characteristics of flux addition. In the study of slag utilization technology, chemical analysis, test on fine aggregate and test on fine particle were carried out to study applicability of the slag discharged from the entrained bed coal gasification power plant to various materials. In the study of a large gas turbine combustor for the demonstrative machine, the demonstrative machine use large gas turbine combustor testing equipment was installed at actual pressure/actual size combustion testing facilities in the pilot plant and further the test use combustor was integrated into them. By using them, the real gas combustion test was made using the adjusted coal for evaluation of combustion performance of the test use combustor. In the simulation study of the total pilot plant system, the comparative study was made between the data on the test using the actual machine and the results of the simulation. (NEDO)

Comparative ethnobotany and in-the-field antibacterial testing of medicinal plants used by the Bulu and inland Kaulong of Papua New Guinea.

Science.gov (United States)

Prescott, Thomas A K; Kiapranis, Robert; Maciver, Sutherland K

2012-01-31

The island of New Britain in Papua New Guinea is an area of great floristic and cultural diversity that has received little attention from ethnobotanists. Here we present the results of a comparative medicinal ethnobotanical survey of the Bulu and inland Kaulong; two distinct people groups inhabiting lowland rainforest on different sides of the island. A high proportion of species are used in the treatment of bacterial infections and plants with antibacterial activity were identified in the field using a specially developed antibacterial assay kit. Follow up testing with human pathogens was used to evaluate active plant material in more detail. Rapid appraisal techniques were used to survey both people groups with all data corroborated by three or more separate sources. Plants from both groups were tested in-the-field with a portable antibacterial test kit based on the agar diffusion assay, using a pressure cooker to sterilise glassware and media. Follow up laboratory based tests were carried out using standardised agar dilution protocols for drug resistant and drug sensitive strains of Staphylococcus aureus and Streptococcus pneumoniae. We find surprisingly little overlap in the plant species used by the two people groups with only 1 out of 70 species used for the same purpose. There is also a difference in emphasis in the conditions treated with 53% of Kaulong medicinal plants dedicated to treating tropical ulcers compared with only 8% of in the Bulu group. In-the-field testing identified Garcinia dulcis bark (a Kaulong tropical ulcer treatment) to have antibacterial activity and follow up tests against a drug resistant strain of Staphylococcus aureus (a pathogen implicated in tropical ulcer pathogenesis) revealed the crude bark extract to be potently active with an MIC of just 1 mg/ml. The results demonstrate extreme differences in medicinal plant use between two people groups living a mere 100 km apart and suggests the two medicinal plant systems have developed

Implementation and test of proposals to integrate human factors in reporting and causal analysis in nuclear power plants

International Nuclear Information System (INIS)

Wilpert; Maimer, H.; Miller, R.; Fahlbruch, B.; Leiber, I.; Szameitat, S.; Baggen, R.; Gans, A.; Becker, G.

1998-01-01

The research project 'Implementation and Test of Proposals to integrate Human Factors in Reporting and Causal Analysis in Nuclear Power Plants' ('Implementation and Test', SR 2039/8) is based on two antecedent projects: 'Reporting System' (SR 2039/1) and 'Causal Analysis' (SR 2039/2). The project 'Implementation and Test' conducted various tests and introductory programs in cooperation with different target groups concerning the event analysis methodology 'SOL - Safety through Organizational Learning': Regulators, consultant organizations, union/works councillors and utilities. Thus, SOL was concurrently optimized and [apted for the practice in the German nuclear power industry. SOL was also validated in a German nuclear power plant using a concrete event. Results of the 'Implementation and Test' project demonstrate that SOL is fit to conduct event analyses practicably and economically with appropriate comprehensiveness and depth. SOL facilitates the identification of relevant contributing factors of events. This report concludes with various concrete proposals for the further development of the Program of the Federal Ministry of Environment, Nature Protection and Reactor Safety (BMU) and the Federal Agency of R[iation Protection (BfS) concerning 'The Contribution of Humans to Safety of Nuclear Power Plants'. (orig.) [de

Guidelines for inservice testing at nuclear power plants

International Nuclear Information System (INIS)

Campbell, P.

1995-04-01

The staff of the U.S. Nuclear Regulatory Commission (NRC) gives licensees guidelines and recommendations for developing and implementing programs for the inservice testing of pumps and valves at commercial nuclear power plants. The staff discusses the regulations; the components to be included in an inservice testing program; and the preparation and content of cold shutdown justifications, refueling outage justifications, and requests for relief from the American Society of Mechanical Engineers Code requirements. The staff also gives specific guidance on relief acceptable to the NRC and advises licensees in the use of this information at their facilities. The staff discusses the revised standard technical specifications for the inservice testing program requirements and gives guidance on the process a licensee may follow upon finding an instance of noncompliance with the Code

Sensitivity analysis of FDS 6 results for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Alvear, Daniel; Puente, Eduardo; Abreu, Orlando [Cantabria Univ., Santander (Spain). Group GIDAI - Fire Safety-Research and Technology; Peco, Julian [Consejo de Seguridad Nuclear, Madrid (Spain)

2015-12-15

The Spanish standard ''Instruction IS-30, Rev. 1'' (February 21, 2013) allows the new approaches of risk informed performance based design (PBD) The Spanish standard ''Instruction IS-30, rev. 1'' (February 21, 2013) for demonstrating the safe shutdown capability in case of fire in nuclear power plants. In this sense, fire computer models have become an interesting tool to study real fire scenarios. Such models use a set of input parameters that define the features of the physical domain, material, radiation, turbulence, etc. This paper analyses the impact of the grid size and different sub-models of the fire simulation code FDS, version 6 with the objective to evaluate and define their relative weight in the final simulation results. For the grid size analysis, two different scale scenarios were selected, the bench scale test PENLIGHT and a large-scale test similar to Appendix B of NUREG - 1934 (17 m x 10 m x 4.6 m, with an ignition source of 2 MW and 16 cable trays). For the sub-model analysis, the PRS-INT4 real scale configuration of the INTEGRAL experimental campaign of the international OECD PRISME Project has been used. The results offer relevant data for users and show the critical parameters that must be selected properly to guarantee the quality of the simulations.

A comprehensive test of evolutionarily increased competitive ability in a highly invasive plant species

Science.gov (United States)

Joshi, Srijana; Gruntman, Michal; Bilton, Mark; Seifan, Merav; Tielbörger, Katja

2014-01-01

Background and Aims A common hypothesis to explain plants' invasive success is that release from natural enemies in the introduced range selects for reduced allocation to resistance traits and a subsequent increase in resources available for growth and competitive ability (evolution of increased competitive ability, EICA). However, studies that have investigated this hypothesis have been incomplete as they either did not test for all aspects of competitive ability or did not select appropriate competitors. Methods Here, the prediction of increased competitive ability was examined with the invasive plant Lythrum salicaria (purple loosestrife) in a set of common-garden experiments that addressed these aspects by carefully distinguishing between competitive effect and response of invasive and native plants, and by using both intraspecific and interspecific competition settings with a highly vigorous neighbour, Urtica dioica (stinging nettle), which occurs in both ranges. Key Results While the intraspecific competition results showed no differences in competitive effect or response between native and invasive plants, the interspecific competition experiment revealed greater competitive response and effect of invasive plants in both biomass and seed production. Conclusions The use of both intra- and interspecific competition experiments in this study revealed opposing results. While the first experiment refutes the EICA hypothesis, the second shows strong support for it, suggesting evolutionarily increased competitive ability in invasive populations of L. salicaria. It is suggested that the use of naturally co-occurring heterospecifics, rather than conspecifics, may provide a better evaluation of the possible evolutionary shift towards greater competitive ability. PMID:25301818

Irradiation effects test Series Scoping Test 1: test results report

International Nuclear Information System (INIS)

Quapp, W.J.; Allison, C.M.; Farrar, L.C.

1977-09-01

The report describes the results of the first scoping test in the Irradiation Effects Test Series conducted by the Thermal Fuels Behavior Program, which is part of the Water Reactor Research Program of EG and G Idaho, Inc. The research is sponsored by the United States Nuclear Regulatory Commission. This test used an unirradiated, three-foot-long, PWR-type fuel rod. The objective of this test was to thoroughly evaluate the remote fabrication procedures to be used for irradiated rods in future tests, handling plans, and reactor operations. Additionally, selected fuel behavior data were obtained. The fuel rod was subjected to a series of preconditioning power cycles followed by a power increase which brought the fuel rod power to about 20.4 kW/ft peak linear heat rating at a coolant mass flux of 1.83 x 10 6 lb/hr-ft 2 . Film boiling occurred for a period of 4.8 minutes following flow reductions to 9.6 x 10 5 and 7.5 x 10 5 lb/hr-ft 2 . The test fuel rod failed following reactor shutdown as a result of heavy internal and external cladding oxidation and embrittlement which occurred during the film boiling operation

Models test on dynamic structure-structure interaction of nuclear power plant buildings

International Nuclear Information System (INIS)

Kitada, Y.; Hirotani, T.

1999-01-01

A reactor building of an NPP (nuclear power plant) is generally constructed closely adjacent to a turbine building and other buildings such as the auxiliary building, and in increasing numbers of NPPs, multiple plants are being planned and constructed closely on a single site. In these situations, adjacent buildings are considered to influence each other through the soil during earthquakes and to exhibit dynamic behaviour different from that of separate buildings, because those buildings in NPP are generally heavy and massive. The dynamic interaction between buildings during earthquake through the soil is termed here as 'dynamic cross interaction (DCI)'. In order to comprehend DCI appropriately, forced vibration tests and earthquake observation are needed using closely constructed building models. Standing on this background, Nuclear Power Engineering Corporation (NUPEC) had planned the project to investigate the DCI effect in 1993 after the preceding SSI (soil-structure interaction) investigation project, 'model tests on embedment effect of reactor building'. The project consists of field and laboratory tests. The field test is being carried out using three different building construction conditions, e.g. a single reactor building to be used for the comparison purposes as for a reference, two same reactor buildings used to evaluate pure DCI effects, and two different buildings, reactor and turbine building models to evaluate DCI effects under the actual plant conditions. Forced vibration tests and earthquake observations are planned in the field test. The laboratory test is planned to evaluate basic characteristics of the DCI effects using simple soil model made of silicon rubber and structure models made of aluminum. In this test, forced vibration tests and shaking table tests are planned. The project was started in April 1994 and will be completed in March 2002. This paper describes an outline and the summary of the current status of this project. (orig.)

Testing methodology of embedded software in digital plant protection system

International Nuclear Information System (INIS)

Seong, Ah Young; Choi, Bong Joo; Lee, Na Young; Hwang, Il Soon

2001-01-01

It is necessary to assure the reliability of software in order to digitalize RPS(Reactor Protection System). Since RPS causes fatal damage on accidental cases, it is classified as Safety 1E class. Therefore we propose the effective testing methodology to assure the reliability of embedded software in the DPPS(Digital Plant Protection System). To test the embedded software effectively in DPPS, our methodology consists of two steps. The first is the re-engineering step that extracts classes from structural source program, and the second is the level of testing step which is composed of unit testing, Integration Testing and System Testing. On each testing step we test the embedded software with selected test cases after the test item identification step. If we use this testing methodology, we can test the embedded software effectively by reducing the cost and the time

Standard practice for analysis and interpretation of physics dosimetry results for test reactors

International Nuclear Information System (INIS)

Anon.

1984-01-01

This practice describes the methodology summarized in Annex Al to be used in the analysis and interpretation of physics-dosimetry results from test reactors. This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods that are in various stages of completion (see Fig. 1). Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, exposure units, and neutron spectrum adjustment methods. This practice is directed towards the development and application of physics-dosimetrymetallurgical data obtained from test reactor irradiation experiments that are performed in support of the operation, licensing, and regulation of LWR nuclear power plants. It specifically addresses the physics-dosimetry aspects of the problem. Procedures related to the analysis, interpretation, and application of both test and power reactor physics-dosimetry-metallurgy results are addressed in Practice E 853, Practice E 560, Matrix E 706(IE), Practice E 185, Matrix E 706(IG), Guide E 900, and Method E 646

Pipe damping: experimental results from laboratory tests in the seismic frequency range

International Nuclear Information System (INIS)

Ware, A.G.; Arendts, J.G.

1986-06-01

The Idaho National Engineering Laboratory (INEL) has been conducting a research program to assist the United States Nuclear Regulatory Commission (USNRC) in determining best-estimate damping values for the seismic analysis of nuclear piping systems. As part of this program, a 5-in. piping system was tested by the INEL, and data from USNRC/EPRI piping vibration tests at the ANCO Engineers facility were evaluated. These systems were subjected to various types of excitation methods and magnitudes, the support configurations were varied, and the effects of pipe insulation and internal pressure were investigated on the INEL system. The INEL has used several different methods to reduce the data to determine the damping in both these piping systems under the various test conditions. It was concluded that at representative seismic excitation levels, pressure was not a contributing factor, but the supports, insulation, and magnitude of response all were major influences contributing to damping. These tests are part of the ongoing program to determine how various parameters and data reduction methods affect piping system damping. The evaluation of all relevant test results, including these two series, will potentially lead to revised damping guidelines for the seismic analysis of nuclear plants, making them safer, less costly, and easier to inspect and maintain. The test results as well as accompanying evaluations and recommendations are presented in this report. 27 refs., 72 figs., 13 tabs

Fission Surface Power Technology Demonstration Unit Test Results

Science.gov (United States)

Briggs, Maxwell H.; Gibson, Marc A.; Geng, Steven M.; Sanzi, James L.

2016-01-01

The Fission Surface Power (FSP) Technology Demonstration Unit (TDU) is a system-level demonstration of fission power technology intended for use on manned missions to Mars. The Baseline FSP systems consists of a 190 kWt UO2 fast-spectrum reactor cooled by a primary pumped liquid metal loop. This liquid metal loop transfers heat to two intermediate liquid metal loops designed to isolate fission products in the primary loop from the balance of plant. The intermediate liquid metal loops transfer heat to four Stirling Power Conversion Units (PCU), each of which produce 12 kWe (48 kW total) and reject waste heat to two pumped water loops, which transfer the waste heat to titanium-water heat pipe radiators. The FSP TDU simulates a single leg of the baseline FSP system using an electrically heater core simulator, a single liquid metal loop, a single PCU, and a pumped water loop which rejects the waste heat to a Facility Cooling System (FCS). When operated at the nominal operating conditions (modified for low liquid metal flow) during TDU testing the PCU produced 8.9 kW of power at an efficiency of 21.7 percent resulting in a net system power of 8.1 kW and a system level efficiency of 17.2 percent. The reduction in PCU power from levels seen during electrically heated testing is the result of insufficient heat transfer from the NaK heater head to the Stirling acceptor, which could not be tested at Sunpower prior to delivery to the NASA Glenn Research Center (GRC). The maximum PCU power of 10.4 kW was achieved at the maximum liquid metal temperature of 875 K, minimum water temperature of 350 K, 1.1 kg/s liquid metal flow, 0.39 kg/s water flow, and 15.0 mm amplitude at an efficiency of 23.3 percent. This resulted in a system net power of 9.7 kW and a system efficiency of 18.7 percent.

Secretory structure and histochemistry test of some Zingiberaceae plants

Science.gov (United States)

Indriyani, Serafinah

2017-11-01

A secretory structure is a structure that produces a plant's metabolite substances. Secretory structures are grouped into an internal and external. Zingiberaceae plants are known as traditional medicine plants and as spice plants due to secretory structures in their tissues. The objective of the research were to describe the secretory structure of Zingiberaceae plants and to discover the qualitatively primary metabolite substances in plant's tissues via histochemistry test. The research was conducted by observation descriptive design, quantitative data including the density of secretory cells per mm². The quantitative data were analyzed by ANOVA and continued by Duncan at α = 5 %. The results showed that the secretory structures in leaves, rhizome, and the root of 14 species of Zingiberaceae plants are found in the mesophyll of leaves and cortex, and also pith in rhizome and roots. The type of secretory structure is internal. Within the root of Zingiber cassumunar Roxb.(bengle), Curcuma domestica Val. (kunyit), Curcuma zedoaria (Berg.) Roscoe (kunyit putih), Zingiber zerumbet (L.) J.E. Smith (lempuyang), Alpiniapurpurata K. Schum (lengkuas merah), and Curcuma aeruginosa Val. (temu ireng) were found amylum grains, while in Kaemferia galanga L. (kencur), Boesen bergiapandurata L. (temu kunci), and Curcuma xanthorrhiza Roxb. (temulawak) there were no amylum grains in the root as well as in the leaves. The roots of bengle had the greatest density of amylum grain, it had 248.1 ± 9.8 secretory cells of amylum grains per mm². Lipids (oil droplets) were found in the root of bengle, Zingiber officinale Roxb. Var. emprit (jahe emprit), Zingiber officinale Roxb. Var. Gajah (jahe gajah), Zingiber officinale Roxb. Var. Rubrum (jahe merah), Keampferia angustifolia L. (kunci pepet), kunyit, kunyit putih, lempuyang, lengkua smerah, Curcuma aeruginosa Val. (temu ireng), and Curcuma mangga Val. and van Zijp (temu mangga); the root of lempuyang had the greatest density of oil

Horizontal loading test by whole model specimen simulating inner concrete structure of PWR type nuclear power plant

International Nuclear Information System (INIS)

Furuya, Noriyuki; Sekine, Masataka; Kimura, Kozo; Yamaguchi, Yoshihiro; Yamaguchi, Tsuneo; Takeda, Toshikazu

1985-01-01

The Nuclear Power Engineering Test Center has performed a horizontal loading test by a whole model specimen simulating the inner concrete structure of a PWR type nuclear power plant in order to investigate restoring force characteristics of reactor buildings. This report describes the results of examination of applicability to the test results of analysis methods based on elastic theory. The analysis results of elastic stiffness, concrete cracking load, rebar yielding loads and ultimate strength were compared with the test results. According to this examination, it is recognized that even these analysis methods based on elastic theory are comparatively effective for analysis of an inner concrete structure of fairly complex configuration, although there are limits of the scope of applicability. (author)

Oregon state university's advanced plant experiment (APEX) AP1000 integral facility test program

International Nuclear Information System (INIS)

Reyes, J.N.; Groome, J.T.; Woods, B.G.; Young, E.; Abel, K.; Wu, Q.

2005-01-01

Oregon State University (OSU) has recently completed a three year study of the thermal hydraulic behavior of the Westinghouse AP1000 passive safety systems. Eleven Design Basis Accident (DBA) scenarios, sponsored by the U.S. Department of Energy (DOE) with technical support from Westinghouse Electric, were simulated in OSU's Advanced Plant Experiment (APEX)-1000. The OSU test program was conducted within the purview of the requirements of 10CFR50 Appendix B, NQA-1 and 10 CFR 21 and the test data was used to provide benchmarks for computer codes used in the final design approval of the AP1000. In addition to the DOE certification testing, OSU conducted eleven confirmatory tests for the U.S. Nuclear Regulatory Commission. This paper presents the test program objectives, a description of the APEX-1000 test facility and an overview of the test matrix that was conducted in support of plant certification. (authors)

Improving motor reliability in nuclear power plants: Volume 2, Functional indicator tests on a small electric motor subjected to accelerated aging

International Nuclear Information System (INIS)

Subudhi, M.; Taylor, J.H.; Lofaro, R.; Sugarman, A.C.; Sheets, M.W.; Skreiner, K.M.

1987-11-01

A ten horsepower electric motor was artificially aged by plug reverse cycling for test purposes. The motor was manufactured in 1967 and was in service at a commercial nuclear power plant for twelve years. Various tests were performed on the motor throughout the aging process. The motor failed after 3.79 million reversals (3 seconds per reversal) over seven months of testing. Each test parameter was trended to assess its suitability in monitoring aging and service wear degradation in motors. Results and conclusions are discussed relative to the applicability of the tests performed to nuclear power plant motor maintenance programs. 15 refs., 28 figs., 1 tab

Irradiation effects test series, test IE-5. Test results report

International Nuclear Information System (INIS)

Croucher, D.W.; Yackle, T.R.; Allison, C.M.; Ploger, S.A.

1978-01-01

Test IE-5, conducted in the Power Burst Facility at the Idaho National Engineering Laboratory, employed three 0.97-m long pressurized water reactor type fuel rods, fabricated from previously irradiated zircaloy-4 cladding and one similar rod fabricated from unirradiated cladding. The objectives of the test were to evaluate the influence of simulated fission products, cladding irradiation damage, and fuel rod internal pressure on pellet-cladding interaction during a power ramp and on fuel rod behavior during film boiling operation. The four rods were subjected to a preconditioning period, a power ramp to an average fuel rod peak power of 65 kW/m, and steady state operation for one hour at a coolant mass flux of 4880 kg/s-m 2 for each rod. After a flow reduction to 1800 kg/s-m 2 , film boiling occurred on one rod. Additional flow reductions to 970 kg/s-m 2 produced film boiling on the three remaining fuel rods. Maximum time in film boiling was 80s. The rod having the highest initial internal pressure (8.3 MPa) failed 10s after the onset of film boiling. A second rod failed about 90s after reactor shutdown. The report contains a description of the experiment, the test conduct, test results, and results from the preliminary postirradiation examination. Calculations using a transient fuel rod behavior code are compared with the test results

Development of a wind energy converter and investigation of its operational function. Part 4: Test setup and results of measurement

Science.gov (United States)

Armbrust, S.; Molly, J. P.

1982-12-01

Measurements made during test operations at the MODA.10 plant as well as at a 25 years old 6 kW wind energy converter are presented. The test arrangements, measurement results of both wind energy converters, and the experience gained are described.

Experimental results and thermodynamic analysis of a natural gas small scale cogeneration plant for power and refrigeration purposes

International Nuclear Information System (INIS)

Bazzo, Edson; Nacif de Carvalho, Alvaro; Matelli, José Alexandre

2013-01-01

In this work, experimental results are reported for a small scale cogeneration plant for power and refrigeration purposes. The plant includes a natural gas microturbine and an ammonia/water absorption chiller fired by steam. The system was tested under different turbine loads, steam pressures and chiller outlet temperatures. An evaluation based on the 1st and 2nd Laws of Thermodynamics was also performed. For the ambient temperature around 24 °C and microturbine at full load, the plant is able to provide 19 kW of saturated steam at 5.3 bar (161 °C), corresponding to 9.2 kW of refrigeration at −5 °C (COP = 0.44). From a 2nd law point-of-view, it was found that there is an optimal chiller outlet temperature that maximizes the chiller exergetic efficiency. As expected, the microturbine presented the highest irreversibilities, followed by the absorption chiller and the HRSG. In order to reduce the plant exergy destruction, it is recommended a new design for the HRSG and a new insulation for the exhaust pipe. -- Highlights: • A small scale cogeneration plant for power and refrigeration is proposed and analyzed. • The plant is based on a microturbine and a modified absorption chiller. • The plant is analysed based on 1st and 2nd laws of thermodynamics. • Experimental results are found for different power and refrigeration conditions. • The plant proved to be technically feasible

A comparison of geostatistically based inverse techniques for use in performance assessment analysis at the Waste Isolation Pilot Plant Site: Results from Test Case No. 1

International Nuclear Information System (INIS)

Zimmerman, D.A.; Gallegos, D.P.

1993-10-01

The groundwater flow pathway in the Culebra Dolomite aquifer at the Waste Isolation Pilot Plant (WIPP) has been identified as a potentially important pathway for radionuclide migration to the accessible environment. Consequently, uncertainties in the models used to describe flow and transport in the Culebra need to be addressed. A ''Geostatistics Test Problem'' is being developed to evaluate a number of inverse techniques that may be used for flow calculations in the WIPP performance assessment (PA). The Test Problem is actually a series of test cases, each being developed as a highly complex synthetic data set; the intent is for the ensemble of these data sets to span the range of possible conceptual models of groundwater flow at the WIPP site. The Test Problem analysis approach is to use a comparison of the probabilistic groundwater travel time (GWTT) estimates produced by each technique as the basis for the evaluation. Participants are given observations of head and transmissivity (possibly including measurement error) or other information such as drawdowns from pumping wells, and are asked to develop stochastic models of groundwater flow for the synthetic system. Cumulative distribution functions (CDFs) of groundwater flow (computed via particle tracking) are constructed using the head and transmissivity data generated through the application of each technique; one semi-analytical method generates the CDFs of groundwater flow directly. This paper describes the results from Test Case No. 1

Fuel staging tests at the Kymijaervi power plant

International Nuclear Information System (INIS)

Kivelae, M.; Rotter, H.; Virkki, J.

1990-01-01

The aim of this study was to measure nitrogen oxide (NO x ) emissions and find the methods to reduce them in plants using coal and natural gas as fuel. The tests involved were made at the Kymijaervi Power Plant, Lahti, Finland. Coal and natural gas was used alone or mixed. With natural gas when using flue gas recirculation, the NO x emission level dropped from 330 mg/m 3 down to 60 mg/m 3 . A negative side effect was that the flue gas temperature increased. At coal combustion and staged combustion, the flue gas recirculation had no significant effect on the NO x emission level. At coal combustion, the staging of combustion air halved the NO x emission but the combustibles increased strongly. With fuel staging, using coal as main fuel and gas as staging fuel, the NO x emission level was decreased from 340 mg/m 3 to 170 mg/m 3 . At the same time the combustibles increased 2 %- units. Also the flue gas temperature increased a little. At the tests, the proportion of natural gas was rather high, one third of the fuel energy input, but it could not be decreased, because the gas flow ratio was already too low to ensure good mixing

Culinary Spice Plants in Dietary Supplement Products and Tested in Clinical Trials123

Science.gov (United States)

Saldanha, Leila G; Dwyer, Johanna T; Betz, Joseph M

2016-01-01

Dried plant parts used as culinary spices (CSs) in food are permitted as dietary ingredients in dietary supplements (DSs) within certain constraints in the United States. We reviewed the amounts, forms, and nutritional support (structure/function) claims of DSs that contain CS plants listed in the Dietary Supplement Label Database (DSLD) and compared this label information with trial doses and health endpoints for CS plants that were the subject of clinical trials listed in clinicaltrials.gov. According to the DSLD, the CS plants occurring most frequently in DSs were cayenne, cinnamon, garlic, ginger, pepper, rosemary, and turmeric. Identifying the botanical species, categorizing the forms used, and determining the amounts from the information provided on DS labels was challenging. CS plants were typically added as a component of a blend, as the powered biomass, dried extracts, and isolated phytochemicals. The amounts added were declared on about 55% of the labels, rendering it difficult to determine the amount of the CS plant used in many DSs. Clinicaltrials.gov provided little information about the composition of test articles in the intervention studies. When plant names were listed on DS labels and in clinical trials, generally the common name and not the Latin binomial name was given. In order to arrive at exposure estimates and enable researchers to reproduce clinical trials, the Latin binomial name, form, and amount of the CS plant used in DSs and tested in clinical trials must be specified. PMID:26980817

Power-hardware-in-the-loop test of VSC-HVDC connection for off-shore wind power plants

Energy Technology Data Exchange (ETDEWEB)

Sharma, Ranjan [Siemens Wind Power A/S, Brande (Denmark); Technical Univ. of Denmark (Denmark). Center for Electric Technology; Cha, Seung T.; Wu, Qiuwei; Rasmussen, Tonny W.; Oestergaard, Jacob [Technical Univ. of Denmark (Denmark). Center for Electric Technology; Jensen, Kim H. [Siemens Wind Power A/S, Brande (Denmark)

2011-07-01

This paper present a power-hardware-in-the-loop (PHIL) test of an off-shore wind power plant (WPP) interconnected to the on-shore grid via a VSC-HVDC connection. The intention of the PHIL test is to verify the hardware interaction and the control co-ordination between the plant side VSC of the HVDC system and the wind turbines within the WPP in order to ensure smooth operation of the WPP under both normal and fault operating condition. The PHIL test platform is comprised of a real time digital simulator (RTDS), a Spitzenberger Spies three phase 7,5 kW power amplifier, a purpose built VSC and a DC chopper. The WPP is simulated in the RTDS as a single full-scale wind turbine. The simulated WPP interacts with the WPP side VSC through the power amplifier. The interface between the RTDS and the power amplifier is done via an analogue GTAO I/O card of the RTDS and the input channel of the amplifier. The amplifier scales up the voltages at the point of connection of the WPP in the RTDS to the voltage level for the WPP side VSC. The WPP side VSC converter is equipped with a DC chopper. The test results show the successful control coordination between the WPP and the plant side VSC converter of the HVDC connection of the WPP. (orig.)

Development of testing system for the thermo-mechanical fatigue crack analysis of nuclear power plant pipes

International Nuclear Information System (INIS)

Lee, Ho Jin; Kim, Maan Won; Lee, Bong Sang

2003-12-01

Fatigue crack growth analysis plays an important role in the structural integrity assessment or the service life calculation of the nuclear power plant pipes. To obtain the material properties as a basic data to achieve an accurate crack growth analysis, a lot of tests and numerical crack growth simulations have been done for decades. The BS 7910 or the ASME Boiler and Pressure Vessel Code Section XI, generally used to evaluate crack growth behavior, were made under the based on simple stress states or at the evaluated isothermal temperature. It is well known that the ASME code could sometimes give so conservative results in some cases of which the cracked components are experiencing with cyclic thermal shock. In this report, we suggested a method for the life assessment of a crack embedded in nuclear power plant pipes under the thermal-mechanical fatigue loads. We here use the numerical method to get the temperature history for thermal- mechanical fatigue crack growth test. And then we can calculate the remaining life time of the pipe by using the fracture mechanics and the test results together. For this purpose, we constructed a thermal-mechanical fatigue crack growth testing system. We also gave a lot of review about recent researches in the experimental field of thermal-mechanical fatigue analysis

Summary revaluation of cold testing of the first block of nuclear power plant Mochovce; Suhrnne zhodnotenie neaktivneho vyskusania 1. bloku jadrovej elektrarne Mochovce

Energy Technology Data Exchange (ETDEWEB)

Miskolci, M; Sarvaic, I [Nuclear Power Plants Research Institute Trnava, Inc., Okruzna 5, 918 64 Trnava (Slovakia)

1998-04-03

The document contents summary revaluation of the stage of cold testing of the first unit of nuclear power plant Mochovce. The valuation is processed in individual systems with safety significance. The process and results of system testing and their conclusions for the block readiness for active testing are summarized in the document. The valuation has been elaborated by a scientific management for start-up of nuclear power plant Mochovce as an independent conductor assistance for activation check from the nuclear safety point of view. The valuation of the activation results of systems in the first unit of nuclear power plant was processed as of 15.3.1998 61 figs., 44 tabs., `Translation to english is available from OMEGA INFO, Vysehradska 33, SK-84215 Bratislava, Slovak Republic, E-Mail: kuruc at fns.uniba.sk`

Methods for Quantifying the Uncertainties of LSIT Test Parameters, Test Results, and Full-Scale Mixing Performance Using Models Developed from Scaled Test Data

Energy Technology Data Exchange (ETDEWEB)

Piepel, Gregory F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cooley, Scott K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kuhn, William L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rector, David R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Heredia-Langner, Alejandro [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-05-01

This report discusses the statistical methods for quantifying uncertainties in 1) test responses and other parameters in the Large Scale Integrated Testing (LSIT), and 2) estimates of coefficients and predictions of mixing performance from models that relate test responses to test parameters. Testing at a larger scale has been committed to by Bechtel National, Inc. and the U.S. Department of Energy (DOE) to “address uncertainties and increase confidence in the projected, full-scale mixing performance and operations” in the Waste Treatment and Immobilization Plant (WTP).

A nuclear power plant certification test plan and checklist

International Nuclear Information System (INIS)

Halverson, S.M.

1989-01-01

Regulations within the nuclear industry are requiring that all reference plant simulators be certified prior to or during 1991. A certification test plan is essential to ensure that this goal is met. A description of each step in the certification process is provided in this paper, along with a checklist to help ensure completion of each item

Decontamination and decommissioning of the EBR-I complex. Topical report No. 3. NAK disposal pilot plant test

International Nuclear Information System (INIS)

Commander, J.C.; Lewis, L.; Hammer, R.

1975-06-01

Decontamination and decommissioning of the Experimental Breeder Reactor No. 1 (EBR-I) requires processing of the primary coolant, an eutectic solution of sodium and potassium (NaK), remaining in the EBR-I primary and secondary coolant systems. While developing design criteria for the NaK processing system, reasonable justification was provided for the development of a pilot test plant for field testing some of the process concepts and proposed hardware. The objective of this activity was to prove the process concept on a low-cost, small-scale test bed. The pilot test plant criteria provided a general description of the test including: the purpose, location, description of test equipment available, waste disposal requirements, and a flow diagram and conceptual equipment layout. The pilot plant test operations procedure provided a detailed step-by-step procedure for operation of the pilot plant to obtain the desired test data and operational experience. It also spelled out the safety precautions to be used by operating personnel, including the requirement for alkali metals training certification, use of protective clothing, availability of fire protection equipment, and caustic handling procedures. The pilot plant test was performed on May 16, 1974. During the test, 32.5 gallons or 240 lb of NaK was successfully converted to caustic by reaction with water in a caustic solution. (auth)

Radioactive Waste Treatment and Conditioning Using Plasma Technology Pilot Plant: Testing and Commissioning

International Nuclear Information System (INIS)

Rafizi Salihuddin; Rohyiza Baan; Norasalwa Zakaria

2016-01-01

Plasma pilot plant was commissioned for research and development program on radioactive waste treatment. The plant is equipped with a 50 kW direct current of non-transferred arc plasma torch which mounted vertically on top of the combustion chamber. The plant also consists of a dual function chamber, a water cooling system, a compress air supply system and a control system. This paper devoted the outcome after testing and commissioning of the plant. The problems arise was discussed in order to find the possible suggestion to overcome the issues. (author)

First Biogenic VOC Flux Results from the UCI Fluxtron Plant Chamber Facility

Science.gov (United States)

Seco, R.; Gu, D.; Joo, E.; Nagalingam, S.; Aristizabal, B. H.; Basu, C.; Kim, S.; Guenther, A. B.

2017-12-01

Atmospheric biogenic volatile organic compounds (BVOCs) have key environmental, ecological and biological roles, and can influence atmospheric chemistry, secondary aerosol formation, and regional climate. Quantifying BVOC emission rates and their impact on atmospheric chemistry is one of the greatest challenges with respect to predicting future air pollution in the context of a changing climate. A new facility, the UCI Fluxtron, has been developed at the Department of Earth System Science at the University of California Irvine to study the response of BVOC emissions to extreme weather and pollution stress. The UCI Fluxtron is designed for automated, continuous measurement of plant physiology and multi-modal BVOC chemical analysis from multiple plants. It consists of two controlled-environment walk-in growth chambers that contain several plant enclosures, a gas make-up system to precisely control the composition (e.g., H2O, CO2, O3 and VOC concentrations) of the air entering each enclosure. A sample manifold with automated inlet switching is used for measurements with in-situ and real-time VOC analysis instruments: H2O, CO2 fluxes can be measured continually with an infrared gas analyzer (IRGA) and BVOCs with a proton transfer reaction -time of flight- mass spectrometer (PTR-TOF-MS). Offline samples can also be taken via adsorbent cartridges to be analyzed in a thermal desorption gas chromatograph coupled to a TOF-MS detector. We present the first results of H2O, CO2 and BVOC fluxes, including the characterization and testing of the Fluxtron system. For example, measurements of young dragon tree (Paulownia elongata) individuals using whole-plant enclosures.

Creep property testing of energy power plant component material

International Nuclear Information System (INIS)

Nitiswati, Sri; Histori; Triyadi, Ari; Haryanto, Mudi

1999-01-01

Creep testing of SA213 T12 boiler piping material from fossil plant, Suralaya has been done. The aim of the testing is to know the creep behaviour of SA213 T12 boiler piping material which has been used more than 10 yeas, what is the material still followed ideal creep curve (there are primary stage, secondary stage, and tertiary stage). This possibility could happened because the material which has been used more than 10 years usually will be through ageing process because corrosion. The testing was conducted in 520 0C, with variety load between 4% until 50% maximum allowable load based on strength of the material in 520 0C

Effects of 60 Hz electromagnetic fields on early growth in three plant species and a replication of previous results

Energy Technology Data Exchange (ETDEWEB)

Davis, M.S. [Univ. of Sunderland (United Kingdom). Ecology Centre

1996-05-01

In an attempt to replicate the findings of Smith et al., seeds of Raphanus sativus L. (radish), Sinapsis alba L. (mustard), and Hordeum vulgare L. (barley) were grown for between 9 and 21 days in continuous electromagnetic fields (EMFs) at ion-cyclotron resonance conditions for stimulation of Ca{sup 2+} (B{sub H} = 78.3 {micro}T, B{sub HAC} = 40 {micro}T peak-peak at 60 Hz, B{sub v} = 0). On harvesting, radish showed results similar to those of Smith et al. Dry stem weight and plant height were both significantly greater (Mann-Whitney tests, Ps < 0.05) in EMF-exposed plants than in control plants in each EMF experiment. Wet root weight was significantly greater in EMF-exposed plants in two out of three experiments, as were dry leaf weight, dry whole weight, and stem diameter. Dry root weight, wet leaf weight, and wet whole weight were significantly greater in EMF-exposed plants in one of three experiments. All significant differences indicated an increase in weight or size in the EMF-exposed plants. In each of the sham experiments, no differences between exposed and control plants were evident. Mustard plants failed to respond to the EMFs in any of the plant parameters measured. In one experiment, barley similarly failed to respond; but in another showed significantly greater wet root weight and significantly smaller stem diameter and dry seed weight at the end of the experiment in exposed plants compared to control plants. Although these results give no clue about the underlying bioelectromagnetic mechanism, they demonstrate that, at least for one EMF-sensitive biosystem, results can be independently replicated in another laboratory. Such replication is crucial in establishing the validity of bioelectromagnetic science.

The remaining risk to be accepted with test facilities and prototype plants, and the relevant legal provisions of nuclear law

International Nuclear Information System (INIS)

Mayinger, T.

1995-01-01

The first chapter explains the provisions laid down in nuclear law to assure that precaution is taken to prevent damage resulting from the operation of nuclear power reactors, in order to set a line for comparison with the relevant legal provisions relating to test facilities and prototype plants. The comparative analysis shows that the means and methods of precaution are defined to comprise three approaches, namely measures taken to avert danger, measures taken to prevent danger, and measures for (remaining) risk minimization. All three approaches are intended to prevent occurrence of specifically nuclear events. The second chapter characterizes power reactors, prototype plant and test facilities and develops criteria for distinction. The third chapter establishes the systematics for comparison, showing whether and how the mandatory precaution to prevent damage defined for power reactors, prototype plant, and test facilities can be distinguished from each other, the results being represented in a systematic survey of licensing requirements as laid down in section 7, sub-section 2 ATG (Atomic Energy Act). (orig./HP) [de

Acute oral toxicity test and phytochemistry of some West African medicinal plants.

Science.gov (United States)

Awobajo, F O; Omorodion-Osagie, E; Olatunji-Bello, I I; Adegoke, O A; Adeleke, T I

2009-01-01

Although there is increased acceptance and utilization of medicinal plants worldwide, many are used indiscriminately without recourse to any safety test. Thus, the need for toxicity tests to determine the safe dose for oral consumption. LD50 and phytochemistry of four medicinal plants of West Africa were investigated. Thirty male and non pregnant female Swiss albino mice weighing 20grams each were used for this study. They were divided into the Control (C), Oldenlandia corymbosa L. aqueous leaf-extract treated (OCG), Parquetina nigrescens aqueous leaf extract treated (PNG), Hybanthus enneaspermus aqueous leaf extract treated (HEG), Ficus carica leaf extract treated (FCG) and Sesamum indicum aqueous seeds extract treated group (SIG). Each group except the control was further divided into four sub-groups of six mice each, and were administered orally, graded doses (SI; 1, 2, 4 and 8, PN; 2.5, 5, 10 and 20, OC; 5, 10, 20 and 40, FC; 1, 2, 4 and 8, HE; 4, 8, 16, 32) of the aqueous extract of each plant (g/kg body weight) after 12 hours fasting. The dry aqueous leaf extracts of HE, OC, PN, FC all have dark brown colour and pH ranging from 6.1 to 7.2 while the seed extract of SI has a light brown color with pH of 7.0. Flavonoids, cardiac glycosides, anthocyanosides, saponin, and reducing sugar were present in all extracts, while cyanogenic glycoside was present only in HE. LD50 determination results obtained using Thompson and Finney methods were as follows; OC; 14.14 +/- 0.27 and 10.56 +/- 0.20, PN; 12.60 +/- 0.10 and 13.10 +/- 0.10, HE; 8.14 +/- 0.30 and 8.24 +/- 0.35, FC; 3.36 +/- 0.26 and 4.00 +/- 0.04, SI; 4.00 +/- 0.10 and 3.10 +/- 0.22 respectively (LD50 values are in g/kg body weight. The results of this study have provided an oral LD50 from where a safe dose can be chosen for further research into the merits of the consumption of these medicinal plants.

Response Time Test for The Application of the Data Communication Network to Nuclear Power Plant

International Nuclear Information System (INIS)

Shin, Y.C.; Lee, J.Y.; Park, H.Y.; Seong, S.H.; Chung, H.Y.

2002-01-01

This paper discusses the response time test for the application of the Data Communication Network (DCN) to Nuclear Power Plant (NPP). Conventional Instrumentation and Control (I and C) Systems using the analog technology in NPP have raised many problems regarding the lack of spare parts, maintenance burden, inaccuracy, etc.. In order to solve the problems, the Korean Next Generation Reactor (KNGR) I and C system has adopted the digital technology and new design features of using the data communication networks. It is essential to prove the response time requirements that arise from the introduction of digital I and C technology and data communication networks to nuclear power plant design. For the response time test, a high reliable data communication network structure has been developed to meet the requirements of redundancy, diversity, and segmentation. This paper presents the results of network load analysis and response time test for the KNGR DCN prototype. The test has been focused on the response time from the field components to the gateway because the response times from the gateway to the specific systems are similar to those of the existing design. It is verified that the response time requirements are met through the prototype test for KNGR I and C systems. (authors)

Methods for Quantifying the Uncertainties of LSIT Test Parameters, Test Results, and Full-Scale Mixing Performance Using Models Developed from Scaled Test Data

International Nuclear Information System (INIS)

Piepel, Gregory F.; Cooley, Scott K.; Kuhn, William L.; Rector, David R.; Heredia-Langner, Alejandro

2015-01-01

This report discusses the statistical methods for quantifying uncertainties in 1) test responses and other parameters in the Large Scale Integrated Testing (LSIT), and 2) estimates of coefficients and predictions of mixing performance from models that relate test responses to test parameters. Testing at a larger scale has been committed to by Bechtel National, Inc. and the U.S. Department of Energy (DOE) to ''address uncertainties and increase confidence in the projected, full-scale mixing performance and operations'' in the Waste Treatment and Immobilization Plant (WTP).

Preliminary field tests of near-real-time materials accountancy system at the Tokai Reprocessing Plant (TASK F)

International Nuclear Information System (INIS)

Tsutsumi, Masayori; Sawahata, Toshio; Sugiyama, Toshihide; Tanaka, Kazuhiko; Suyama, Naohiro

1982-01-01

A study of applying the proposed near-real-time material accountancy model to the Tokai Reprocessing Plant, PNC (Power Reactor and Nuclear Fuel Development Corp.), showed that the model was feasible and effective to meet the IAEA (International Atomic Energy Agency) safeguards criteria in terms of detection timeliness and sensitivity. This study using the computer simulation technique is shown in this paper. In order to investigate the applicability of the model to the actual plant, the field test was carried out on the process in the material balance area (MBA) which covers the area from the input accountability vessel (IAV) to the product accountability vessel (PAV), in cooperation with JAERI. The key measuring points for dynamic physical inventory counts (D-PIT) are shown. The results of test evaluation are as follows: For timely detection, it will be able to evaluate an abnoumal accountancy in process by using the MUFd (material unaccounted for) obtained by the D-PIT about once every week. Therefore, this seems to satisfy the timely detection of IAEA safeguards criteria. As for detection, sensitivity and verification procedures, in order to clarify these criteria for a large scale reprocessing plant, further research and development will be required. In addition, since the field test was carried out along with normal plant operation, additional man-power problem was also considered. (Wakatsuki, Y.)

Orimulsion{reg_sign} new generation: New commercial tests results

Energy Technology Data Exchange (ETDEWEB)

Marruffo, F.; Sarmiento, W.

2000-07-01

The new generation of Orimulsion{reg_sign} has been tested since September 1998 and was released to the market on January 1999 by PDVSA-Bitor. Main results from plants will be treated in detail within this paper. Boiler performance has been considerably improved by fuel switching. Operation changes could be summarized as follows: (1) Better furnace wall heat absorption; (2) Lower backend temperature; and (3) Lower sootblowing frequency. In Fuel Oil designed units, the new Orimulsion{reg_sign} firing current performance is similar to Fuel Oil, which indicates a similarity in combustion characteristics between these two fuels. This permits the switching to Orimulsion{reg_sign} in these units with only very minor modification. When compared to its predecessor, the new generation of Orimulsion{reg_sign} has proven to be a better product also from the environmental point of view. Due to its completely new designed surfactant package and its new architecture, the following results have consistently been shown: (1) Lower SO{sub 3} emissions; (2) Less CO production; and (3) Lower particulate levels.

Results obtained by Trillo Nuclear Power Plant in 2001

International Nuclear Information System (INIS)

Solar, V.

2002-01-01

Unification of the Almaraz and Trillo plant managements was given a strong boost with unification of the central offices and management computer platforms (economic and technical) and with signature of an agreement with Madrid's trade unions providing for appropriate management of surpluses and the transfer of knowledge between people leaving the company and those taking their place. The operating results have been maintained at a high level although they were lower than last year, which was exceptional. The plant was on line for 7,979 hours, most of them at full load. (Author)

Experimental results from containment piping bellows subjected to severe accident conditions: Results from bellows tested in corroded conditions. Volume 2

International Nuclear Information System (INIS)

Lambert, L.D.; Parks, M.B.

1995-10-01

Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall, while minimizing the load imposed on the piping and wall. Piping bellows are primarily used in steel containments; however, they have received limited use in some concrete (reinforced and prestressed) containments. In a severe accident they may be subjected to pressure and temperature conditions that exceed the design values, along with a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted at Sandia National Laboratories under the sponsorship of the US Nuclear Regulatory Commission. Several different bellows geometries, representative of actual containment bellows, have been subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of nineteen bellows have been tested. Thirteen bellows were tested in ''like-new'' condition (results reported in Volume 1), and six were tested in a corroded condition. The tests showed that bellows in ''like-new'' condition are capable of withstanding relatively large deformations, up to, or near, the point of full compression or elongation, before developing leakage, while those in a corroded condition did not perform as well, depending on the amount of corrosion. The corroded bellows test program and results are presented in this report

Does mycorrhizal inoculation benefit plant survival, plant development and small-scale soil fixation? Results from a perennial eco-engineering field experiment in the Swiss Alps.

Science.gov (United States)

Bast, Alexander; Grimm, Maria; Graf, Frank; Baumhauer, Roland; Gärtner, Holger

2015-04-01

aggregate stabilization relative to the non-inoculated site but resulted in a significantly higher aggregate stability compared to the control and the non-inoculated site at the end of the third growing season. (ii) Plant survival was significantly improved by the inoculation. Fine-root development was stimulated but not immediately. At the end of the third growing season, root length density tended to be higher and mean root diameter was significantly increased at the mycorrhizal treated site. (iii) Analyses on plant performance of Alnus and Salix demonstrated that the inoculated saplings achieved significantly higher survival rates. There was no treatment effect on plant growth properties except in 2010, where plant height and main stem diameter of Alnus was increased at the mycorrhizal treated site. The estimated total biomass of Alnus and Salix was higher at the mycorrhizal treated site. (iv) There was a positive correlation between root length density and aggregate stability, whereas roots stability. (v) Interannual climatic variations seem to have a crucial influence on root development and, hence, on slope stability. There is a temporal offset of two growing seasons between inoculation effects tested in greenhouse/laboratory and the presented field experiment. However, the application of a commercial mycorrhizal inoculum in eco-engineering measures is a beneficial promoter to mitigate slope instability and surface erosion but needs to be tested at other sites. The contribution is mainly based on Bast (2014) and was funded by the Wolfermann Nägeli Stiftung Zürich and the Swiss Federal Office for Environment (BAFU No.: 09.0027.PJ/I211-3446). Bast, A. (2014): Mycorrhizal inoculation as a promoter for sustainable eco-engineering measures in steep alpine environments? Results of a three-year field experiment in the Arieschbach catchment, Fideris, eastern Swiss Alps. PhD Thesis. University of Berne: 149pp.

MITG test procedure and results

International Nuclear Information System (INIS)

Eck, M.E.; Mukunda, M.

1983-01-01

Elements and modules for Radioisotope Thermoelectric Generator have been performance tested since the inception of the RTG program. These test articles seldom resembled flight hardware and often lacked adequate diagnostic instrumentation. Because of this, performance problems were not identified in the early stage of program development. The lack of test data in an unexpected area often hampered the development of a problem solution. A procedure for conducting the MITG Test was developed in an effort to obtain data in a systematic, unambiguous manner. This procedure required the development of extensive data acquisition software and test automation. The development of a facility to implement the test procedure, the facility hardware and software requirements, and the results of the MITG testing are the subject of this paper

Operational aspects, results and problems associated with R/B testing at Gentilly 2

International Nuclear Information System (INIS)

Garceau, N.; Beaudoin, R.

1991-01-01

There are many methods, some more complex or difficult to deal with than others, to verify the containment building integrity. At G-2, we chose the temperature compensation method. Our selection criteria were: 1) the greater precision of this method; 2) the possibility of executing the test with the plant running at full power; 3) short period required for the test; 4) after the technique is understood, its simplicity of execution; 5) can be easily inserted in the normal operating test program with a minimum of personnel; 6) this technique can be used at both low and high pressure. In this presentation we will succinctly discuss the different phases of the technique such as: the background, the prerequisite, the problems, the results and, finally, we will give some recommendations to facilitate the use of this method

Toothpick test: a methodology for the detection of RR soybean plants1

Directory of Open Access Journals (Sweden)

Fabiana Mota da Silva

Full Text Available Due to the large increase in the area cultivated with genetically modified soybean in Brazil, it has become necessary to determine methods that are fast and efficient for detecting these cultivars. The aim of this work was to test the efficiency of the toothpick method in the detection of RR soybean plants, as well as to distinguish between cultivars, for sensitivity caused by herbicide. Ten transgenic soybean cultivars, resistant to the active ingredient glyphosate, and ten conventional soybean cultivars were used. Toothpicks soaked in glyphosate were applied to all the plants at stage V6 and evaluations were made at 2, 4, 6, 8 and 10 days after application (DAA. The effects of the glyphosate on the cultivars, and the symptoms of phytotoxicity caused in the transgenic plants were evaluated by means of grading scales. The toothpick test is effective in identifying RR soybean cultivars and also in separating them into groups by sensitivity to the symptoms caused by the glyphosate.

Evaluation tests on controbloc, a programmable automaton for nuclear power plants

International Nuclear Information System (INIS)

Pralus, B.; Bourassin, J.L.; Varaldi, G.

1983-01-01

Controbloc is the programmable automaton used by Electricite de France (EDF) to equip its 1300 MW range of power plants. EDF and the designer CGEE Alsthom have conducted a large number of tests on prototype series equipment to determine whether it matches its specifications. These tests were performed in various laboratories in the Paris area and were concerned with: (1) verifying performance (acquisition, processing and retrieval cycle times - multiplexed exchange performance); (2) the robustness of equipment and its behaviour in industrial conditions (climatic tests in line with IEC regulations - earthquake resistance in particularly severe conditions - resistance to industrial interference); (3) behaviour when internal faults occur, an aspect of particular importance in view of the role played by Controbloc in nuclear power plants; in the event of a failure, safety (incorrect commands must not be given) and availability (resumption of operation) must be guaranteed. A large number of checks have been carried out both by the designer and EDF, the latter having put into operation an original method which is described in the paper. Controbloc has been thoroughly tested and its modular construction has facilitated the performance of these tests. Some shortcomings have been revealed which the designer or EDF has remedied. (author)

Irradiation Effects Test Series: Test IE-3. Test results report

International Nuclear Information System (INIS)

Farrar, L.C.; Allison, C.M.; Croucher, D.W.; Ploger, S.A.

1977-10-01

The objectives of the test reported were to: (a) determine the behavior of irradiated fuel rods subjected to a rapid power increase during which the possibility of a pellet-cladding mechanical interaction failure is enhanced and (b) determine the behavior of these fuel rods during film boiling following this rapid power increase. Test IE-3 used four 0.97-m long pressurized water reactor type fuel rods fabricated from previously irradiated fuel. The fuel rods were subjected to a preconditioning period, followed by a power ramp to 69 kW/m at a coolant mass flux of 4920 kg/s-m 2 . After a flow reduction to 2120 kg/s-m 2 , film boiling occurred on the fuel rods. One rod failed approximately 45 seconds after the reactor was shut down as a result of cladding embrittlement due to extensive cladding oxidation. Data are presented on the behavior of these irradiated fuel rods during steady-state operation, the power ramp, and film boiling operation. The effects of a power ramp and power ramp rates on pellet-cladding interaction are discussed. Test data are compared with FRAP-T3 computer model calculations and data from a previous Irradiation Effects test in which four irradiated fuel rods of a similar design were tested. Test IE-3 results indicate that the irradiated state of the fuel rods did not significantly affect fuel rod behavior during normal, abnormal (power ramp of 20 kW/m per minute), and accident (film boiling) conditions

78 FR 71676 - NUREG-1482, Revision 2, “Guidelines for Inservice Testing at Nuclear Power Plants, Final Report”

Science.gov (United States)

2013-11-29

... Testing at Nuclear Power Plants, Final Report'' AGENCY: Nuclear Regulatory Commission. ACTION: Notice of... entitled: NUREG-1482, Revision 2, ``Guidelines for Inservice Testing at Nuclear Power Plants,'' and... Restraints (Snubbers) at Nuclear Power Plants.'' In the previous Revisions 0 and 1 of NUREG-1482, the NRC...

Aboveground mechanical stimuli affect belowground plant-plant communication.

Science.gov (United States)

Elhakeem, Ali; Markovic, Dimitrije; Broberg, Anders; Anten, Niels P R; Ninkovic, Velemir

2018-01-01

Plants can detect the presence of their neighbours and modify their growth behaviour accordingly. But the extent to which this neighbour detection is mediated by abiotic stressors is not well known. In this study we tested the acclimation response of Zea mays L. seedlings through belowground interactions to the presence of their siblings exposed to brief mechano stimuli. Maize seedling simultaneously shared the growth solution of touched plants or they were transferred to the growth solution of previously touched plants. We tested the growth preferences of newly germinated seedlings toward the growth solution of touched (T_solution) or untouched plants (C_solution). The primary root of the newly germinated seedlings grew significantly less towards T_solution than to C_solution. Plants transferred to T_solution allocated more biomass to shoots and less to roots. While plants that simultaneously shared their growth solution with the touched plants produced more biomass. Results show that plant responses to neighbours can be modified by aboveground abiotic stress to those neighbours and suggest that these modifications are mediated by belowground interactions.

Aboveground mechanical stimuli affect belowground plant-plant communication.

Directory of Open Access Journals (Sweden)

Ali Elhakeem

Full Text Available Plants can detect the presence of their neighbours and modify their growth behaviour accordingly. But the extent to which this neighbour detection is mediated by abiotic stressors is not well known. In this study we tested the acclimation response of Zea mays L. seedlings through belowground interactions to the presence of their siblings exposed to brief mechano stimuli. Maize seedling simultaneously shared the growth solution of touched plants or they were transferred to the growth solution of previously touched plants. We tested the growth preferences of newly germinated seedlings toward the growth solution of touched (T_solution or untouched plants (C_solution. The primary root of the newly germinated seedlings grew significantly less towards T_solution than to C_solution. Plants transferred to T_solution allocated more biomass to shoots and less to roots. While plants that simultaneously shared their growth solution with the touched plants produced more biomass. Results show that plant responses to neighbours can be modified by aboveground abiotic stress to those neighbours and suggest that these modifications are mediated by belowground interactions.

Models of cognitive behavior in nuclear power plant personnel. A feasibility study: summary of results. Volume 1

International Nuclear Information System (INIS)

Woods, D.D.; Roth, E.M.; Hanes, L.F.

1986-07-01

This report summarizes the results of a feasibility study to determine if the current state of models of human cognitive activities can serve as the basis for improved techniques for predicting human error in nuclear power plants emergency operations. Based on the answer to this question, two subsequent phases of research are planned. Phase II is to develop a model of cognitive activities, and Phase III is to test the model. The feasibility study included an analysis of the cognitive activities that occur in emergency operations and an assessment of the modeling concepts/tools available to capture these cognitive activities. The results indicated that a symbolic processing (or artificial intelligence) model of cognitive activities in nuclear power plants is both desirable and feasible. This cognitive model can be built upon the computational framework provided by an existing artificial intelligence system for medical problem solving, called Caduceus. The resulting cognitive model will increase the capability to capture the human contribution to risk in probabilistic risk assessment studies. Volume 1 summarizes the major findings and conclusions of the study. Volume 2 provides a complete description of the methods and results, including a synthesis of the cognitive activities that occur during emergency operations, and a literature review on cognitive modeling relevant to nuclear power plants. 19 refs

Alternative filtration testing program: Pre-evaluation of test results

International Nuclear Information System (INIS)

Georgeton, G.K.; Poirier, M.R.

1990-01-01

Based on results of testing eight solids removal technologies and one pretreatment option, it is recommended that a centrifugal ultrafilter and polymeric ultrafilter undergo further testing as possible alternatives to the Norton Ceramic filters. Deep bed filtration should be considered as a third alternative, if a backwashable cartridge filter is shown to be inefficient in separate testing

Alternative filtration testing program: Pre-evaluation of test results

Energy Technology Data Exchange (ETDEWEB)

Georgeton, G.K.; Poirier, M.R.

1990-09-28

Based on results of testing eight solids removal technologies and one pretreatment option, it is recommended that a centrifugal ultrafilter and polymeric ultrafilter undergo further testing as possible alternatives to the Norton Ceramic filters. Deep bed filtration should be considered as a third alternative, if a backwashable cartridge filter is shown to be inefficient in separate testing.

Testing of adsorbents used in nuclear power plant air cleaning systems using the open-quotes Newclose quotes standards

International Nuclear Information System (INIS)

Freeman, W.P.

1993-01-01

Ever since the publication of the NRC Information Notice No. 87-32: Deficiencies in the Testing of Nuclear-Grade Activated Charcoal, nuclear power facilities in the US have struggled in their efforts to open-quotes...review the information for applicability to their facilities and consider action, if appropriate ...close quotes as stated in the notice. The encouragement of resident NRC inspectors at some nuclear power facilities has prompted a variety of responses ranging from no change at all in testing requirements to contemplated changes in plant technical specifications. This confusion is the result of a couple factors. The first factor is the lack of a current revision to NRC Regulatory Guide 1.52, the basic document used in nuclear power plant technical specifications for the testing of engineered-safety feature (ESF) post accident air cleaning systems. The second factor is the standards that have been written since the last revision of Reg. Guide 1.52 which include two revision of ANSI N509 and N510, two revisions of RDT M16-1T, two version of ASTM D3803, two versions of ASTM D4069, and three versions of an SME code AG-1. Few of the standards and codes listed above are commensurate with each other and, thus, present a nearly insolvable maze to the HVAC engineer asked to upgrade adsorbent testing requirements following the standards. This paper describes the authors experience with a number of nuclear power facilities in their efforts to meet the requirements of the new standards of testing adsorbents from nuclear power plant air cleaning systems. The existing standards are discussed in light of the current state of the art for adsorbent testing of adsorbent media from nuclear air treatment systems. Test results are presented showing the impact of new test requirements on acceptance criteria when compared to the old test requirements and recommendations are offered for solution of this testing problem in the future. 12 refs., 5 tabs

Plant management tools tested with a small-scale distributed generation laboratory

International Nuclear Information System (INIS)

Ferrari, Mario L.; Traverso, Alberto; Pascenti, Matteo; Massardo, Aristide F.

2014-01-01

Highlights: • Thermal grid innovative layouts. • Experimental rig for distributed generation. • Real-time management tool. • Experimental results for plant management. • Comparison with results from an optimization complete software. - Abstract: Optimization of power generation with smart grids is an important issue for extensive sustainable development of distributed generation. Since an experimental approach is essential for implementing validated optimization software, the TPG research team of the University of Genoa has installed a laboratory facility for carrying out studies on polygeneration grids. The facility consists of two co-generation prime movers based on conventional technology: a 100 kWe gas turbine (mGT) and a 20 kWe internal combustion engine (ICE). The rig high flexibility allows the possibility of integration with renewable-source based devices, such as biomass-fed boilers and solar panels. Special attention was devoted to thermal distribution grid design. To ensure the possibility of application in medium-large districts, composed of several buildings including energy users, generators or both, an innovative layout based on two ring pipes was examined. Thermal storage devices were also included in order to have a complete hardware platform suitable for assessing the performance of different management tools. The test presented in this paper was carried out with both the mGT and the ICE connected to this innovative thermal grid, while users were emulated by means of fan coolers controlled by inverters. During this test the plant is controlled by a real-time model capable of calculating a machine performance ranking, which is necessary in order to split power demands between the prime movers (marginal cost decrease objective). A complete optimization tool devised by TPG (ECoMP program) was also used in order to obtain theoretical results considering the same machines and load values. The data obtained with ECoMP were compared with the

The current state of inservice testing programs at U.S. Nuclear Power Plants - a regulatory overview

International Nuclear Information System (INIS)

Campbell, P.; Colaccino, J.

1994-01-01

Information is provided on inservice testing (IST) of pumps and valves at U.S. nuclear power plants to provide consistency in the implementation of regulatory requirements and to enhance communications among utility licensees who may have, like NSSS vendors, similar kinds and numbers of components or comparable IST programs. Documents discussed include the ASME Operation and Maintenance Standards Parts 6 and 10 (covering inservice testing of pumps and valves in light water reactor power plants), the draft NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants (including review comments by Nuclear Management and Resource Council), and applicable Licensee Event Reports including summaries of several reports relating to IST

Irradiation effects test series test IE-1 test results report

International Nuclear Information System (INIS)

Quapp, W.J.; Allison, C.M.; Farrar, L.C.; Mehner, A.S.

1977-03-01

The report describes the results of the first programmatic test in the Nuclear Regulatory Commission Irradiation Effects Test Series. This test (IE-1) used four 0.97m long PWR-type fuel rods fabricated from previously irradiated Saxton fuel. The objectives of this test were to evaluate the effect of fuel pellet density on pellet-cladding interaction during a power ramp and to evaluate the influence of the irradiated state of the fuel and cladding on rod behavior during film boiling operation. Data are presented on the behavior of irradiated fuel rods during steady-state operation, a power ramp, and film boiling operation. The effects of as-fabricated gap size, as-fabricated fuel density, rod power, and power ramp rate on pellet-cladding interaction are discussed. Test data are compared with FRAP-T2 computer model predictions, and comments on the consequences of sustained film boiling operation on irradiated fuel rod behavior are provided

An evaluation of corrosion resistant alloys by field corrosion test in Japanese refuse incineration plants

International Nuclear Information System (INIS)

Kawahara, Yuuzou; Nakamura, Masanori; Shibuya, Eiichi; Yukawa, Kenichi

1995-01-01

As the first step for development of the corrosion resistant superheater tube materials of 500 C, 100 ata used in high efficient waste-to-energy plants, field corrosion tests of six conventional alloys were carried out at metal temperatures of 450 C and 550 C for 700 and 3,000 hours in four typical Japanese waste incineration plants. The test results indicate that austenitic alloys containing approximately 80 wt% [Cr+Ni] show excellent corrosion resistance. When the corrosive environment is severe, intergranular corrosion of 40∼200 microm depth occurs in stainless steel and high alloyed materials. It is confirmed quantitatively that corrosion behavior is influenced by environmental corrosion factors such as Cl concentration and thickness of deposits on tube surface, metal temperature, and flue gas temperature. The excellent corrosion resistance of high [Cr+Ni+Mo] alloys such as Alloy 625 is explained by the stability of its protective oxide, such that the time dependence of corrosion nearly obeys the parabolic rate law

Field evaluation of a horizontal well recirculation system for groundwater treatment: Pilot test at the Clean Test Site Portsmouth Gaseous Diffusion Plant, Piketon, Ohio

International Nuclear Information System (INIS)

Muck, M.T.; Kearl, P.M.; Siegrist, R.L.

1998-01-01

This report presents the results of field testing a horizontal well recirculation system at the Portsmouth Gaseous Diffusion Plant (PORTS). The recirculation system uses a pair of horizontal wells, one for groundwater extraction and treatment and the other for reinjection of treated groundwater, to set up a recirculation flow field. The induced flow field from the injection well to the extraction well establishes a sweeping action for the removal and treatment of groundwater contaminants. The overall purpose of this project is to study treatment of mixed groundwater contaminants that occur in a thin water-bearing zone not easily targeted by traditional vertical wells. The project involves several research elements, including treatment-process evaluation, hydrodynamic flow and transport modeling, pilot testing at an uncontaminated site, and pilot testing at a contaminated site. The results of the pilot test at an uncontaminated site, the Clean Test Site (CTS), are presented in this report

Effects in Plant Populations Resulting from Chronic Radiation Exposure

Energy Technology Data Exchange (ETDEWEB)

Geras' kin, Stanislav A.; Volkova, Polina Yu.; Vasiliyev, Denis V.; Dikareva, Nina S.; Oudalova, Alla A. [Russian Institute of Agricultural Radiology and Agroecology, 249032, Obninsk (Russian Federation)

2014-07-01

Human industrial activities have left behind a legacy of ecosystems strongly impacted by a wide range of contaminants, including radionuclides. Phyto-toxic effects of acute impact are well known, but the consequences of long-term chronic exposure to low pollutant concentrations is neither well understood nor adequately included in risk assessments. To understand effects of real-world contaminant exposure properly we must pay attention to what is actually going on in the field. However, for many wildlife groups and endpoints, there are no, or very few, studies that link accumulation, chronic exposure and biological effects in natural settings. To fill the gaps, results of field studies carried out on different plant species (winter rye and wheat, spring barley, oats, Scots pine, wild vetch, crested hair-grass) in various radioecological situations (nuclear weapon testing, the Chernobyl accident, uranium and radium processing) to investigate effects of long-term chronic exposure to radionuclides are discussed. Because each impacted site developed in its own way due to a unique history of events, the experience from one case study is rarely directly applicable to another situation. In spite of high heterogeneity in response, we have detected several general patterns. Plant populations growing in areas with relatively low levels of pollution are characterized by the increased level of both cytogenetic alterations and genetic diversity. Accumulation of cellular alterations may afterward influence biological parameters important for populations such as health and reproduction. Presented data provide evidence that in plant populations inhabiting heavily contaminated territories cytogenetic damage were accompanied by decrease in reproductive ability. In less contaminated sites, because of the scarcity of data available, it is impossible to establish exactly the relationship between cytogenetic effects and reproductive ability. Radioactive contamination of the plants

Modification of EBR-II plant to conduct loss-of-flow-without-scram tests

Energy Technology Data Exchange (ETDEWEB)

Messick, N C; Betten, P R; Booty, W F; Christensen, L J; Fryer, R M; Mohr, D; Planchon, H P; Radtke, W H

1987-04-01

This paper describes the details of and the philosophy behind changes made to the EBR-II plant in order to conduct loss-of-flow-without-scram tests. No changes were required to conduct loss-of-heat-sink-without-scram tests.

Modification of EBR-II plant to conduct loss-of-flow-without-scram tests

International Nuclear Information System (INIS)

Messick, N.C.; Betten, P.R.; Booty, W.F.; Christensen, L.J.; Fryer, R.M.; Mohr, D.; Planchon, H.P.; Radtke, W.H.

1987-01-01

This paper describes the details of and the philosophy behind changes made to the EBR-II plant in order to conduct loss-of-flow-without-scram tests. No changes were required to conduct loss-of-heat-sink-without-scram tests. (orig.)

Irradiation effects test series, test IE-5. Test results report. [PWR

Energy Technology Data Exchange (ETDEWEB)

Croucher, D. W.; Yackle, T. R.; Allison, C. M.; Ploger, S. A.

1978-01-01

Test IE-5, conducted in the Power Burst Facility at the Idaho National Engineering Laboratory, employed three 0.97-m long pressurized water reactor type fuel rods, fabricated from previously irradiated zircaloy-4 cladding and one similar rod fabricated from unirradiated cladding. The objectives of the test were to evaluate the influence of simulated fission products, cladding irradiation damage, and fuel rod internal pressure on pellet-cladding interaction during a power ramp and on fuel rod behavior during film boiling operation. The four rods were subjected to a preconditioning period, a power ramp to an average fuel rod peak power of 65 kW/m, and steady state operation for one hour at a coolant mass flux of 4880 kg/s-m/sup 2/ for each rod. After a flow reduction to 1800 kg/s-m/sup 2/, film boiling occurred on one rod. Additional flow reductions to 970 kg/s-m/sup 2/ produced film boiling on the three remaining fuel rods. Maximum time in film boiling was 80s. The rod having the highest initial internal pressure (8.3 MPa) failed 10s after the onset of film boiling. A second rod failed about 90s after reactor shutdown. The report contains a description of the experiment, the test conduct, test results, and results from the preliminary postirradiation examination. Calculations using a transient fuel rod behavior code are compared with the test results.

Thermal performance test for steam turbine of nuclear power plants

International Nuclear Information System (INIS)

Bu Yubing; Xu Zongfu; Wang Shiyong

2014-01-01

Through study of steam turbine thermal performance test of CPR1000 nuclear power plant, we solve the enthalpy calculation problems of the steam turbine in wet steam zone using heat balance method which can help to figure out the real overall heat balance diagram for the first time, and we develop a useful software for thermal heat balance calculation. Ling'ao phase II as an example, this paper includes test instrument layout, system isolation, risk control, data acquisition, wetness measurement, heat balance calculation, etc. (authors)

Digital I and C system pre-tests using plant specific simulators

International Nuclear Information System (INIS)

Holl, B.; Probst, H.; Wischert, W.

2006-01-01

The paper focuses on strategic aspects of the implementation of modern digital instrumentation and control system (I and C) in nuclear power plant (NPP) training simulators and points out the way to identify the most appropriate implementation method of the digital I and C system in the simulator development environment which fulfils the requirement imposed by the nuclear power plants. This regards mainly training aspects, simulator as a test bed for design verification and validation (V and V), and software maintenance aspects with respect to future evolutions of the digital I and C system. (author)

Seismic tests on a reduced scale mock-up of a reprocessing plant cooling pond

International Nuclear Information System (INIS)

Queval, J.C.; Gantenbein, F.; Lebelle, M.

1995-01-01

In conjunction with COGEMA and SGN, CEA has launched an important research program to validate the reprocessing plant cooling pond calculation mainly for the effect of the racks on the fluid-pond interaction. The paper presents the tests performed on a reduced scale mock-up (scale 1/5). The tests are composed by: -random excitations at very low excitation level to measure the natural frequencies, especially the first sloshing mode frequency; -sinusoidal tests to measure the damping; -seismic tests performed with 3 different time reduction scales (1, 1/5, 1/√5) and 3 different synthetic accelerograms. Two types of simplified model with added masses and finite element model were developed. Comparisons of measured and calculated pressure fields against the panels will be presented. The measured frequencies, obtained during tests, are in good agreement with Housner's results. (authors). 2 refs., 4 figs., 5 tabs

APPLICATIONS LIQUID ORGANIC FERTILIZER AND COMPOSITION OF PLANT MEDIA TO RESULT OF SELADA PLANTS (Lactuca sativa L

Directory of Open Access Journals (Sweden)

Sri Hidayati

2017-12-01

Full Text Available Abstract:           Lettuce (lactuca sativa is a vegetable that has a very high economic value. Where this plant can be grown in temperate and tropical regions, Lettuce production is still low, then this plant needs to be given fertilizer treatment. One of the fertilizer that can be used is liquid organic fertilizer. Liquid Organic Fertilizer has several benefits such as to encourage and increase the growth and yield of plants.             Objective: To know the effect of combination of planting media composition and liquid organic fertilizer to growth and yield of lettuce crop; To know the influence of plant plant composition on growth and yield of lettuce plant; To know the effect of liquid organic fertilizer on growth and yield of lettuce plant.            The experiment was conducted in experimental garden of Faculty of Agriculture Universitas Merdeka Surabaya Jl.Ketintang Madya VII / 2 Surabaya, with the space 0-20 meters above sea level.      This research is a pot experiment and is a two factor factorial research with Randomized Block Design (RAK, the first factor is Liquid Organic Fertilizer with 3 levels and the second factor is the composition of planting media with 4 levels. Where Factor I: liquid organic fertilizer consisting of: P1: 1 ml / plant; P2: 2 ml / plant; P3: 3 ml / plant, Factor II: planting medium consisting of 4 (four levels, namely: M1: soil + manure + rice husk: 2: 1: 1; M2: soil + manure + rice husk: 1: 1: 1; M3: ground + manure + sand: 2: 1: 1; M4: ground + manure + sand: 1: 1: 1, treatment repeated 3 times and each treatment there are 2 plant samples, so the number of plants as much 72 or 72 polybag.Based on the results of research conducted, it can be concluded as follows:1. POC concentration factor (P showed significant influence on all variables studied such as leaf number, plant length and wet weight of plant.2. The media composition factor (M showed a nonsignificant effect

Results of cultivation experiment at clean farm (vegetable plant). 1. Clean farm (yasai kojo) ni okeru saibai jikken kekka. 1

Energy Technology Data Exchange (ETDEWEB)

Miyaishi, T; Kawagishi, K; Matsuzaki, O; Nakahara, M [Kyushu Electric Power Co. Ltd., Fukuoka (Japan)

1991-03-31

This paper reports a summary of the facilities in an experimental plant constructed by the Kyushu Electric Power Company in 1988, and a result of experiments on cultivating salad and lettuce. The plant has environmentally controlled cultivating rooms of solar beam combined type and totally artificial light type, each having a floor area of 50 m {sup 2}, disposed with cultivating stages divided into three divisions of seedling culture, growth, and forced culture, nutritious liguid feeding devices of circulation type, and air conditioning equipment. The paper describes results of the tests aimed at realizing an increase in yield and profit, an optimum cultivating system, and economic facilities at the vegetable plant. Included in the tests are that (a) four kinds of salads and five kinds lettuce were selected for cultivation to decide most suitable kinds, based on literature survey and preliminary experiments; (b) varying the environmental conditions for the culture, such as temperature, radiating condition, concentration of the nutritious liquid, and concentration of carbon dioxide, conditions optimum or suitable for the plant growth were selected; (c) the plant was compared with glass green houses with respect to the required cultivating period of time, vitamin C content and color tones of the products; (d) the solar beam combined type room and the totally artificial light rooms were compared with respect to power consumption and heat capacity that passes through the cultivation room walls, and the latter was concluded being superior in economics and stability. 12 refs., 37 figs., 30 tabs.

Thermal tests of large recirculation cooling installations for nuclear power plants

Science.gov (United States)

Balunov, B. F.; Lychakov, V. D.; Il'in, V. A.; Shcheglov, A. A.; Maslov, O. P.; Rasskazova, N. A.; Rakhimov, R. Z.; Boyarov, R. A.

2017-11-01

The article presents the results from thermal tests of some recirculation installations for cooling air in nuclear power plant premises, including the volume under the containment. The cooling effect in such installations is produced by pumping water through their heat-transfer tubes. Air from the cooled room is blown by a fan through a bundle of transversely finned tubes and is removed to the same room after having been cooled. The finning of tubes used in the tested installations was made of Grade 08Kh18N10T and Grade 08Kh18N10 stainless steels or Grade AD1 aluminum. Steel fins were attached to the tube over their entire length by means of high-frequency welding. Aluminum fins were extruded on a lathe from the external tube sheath into which a steel tube had preliminarily been placed. Although the fin extrusion operation was accompanied by pressing the sheath inner part to the steel tube, tight contact between them over the entire surface was not fully achieved. In view of this, the air gap's thermal resistance coefficient was introduced in calculating the heat transfer between the heat-transferring media. The air gap average thickness was determined from the test results taking into account the gap variation with temperature due to different linear expansion coefficients of steel and aluminum. These tests, which are part of the acceptance tests of the considered installations, were carried out at the NPO TsKTI test facility and were mainly aimed at checking if the obtained thermal characteristics were consistent with the values calculated according to the standard recommendations with introduction, if necessary, of modifications to those recommendations.

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-05-23

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-05-23

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

FY 1991 report on the results of the development of an entrained bed coal gasification power plant. Part 4. Operation of pilot plant; 1991 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu - Sono 4. Pilot plant unten sosa hen

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-01-01

A record was summarized of the operation of the 200 t/d entrained bed coal gasification pilot plant that was constructed with the aim of establishing technology of the integrated coal gasification combined cycle power generation. As to the actual results of operation hours, the paper summarized the records of gasifier facilities, gas refining facilities, gas turbine facilities and safety environment facilities which were collected from April 1991 to January 1993. Relating to the actual results of start-up/stop, the paper summarized the records of gasifier facilities, gas refining facilities (desulfurization), gas refining facilities (dedusting), gas turbine facilities and safety environment facilities. Further, operation manuals were made for the schedule of plant start-up/stop, generalization, gasifier facilities, gas refining facilities (desulfurization), gas refining facilities (dedusting), gas turbine facilities, actual pressure/actual size combustor testing facilities and safety environment facilities. (NEDO)

Verification test of advanced LWR fuel components of Westinghouse type nuclear power plants

International Nuclear Information System (INIS)

Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Young Ho

2004-08-01

The purpose of this project is to independently conduct the performance test of the spacer grids and the cladding material of the 16x16 and 17x17 advanced fuels for Westinghouse type plants, and to improve the relevant test technology. Major works and results of the present research are as follows. 1. The design and structural features of the spacer grids were investigated, especially the finally determined I-spring was thoroughly analyzed in the point of the mechanical damage and characteristic. 2. As for the mechanical tests of the space grids, the characterization, the impact and the fretting wear tests were carried out. The block as well as the in-grid tests were conducted for the spring/dimple characterization, from which a simple method was developed that simulated the boundary conditions of the assembled grid straps. The impact tester was modified and improved to accommodate a full size grid assembly. The impact result showed that the grid assembly fulfilled the design criteria. As for the fretting wear tests, a sliding test under the room temperature air/water, a sliding/impact test under the room temperature air and a sliding/impact tests under the high temperature and pressure environments were carried out. To this end, a high temperature and pressure fretting wear tester was newly developed. The wear characteristic and the resistibility of the advanced grid spring/dimple were analyzed in detail. The test results were verified through comparing those with the test results by the Westinghouse company. 3. The properties and performance of the newly adopted material for the cladding, Low Sn Zirlo was investigated by a room and high temperature tensile tests and a corrosion tests under the environments of 360 .deg. C water, 400 steam and 360 .deg. C 70ppm LiOH. Through the present project, all the test equipment and technologies for the fuel components were procured, which will be used for future domestic development of a new fuel

A test of plant-aided petroleum hydrocarbon degradation

Energy Technology Data Exchange (ETDEWEB)

Hosler, K R [Water Technology International Corp., Burlington, ON (Canada); Drake, E N [Exxon Research Engineering Co., Annandale, NJ (United States)

1999-12-31

A research program was established to develop environmental restoration technologies which apply to contaminated industrial sites. The program involved two separate but related parts. Part One involved a multi-year field study, Part Two a greenhouse potted plant study. This paper presents the results of the greenhouse-based phytoremediation experiment which assessed the potential impacts of three treatment factors on the degradation of total petroleum hydrocarbons (TPH) in contaminated soils for use in those cases where the use of plants for restoring contaminated environments might be a simple and cost-effective clean-up alternative. This study showed that biologically-aided contaminant degradation can be enhanced by various treatments such as adding nutrients in the form of inorganic fertilizers, adding oxygen or modifying soil conditions. The study also showed that contaminant degradation can be enhanced in the rhizosphere of various plant species and that remediation of some contaminants can be achieved by exploiting the unique symbiotic relationship between some fungal species and plant roots. 22 refs., 3 tabs., 1 fig.

A test of plant-aided petroleum hydrocarbon degradation

International Nuclear Information System (INIS)

Hosler, K.R.; Drake, E.N.

1998-01-01

A research program was established to develop environmental restoration technologies which apply to contaminated industrial sites. The program involved two separate but related parts. Part One involved a multi-year field study, Part Two a greenhouse potted plant study. This paper presents the results of the greenhouse-based phytoremediation experiment which assessed the potential impacts of three treatment factors on the degradation of total petroleum hydrocarbons (TPH) in contaminated soils for use in those cases where the use of plants for restoring contaminated environments might be a simple and cost-effective clean-up alternative. This study showed that biologically-aided contaminant degradation can be enhanced by various treatments such as adding nutrients in the form of inorganic fertilizers, adding oxygen or modifying soil conditions. The study also showed that contaminant degradation can be enhanced in the rhizosphere of various plant species and that remediation of some contaminants can be achieved by exploiting the unique symbiotic relationship between some fungal species and plant roots. 22 refs., 3 tabs., 1 fig

Plant data evaluation of performance confirmation test in HTTR after Tohoku-Pacific Ocean Earthquake

International Nuclear Information System (INIS)

Ono, Masato; Tochio, Daisuke; Shinohara, Masanori; Shimazaki, Yosuke; Yanagi, Shunki; Iigaki, Kazuhiko

2012-03-01

Tohoku-Pacific Ocean Earthquake occurred on March 11th 2011 and the earthquake intensity of an upper 5 on the Japanese scale was observed in Oarai town. HTTR conducted the confirmation test on cold state in order to ensure the facilities/instruments of reactor building operate normally. In this test, the plant data in the facilities/instruments start-up phase and continue steady operation phase were measured and compared with the previous operation data, and the soundness of facilities/instruments is evaluated. As a result, in after the earthquake, the facilities/instruments operate normally and the reactor cooling function of the HTTR were ensured. (author)

Plant data evaluation of performance confirmation test in HTTR after Tohoku-Pacific Ocean Earthquake

International Nuclear Information System (INIS)

Ono, Masato; Tochio, Daisuke; Shinohara, Masanori; Shimazaki, Yosuke; Yanagi, Shunki; Iigaki, Kazuhiko

2012-01-01

Tohoku-Pacific Ocean Earthquake occurred on March 11th 2011 and the earthquake intensity of an upper 5 on the Japanese scale was observed in Oarai town. HTTR conducted the confirmation test on cold state in order to ensure the facilities/instruments of reactor building operate normally. In this test, the plant data in the facilities/instruments start-up phase and continue steady operation phase were measured and compared with the previous operation data, and the soundness of facilities/instruments is evaluated. As a result, in after the earthquake, the facilities/instruments operate normally and the reactor cooling function of the HTTR were ensured. (author)

Test and evaluation of the in-line plutonium solution K-absorption-edge densitometer at the Savannah River Plant. Phase I. Off-line testing results

International Nuclear Information System (INIS)

Smith, H.A. Jr.; Marks, T.; Johnson, S.S.

1982-04-01

An in-line, plutonium-solution, K-edge absorption densitometer has been developed at Los Alamos and is currently undergoing test and evaluation at the Savannah River Plant (SRP). The first phase of the test and evaluation (off-line instrument calibration and solution assays) was completed, and preparations are under way to install the instrument in-line, as soon as process schedules permit. Calibration data in the design concentration range of 25 to 40 g Pu/L demonstrate routine achievement of densitometry assay precisions of 0.5% or better in 40 min. Plutonium assays at concentrations outside the calibration range were investigated in an effort to define better the limitations of the instrument and address other possible assay situations at SRP. Densitometry precisions obtained for 40-min assays range from 3% to 5 g Pu/L down to 0.4% at 70 g Pu/L. At higher plutonium concentrations, the precision deteriorated due to increasing gamma-ray absorption by the solution. In addition, with actinide concentrations above approximately 100 g/L, the assay accuracy also suffered because of enhanced small-angle scattering effects in the large sample cell. Measurements on mixed U/Pu solutions demonstrated the feasibility of accurate plutonium assays with correction for the large uranium matrix contributions being determined from the measurement data. The 239 240 Pu weight fractions and 241 Pu/ 239 Pu and 238 Pu/ 239 Pu isotopic ratios can be determined. In a mockup of the in-line solution plumbing system, all assay sequences, error conditions, and interlock criteria were exercised and verified to be working properly

Human errors in test and maintenance of nuclear power plants. Nordic project work

International Nuclear Information System (INIS)

Andersson, H.; Liwaang, B.

1985-08-01

The present report is a summary of the NKA/LIT-1 project performed for the period 1981-1985. The report summarizes work on human error influence in test and calibration activities in nuclear power plants, reviews problems regarding optimization of the test intervals, organization of test and maintenance activities, and the analysis of human error contribution to the overall risk in test and mainenace tasks. (author)

Review of scientific Research results in identification of plant raw materials in food products

OpenAIRE

GOLUBTSOVA YU. V.

2016-01-01

Currently, the science-based capabilities have been generated to develop and test various identification methods of food products and reveal adulteration using advanced technique and processes. This article reviews researches and developments to identify the plant raw materials in food products based on morphological, anatomic, physical and chemical test methods and the latest DNA-technologies. Review of physical, chemical, anatomic and morphological test methods to identify raw materials bot...

Supercritical CO2 test loop operation and first test results

International Nuclear Information System (INIS)

Wright, Steven A.; Pickard, Paul S.

2009-01-01

The DOE Office of Nuclear Energy is investigating advanced Brayton cycles for use with next generation nuclear power plants. The focus of this work is on the supercritical CO 2 Brayton cycle which has the potential for high efficiency, and for reduced capital costs due to very compact turbomachinery. Sandia has fabricated and is operating a supercritical CO 2 (S-CO 2 ) test loop to investigate the key technology issues associated with this cycle. This loop is part of a multi-year phased development program to develop a megawatt (MW) class closed S-CO 2 Brayton cycle to demonstrate the applicability of this cycle for DOE Gen-IV program. The current loop has been configured as both a compression loop and as simple heated but unrecuperated Brayton cycle. A second split-flow or re-compression Brayton cycle is currently under development that will use approximately 1 MW of heat to run the Brayton cycle. Early configurations of this split-flow Brayton cycle will be operational later this fiscal year. The key issues for this cycle include the fundamental issues of compressor fluid performance and system control near the critical point, but also the supporting technology issues of bearings, sealing technologies, and rotor windage losses which are also essential to achieving efficiency and cost objectives. These tests are providing the first measurements and information on these key supercritical CO 2 power conversion systems questions. Important data for all these issues has been obtained. This report presents the major results of the testing by showing and comparing the measured compressor performance map with the predicted performance. The compression loop uses a ∼50 kWe motor driven compressor to spin a 37 mm OD compressor at design speeds up to 75,000 rpm with a pressure ratio of 1.8 and a flow rate of 3.53 kg/s for a compressor inlet condition of 305.3 K and 7690 kPa. The most recent configuration of this loop has added a small turbine and 260 kW of heater power is

The software testing of PPS for shin Ulchin nuclear power plant units 1 and 2

International Nuclear Information System (INIS)

Kang, Dong Pa; Park, Cheol Lak; Cho, Chang Hui; Sohn, Se Do; Baek, Seung Min

2012-01-01

The testing of software (S/W) is the process of analyzing a software item to detect the differences between existing and required conditions to evaluate the features of the software items. This paper introduces the S/W testing of Plant Protection System (PPS), as a safety system which actuate Reactor Trip (RT) and Engineered Safety Features (ESF) for Shin Ulchin Nuclear Power Plant Units 1 and 2 (SUN 1 and 2)

FY 1989 report on the results of the development of the entrained bed coal gasification power plant. Support study for the development of the entrained bed coal gasification power plant; 1989 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu (Funryusho sekitan gaska hatsuden plant kaihatsu no shien kenkyu)

Energy Technology Data Exchange (ETDEWEB)

NONE

1990-05-01

For the purpose of supplying the data necessary for construction/operation/maintenance of system, gas turbine, etc. of the desulfurization/dust removal process to be adopted to a pilot plant of 200t/d entrained bed coal gasification power generation, support study was made using the Yubari 40t/d test equipment, and the FY 1989 results were summarized. As to the operation of the 40t/d gasification system, operation was stably continued, and also the system was able to feed gas to the wake device. In the high temperature dry desulfurization test, test was made on the following: structure of the sampling system of the continuous analyzer installed in the coal gas system and circulation gas system, load response corresponding control system, sequence control of the system for supplying reducing agent to SO{sub 2} reduction tower and the system for ash discharge, back washing of D-42 lift gas filter, etc. In the high temperature dry dust removal test, improvement in filtration material/dust separation performance, sequence control of the filtration material supply/discharge system, continuous dust densitometer, operational automation, etc. Through the tests, obtained were the results that are useful for the 200t/d plant. (NEDO)

Modelling of wind power plant controller, wind speed time series, aggregation and sample results

DEFF Research Database (Denmark)

Hansen, Anca Daniela; Altin, Müfit; Cutululis, Nicolaos Antonio

This report describes the modelling of a wind power plant (WPP) including its controller. Several ancillary services like inertial response (IR), power oscillation damping (POD) and synchronising power (SP) are implemented. The focus in this document is on the performance of the WPP output...... and not the impact of the WPP on the power system. By means of simulation tests, the capability of the implemented wind power plant model to deliver ancillary services is investigated....

Demonstration test for reliability of valves for atomic power plants

International Nuclear Information System (INIS)

Hosaka, Shiro

1978-01-01

The demonstration test on the reliability of valves for atomic power plants being carried out by the Nuclear Engineering Test Center is reported. This test series is conducted as six-year project from FY 1976 to FY 1981 at the Isogo Test Center. The demonstration test consists of (1) environmental test, (2) reaction force test, (3) vibration test, (4) stress measurement test, (5) operational characteristic test, (6) flow resistance coefficient measuring test, (7) leakage test and (8) safety valve and relief valve test. These contents are explained about the special requirements for nuclear use, for example, the enviornmental condition after the design base accident of PWRs and BWRs, the environmental test sequence for isolation valves of containment vessels under the emergency condition, the seismic test condition for valves of nuclear use, the various stress measurements under thermal transient conditions, the leak test after 500 cycles between the normal operating conditions for PWRs and BWRs and the start up conditions and so on. As for the testing facilities, the whole flow diagram is shown, in which the environmental test section, the vibration test section, the steam test section, the hot water test section, the safety valve test section and main components are included. The specifications of each test section and main components are presented. (Nakai, Y.)

Reality testing a plant design 'virtually' anywhere

International Nuclear Information System (INIS)

Anon.

1997-01-01

The development of a new world-wide-web compatible information system known as HyperPlant will allow users to navigate real-time three-dimensional plant design and contraction software. It is anticipated that corporate Intranets will be created to facilitate computer-aided design of industrial plants such as piping routes, process schematics, fabrication drawings, and allow use of PDMS (the Plant Design Management System). HyperPlant can also assist in plant commissioning and operation as well as for planning operation and maintenance procedures. (UK)

The application of plant tests for sediment evaluation; Der Einsatz von Pflanzentests bei der Sedimentbewertung

Energy Technology Data Exchange (ETDEWEB)

Feiler, U. [Bundesanstalt fuer Gewaesserkunde, Koblenz (Germany); Claus, E. [Bundesanstalt fuer Gewaesserkunde, Berlin (Germany); Heininger, P. [Bundesanstalt fuer Gewaesserkunde, Koblenz (Germany); Bundesanstalt fuer Gewaesserkunde, Berlin (Germany)

2002-07-01

The aim of the present study is to demonstrate that the use of higher plants in biotests for analyses of anthropogenically contaminated sediments yields valuable results, which may be included in a concept for the integrated assessment of waters. The results of this study prove that the selected aquatic plant, Lemna minor, is basically able to indicate contamination. In the aquatic test of the sediment extracts, it showed weak, but very selective, responses to certain classes of contaminants. Fractionating of the sample and subsequent chemical analysis combined with toxicity tests allow to narrow down the groups of substances causing toxic effects. This toxicity was confirmed by analyses of the pore waters and whole sediment samples. Together with other toxicity tests (e.g. standardized bioassays) and combined with biological benthos examinations, an overall judgment can be given for the integrated assessment of waters. (orig.) [German] Ziel der hier vorgestellten Untersuchungen war es zu zeigen, dass der Einsatz von hoeheren Pflanzen in Biotests zur Untersuchung anthropogenen belasteter Sedimente wertvolle Ergebnisse liefert, die in einem Konzept zur integrierten Gewaesserbewertung verwendet werden koennen. Die Ergebnisse dieser Arbeit machen deutlich, dass die ausgewaehlte Wasserpflanze Lemna minor Schadstoffbelastungen grundsaetzlich anzeigt. Im aquatischen Test der Sedimentextrakte weist sie eine zwar schwache, aber sehr selektive Reaktion auf bestimmte Schadstoffklassen auf. Die Fraktionierung der Probe mit anschliessender Stoffanlayse kombiniert mit Toxizitaetstests erlaubt die Eingrenzung der toxisch wirksamen Stoffgruppen. Diese toxische Belastung wurde durch die Porenwasser- und Gesamtsedimentuntersuchung bestaetigt. Zusammen mit weiteren Toxizitaetstests (z.B. standardisierte Biotests) und in Kombination mit benthosbiologischen Untersuchngen ergibt sich eine Gesamtaussage zur integrierten Gewaesserbewertung. (orig.)

A comprehensive test of evolutionarily increased competitive ability in a highly invasive plant species.

Science.gov (United States)

Joshi, Srijana; Gruntman, Michal; Bilton, Mark; Seifan, Merav; Tielbörger, Katja

2014-12-01

A common hypothesis to explain plants' invasive success is that release from natural enemies in the introduced range selects for reduced allocation to resistance traits and a subsequent increase in resources available for growth and competitive ability (evolution of increased competitive ability, EICA). However, studies that have investigated this hypothesis have been incomplete as they either did not test for all aspects of competitive ability or did not select appropriate competitors. Here, the prediction of increased competitive ability was examined with the invasive plant Lythrum salicaria (purple loosestrife) in a set of common-garden experiments that addressed these aspects by carefully distinguishing between competitive effect and response of invasive and native plants, and by using both intraspecific and interspecific competition settings with a highly vigorous neighbour, Urtica dioica (stinging nettle), which occurs in both ranges. While the intraspecific competition results showed no differences in competitive effect or response between native and invasive plants, the interspecific competition experiment revealed greater competitive response and effect of invasive plants in both biomass and seed production. The use of both intra- and interspecific competition experiments in this study revealed opposing results. While the first experiment refutes the EICA hypothesis, the second shows strong support for it, suggesting evolutionarily increased competitive ability in invasive populations of L. salicaria. It is suggested that the use of naturally co-occurring heterospecifics, rather than conspecifics, may provide a better evaluation of the possible evolutionary shift towards greater competitive ability. © The Author 2014. Published by Oxford University Press on behalf of the Annals of Botany Company. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Lead test assembly irradiation and analysis Watts Bar Nuclear Plant, Tennessee and Hanford Site, Richland, Washington

International Nuclear Information System (INIS)

1997-07-01

The U.S. Department of Energy (DOE) needs to confirm the viability of using a commercial light water reactor (CLWR) as a potential source for maintaining the nation's supply of tritium. The Proposed Action discussed in this environmental assessment is a limited scale confirmatory test that would provide DOE with information needed to assess that option. This document contains the environmental assessment results for the Lead test assembly irradiation and analysis for the Watts Bar Nuclear Plant, Tennessee, and the Hanford Site in Richland, Washington

Decontamination and remote dismantling tests in the Itrec reprocessing plant

International Nuclear Information System (INIS)

Candelieri, T.; Gerardi, A.; Soffietto, G.

1993-01-01

The scope of this research is to evaluate the advantages of the rack removal system in the dismantling of reprocessing installations. The objective of this work is to verify experimentally the possibility of the decontamination of any particular module and the capability of the remote dismantling of components installed in the mobile rack. In particular, the main objective is to develop remotely operated equipment for the dismantling of centrifugal contactors. The decontamination of the equipment which represents the most important preliminary phase of the decommissioning operation, allowed to obtain low-level radioactivity. A supporting programme has been performed in order to collect sufficient data for the project and design of the remote dismantling machine. On the basis of technological cold test results, the project of the dismantling machine's construction has been optimized. Positive results obtained during the hot dismantling operations on the Rack 6 bis attested the effectiveness of the rack removal system as an original design which facilitates decommissioning of reprocessing plants. 2 tabs., 18 figs

Engineering model cryocooler test results

International Nuclear Information System (INIS)

Skimko, M.A.; Stacy, W.D.; McCormick, J.A.

1992-01-01

This paper reports that recent testing of diaphragm-defined, Stirling-cycle machines and components has demonstrated cooling performance potential, validated the design code, and confirmed several critical operating characteristics. A breadboard cryocooler was rebuilt and tested from cryogenic to near-ambient cold end temperatures. There was a significant increase in capacity at cryogenic temperatures and the performance results compared will with code predictions at all temperatures. Further testing on a breadboard diaphragm compressor validated the calculated requirement for a minimum axial clearance between diaphragms and mating heads

Irradiation Effects Test Series: Test IE-2. Test results report

International Nuclear Information System (INIS)

Allison, C.M.; Croucher, D.W.; Ploger, S.A.; Mehner, A.S.

1977-08-01

The report describes the results of a test using four 0.97-m long PWR-type fuel rods with differences in diametral gap and cladding irradiation. The objective of this test was to provide information about the effects of these differences on fuel rod behavior during quasi-equilibrium and film boiling operation. The fuel rods were subjected to a series of preconditioning power cycles of less than 30 kW/m. Rod powers were then increased to 68 kW/m at a coolant mass flux of 4900 kg/s-m 2 . After one hour at 68 kW/m, a power-cooling-mismatch sequence was initiated by a flow reduction at constant power. At a flow of 2550 kg/s-m 2 , the onset of film boiling occurred on one rod, Rod IE-011. An additional flow reduction to 2245 kg/s-m 2 caused the onset of film boiling on the remaining three rods. Data are presented on the behavior of fuel rods during quasiequilibrium and during film boiling operation. The effects of initial gap size, cladding irradiation, rod power cycling, a rapid power increase, and sustained film boiling are discussed. These discussions are based on measured test data, preliminary postirradiation examination results, and comparisons of results with FRAP-T3 computer model calculations

Summary of CCTF test results

International Nuclear Information System (INIS)

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

1987-01-01

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

Possibilities for using plant extracts added to ruminant feed aimed at improving production results

Directory of Open Access Journals (Sweden)

Grdović Svetlana

2010-01-01

Full Text Available The use of plant extracts with the objective of improving production results and the quality of food articles of animal origin is an area which is acquiring increasing scientific importance. Numerous investigations carried out so far on ruminants and other species of domestic animals have been aimed at examining specific bioactive matter of plants. The results of these investigations have demonstrated a positive influence on the production results. A large number of data indicate that plant extracts added to animal feed contribute to increasing overall productivity. Furthermore, plant extracts as additives in animal feed have a positive effect also on the health condition of the animals. A large number of plants have characteristics which potentially improve consumption, digestibility and conversion of food, and also growth. Examinations have been performed of the effects of different plant extracts on food consumption, wool growth, growth and composition of the trunk, milk production, reproductive parameters, agents for wool shearing, preventing bloat, methane production, as well as the influence of plants on curbing nematode infestations of ruminants. This work presents a review of scientific investigations of different plant species and their effects on the production characteristics of ruminants. .

Plant critical concept

International Nuclear Information System (INIS)

O'Regan, P.J.

1995-01-01

The achievement of operation and maintenance (O ampersand M) cost reductions is a prime concern for plant operators. Initiatives by the nuclear industry to address this concern are under way and/or in development. These efforts include plant reliability studies, reliability-centered maintenance, risk ranking and testing philosophies, performance-based testing philosophies, graded quality assurance, and so forth. This paper presents the results of an effort to develop a methodology that integrates and applies the common data and analysis requirements for a number of risk-based and performance-based initiatives. This initial phase of the effort applied the methodology and its results to two initiatives. These were the procurement function and the preventive maintenance function. This effort integrated multiple programs and functions to identify those components that are truly critical from an integrated plant performance perspective. The paper describes the scope of the effort, the development of a methodology to identify plant critical components, and the application of these results to the maintenance rule compliance, preventive maintenance, and procurement functions at the candidate plant

Screening potential genotoxic effect of aquatic plant extracts using the mussel micronucleus test

Directory of Open Access Journals (Sweden)

Bettina Eck-Varanka

2016-01-01

Full Text Available Objective: To assess the genotoxic potential of selected aquatic macrophytes: Ceratophyllum demersum L. (hornwort, family Ceratophyllaceae, Typha angustifolia L. (narrowleaf cattail, family Typhaceae, Stratiotes aloides L. (water soldier, family Butomaceae, and Oenanthe aquatica (L. Poir. (water dropwort, family Umbelliferae. Methods: For genotoxicity assessment, the mussel micronucleus test was applied. Micronucleus frequency was determined from the haemolymph of Unio pictorum L. (painter’s mussel. In parallel, total and hydrolisable tannin contents were determined. Results: All plant extracts elucidated significant mutagenic effect. Significant correlation was determined between tannin content and mutagenic capacity. Conclusions: The significant correlation between genotoxicity as expressed by micronucleus frequency and tannin content (both total and hydrolisable tannins indicate that tannin is amongst the main compounds being responsible for the genotoxic potential. It might be suggested that genotoxic capacity of these plants elucidate a real ecological effect in the ecosystem.

Application of FT-IR Classification Method in Silica-Plant Extracts Composites Quality Testing

Science.gov (United States)

Bicu, A.; Drumea, V.; Mihaiescu, D. E.; Purcareanu, B.; Florea, M. A.; Trică, B.; Vasilievici, G.; Draga, S.; Buse, E.; Olariu, L.

2018-06-01

Our present work is concerned with the validation and quality testing efforts of mesoporous silica - plant extracts composites, in order to sustain the standardization process of plant-based pharmaceutical products. The synthesis of the silica support were performed by using a TEOS based synthetic route and CTAB as a template, at room temperature and normal pressure. The silica support was analyzed by advanced characterization methods (SEM, TEM, BET, DLS and FT-IR), and loaded with Calendula officinalis and Salvia officinalis standardized extracts. Further desorption studies were performed in order to prove the sustained release properties of the final materials. Intermediate and final product identification was performed by a FT-IR classification method, using the MID-range of the IR spectra, and statistical representative samples from repetitive synthetic stages. The obtained results recommend this analytical method as a fast and cost effective alternative to the classic identification methods.

The subsurface hydrology around Building 9201-2: Results of the July 1994 water level recovery test, Oak Ridge Y-12 plant, Oak Ridge, Tennessee

International Nuclear Information System (INIS)

1996-06-01

A water level recovery test was conducted at Building 9201-2 at the Oak Ridge Y-12 Plant in Oak Ridge, Tennessee, from 12:45 p.m. on July 29 until 8:22 a.m. on July 31, 1994. The purpose of the test was to improve the general understanding of the subsurface hydrology around the building. The information is needed to determine the minimum pumping capacity necessary to maintain safe water levels in the basement of the building and to assist in designing systems for treating mercury-bearing waters in the basement. The test was initiated by shutting off the three main sump pumps in Building 9201-2 (i.e., O-12, E-13, and E-22) for 43.5 hr and allowing the water in the basement to approach a static level. The pumps in sumps F-3 and P-6 were also not operating during the test. During the test, water levels were monitored in 5 sumps (P-6, O-12, F-3, E-13, and E-22); a pit near sump K-22; 4 monitoring wells or piezometers in the basement near the O-12 sump, and 16 wells outside of the building. Sump K-22 was dry during the entire test

On-line testing of response time and calibration of temperature and pressure sensors in nuclear power plants

International Nuclear Information System (INIS)

Hashemian, H.M.

1995-01-01

Periodic calibrations and response time measurements are necessary for temperature and pressure sensors in the safety systems of nuclear power plants. Conventional measurement methods require the test to be performed at the sensor location or involve removing the sensor from the process and performing the tests in a laboratory or on the bench. The conventional methods are time consuming and have the potential of causing wear and tear on the equipment, can expose the test personnel to radiation and other harsh environments, and increase the length of the plant outage. Also, the conventional methods do not account for the installation effects which may have an influence on sensor performance. On-line testing methods alleviate these problems by providing remote sensor response time and calibration capabilities. For temperature sensors such as Resistance Temperature Detectors (RTDs) and thermocouples, an on-line test method called the Loop Current Step Response (LCSR) technique has been developed, and for pressure transmitters, an on-line method called noise analysis which was available for reactor diagnostics was validated for response time testing applications. Both the LCSR and noise analysis tests are performed periodically in U.S. nuclear power plants to meet the plant technical specification requirements for response time testing of safety-related sensors. Automated testing of the calibration of both temperature and pressure sensors can be accomplished through an on-line monitoring system installed in the plant. The system monitors the DC output of the sensors over the fuel cycle to determine if any calibration drift has occurred. Changes in calibration can be detected using signal averaging and intercomparison methods and analytical redundancy techniques. (author)

Tests of the heat transfer characteristic of air cooler during cooling by natural convection of the Fast Breeder Reactor plants

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

The purpose of this study is to confirm the heat transfer characteristics of the air cooler (AC) of the Fast Breeder Reactor(FBR) which has a function to remove the residual heat of the reactor by heat exchange between sodium and air in natural convection region if electric power would be lost. In order to confirm the characteristics of the AC installed in the FBR plant, the heat transfer test by using the AC which is installed in the sodium test loop owned by Toshiba Corporation has been planned. In this study, the heat transfer characteristic tests were performed by using the AC in sodium test loop, and the CFD analyses were conducted to evaluate the test results and the heat transfer characteristics of the plant scale AC at the condition of natural convection. In addition, the elemental tests to confirm the influence of the heat transfer tube placement by using the heat transfer tube of the same specification as the AC of Monju were performed. (author)

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3B. Kozloduy NPP units 5/6: Analysis/testing. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-07-01

The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. This volume of Working material contains reports related analyses and testing of Kozloduy nuclear power plant, units 5 and 6.

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3A. Kozloduy NPP units 5/6: Analysis/testing. Working material

International Nuclear Information System (INIS)

1995-01-01

The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. This volume of Working material contains reports related analyses and testing of Kozloduy nuclear power plant, units 5 and 6

FY 1974 Report on results of Sunshine Project. Research and development of binary cycle geothermal power generation plant (Part I); 1974 nendo binary cycle chinetsu hatsuden plant no kenkyu kaihatsu seika hokokusho. 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1975-05-30

The power generation system survey/research project surveys properties and the like of secondary fluids (e.g., n-butane and chlorofluorohydrocarbons); prepares the heat balances after taking into consideration the system concepts and thermodynamic characteristics of these fluids; and completes, based on the heat balances, basic designs of the major components (e.g., 10 MW plant turbine, evaporator and condenser), equipment layouts for the power plant, piping plans, and plant control system plans. For development of the turbine, the preliminary designs are drawn, based on the existing steam and gas turbine techniques, to complete the preparations for the detailed designs. For shaft sealing devices, the plan for the test apparatus is completed, the test procedures are drawn, and the preparations for the tests are partly completed. For the heat exchangers, the preliminary designs are completed for the optimum types. It is planned that the heating and cooling tubes for the heat exchangers are surface-treated to improve heat transfer coefficients. The surface treatment and surface patterns are studied, and the treated tubes are developed on a trial basis. The test unit for evaluating their performance is designed and constructed, thus completing the preparations for the tests. The corrosion test unit is installed, and the small-size corrosion simulation unit is completed. This report covers the results up to Chapter 2, Section 3, the remainder being described in JN0040364. (NEDO)

FY 1974 Report on results of Sunshine Project. Research and development of binary cycle geothermal power generation plant (Part II); 1974 nendo binary cycle chinetsu hatsuden plant no kenkyu kaihatsu seika hokokusho. 2

Energy Technology Data Exchange (ETDEWEB)

NONE

1975-05-30

The power generation system survey/research project surveys properties and the like of secondary fluids (e.g., n-butane and chlorofluorohydrocarbons); prepares the heat balances after taking into consideration the system concepts and thermodynamic characteristics of these fluids; and completes, based on the heat balances, basic designs of the major components (e.g., 10 MW plant turbine, evaporator and condenser), equipment layouts for the power plant, piping plans, and plant control system plans. For development of the turbine, the preliminary designs are drawn, based on the existing steam and gas turbine techniques, to complete the preparations for the detailed designs. For shaft sealing devices, the plan for the test apparatus is completed, the test procedures are drawn, and the preparations for the tests are partly completed. For the heat exchangers, the preliminary designs are completed for the optimum types. It is planned that the heating and cooling tubes for the heat exchangers are surface-treated to improve heat transfer coefficients. The surface treatment and surface patterns are studied, and the treated tubes are developed on a trial basis. The test unit for evaluating their performance is designed and constructed, thus completing the preparations for the tests. The corrosion test unit is installed, and the small-size corrosion simulation unit is completed. This report covers the results in and after Chapter 2, Section 4, those before being described in JN0040363. (NEDO)

Climax granite test results

Energy Technology Data Exchange (ETDEWEB)

Ramspott, L.D.

1980-01-15

The Lawrence Livermore Laboratory (LLL), as part of the Nevada Nuclear Waste Storage Investigations (NNWSI) program, is carrying out in situ rock mechanics testing in the Climax granitic stock at the Nevada Test Site (NTS). This summary addresses only those field data taken to date that address thermomechanical modeling for a hard-rock repository. The results to be discussed include thermal measurements in a heater test that was conducted from October 1977 through July 1978, and stress and displacement measurements made during and after excavation of the canister storage drift for the Spent Fuel Test (SFT) in the Climax granite. Associated laboratory and field measurements are summarized. The rock temperature for a given applied heat load at a point in time and space can be adequately modeled with simple analytic calculations involving superposition and integration of numerous point source solutions. The input, for locations beyond about a meter from the source, can be a constant thermal conductivity and diffusivity. The value of thermal conductivity required to match the field data is as much as 25% different from laboratory-measured values. Therefore, unless we come to understand the mechanisms for this difference, a simple in situ test will be required to obtain a value for final repository design. Some sensitivity calculations have shown that the temperature field is about ten times more sensitive to conductivity than to diffusivity under the test conditions. The orthogonal array was designed to detect anisotropy. After considering all error sources, anisotropic efforts in the thermal field were less than 5 to 10%.

Optimal testing input sets for reduced diagnosis time of nuclear power plant digital electronic circuits

International Nuclear Information System (INIS)

Kim, D.S.; Seong, P.H.

1994-01-01

This paper describes the optimal testing input sets required for the fault diagnosis of the nuclear power plant digital electronic circuits. With the complicated systems such as very large scale integration (VLSI), nuclear power plant (NPP), and aircraft, testing is the major factor of the maintenance of the system. Particularly, diagnosis time grows quickly with the complexity of the component. In this research, for reduce diagnosis time the authors derived the optimal testing sets that are the minimal testing sets required for detecting the failure and for locating of the failed component. For reduced diagnosis time, the technique presented by Hayes fits best for the approach to testing sets generation among many conventional methods. However, this method has the following disadvantages: (a) it considers only the simple network (b) it concerns only whether the system is in failed state or not and does not provide the way to locate the failed component. Therefore the authors have derived the optimal testing input sets that resolve these problems by Hayes while preserving its advantages. When they applied the optimal testing sets to the automatic fault diagnosis system (AFDS) which incorporates the advanced fault diagnosis method of artificial intelligence technique, they found that the fault diagnosis using the optimal testing sets makes testing the digital electronic circuits much faster than that using exhaustive testing input sets; when they applied them to test the Universal (UV) Card which is a nuclear power plant digital input/output solid state protection system card, they reduced the testing time up to about 100 times

Germination and root elongation bioassays in six different plant species for testing Ni contamination in soil.

Science.gov (United States)

Visioli, Giovanna; Conti, Federica D; Gardi, Ciro; Menta, Cristina

2014-04-01

In vitro short-term chronic phytotoxicity germination and root elongation test were applied to test the effects of nickel (Ni) in seed germination and root elongation in six plants species: Cucumis sativus (Cucurbitaceae), Lepidium sativum and Brassica nigra (Brassicaceae), Trifolium alexandrinum and Medicago sativa (Fabaceae), Phacelia tanacetifolia (Boraginaceae). A naturally Ni rich soil was used to compare the results obtained. Unlike root elongation, germination was not affected by Ni in any of the six species tested. EC50 values, calculated on the root elongation, showed that Ni toxicity decreases in the following order: P. tanacetifolia > B. nigra > C. sativus > L. sativum > M. sativa > T. alexandrinum. The test conducted using soil elutriate revealed a significantly lower effect in both seed germination and root elongation when compared to the results obtained using untreated soil. Conversely, the test performed on soil confirmed the high sensitivity of C. sativus, P. tanacetifolia and L. sativum to Ni.

Abnormal Cervical Cancer Screening Test Results

Science.gov (United States)

... AQ FREQUENTLY ASKED QUESTIONS FAQ187 GYNECOLOGIC PROBLEMS Abnormal Cervical Cancer Screening Test Results • What is cervical cancer screening? • What causes abnormal cervical cancer screening test ...

Higher plant modelling for life support applications: first results of a simple mechanistic model

Science.gov (United States)

Hezard, Pauline; Dussap, Claude-Gilles; Sasidharan L, Swathy

2012-07-01

In the case of closed ecological life support systems, the air and water regeneration and food production are performed using microorganisms and higher plants. Wheat, rice, soybean, lettuce, tomato or other types of eatable annual plants produce fresh food while recycling CO2 into breathable oxygen. Additionally, they evaporate a large quantity of water, which can be condensed and used as potable water. This shows that recycling functions of air revitalization and food production are completely linked. Consequently, the control of a growth chamber for higher plant production has to be performed with efficient mechanistic models, in order to ensure a realistic prediction of plant behaviour, water and gas recycling whatever the environmental conditions. Purely mechanistic models of plant production in controlled environments are not available yet. This is the reason why new models must be developed and validated. This work concerns the design and test of a simplified version of a mathematical model coupling plant architecture and mass balance purposes in order to compare its results with available data of lettuce grown in closed and controlled chambers. The carbon exchange rate, water absorption and evaporation rate, biomass fresh weight as well as leaf surface are modelled and compared with available data. The model consists of four modules. The first one evaluates plant architecture, like total leaf surface, leaf area index and stem length data. The second one calculates the rate of matter and energy exchange depending on architectural and environmental data: light absorption in the canopy, CO2 uptake or release, water uptake and evapotranspiration. The third module evaluates which of the previous rates is limiting overall biomass growth; and the last one calculates biomass growth rate depending on matter exchange rates, using a global stoichiometric equation. All these rates are a set of differential equations, which are integrated with time in order to provide

A test of the herbivore optimization hypothesis using muskoxen and a graminoid meadow plant community

Directory of Open Access Journals (Sweden)

David L. Smith

1996-01-01

Full Text Available A prediction from the herbivore optimization hypothesis is that grazing by herbivores at moderate intensities will increase net above-ground primary productivity more than at lower or higher intensities. I tested this hypothesis in an area of high muskox {Ovibos moschatus density on north-central Banks Island, Northwest Territories, Canada (73°50'N, 119°53'W. Plots (1 m2 in graminoid meadows dominated by cottongrass (Eriophorum triste were either clipped, exposed to muskoxen, protected for part of one growing season, or permanently protected. This resulted in the removal of 22-44%, 10-39%, 0-39% or 0%, respectively, of shoot tissue during each growing season. Contrary to the predictions of the herbivore optimization hypothesis, productivity did not increase across this range of tissue removal. Productivity of plants clipped at 1.5 cm above ground once or twice per growing season, declined by 60+/-5% in 64% of the tests. The productivity of plants grazed by muskoxen declined by 56+/-7% in 25% of the tests. No significant change in productivity was observed in 36% and 75% of the tests in clipped and grazed treatments, respecrively. Clipping and grazing reduced below-ground standing crop except where removals were small. Grazing and clipping did not stimulate productivity of north-central Banks Island graminoid meadows.

Secondary water chemistry control practices and results of the Japanese PWR plants

International Nuclear Information System (INIS)

Maeda, Akihiro; Shoda, Yasuhiko; Ishihara, Nobuo; Murata, Kazutoyo; Fujiwara, Hiroyuki; Hayakawa, Hitoshi; Matsuda, Tadashi

2012-09-01

the concept of the Japanese water chemistry specification, the status and the results of the chemical treatment of all the Japanese PWR plants, and examples of the water chemistry improvement program and the results at the representative plants. (authors)

Irradiation Effects Test Series: Test IE-3. Test results report. [PWR

Energy Technology Data Exchange (ETDEWEB)

Farrar, L. C.; Allison, C. M.; Croucher, D. W.; Ploger, S. A.

1977-10-01

The objectives of the test reported were to: (a) determine the behavior of irradiated fuel rods subjected to a rapid power increase during which the possibility of a pellet-cladding mechanical interaction failure is enhanced and (b) determine the behavior of these fuel rods during film boiling following this rapid power increase. Test IE-3 used four 0.97-m long pressurized water reactor type fuel rods fabricated from previously irradiated fuel. The fuel rods were subjected to a preconditioning period, followed by a power ramp to 69 kW/m at a coolant mass flux of 4920 kg/s-m/sup 2/. After a flow reduction to 2120 kg/s-m/sup 2/, film boiling occurred on the fuel rods. One rod failed approximately 45 seconds after the reactor was shut down as a result of cladding embrittlement due to extensive cladding oxidation. Data are presented on the behavior of these irradiated fuel rods during steady-state operation, the power ramp, and film boiling operation. The effects of a power ramp and power ramp rates on pellet-cladding interaction are discussed. Test data are compared with FRAP-T3 computer model calculations and data from a previous Irradiation Effects test in which four irradiated fuel rods of a similar design were tested. Test IE-3 results indicate that the irradiated state of the fuel rods did not significantly affect fuel rod behavior during normal, abnormal (power ramp of 20 kW/m per minute), and accident (film boiling) conditions.

Summary of the last step of active test at separation facility and purification facility in Rokkasho Reprocessing Plant

International Nuclear Information System (INIS)

Kuroishi, Yuuki; Iseki, Tadahiro; Mitani, Akira; Takahashi, Naoki; Tsujimura, Akino; Sato, Nobuharu; Inaba, Makoto; Itagaki, Takashi

2008-01-01

During the last step of Active Test (AT) at Rokkasho Reprocessing Plant (RRP), the performance of the Separation Facility, mainly for pulsed columns and mixer-settlers were tested; Diluent washing efficiency, Plutonium extraction and stripping efficiency, Decontamination factors of fission products and Uranium and plutonium losses into wastes. Also, those of the Plutonium purification unit in the Purification Facility have been checked; Diluent washing efficiency, Plutonium extraction and stripping efficiency and Plutonium losses into wastes. Test results were equivalent to or better than expected values. (author)

78 FR 38411 - Vogtle Electric Generating Plant, Unit 4; Inspections, Tests, Analyses, and Acceptance Criteria

Science.gov (United States)

2013-06-26

... Plant, Unit 4; Inspections, Tests, Analyses, and Acceptance Criteria AGENCY: Nuclear Regulatory Commission. ACTION: Determination of inspections, tests, analyses, and acceptance criteria completion. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) staff has determined that the inspections, tests...

Performance testing of self-powered detector signal converters at Dukovany nuclear power plant - stage 1

International Nuclear Information System (INIS)

Erben, O.; Hajek, P.; Zerola, L.; Karsulin, M.

1990-11-01

The converters were manufactured at the Institute of Nuclear Research, Rez. Dynamic functions of the converters were tested during the start-up of reactor unit 4, Dukovany nuclear power plant, and their stability during its normal operation. The results and evaluation of the measurements show a good performance of converters. They have a low offset, good stability and the values of current are in a good agreement with the values obtained using other methods. The values of insulation resistance are in a good agreement with the values obtained manually using the method of additional resistance. These converters are planned to be used in the upgraded in-service inspection system in WWER-440 nuclear power plants. (Z.S.) 9 tabs., 22 figs., 1 ref

Linear Array Ultrasonic Test Results from Alkali-Silica Reaction (ASR) Specimens

Energy Technology Data Exchange (ETDEWEB)

Clayton, Dwight A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Khazanovich, Dr. Lev [Univ. of Minnesota, Minneapolis, MN (United States); Salles, Lucio [Univ. of Minnesota, Minneapolis, MN (United States)

2016-04-01

The purpose of the U.S. Department of Energy Office of Nuclear Energy’s Light Water Reactor Sustainability (LWRS) Program is to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the operating lifetimes of nuclear power plants (NPPs) beyond 60 years. Since many important safety structures in an NPP are constructed of concrete, inspection techniques must be developed and tested to evaluate the internal condition. In-service containment structures generally do not allow for the destructive measures necessary to validate the accuracy of these inspection techniques. This creates a need for comparative testing of the various nondestructive evaluation (NDE) measurement techniques on concrete specimens with known material properties, voids, internal microstructure flaws, and reinforcement locations.This report presents results of the ultrasound evaluation of four concrete slabs with varying levels of ASR damage present. This included an investigation of the experimental results, as well as a supplemental simulation considering the effect of ASR damage by elasto-dynamic wave propagation using a finite integration technique method. It was found that the Hilbert Transform Indicator (HTI), developed for quantification of freeze/thaw damage in concrete structures, could also be successfully utilized for quantification of ASR damage. internal microstructure flaws, and reinforcement locations.

Radiation mutagenesis of subtropic plants

International Nuclear Information System (INIS)

Kerkadze, I.G.

1987-01-01

Possibilities of expansion of subtropic plant changeability and development of new gene bank for future selection-genetic studies are detected. New trends of radiation mutagenesis of subtropic plants are formulated as results of studies during many years. A lot of mutants is subjected to sufficient tests, and concrete results are obtained with the help of these tests for definite species. Summing genetic and selection estimations of the results, it is possible to make the conclusion that mutant selection represents one of the powerful methods of preparation of productive and qualitative species of subtropic plants, which are successfully introduced into practice

Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant. Design Requirements Document (DRD)

Science.gov (United States)

Rigo, H. S.; Bercaw, R. W.; Burkhart, J. A.; Mroz, T. S.; Bents, D. J.; Hatch, A. M.

1981-01-01

A description and the design requirements for the 200 MWe (nominal) net output MHD Engineering Test Facility (ETF) Conceptual Design, are presented. Performance requirements for the plant are identified and process conditions are indicated at interface stations between the major systems comprising the plant. Also included are the description, functions, interfaces and requirements for each of these major systems. The lastest information (1980-1981) from the MHD technology program are integrated with elements of a conventional steam electric power generating plant.

Verification of neutron pad and 17 x 7 guide tube designs by preoperational tests on the Trojan I power plant

International Nuclear Information System (INIS)

Bloyd, C.N.; Singleton, N.R.; Ciaramitaro, W.

1976-05-01

The internals vibration measurement program carried out on the Trojan-1 reactor during preoperational testing is described. The flow induced response of a 17 x 17 guide tube and the neutron pad core barrel were deduced from the plant test data and compared with the expected responses. The results showed good agreement with expected vibration levels

Innovative test method for the estimation of the foaming tendency of substrates for biogas plants.

Science.gov (United States)

Moeller, Lucie; Eismann, Frank; Wißmann, Daniel; Nägele, Hans-Joachim; Zielonka, Simon; Müller, Roland A; Zehnsdorf, Andreas

2015-07-01

Excessive foaming in anaerobic digestion occurs at many biogas plants and can cause problems including plugged gas pipes. Unfortunately, the majority of biogas plant operators are unable to identify the causes of foaming in their biogas reactor. The occurrence of foaming is often related to the chemical composition of substrates fed to the reactor. The consistency of the digestate itself is also a crucial part of the foam formation process. Thus, no specific recommendations concerning substrates can be given in order to prevent foam formation in biogas plants. The safest way to avoid foaming is to test the foaming tendency of substrates on-site. A possible solution is offered by an innovative foaming test. With the help of this tool, biogas plant operators can evaluate the foaming disposition of new substrates prior to use in order to adjust the composition of substrate mixes. Copyright © 2015 Elsevier Ltd. All rights reserved.

Space flight research leading to the development of enhanced plant products: Results from STS-94

Science.gov (United States)

Stodieck, Louis S.; Hoehn, Alex; Heyenga, A. Gerard

1998-01-01

Products derived from plants, such as foods, pharmaceuticals, lumber, paper, oils, etc., are pervasive in everyday life and generate revenues in the hundreds of billions of dollars. Research on space-grown plants has the potential to alter quantities, properties and types of plant-derived products in beneficial ways. Research on space grown plants may help expand the utilization of this resource for Earth based benefit to an even greater extent. The use of space flight conditions may help provide a greater understanding and ultimate manipulation of the metabolic and genetic control of commercially important plant products. Companies that derive and sell plant products could significantly benefit from investing in space research and development. A flight investigation was conducted on the Shuttle mission STS-94 to establish the initial experimental conditions necessary to test the hypothesis that the exposure of certain plant forms to an adequate period of microgravity may divert the cell metabolic expenditure on structural compounds such as lignin to alternative secondary metabolic compounds which are of commercial interest. Nine species of plants were grown for 16 days in the Astro/Plant Generic Bioprocessing Apparatus (Astro/PGBA) under well-controlled environmental conditions. Approximately half of the plant species exhibited significant growth comparable with synchronous ground controls. The other flight plant species were stunted and showed signs of stress with the cause still under investigation. For the plants that grew well, analyses are underway and are expected to demonstrate the potential for space flight biotechnology research.

Prediction of high level vibration test results by use of available inelastic analysis techniques

International Nuclear Information System (INIS)

Hofmayer, C.H.; Park, Y.J.; Costello, J.F.

1991-01-01

As part of a cooperative study between the United States and Japan, the US Nuclear Regulatory Commission and the Ministry of International Trade and Industry of Japan agreed to perform a test program that would subject a large scale piping model to significant plastic strains under excitation conditions much greater than the design condition for nuclear power plants. The objective was to compare the results of the tests with state-of-the-art analyses. Comparisons were done at different excitation levels from elastic to elastic-plastic to levels where cracking was induced in the test model. The program was called the high Level Vibration Test (HLVT). The HLVT was performed on the seismic table at the Tadotsu Engineering Laboratory of Nuclear Power Engineering Test Center in Japan. The test model was constructed by modifying the 1/2.5 scale model of one loop of a PWR primary coolant system which was previously tested by NUPEC as part of their seismic proving test program. A comparison of various analysis techniques with test results shows a higher prediction error in the detailed strain values than in the overall response values. This prediction error is magnified as the plasticity in the test model increases. There is no significant difference in the peak responses between the simplified and the detailed analyses. A comparison between various detailed finite element model runs indicates that the material properties and plasticity modeling have a significant impact on the plastic strain responses under dynamic loading reversals. 5 refs., 12 figs

Power-Hardware-In-the-Loop (PHIL) Test of VSC-based HVDC connection for Offshore Wind Power Plants (WPPs)

DEFF Research Database (Denmark)

Sharma, Ranjan; Cha, Seung-Tae; Wu, Qiuwei

2011-01-01

This paper presents a power-hardware-in-the-loop (PHIL) test for an offshore wind power plant (WPP) interconnected to the onshore grid by a VSC-based HVDC connection. The intention of the PHIL test is to verify the control coordination between the plant side converter of the HVDC connection...... the successful control coordination between the WPP and the plant side VSC converter of the HVDC connection of the WPP....

United States Advanced Ultra-Supercritical Component Test Facility for 760°C Steam Power Plants ComTest Project

Energy Technology Data Exchange (ETDEWEB)

Hack, Horst [Electric Power Research Institute (EPRI); Purgert, Robert Michael [Energy Industries of Ohio

2017-12-13

Following the successful completion of a 15-year effort to develop and test materials that would allow coal-fired power plants to be operated at advanced ultra-supercritical (A-USC) steam conditions, a United States-based consortium is presently engaged in a project to build an A-USC component test facility (ComTest). A-USC steam cycles have the potential to improve cycle efficiency, reduce fuel costs, and reduce greenhouse gas emissions. Current development and demonstration efforts are focused on enabling the construction of A-USC plants, operating with steam temperatures as high as 1400°F (760°C) and steam pressures up to 5000 psi (35 MPa), which can potentially increase cycle efficiencies to 47% HHV (higher heating value), or approximately 50% LHV (lower heating value), and reduce CO₂ emissions by roughly 25%, compared to today’s U.S. fleet. A-USC technology provides a lower-cost method to reduce CO₂ emissions, compared to CO₂ capture technologies, while retaining a viable coal option for owners of coal generation assets. Among the goals of the ComTest facility are to validate that components made from advanced nickel-based alloys can operate and perform under A-USC conditions, to accelerate the development of a U.S.-based supply chain for the full complement of A-USC components, and to decrease the uncertainty of cost estimates for future A-USC power plants. The configuration of the ComTest facility would include the key A-USC technology components that were identified for expanded operational testing, including a gas-fired superheater, high-temperature steam piping, steam turbine valve, and cycling header component. Membrane walls in the superheater have been designed to operate at the full temperatures expected in a commercial A-USC boiler, but at a lower (intermediate) operating pressure. This superheater has been designed to increase the temperature of the steam supplied by the host utility boiler up to 1400°F (760�

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4C. Paks NPP: Analysis and testing. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-07-01

The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material involves comparative analysis of the seismic analysis results of the reactor building for soft soil conditions, derivation of design response spectra for components and systems; and upper range design response spectra for soft soil site conditions at Paks NPP.

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4C. Paks NPP: Analysis and testing. Working material

International Nuclear Information System (INIS)

1996-01-01

The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material involves comparative analysis of the seismic analysis results of the reactor building for soft soil conditions, derivation of design response spectra for components and systems; and upper range design response spectra for soft soil site conditions at Paks NPP

[Replacement of dogs as research animals for the approval testing of plant protection products].

Science.gov (United States)

Box, Rainer J

2006-01-01

The replacement of animal testing using dogs for the registration of plant protection products requires a long-term step-by-step procedure. The first goal should be to achieve international agreement on using only one single study in dogs. This would result in a significant short-term reduction of the use of dogs for this purpose. The competent working groups both in the EU and the United States EPA have declared this to be their intended aim. In this context, the 90-day study is to be the preferred study from the scientific as well as the animal welfare points of view. It is proposed to set up an international expert task force within the next 12 months, which should seek to initiate a process of international harmonization of the testing requirements following the example of the International Conference of Harmonization of Technical Requirements for Medical Products, ICH. The goal should be to achieve international agreement on only one single study with dogs within the next 2 to 3 years. In addition, other valid scientific procedures, with which the use of dogs for testing can be reduced, should be critically assessed. A complete replacement of the use of dogs for plant protection product testing is suggested to take place at a later stage. This may be achieved by either deriving safety threshold values by applying a safety factor to chronic NOAEL values obtained in studies using rats for those groups of substances, for which there is evidence that the dog is the more sensitive species, or by combining the chronic rat study with other animal tests stipulated for the registration of pesticides.

Development of Open Test-bed for Autonomous Operation in Nuclear Power Plants

International Nuclear Information System (INIS)

Kim, Seungmin; Heo, Gyunyoung

2017-01-01

Nuclear power plants also recognize the need for automation. However, it is dangerous technology to have a significant impact on human society. In addition, due to the uncertain legal responsibility for autonomous operation, the application and development speed of nuclear energy related automation technology will be significantly decrease compared to other industries. It is argued that the application of AI and automation technology to power plants should not be prematurely applied or not based on the principle of applying proven technology since nuclear power plants are the highest level security operated facilities. As described above, the overall algorithm of the Test Bed is an autonomous operation algorithm (rulebased algorithm, learning-based algorithm, semiautonomous operation algorithm) to judge the entry condition of the procedure through condition monitoring and to enter the appropriate operating procedure. In order to make a test bed, the investigation for the heuristic part of the existing procedures and the heuristic part from the circumstance which is not specified in the procedure is needed. In the learning based and semi-autonomous operation algorithms, using MARS to extract its operating data and operational logs and try out various diagnostic algorithms as described above. Through the completion of these future tasks, the test bed which can compared with actual operators will be constructed and that it will be able to check its effectiveness by improving competitively with other research teams through the characteristics of shared platform.

BWR Full Integral Simulation Test (FIST). Phase I test results

International Nuclear Information System (INIS)

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

1984-09-01

A new full height BWR system simulator has been built under the Full-Integral-Simulation-Test (FIST) program to investigate the system responses to various transients. The test program consists of two test phases. This report provides a summary, discussions, highlights and conclusions of the FIST Phase I tests. Eight matrix tests were conducted in the FIST Phase I. These tests have investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. Results and governing phenomena of each test have been evaluated and discussed in detail in this report. One of the FIST program objectives is to assess the TRAC code by comparisons with test data. Two pretest predictions made with TRACB02 are presented and compared with test data in this report

Safety review on unit testing of safety system software of nuclear power plant

International Nuclear Information System (INIS)

Liu Le; Zhang Qi

2013-01-01

Software unit testing has an important place in the testing of safety system software of nuclear power plants, and in the wider scope of the verification and validation. It is a comprehensive, systematic process, and its documentation shall meet the related requirements. When reviewing software unit testing, attention should be paid to the coverage of software safety requirements, the coverage of software internal structure, and the independence of the work. (authors)

Status of the flora and fauna on the Nevada Test Site, 1993. Results of continuing basic environmental monitoring, January through December 1993

International Nuclear Information System (INIS)

Hunter, R.B.

1994-09-01

This report provides the results of monitoring of plants and animals on the Nevada Test Site during calendar year 1993. Monitoring was accomplished under the Department of Energy's Basic Environmental Compliance and Monitoring Program, initiated in 1987. The program looks at both baseline study areas, chosen to represent undisturbed conditions as much as possible, and areas disturbed by Department of energy (DOE) activities or natural phenomena. DOE disturbances studied include areas blasted by above-ground nuclear tests before 1962, subsidence craters created by underground nuclear tests, road maintenance activities, areas cleared for drilling, and influences of man-made water sources. Natural phenomena studied include recovery from range fires, effects of introduced species, damage to plants by insect outbreaks, and effects of weather fluctuations. In 1993 disturbances examined included several burned areas and roadsides, a drill pad on Pahute Mesa, introduced grasses and shrub removal effects on ephemeral plants, and effects on pine trees of an infestation of pinyon needle scale insects

Status of the flora and fauna on the Nevada Test Site, 1993. Results of continuing basic environmental monitoring, January through December 1993

Energy Technology Data Exchange (ETDEWEB)

Hunter, R.B. [comp.

1994-09-01

This report provides the results of monitoring of plants and animals on the Nevada Test Site during calendar year 1993. Monitoring was accomplished under the Department of Energy`s Basic Environmental Compliance and Monitoring Program, initiated in 1987. The program looks at both baseline study areas, chosen to represent undisturbed conditions as much as possible, and areas disturbed by Department of energy (DOE) activities or natural phenomena. DOE disturbances studied include areas blasted by above-ground nuclear tests before 1962, subsidence craters created by underground nuclear tests, road maintenance activities, areas cleared for drilling, and influences of man-made water sources. Natural phenomena studied include recovery from range fires, effects of introduced species, damage to plants by insect outbreaks, and effects of weather fluctuations. In 1993 disturbances examined included several burned areas and roadsides, a drill pad on Pahute Mesa, introduced grasses and shrub removal effects on ephemeral plants, and effects on pine trees of an infestation of pinyon needle scale insects.

Veggie ISS Validation Test Results and Produce Consumption

Science.gov (United States)

Massa, Gioia; Hummerick, Mary; Spencer, LaShelle; Smith, Trent

2015-01-01

The Veggie vegetable production system flew to the International Space Station (ISS) in the spring of 2014. The first set of plants, Outredgeous red romaine lettuce, was grown, harvested, frozen, and returned to Earth in October. Ground control and flight plant tissue was sub-sectioned for microbial analysis, anthocyanin antioxidant phenolic analysis, and elemental analysis. Microbial analysis was also performed on samples swabbed on orbit from plants, Veggie bellows, and plant pillow surfaces, on water samples, and on samples of roots, media, and wick material from two returned plant pillows. Microbial levels of plants were comparable to ground controls, with some differences in community composition. The range in aerobic bacterial plate counts between individual plants was much greater in the ground controls than in flight plants. No pathogens were found. Anthocyanin concentrations were the same between ground and flight plants, while antioxidant and phenolic levels were slightly higher in flight plants. Elements varied, but key target elements for astronaut nutrition were similar between ground and flight plants. Aerobic plate counts of the flight plant pillow components were significantly higher than ground controls. Surface swab samples showed low microbial counts, with most below detection limits. Flight plant microbial levels were less than bacterial guidelines set for non-thermostabalized food and near or below those for fungi. These guidelines are not for fresh produce but are the closest approximate standards. Forward work includes the development of standards for space-grown produce. A produce consumption strategy for Veggie on ISS includes pre-flight assessments of all crops to down select candidates, wiping flight-grown plants with sanitizing food wipes, and regular Veggie hardware cleaning and microbial monitoring. Produce then could be consumed by astronauts, however some plant material would be reserved and returned for analysis. Implementation of

78 FR 15753 - Maintenance, Testing, and Replacement of Vented Lead-Acid Storage Batteries for Nuclear Power Plants

Science.gov (United States)

2013-03-12

...-Acid Storage Batteries for Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Draft...-Acid Storage Batteries for Nuclear Power Plants.'' The draft guide describes methods that the NRC staff..., testing, and replacement of vented lead-acid storage batteries in nuclear power plants. DATES: Submit...

Results of steel containment vessel model test

International Nuclear Information System (INIS)

Luk, V.K.; Ludwigsen, J.S.; Hessheimer, M.F.; Komine, Kuniaki; Matsumoto, Tomoyuki; Costello, J.F.

1998-05-01

A series of static overpressurization tests of scale models of nuclear containment structures is being conducted by Sandia National Laboratories for the Nuclear Power Engineering Corporation of Japan and the US Nuclear Regulatory Commission. Two tests are being conducted: (1) a test of a model of a steel containment vessel (SCV) and (2) a test of a model of a prestressed concrete containment vessel (PCCV). This paper summarizes the conduct of the high pressure pneumatic test of the SCV model and the results of that test. Results of this test are summarized and are compared with pretest predictions performed by the sponsoring organizations and others who participated in a blind pretest prediction effort. Questions raised by this comparison are identified and plans for posttest analysis are discussed

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

Superconducting solenoid model magnet test results

International Nuclear Information System (INIS)

Carcagno, R.; Dimarco, J.; Feher, S.; Ginsburg, C.M.; Hess, C.; Kashikhin, V.V.; Orris, D.F.; Pischalnikov, Y.; Sylvester, C.; Tartaglia, M.A.; Terechkine, I.; Tompkins, J.C.; Wokas, T.; Fermilab

2006-01-01

Superconducting solenoid magnets suitable for the room temperature front end of the Fermilab High Intensity Neutrino Source (formerly known as Proton Driver), an 8 GeV superconducting H- linac, have been designed and fabricated at Fermilab, and tested in the Fermilab Magnet Test Facility. We report here results of studies on the first model magnets in this program, including the mechanical properties during fabrication and testing in liquid helium at 4.2 K, quench performance, and magnetic field measurements. We also describe new test facility systems and instrumentation that have been developed to accomplish these tests

Superconducting solenoid model magnet test results

Energy Technology Data Exchange (ETDEWEB)

Carcagno, R.; Dimarco, J.; Feher, S.; Ginsburg, C.M.; Hess, C.; Kashikhin, V.V.; Orris, D.F.; Pischalnikov, Y.; Sylvester, C.; Tartaglia, M.A.; Terechkine, I.; /Fermilab

2006-08-01

Superconducting solenoid magnets suitable for the room temperature front end of the Fermilab High Intensity Neutrino Source (formerly known as Proton Driver), an 8 GeV superconducting H- linac, have been designed and fabricated at Fermilab, and tested in the Fermilab Magnet Test Facility. We report here results of studies on the first model magnets in this program, including the mechanical properties during fabrication and testing in liquid helium at 4.2 K, quench performance, and magnetic field measurements. We also describe new test facility systems and instrumentation that have been developed to accomplish these tests.

Preliminary results of ecotoxicological assessment of an Acid Mine Drainage (AMD) passive treatment system testing water quality of depurated lixiviates

OpenAIRE

Miguel Sarmiento, Aguasanta; Bonnail, Estefanía; Nieto Liñán, José Miguel; Valls Casillas, Tomás Ángel del

2017-01-01

The current work reports on the preliminary results of a toxicity test using screening experiments to check the efficiency of an innovative passive treatment plant designed for acid mine drainage purification. Bioassays took place with water samples before and after the treatment system and in the river, once treated water is discharged. Due to the high toxicity of the water collected at the mouth of the mine (before the treatment plant), the bioassay was designed and developed with respect t...

Failure of PWR-RHRS under cold shutdown conditions: Experimental results from the PKL test facility

International Nuclear Information System (INIS)

Mandl, R.M.; Umminger, K.J.; Logt, J.V.D.

1991-01-01

The Residual Heat Removal System (RHRS) of a PWR is designed to transfer thermal energy from the core after plant shutdown and maintain the plant in cold shutdown or refuelling conditions for extended periods of time. Initial reactor cooling after shutdown is achieved by dissipating heat through the steam generators (SGs) and discharging steam to the condenser by means of the Turbine Bypass System (TBS). When the reactor coolant temperature has dropped to about 160C and pressure has been reduced to 30 bar the RHRS is placed into operation. it reduces the coolant temperature to 50C within 20 hours after shutdown. The time margin for establishing alternate methods of heat removal following a failure of the RHRS depends on the Reactor Coolant System (RCS) temperature, the decay heat rate and the amount of RCS inventory. During some shutdown operations the RCS may be partially drained (e. g. to perform SG inspections). Decreased primary system inventory can significantly reduce the time available to recover the RHRS's function prior to bulk boiling and possible core uncovery. In the PKL test facility, which simulates a 1,300 MWe 4-loop PWR on a scale 1:145, a failure of RHRS under cold shutdown conditions was performed. This presentation gives a brief description of the test facility followed by the test objectives and results of this experiment

Test results from the GA Technologies engineering-scale off-gas treatment system

International Nuclear Information System (INIS)

Jensen, D.D.; Olguin, L.J.; Wilbourn, R.G.

1985-01-01

Test results are available from the GA Technologies (GA) off-gas treatment facilities using gas streams from both the graphite fuel element burner system and from the spent fuel dissolver. The off-gas system is part of a pilot plant for development of processes for treating spent fuel from high temperature gas-cooled reactors (HTGRs). One method for reducing the volume of HTGR fuel prior to reprocessing or spent fuel storage is to crush and burn the graphite fuel elements. The burner off-gas (BOG) contains radioactive components, principally H-3, C-14, Kr-85, I-129, and Rn-220, as well as chemical forms such as CO 2 , CO, O 2 , and SO 2 . The BOG system employs components designed to remove these constituents. Test results are reported for the iodine and SO 2 adsorbers and the CO/HT oxidizer. Integrated testing of major BOG system components confirmed the performance of units evaluated in individual tests. Design decontamination and conversion factors were maintained for up to 72 h. In a reprocessing flowsheet, the solid product from the burners is dissolved in nitric or Thorex acid. The dissolver off-gas (DOG) contains radioactive components H-3, Kr-85, I-129, Rn-220 plus chemical forms such as nitrogen oxides (NO/sub x/). In the pilot-scale system iodine is removed from the DOG by adsorption. Tests of iodine removal have been conducted using either silver-exchanged mordenite (AgZ) or AgNO 3 -impregnated silica gel (AC-6120). Although each sorbent performed well in the presence of NO/sub x/, the silica gel adsorbent proved more efficient in silver utilization and, thus, more cost effective

Qualification of nuclear power plant inspection, examination, and testing personnel. Revision 1, September 1980

International Nuclear Information System (INIS)

Anon.

1981-01-01

Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, requires that the quality assurance program provide for indoctrination and training of personnel performing activities affecting quality as necessary to ensure that quality assurance personnel achieve and maintain suitable proficiency. This guide describes a method acceptable to the NRC staff for complying with the Commission's regulations with regard to qualification of inspection, examination, and testing personnel for all types of nuclear power plants

On-plant test of TUV HCM12 and ASME T23 alloys for use as waterwall materials

DEFF Research Database (Denmark)

Karlsson, Asger; Rasmussen, Frands; Montgomery, Melanie

2000-01-01

to the stress' which appear in the water walls of an USC plant. The exposure was performed in the stress region close to the yield strength limit and cracks appeared in one of the materials. The tests have revealed that HCM12 is difficult to weld but develops no thermal fatigue crack whereas HCM2S is easy......The development of Ultra Super Critical power plants has emphasised the need for stronger alloys for use as water walls. For this purpose HCM12 and HCM2S have been tested at 2 Danish power plants. HCM12 and HCM2S have been tested for their weldability, for thermal fatigue properties...... and for their oxidation properties under steam conditions 435-520º C, 26-27 MPa. The test included application of different weld methods and testing of the mechanical properties of the weld seams. The weld parameters giving the optimal properties were chosen to manufacture two pieces of water wall (1 times 1.5 meter...

Impact of closed Brayton cycle test results on gas cooled reactor operation and safety

International Nuclear Information System (INIS)

Wright, St.A.; Pickard, P.S.

2007-01-01

This report summarizes the measurements and model predictions for a series of tests supported by the U.S. Department of Energy that were performed using the recently constructed Sandia Brayton Loop (SBL-30). From the test results we have developed steady-state power operating curves, controls methodologies, and transient data for normal and off-normal behavior, such as loss of load events, and for decay heat removal conditions after shutdown. These tests and models show that because the turbomachinery operates off of the temperature difference (between the heat source and the heat sink), that the turbomachinery can continue to operate (off of sensible heat) for long periods of time without auxiliary power. For our test hardware, operations up to one hour have been observed. This effect can provide significant operations and safety benefits for nuclear reactors that are coupled to a Brayton cycles because the operating turbomachinery continues to provide cooling to the reactor. These capabilities mean that the decay-heat removal can be accommodated by properly managing the electrical power produced by the generator/alternator. In some conditions, it may even be possible to produce sufficient power to continue operating auxiliary systems including the waste heat circulatory system. In addition, the Brayton plant impacts the consequences of off-normal and accident events including loss of load and loss of on-site power. We have observed that for a loss of load or a loss of on-site power event, with a reactor scram, the transient consists initially of a turbomachinery speed increase to a new stable operating point. Because the turbomachinery is still spinning, the reactor is still being cooled provided the ultimate heat sink remains available. These highly desirable operational characteristics were observed in the Sandia Brayton loop. This type of behavior is also predicted by our models. Ultimately, these results provide the designers the opportunity to design gas

Test of job performance aids for power plants. Final report

International Nuclear Information System (INIS)

Shriver, E.L.; Zach, S.E.; Foley, J.P. Jr.

1982-10-01

The objective of EPRI Research Project 1396-1 was to evaluate the applicability and effectiveness of Job Performance Aids (JPAs) in nuclear power plant situations. For over twenty years, JPAs have been developed in military situations to meet the problems of confusing, incomplete, and inaccurate procedures on maintenance jobs. Kinton, Incorporated of Alexandria, Virginia applied the military experience with JPAs to nuclear power plant situations and identified potential benefits in terms of cost reductions and improved performance. Sample JPAs were developed for Control Room Operations, Maintenance, Plant Operations, Instrumentation and Control, Health Physics, and Quality Assurance tasks (procedures) in selected nuclear plants. JPAs were also developed for a prototype condenser tube leak detection system in the design stage, as well as for generic classes of circuit breaker equipment. Based on the results of the study, the use of JPAs is recommended for plant procedures of medium to high difficulty and for those tasks performed infrequently, even if fairly simple

FY 1980 Report on results of Sunshine Project. Development of coal liquefaction techniques (Development of materials for the coal liquefaction plant); 1980 nendo sekitan ekika gijutsu no kaihatsu seika hokokusho. Sekitan ekika plant zairyo no kaihatsu

Energy Technology Data Exchange (ETDEWEB)

NONE

1981-03-01

A 1 T/D solvent-extraction type coal liquefaction test plant was constructed and operated to obtain the technical data for the design of, and establish the techniques for, an efficient coal liquefaction plant. The FY 1980 program includes surveys on the materials for coal liquefaction plants, covering those already developed and under development, to clarify the problematical points; drafting the test schedules; and conceptual designs of the material testing facilities. The major problems involved in the materials for coal liquefaction plants include erosion by fluidizing coal slurry, hydrogen embrittlement of the reactor materials, and corrosion by the liquefaction products (e.g., stress-corrosion cracking of austenitic steel, and corrosion by organic acids). The surveys on materials research trends suggest that USA seems to concentrate their research efforts on the reactor materials. The corrosion tests are mostly of in-plant tests, but the stress corrosion and slurry erosion tests are conducted on a laboratory scale. The conceptual designs are drawn for some testing units, e.g., the loop type material testing unit and basic testing unit for jet-spray type slurry erosion. (NEDO)