Design, Fabrication, Test Report of the Material Capsule(08M-10K) with Double Thermal Media for High-temperature Irradiation

International Nuclear Information System (INIS)

Cho, Man Soon; Choo, K. N.; Kang, Y. H.; Sohn, J. M.; Shin, Y. T.; Park, S. J.; Kim, B. G.; Oh, S. Y.

2010-01-01

To overcome the restriction of the irradiation test at a high temperature of the existing material capsule with Al thermal media, a capsule suitable for the irradiation at the high temperature was developed and the performance test was undertaken. The 08M-10K capsule was designed and fabricated as that with double thermal media to verify the structural and external integrity in the high-temperature irradiation higher than 500 .deg. C. The thermal performance test was undertaken at the out-pile test facility, and the soundness of the double thermal media was confirmed with the naked eye after disassembling the capsule. Though the temperatures of the specimens reach 500±20 .deg. C as a result maintaining the capsule during 5 hours after setting the specimens temperatures in the target range, the high-temperature thermal media with double structure was confirmed to maintain the soundness. And the specimens and the thermal media were heated to 600 .deg. C for about 3 minutes, but the thermal media were maintained sound. Especially, the Al thermal media were heated for 5 hours in range of 500±20 .deg. C and for 3 minutes at the temperature of 600 .deg. C. However, the thermal media were confirmed to maintain the soundness. Whether a capsule has only Al thermal media or the high-temperature thermal media with double structure, any capsule can be used in the range of 500±20 .deg. C as the result of this experiment maintaining the specimens high-temperature

Post-irradiation examinations of inert matrix nitride fuel irradiated in JMTR (01F-51A capsule)

International Nuclear Information System (INIS)

Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Honda, Junichi; Hatakeyama, Yuichi; Ono, Katsuto; Matsui, Hiroki; Arai, Yasuo

2007-03-01

A plutonium nitride fuel pin containing inert matrix such as ZrN and TiN was encapsulated in 01F-51A and irradiated in JMTR. Minor actinides are surrogated by plutonium. Average linear powers and burnups were 408W/cm, 30000MWd/t(Zr+Pu) [132000MWd/t-Pu] for (Zr,Pu)N and 355W/cm, 38000MWd/t(Ti+Pu) [153000MWd/t-Pu] for (TiN,PuN). The irradiated capsule was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in the irradiated fuel pin. Very low fission gas release rate of about 1.6% was measured. The inner surface of cladding tube did not show any signs of chemical interaction with fuel pellet. (author)

Performance test of the I and C system (GSF - 2002) for the irradiation tests using a fuel capsule

Energy Technology Data Exchange (ETDEWEB)

Kang, Young Hwan; Park, S. J.; Kim, B. G.; Ahn, D. H

2004-12-01

HANARO is a very important facility in Korea. It offers various types of irradiation tests of nuclear fuels and materials. With the various applications of the HANARO capsule for the academic and industrial applications, new technologies and relevant facilities will become more important especially for the advanced nuclear fuels and materials development. A new I and C system for an irradiation test using an instrumented fuel capsule have been designed and manufactured to provide more qualified data to fuel developer. The performance test which started in 2004, was done to investigate the thermal response of the capsule connected to the gas mixing system of the new I and C system(GSF-2002) in the cold test loop under the HANARO hydraulic operational condition. Main test parameters are mass flow rate of 25, 50 and 100 cc/min of He/Ne gas, gas pressure of 1 to 3 kg/cm{sup 2}, heater power of 1 to 3.4kW and different gas mixing ratios of He to Ne. From the out-pile tests, it was confirmed that the I and C system(GSF-2002) would be feasible for the fuel irradiation tests. Both analytical and test data prepared by this study are directly used for the fuel experiments related to advanced fuel development program.

The development of the neutron flux measurement technology using SPNDs during nuclear fuel irradiation test

Energy Technology Data Exchange (ETDEWEB)

Kim, B. G.; Kang, Y. H.; Cho, M. S.; Joo, K. N.; Choi, M. H.; Park, S. J.; Shin, Y. T.; Oh, J. M.; Kim, Y. J

2004-03-01

As a part of the development of instrumentation technologies for a nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), a study is performed to measure and evaluate the neutron flux at the same position as the nuclear fuel during irradiation test using the SPND(Self Powered Neutron Detector). To perform this study, rhodium type SPNDs and amplifier are selected suitable to irradiation test, and the selected SPNDs are installed in instrumented fuel capsule(02F-11K). The irradiation test using a instrumented fuel capsule are performed in the OR5 vertical hole of HANARO for about 54 days, and SPND output signals are acquired successfully during irradiation test. Acquired SPND signals are analyzed and evaluated as a reliable data by COSMOS Code. This will be utilized for the fuel related research together with fuel center temperature and reactor operation data.

Development status of irradiation devices and instrumentation for material and nuclear fuel irradiation tests in HANARO

Energy Technology Data Exchange (ETDEWEB)

Kim, Bong Goo; Sohn, Jae Min; Choo, Kee Nam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-04-15

The High flux Advanced Neutron Application ReactOr (HANARO), an open-tank-in-pool type reactor, is one of the multi-purpose research reactors in the world. Since the commencement of HANARO's operations in 1995, a significant number of experimental facilities have been developed and installed at HANARO, and continued efforts to develop more facilities are in progress. Owing to the stable operation of the reactor and its frequent utilization, more experimental facilities are being continuously added to satisfy various fields of study and diverse applications. The irradiation testing equipment for nuclear fuels and materials at HANARO can be classified into capsules and the Fuel Test Loop (FTL). Capsules for irradiation tests of nuclear fuels in HANARO have been developed for use under the dry conditions of the coolant and materials at HANARO and are now successfully utilized to perform irradiation tests. The FTL can be used to conduct irradiation testing of a nuclear fuel under the operating conditions of commercial nuclear power plants. During irradiation tests conducted using these capsules in HANARO, instruments such as the thermocouple, Linear Variable Differential Transformer (LVDT), small heater, Fluence Monitor (F/M) and Self-Powered Neutron Detector (SPND) are used to measure various characteristics of the nuclear fuel and irradiated material. This paper describes not only the status of HANARO and the status and perspective of irradiation devices and instrumentation for carrying out nuclear fuel and material tests in HANARO but also some results from instrumentation during irradiation tests

Development of out-of-pile version of instrumented irradiation capsule for determination of online creep deformation

International Nuclear Information System (INIS)

Venkatesu, Sadu; Saxena, Rajesh; Chaurasia, P.K.; Muthuganesh, M.; Murugan, S.; Venugopal, S.

2016-01-01

Materials used for fuel cladding and structural components in fast reactors can undergo significant dimensional and physical changes due to exposure to high energy neutrons. At high temperatures in nuclear environment, material undergoes considerable deformation due to thermal and irradiation creep. Diametral increase of fuel pin due to thermal and irradiation creep, apart from irradiation swelling, reduces the coolant flow area around the fuel pins affecting the effective removal of heat generated in the fuel pins. The changes due to creep can be determined by two types of material irradiation tests in reactor. The first type includes non-instrumented irradiation tests with specimen dimensional evaluations carried out in post-irradiation examinations. The second type includes instrumented irradiation tests with online monitoring and/or controlling of test conditions and real time measurement of changes in dimensions of the specimen. During instrumented irradiation tests, parameters such as specimen temperature, the load exerted on the specimen, specimen elongation, etc. can be monitored and/or controlled using suitable components such as linear variable differential transformers (LVDTs), bellows, thermocouples, etc. Instrumented irradiation experiments in reactors are relatively complex in design but can provide full information on the experimental parameters. Such benefits provide motivation for development of instrumented irradiation capsule to measure creep behavior online during in-pile instrumented irradiation tests. Out-of-pile version of the instrumented irradiation capsule for determination of online creep deformation has been developed and tested in the furnace by raising the temperature gradually up to 330 °C. This paper discusses the details of the design, assembly of experimental set up and experimental results of the out-of-pile version of instrumented capsule developed in our laboratory for determination of online creep deformation. (author)

A study on the measurement and evaluation of neutron flux using SPNDs during nuclear fuel irradiation test

Energy Technology Data Exchange (ETDEWEB)

Son, J. M.; Kim, B. K.; Oh, J. M.; Park, S. J.; Lee, B. H.; Seo, C. G.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

2003-10-01

As a part of the development of instrumentation technologies for a nuclear fuel irradiation test in HANARO(High-Flux Advanced Nuclear Application Reactor), a study is performed to measure and evaluate the neutron flux at the same position as the nuclear fuel during irradiation test using the SPND(Self Powered Neutron Detector). To perform this study, rhodium type SPNDs and amplifier are selected suitable to irradiation test, and the selected SPNDs are installed in instrumented fuel capsule(02F-11K). The irradiation test using a instrumented fuel capsule are performed in the OR5 vertical hole of HANARO for about 54 days, and SPND output signals are acquired successfully during irradiation test. Acquired SPND signals are analyzed and evaluated as a reliable data by COSMOS Code, and this will be utilized for the fuel related research together with fuel center temperature and reactor operation data.

Design Improvements of a Fuel Capsule for Re-irradiation Tests

International Nuclear Information System (INIS)

Kang, Young-Hwan; Choi, Myung-Hwan; Kim, Jong Kiun; Youm, Ki Un; Yoon, Ki Byeong; Kim, Bong Goo

2006-01-01

The development of an advanced reactor system such as the next generation nuclear plant and other generation IV systems require new fuels, claddings, and structural materials. To characterize the performance of these new materials, it is necessary for us to have leading-edge technology to satisfy the specific test requirements of the recent R and D activities such as the high-fluence- and high burnup- related tests. Thus, new capsule assembling technology and re-instrumentation technology has been developed to meet the demands for the high burnup test at HANARO since 2003. In 2003, a mockup of the capsule assembly machine was designed and fabricated. The performance test which started in 2004 was undertaken to determine and present the main performance characteristics of the capsule assembly machine (CAM) including the special tools. In 2005, a series of analyses using a finite element analysis program, ANSYS and full scale tests in air were performed to improve the design of the capsule's components for an effective utilization of the CAM. The handling tools were fully qualified through the performance tests in 2006. KAERI is now reviewing the water flow area in the top region of a fuel capsule main body for re-irradiation tests and optimizing the design of the central region area of a capsule to be joined with special bolts

F2 phenomenological test on fuel motion (Interim report)

International Nuclear Information System (INIS)

Palm, R.G.; Fink, C.L.; Stewart, R.R.; Gehl, S.M.; Rothman, A.B.

1976-09-01

TREAT F-series tests are being conducted to provide data on fuel motion at accident power levels from one to about ten times design for use in development of fuel motion models. Test F2 was conducted to evaluate motion of high power fuel in a hypothetical LMFBR unprotected TUC (transient undercooling) accident. Fuel and fuel-boundary conditions following coolant boiling and dryout under TUC conditions are achieved in each F-series test with a single fuel element surrounded by a nuclear heated wall in a dry test capsule. Test F2 was conducted with a low burnup but restructured fuel element to investigate the effect of fuel vapor pressure on fuel motion. Results are presented and discussed

Design, fabrication, and operation of capsules for the irradiation testing of candidate advanced space reactor fuel pins

International Nuclear Information System (INIS)

Thoms, K.R.

1975-04-01

Fuel irradiation experiments were designed, built, and operated to test uranium mononitride (UN) fuel clad in tungsten-lined T-111 (Ta-8 percent W-2 percent Hf) and uranium dioxide (UO 2 ) fuel clad in both tungsten-lined T-111 and tungsten-lined Nb-1 percent Zr. A total of nine fuel pins was irradiated (four containing porous UN, two containing dense, nonporous UN, and three containing dense UO 2 ) at average cladding temperatures ranging from 931 to 1015 0 C. The UN experiments, capsules UN-4 and -5, operated for 10,480 and 10,037 hr, respectively, at an average linear heat generation rate of 10 kW/ft. The UO 2 experiment, capsule UN-6, operated for 8333 hr at an average linear heat generation rate of approximately 5 kW/ft. Following irradiation, the nine fuel pins were removed from their capsules, externally examined, and sent to the NASA Plum Brook Facility for more detailed postirradiation examination. During visual examination, it was discovered that the cladding of the fuel pin containing dense UN in each of capsules UN-4 and -5 had failed, exposing the UN fuel to the NaK in which the pins were submerged and permitting the release of fission gas from the failed pins. A rough analysis of the fission gas seen in samples of the gas in the fuel pin region indicated fission gas release-to-birth rates from these fuel pins in the range of 10 -5 . (U.S.)

SrF2 capsule design for heat engine applications

International Nuclear Information System (INIS)

Lester, D.H.

1976-04-01

A number of design changes were considered to improve heat transfer characteristics of the WESF capsule. This capsule was evaluated in a design concept for use as a heat source in a helium-working fluid, Stirling heat engine. Throughout the study a heat block concept was used. The helium was assumed to be at 1200 0 F and 200 atm. The upper temperature limit at the fuel-metal interface was assumed to be 800 0 C because of material compatibility considerations. A 0.6-in. thick outer can was considered since it may be required for impact resistance and high pressure accident environments. The modifications considered were: (1) filling all gaps with helium rather than air, (2) filling gaps with powdered metal, and (3) adding a third can to the existing capsule. Also, enhancement of emissivity on metal surfaces was considered as a possible modification

Strontium and cesium radionuclide leak detection alternatives in a capsule storage pool

International Nuclear Information System (INIS)

Larson, D.E.; Crawford, T.W.; Joyce, S.M.

1981-08-01

A study was performed to assess radionuclide leak-detection systems for use in locating a capsule leaking strontium-90 or cesium-137 into a water-filled pool. Each storage pool contains about 35,000 L of water and up to 715 capsules, each of which contains up to 150 kCi strontium-90 or 80 kCi cesium-137. Potential systems assessed included instrumental chemical analyses, radionuclide detection, visual examination, and other nondestructive nuclear-fuel examination techniques. Factors considered in the assessment include: cost, simplicity of maintenance and operation, technology availability, reliability, remote operation, sensitivity, and ability to locate an individual leaking capsule in its storage location. The study concluded that an adaption of the spent nuclear-fuel examination technique of wet sipping be considered for adaption. In the suggested approoch, samples would be taken continuously from pool water adjacent to the capsule(s) being examined for remote radiation detection. In-place capsule isolation and subsequent water sampling would confirm that a capsule was leaking radionuclides. Additional studies are needed before implementing this option. Two other techniques that show promise are ultrasonic testing and eddy-current testing

Preparation of U-Si/U-Me (Me = Fe, Ni, Mn) aluminum-dispersion plate-type fuel (miniplates) for capsule irradiation

International Nuclear Information System (INIS)

Ugajin, Mitsuhiro; Itoh, Akinori; Akabori, Mitsuo

1993-06-01

Details of equipment installed, method adopted and final products were described on the preparation of uranium silicides and other fuels for capsule irradiation. Main emphasis was placed on the preparation of laboratory-scale aluminum-dispersion plate-type fuel (miniplates) loaded to the first and second JMTR silicide capsules. Fuels contained in the capsules are as follows: (A) uranium-silicide base alloys U 3 Si 2 , Mo- added U 3 Si 2 , U 3 Si 2 +U 3 Si, U 3 Si 2 +USi, U 3 Si, U 3 (Si 0.8 Ge 0.2 ), U 3 (Si 0.6 Ge 0.4 ) (B) U 6 Me-type alloys with higher uranium density U 6 Mn, U 6 Ni, U 6 (Fe 0.4 Ni 0.6 ), U 6 (Fe 0.6 Mn 0.4 ) The powder-metallurgical picture-frame method was adopted and laboratory-scale technique was established for the preparation of miniplates. As a result of inspection for capsule irradiation, miniplates were prepared to meet the requirements of specification. (author)

Thermal analysis of an instrumented capsule using an ANSYS program

International Nuclear Information System (INIS)

Choi, Myoung Hwan; Choo, Kee Nam; Kang, Young Hwan; Cho, Man Soon; Sohn, Jae Min; Kim, Bong Goo

2006-01-01

An instrumented capsule has been used for an irradiation test of various nuclear materials in the research reactor, HANARO. To obtain the design data of the instrumented capsule, a thermal analysis is performed using a finite element analysis program, ANSYS. The 2-dimensional model for a cross section of the capsule including the specimens is generated, and a gamma-heating rate of the materials for the HANARO power of 24 or 30 MW is considered as an input force. The effect of the gap size and the control rod position on the temperature of the specimens or other components is discussed. From the analysis it is found that the gap between the thermal media and the external tube has a significant effect on the temperature of the specimen. In the case of the material capsule, the maximum temperature for the reactor power of 24 MW is 255degC for an irradiation test and 257degC for a FE analysis at the center stage of the capsule in the axial direction. It is expected that the analysis models using an ANSYS program will be useful in designing the instrumented capsules for an irradiation test and estimating the test results. (author)

Development of MOX fuel database

International Nuclear Information System (INIS)

Ikusawa, Yoshihisa; Ozawa, Takayuki

2007-03-01

We developed MOX Fuel Database, which included valuable data from several irradiation tests in FUGEN and Halden reactor, for help of LWR MOX use. This database includes the data of fabrication and irradiation, and the results of post-irradiation examinations for seven fuel assemblies, i.e. P06, P2R, E03, E06, E07, E08 and E09, irradiated in FUGEN. The highest pellet peak burn-up reached ∼48GWd/t in MOX fuels, of which the maximum plutonium content was ∼6 wt%, irradiated in E09 fuel assembly without any failure. Also the data from the instrumented MOX fuels irradiated in HBWR to study the irradiation behavior of BWR MOX fuels under the steady state condition (IFA-514/565 and IFA-529), under the load-follow operation condition (IFA-554/555) and under the transit condition (IFA-591) are included in this database. The highest assembly burn-up reached ∼56 GWd/t in IFA-565 steady state irradiation test, and the maximum linear power of MOX fuel rods was 58.3-68.4 kW/m without any failure in IFA-591 ramp test. In addition, valuable instrument data, i.e. cladding elongation, fuel stack elongation, fuel center temperature and rod inner pressure were obtained from IFA-554/555 load-follow test. (author)

Vented fuel experiment for gas-cooled fast reactor application

International Nuclear Information System (INIS)

Longest, A.W.; Gat, U.; Conlin, J.A.; Campana, R.J.

1975-01-01

A pressure-equalized and vented fuel rod is being irradiated in an instrumented capsule designated GB-10 to approximately 100 MWd/kg-heavy metal. The fuel is a sol-gel derived 88 atom-percent uranium (approximately 9 percent 235 U) 12 atom-percent plutonium oxide, and the cladding is 20 percent cold-worked 316 stainless steel. The capsule is being irradiated in the Oak Ridge Research Reactor (ORR) and has exceeded a burnup of 70 MWd/kg. The fuel has been operated at linear power rates of 39 and 44 kW/ m, and peak outer cladding temperature of 565 0 and 630 0 C respectively. A similar fuel rod in a previous capsule (GB-9) was subjected to 48 kW/m (685 0 C). 4 references. (auth)

Final Report on Design, Fabrication and Test of HANARO Instrumented Capsule (07M-13N) for the Researches of Irradiation Performance of Parts of X-Gen Nuclear Fuel Assembly

Energy Technology Data Exchange (ETDEWEB)

Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

2008-08-15

An instrumented capsule of 07M-13N was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of the parts of a X-Gen nuclear fuel assembly for PWR requested by KNF. Some specimens of control rod materials of AP1000 reactor requested by Westinghouse Co. were inserted in this capsule as a preliminary irradiation test and Polyimide specimens requested by Hanyang university were also inserted. 463 specimens such as buckling and spring test specimens of cell spacer grid, tensile, microstructure and tensile of welded parts, irradiation growth, spring test specimens made of HANA tube, Zirlo, Zircaloy-4, Inconel-718, Polyimide, Ag and Ag-In-Cd alloys were placed in the capsule. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 7 sets of neutron fluence monitors installed in the capsule. A new friction welded tube between STS304 and Al1050 alloys was introduced in the capsule to prevent a coolant leakage into a capsule during a capsule cutting process in HANARO. The capsule was irradiated for 95.19 days (4 cycles) in the CT test hole of HANARO of a 30MW thermal output at 230 {approx} 420 .deg. C. The specimens were irradiated up to a maximum fast neutron fluence of 1.27x10{sup 21}(n/cm{sup 2}) (E>1.0MeV) and the dpa of the irradiated specimens were evaluated as 1.21 {approx} 1.97. The irradiated specimens were tested to evaluate the irradiation performance of the parts of an X-Gen fuel assembly in the IMEF hot cell and the obtained results will be very valuable for the related researches of the users.

Ignition capsules with aerogel-supported liquid DT fuel for the National Ignition Facility

Directory of Open Access Journals (Sweden)

Ho D.D.-M.

2013-11-01

Full Text Available For high repetition-rate fusion power plant applications, capsules with aerogel-supported liquid DT fuel can have much reduced fill time compared to β-layering a solid DT fuel layer. The melting point of liquid DT can be lowered once liquid DT is embedded in an aerogel matrix, and the DT vapor density is consequently closer to the desired density for optimal capsule design requirement. We present design for NIF-scale aerogel-filled capsules based on 1-D and 2-D simulations. An optimal configuration is obtained when the outer radius is increased until the clean fuel fraction is within 65 – 75% at peak velocity. A scan (in ablator and fuel thickness parameter space is used to optimize the capsule configurations. The optimized aerogel-filled capsule has good low-mode robustness and acceptable high-mode mix.

Fabrication of the instrumented fuel rods for the 3-Pin Fuel Test Loop at HANARO

International Nuclear Information System (INIS)

Sohn, Jae Min; Park, Sung Jae; Shin, Yoon Tag; Lee, Jong Min; Ahn, Sung Ho; Kim, Soo Sung; Kim, Bong Goo; Kim, Young Ki; Lee, Ki Hong; Kim, Kwan Hyun

2008-09-01

The 3-Pin Fuel Test Loop(hereinafter referred to as the '3-Pin FTL') facility has been installed at HANARO(High-flux Advanced Neutron Application Reactor) and the 3-Pin FTL is under a test operation. The purpose of this report is to fabricate the instrumented fuel rods for the 3-Pin FTL. The fabrication of these fuel rods was based on experiences and technologies of the instrumented fuel rods for an irradiation fuel capsule. The three instrumented fuel rods of the 3-Pin FTL have been designed. The one fuel rod(180 .deg. ) was designed to measure the centerline temperature of the nuclear fuels and the internal pressure of the fuel rod, and others(60 .deg. and 300 .deg. ) were designed to measure the centerline temperature of the fuel pellets. The claddings were made of the reference material 1 and 2 and new material 1 and 2. And nuclear fuel was used UO 2 (2.0w/o) pellet type with large grain and standard grain. The major procedures of fabrication are followings: (1) the assembling and weld of fuel rods with the pellet mockups and the sensor mockups for the qualification tests, (2) the qualification tests(dimension measurements, tensile tests, metallography examinations and helium leak tests) of weld, (3) the assembling and weld of instrumented fuel rods with the nuclear pellets and the sensors for the irradiation test, and (4) the qualification tests(the helium leak test, the dimensional measurement, electric resistance measurements of sensors) of test fuel rods. Satisfactory results were obtained for all the qualification tests of the instrumented fuel rods for the 3-Pin FTL. Therefore the three instrumented fuel rods for the 3-Pin FTL have been fabricated successfully. These will be installed in the In-Pile Section of 3-Pin FTL. And the irradiation test of these fuel rods is planned from the early next year for about 3 years at HANARO

Preliminary Study of the Onset of Nucleate Boiling (ONB) for the Thermal-hydraulic Design of HANARO Irradiation non-instrumented Capsule during the Natural Convection

Energy Technology Data Exchange (ETDEWEB)

Nam, Kyungho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The HANARO reactor is an open-tank-in-pool type for easy access, and the capsules are being utilized for the irradiation test of materials and nuclear fuel in HANARO. The concept of the capsule is the direct contact with the coolant to cool the temperature of specimen down. To successfully accomplish the irradiation test, it is essential that the capsule should be designed considering the thermal margin such as the margin to Onset of Nucleate Boiling (ONB), the margin to Departure from Nucleate Boiling (DNB). In this paper, the preliminary study was performed by focusing on the ONB and the capsule design will be performed using the heat flux and temperature at ONB condition calculated in this paper. In this paper, the temperature and heat flux under ONB condition are simply calculated for the thermal design of fuel capsule for irradiation test. These values will be considered to design the non-instrumented capsule for natural circulation. To confirm the calculated value, detailed calculation will be performed using the one dimensional and multi-dimensional codes.

Irradiation Performance of HTGR Fuel in WWR-K Research Reactor

International Nuclear Information System (INIS)

Ueta, Shohei; Sakaba, Nariaki; Shaimerdenov, Asset; Gizatulin, Shamil; Chekushina, Lyudmila; Chakrov, Petr; Honda, Masaki; Takahashi, Masashi; Kitagawa, Kenichi

2014-01-01

A capsule irradiation test with the high temperature gas-cooled reactor (HTGR) fuel is being carried out using WWR-K research reactor in the Institute of Nuclear Physics of the Republic of Kazakhstan (INP) to attain 100 GWd/t-U of burnup under normal operating condition of a practical small-sized HTGR. This is the first HTGR fuel irradiation test for INP in Kazakhstan collaborated with Japan Atomic Energy Agency (JAEA) in frame of International Science and Technology Center (ISTC) project. In the test, TRISO coated fuel particle with low-enriched UO_2 (less than 10 % of "2"3"5U) is used, which was newly designed by JAEA to extend burnup up to 100 GWd/t-U comparing with that of the HTTR (33 GWd/t-U). Both TRISO and fuel compact as the irradiation test specimen were fabricated in basis of the HTTR fuel technology by Nuclear Fuel Industries, Ltd. in Japan. A helium-gas-swept capsule and a swept-gas sampling device installed in WWR-K were designed and constructed by INP. The irradiation test has been started in October 2012 and will be completed up to the end of February 2015. The irradiation test is in the progress up to 69 GWd/t of burnup, and integrity of new TRISO fuel has been confirmed. In addition, as predicted by the fuel design, fission gas release was observed due to additional failure of as-fabricated SiC-defective fuel. (author)

Instrument for measuring fuel cladding strain

International Nuclear Information System (INIS)

Billeter, T.R.

1976-01-01

Development work to provide instrumentation for the continuous measurement of strain of material specimens such as nuclear fuel cladding has shown that a microwave sensor and associated instrumentation hold promise. The cylindrical sensor body enclosing the specimen results in a coaxial resonator absorbing microwave energy at frequencies dependent upon the diameter of the specimen. Diametral changes of a microinch can be resolved with use of the instrumentation. Very reasonable values of elastic strain were measured at 75 0 F and 1000 0 F for an internally pressurized 20 percent C.W. 316 stainless steel specimen simulating nuclear fuel cladding. The instrument also indicated the creep strain of the same specimen pressurized at 6500 psi and at a temperature of 1000 0 F for a period of 700 hours. Although the indicated strain appears greater than actual, the sensor/specimen unit experienced considerable oxidation even though an inert gas purge persisted throughout the test duration. By monitoring at least two modes of resonance, the measured strain was shown to be nearly independent of sensor temperature. To prevent oxidation, a second test was performed in which the specimen/sensor units were contained in an evacuated enclosure. The strain of the two prepressurized specimens as indicated by the microwave instrumentation agreed very closely with pre- and post-test measurements obtained with use of a laser interferometer

Irradiation data analysis and thermal analysis of the 02M-02K capsule for material irradiation test

International Nuclear Information System (INIS)

Choi, Myoung Hwan; Choo, K. N.; Kang, Y. H.; Kim, B. G.; Cho, M. S.; Sohn, J. M.; Shin, Y. T.; Park, S. J.; Kim, Y. J.

2004-11-01

In order to evaluate the fracture toughness of RPV materials, the material irradiation test using the instrumented capsule (02M-02K) were carried out in the HANARO in August 2003. Based on the user's requirements the thermal design analysis of the capsule 02M-02K was performed, and the specimens were suitably arranged in each step of the capsule main body. In this report, both the temperature data of specimens measured during irradiation test and the calculated data from the thermal analysis are compared and evaluated. Also, the temperature profile in each step with the HANARO reactor power and helium pressure is reviewed and evaluated. The effects of the gap size such as theoretically calculated from thermal expansion during irradiation test and measured one in the manufacturing of the capsule on the specimen temperature were reviewed. The thermal analysis was performed by using a Finite Element (FE) analysis program, ANSYS. Two-dimensional model for the 1/4 section of the capsule is generated, and the γ-heating rate of the materials used in the capsule at the control rod position of 430 mm is used as input data. The thermal analysis using a 3-dimensional model, which is quite similar to the actual shape of the capsule, is also conducted to obtain the temperature distribution in the axial direction. The analysis results show that the temperature difference between the top and bottom positions of a specimen is found to be smaller than 13.2 .deg. C. The maximum measured and calculated temperature in the step 3 of the capsule is 256 .deg. C and 264 .deg. C, respectively. The measured temperature data are obtained at the reactor power of 24 MW, the heater power of 0 W and the helium pressure of 760 torr. Generally, the temperature data obtained by the FE analysis are slightly lower than those of the measured except the step 1 of the capsule. However, the temperature difference between the measured and the calculated shows a good agreement within 9 percent. It is

Estimation of burnup with cesium isotopes based on gamma-scanning of a instrumented fuel capsule(02F-11K) in hot-cell

International Nuclear Information System (INIS)

Song, Ung Sup; Kim, Hee Moon; Park, Dae Gyu; Paik, Seung Je; Lee, Hong Gi; Choo, Yong Sun; Hong Kwon Pyo

2004-01-01

Many experimental inspection have been performed to obtain the burnup of fuel. In the case, chemical analysis were popular with high reliability. High radioactivity of fuel was severe problem during destructive procedure. Afterward, many researchers have studied calculation of burnup using gamma detector as the non-destructive method. methodologies of gamma-scanning test have been developed as well as higher accuracy of detector. Generally, Cs-137 and Cs-134 are standard isotopes for long-term cooling spent fuel to estimate burnup, because atomic ratio of them follows the linearity with burnup

The Shielding Analysis for the I and C System(GSF-2002) Development of an Instrumented Fuel Capsule

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong Su; Kang, Young Hwan; Kim, Bong Goo; Park, Chang Je

2006-09-15

The Origen-2 code have been used to calculate the amount of nuclides generated in this study. It was validated by a comparison with the values calculated by this code and the measured ones by gathering the fission gases from two spent fuel rod (burned PWR fuel rods in Kori and Yungkwang). Error rate the between two values was estimated to be less than 10% in this study. The quantities of the nuclides generated by fission, calculated by the Origen-2, were larger than that by Origen-ARP. These differences of the two values are 1Ci to 40Ci. If the fuel is not intact, even if operated in a closed system, radioactive gases will be released to the gas flow tube from the instrumented capsule and the I and C system (GSF-2002) In a previous case study after 27 nuclides of Kr, Xe, I and Br series have been chosen, source term was generated. R/B ratio, used in this case, was calculated to serve this analysis with a reference to the ANS-5.4 model. And the leakage ratio adopted the result of the leak test in the PLUTO reactor. Specific points in the work place were determined and the radioactivity dose rates were calculated by running the Microshield 6.20. code at these points. Equivalent dose rates were 4.82X10-2{approx}3.74 {mu}Sv/hr at each point. These were distributed within the safety guide(6.25{mu}Sv/hr). In the failure case, almost all of the fission gases in the flow tube will be sucked into the delay tank, thus the condition of this tank will be a high radioactive area. Therefore the delay tank should be shielded by heavy materials and the thicknesses of them should be calculated with 9 cm and 20.5cm depending on performing conditions.

Irradiation capsules VISA-2a-f, chapter VI

International Nuclear Information System (INIS)

Pavicevic, M.

1962-01-01

Irradiation capsules VISA-2a, b,c,d, and e were constructed in Saclay according to the drawings from Vinca and according to the demand of the experimentators. This chapter VI includes documentation for each type of capsule, review about each experiment within the VISA-2 project, the objective and purpose of the experiment as well as experimental device. Irradiation capsule VISA-2f was placed in the RA reactor core in September 1962. It was completely manufactured in Vinca including sample holders and leak tight shells. It will remain in the reactor core for about month in order to obtain the integral fast neutron flux [sr

Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

Science.gov (United States)

Mcdonald, G. E.

1973-01-01

A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.

VHTR-fuel irradiation capsules for VT-1 hole of JRR-2

International Nuclear Information System (INIS)

Kikuchi, Teruo; Kikuchi, Akira; Tobita, Tsutomu; Kashimura, Satoru; Miyasaka, Yasuhiko

1977-02-01

Irradiations of VHTR fuels were made in the VT-1 irradiation hole of JRR-2. Three capsules, VP-1, VP-2 and VP-4, which contained fuel compacts, were irradiated for 300 hr at temperatures of 950 0 , 1370 0 and 1500 0 C up to the estimated burn-ups of 0.74, 0.87 and 0.80%FIMA, respectively. And, to study the amoeba effect of fuel particles, two capsules, VP-3 and VP-5, were irradiated for 300 hr at temperatures of 1650 0 and 1670 0 C up to the estimated burn-ups of 0.38 and 0.33%FIMA, respectively. (auth.)

Modular nuclear fuel element, modular capsule for a such element and fabrication process for a modular capsule

International Nuclear Information System (INIS)

Chotard, A.

1988-01-01

The nuclear fuel rod is made by a tubular casing closed at both ends and containing a series of modular capsules with little play with the casing and made by a jacket closed by porous plugs at both ends and containing a stack of fuel pellets [fr

Final report on development and operation of instrumented irradiation capsules for creep experiments on nuclear fuels at FR2

International Nuclear Information System (INIS)

Haefner, H.E.; Philipp, K.; Blumhofer, M.

1980-02-01

The capsule test rig No. 154 removed from FR2 in April 1979 was the last irradiation rig in a long series of creep experiments. The target of the irradiation tests, started exactly ten years ago, was to investigate the creep behaviour of various ceramic nuclear fuels under different in-pile irradiation conditions. An irradiation test rig had been developed for this purpose which allowed the continuous measurement of changes in length of fuel specimens. A total of 28 capsule test rigs each containing two packages of creep specimens have been irradiated in FR2 during this decade. They included 23 specimen stacks of UO 2 , 16 specimen stacks of UO 2 -PuO 2 , 4 specimen stacks of UN, 10 specimen stacks of (U,Pu) C, and 13 reference specimens of molybdenum. Besides the description of the test facility, the report provides above all a survey of the operation data applicable to the specimens and of the operating experience gathered as well as of the findings obtained in post-irradiation examinations. (orig.) [de

Capsule Development and Utilization for Material Irradiation Tests

International Nuclear Information System (INIS)

Kang, Young Hwan; Kim, B. G.; Joo, K. N.

2003-05-01

The objective of this project was to establish basic capsule irradiation technology using the multi-purpose research reactor [HANARO] to eventually support national R and D projects of advanced fuel and materials related to domestic nuclear power plants and next generation reactors. There are several national nuclear projects in KAERI, which require several irradiation tests to investigate in-pile behavior of nuclear reactor fuel and materials for the R and D of several types of fuels such as advanced PWR and DUPIC fuels and for the R and D of structural materials such as RPV(reactor pressure vessel) steel, Inconel, zirconium alloy, and stainless steel. At the moment, internal and external researchers in institutes, industries and universities are interested in investigating the irradiation characteristics of materials using the irradiation facilities of HANARO. For these kinds of material irradiation tests, it is important to develop various capsules using our own techniques. The development of capsules requires several leading-edge technologies and our own experiences related to design and fabrication. In the second phase from April 1,2000 to March 31, 2003, the utilization technologies were developed using various sensors for the measurements of temperature, pressure and displacement, and instrumented capsule technologies for the required fuel irradiation tests were developed. In addition, the improvement of the existing capsule technologies and the development of an in-situ measurable creep capsule for specific purposes were done to meet the various requirements of users

Endurance test for DUPIC capsule

International Nuclear Information System (INIS)

Chung, Heung June; Bae, K. K.; Lee, C. Y.; Park, J. M.; Ryu, J. S.

1999-07-01

This report presents the pressure drop, vibration and endurance test results for mini-plate fuel rig which were designed fabricately by KAERI. From the pressure drop test results, it is noted that the flow rate across the capsule corresponding to the pressure drop of 200 kPa is measured to be about 9.632 kg/sec. Vibration frequency for the capsule ranges from 14 to 18.5 Hz. RMS (Root Mean Square) displacement for the fuel rig is less than 14 μm, and the maximum displacement is less than 54 μm. Based on the endurance test results, the appreciable fretting wear for the DUPIC capsule was not detected. Oxidation on the support tube is observed, also tiny trace of wear between contact points observed. (author). 4 refs., 10 tabs., 45 figs

Design, Fabrication and Test Report on a Verification Capsule (05M-06K) for the Control of a Neutron Irradiation Fluence of Specimens in HANARO

Energy Technology Data Exchange (ETDEWEB)

Choo, K. N.; Kim, B. G.; Kang, Y. H.; Cho, M. S.; Son, J. M.; Shin, Y. T.; Park, S. J.; Choi, M. H.; Lee, D. S.

2007-02-15

As a part of a project for a capsule development and utilization for an irradiation test, a verification capsule (05M-06K) was designed, fabricated and tested for the development of new instrumented capsule technology for a more precise control of the irradiation fluence of a specimen, irrespective of the reactor operation condition. The basic structure of the 05M-06K capsule was based on the 04M-22K mock-up capsule which was successfully designed and out-pile tested to confirm the various key technologies necessary for the fluence control of a specimen. 21 square and round shaped specimens made of STS 304 were inserted into the capsule. The capsule was constructed in 5 stages with specimens and an independent electric heater at each stage. Each of the five specimens which were accommodated in the 1st stage (top) of the capsule can be taken out of the HANARO core during a normal reactor operation. The specimen is extracted by a specimen extraction mechanism using a steel wire. During the out-pile test, the temperatures of the specimens were measured by 12 thermocouples installed in the capsule. The capsule was successfully out-pile tested in a single channel test loop. The obtained results will be used for a safety evaluation of the new irradiation capsule for controlling the irradiation fluence of specimens in HANARO.

Investigation of special capsule technologies for material in-pile irradiation test and development plan in HANARO

Energy Technology Data Exchange (ETDEWEB)

Cho, M. S.; Son, J. M.; Kim, D. S.; Park, S. J.; Cho, Y. G.; Seo, C. K.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

2002-10-01

In-pile test for several materials such as Zr alloy, stainless steel, Cr-Ni steel etc. which are used as structural material of the advanced reactor and KNGR(Korea Next Generation Reactor) like SMART, is necessary to produce the design data for developing new reactor materials. Advanced countries like USA, Europe and Japan etc. are not only performing the simple irradiation test for materials, but developing many kinds of special capsule to perform in-pile test having special purpose. For the special test items of fuel rod, fission products, total heat generation, swelling, deformation, sweep gas, temperature ramping and BOCA etc. are being actively concerned. There are capsules measuring creep, fatigue, crack growth, and controlling fluence etc. for special irradiation test of materials. In addition, the advanced countries are developing several instrument technologies suitable for the special capsules. In HANARO, non-instrumented, instrumented material capsules and non-instrumented fuel capsule have been developed and they have been utilized in the irradiation test for users, and creep capsule loading single specimen was made and is planned to test in the reactor soon. For some forthcoming years, special capsules not only measuring creep deformation with multi-specimens, fatigue, controlling fluence but crack propagation and gas sweep considering the requirements of users will be developed in HANARO.

Capsule Development and Utilization for Material Irradiation Tests

Energy Technology Data Exchange (ETDEWEB)

Kim, Bong Goo; Kang, Y H; Cho, M S [and others

2007-06-15

The essential technology for an irradiation test of materials and nuclear fuel has been successively developed and utilized to meet the user's requirements in Phase I(July 21, 1997 to March 31, 2000). It enables irradiation tests to be performed for a non-fissile material under a temperature control(300{+-}10 .deg. C) in a He gas environment, and most of the irradiation tests for the internal and external users are able to be conducted effectively. The basic technology was established to irradiate a nuclear fuel, and a creep capsule was also developed to measure the creep property of a material during an irradiation test in HANARO in Phase II(April 1, 2000 to March 31, 2003). The development of a specific purpose capsule, essential technology for a re-irradiation of a nuclear fuel, advanced technology for an irradiation of materials and a nuclear fuel were performed in Phase III(April 1, 2003 to February 28, 2007). Therefore, the technology for an irradiation test was established to support the irradiation of materials and a nuclear fuel which is required for the National Nuclear R and D Programs. In addition, an improvement of the existing capsule design and fabrication technology, and the development of an instrumented capsule for a nuclear fuel and a specific purpose will be able to satisfy the user's requirements. In order to support the irradiation test of materials and a nuclear fuel for developing the next generation nuclear system, it is also necessary to continuously improve the design and fabrication technology of the existing capsule and the irradiation technology.

HRB-22 capsule irradiation test for HTGR fuel. JAERI/USDOE collaborative irradiation test

Energy Technology Data Exchange (ETDEWEB)

Minato, Kazuo; Sawa, Kazuhiro; Fukuda, Kousaku [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; and others

1998-03-01

As a JAERI/USDOE collaborative irradiation test for high-temperature gas-cooled reactor fuel, JAERI fuel compacts were irradiated in the HRB-22 irradiation capsule in the High Flux Isotope Reactor at the Oak Ridge National Laboratory (ORNL). Postirradiation examinations also were performed at ORNL. This report describes 1) the preirradiation characterization of the irradiation samples of annular-shaped fuel compacts containing the Triso-coated fuel particles, 2) the irradiation conditions and fission gas releases during the irradiation to measure the performance of the coated particle fuel, 3) the postirradiation examinations of the disassembled capsule involving visual inspection, metrology, ceramography and gamma-ray spectrometry of the samples, and 4) the accident condition tests on the irradiated fuels at 1600 to 1800degC to obtain information about fuel performance and fission product release behavior under accident conditions. (author)

Design, fabrication and irradiation test report on HANARO instrumented capsule (05M-07U) for the researches of universities in 2005

Energy Technology Data Exchange (ETDEWEB)

Choo, K. N.; Kim, B. G.; Kang, Y. H.; Choi, M. H.; Cho, M. S.; Son, J. M.; Choi, M. H.; Shin, Y. T.; Park, S. J.

2006-09-15

As a part of the 2005 project for an active utilization of HANARO, an instrumented capsule (05M-07U) was designed, fabricated and irradiated for an irradiation test of various unclear materials under irradiation conditions which was requested by external researchers from universities. The basic structure of the 05M-07U capsule was based on the 00M-01U, 01M-05U, 02M-05U, 03M-06U and 04M-07U capsules which had been successfully irradiated in HANARO as part of the 2000, 2001, 2002, 2003 and 2004 projects. However, because of a limited number of specimens and the budget of one university, the remaining space in the capsule was filled with various KAERI specimens for researches on a nuclear core and SMART materials, and parts of a nuclear fuel assembly of KNFC. Various types of specimens such as tensile, Charpy, TEM, hardness, compression and growth specimens made of Zr 702, Ti and Ni alloys, Zirlo, Inconel, STS 316L and Cr-Mo alloys were placed in the capsule. Especially, this capsule was designed to evaluate the nuclear characteristics of the parts of a nuclear fuel assembly and the Ti tubes in HANARO. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 5 sets of Ni-Ti-Fe neutron fluence monitors installed in the capsule. The capsule was irradiated in the CT test hole of HANARO of a 30MW thermal output at 270 ∼ 400 .deg. C up to a fast neutron fluence of 5.7 x 10{sup 20} (n/cm{sup 2}) (E >1.0MeV). The obtained results will be very valuable for the related research of the users.

Development of target capsules for muon catalyzed fusion experiments

International Nuclear Information System (INIS)

Watts, K.D.; Jones, S.E.; Caffrey, A.J.

1983-01-01

A series of Muon Catalyzed Fusion experiments has been conducted at the Los Alamos Meson Physics Facility to determine how many fusion reactions one muon would catalyze under various temperature, pressure, contamination, and tritium concentration conditions. Target capsules to contain deuterium and tritium at elevated temperatures and pressures were engineered for a maximum temperature of 540 K (512 0 F) and a maximum pressure of 103 MPa (15,000 psig). Experimental data collected with these capsules indicated that the number of fusion reactions per muon continued to increase with temperature up to the 540-K design limit. Theory had indicated that the reaction rate should peak at approximately 540 K, but this was not confirmed during the experiments. A second generation of capsules which have a maximum design temperature of 800 K (980 0 F) and a maximum design pressure of 103 MPa (15,000 psig) has now been engineered. These new capsules will be used to further study the muon catalysis rate versus deuterium-tritium mixture temperature

Membrane support of accelerated fuel capsules for inertial fusion energy reactors

International Nuclear Information System (INIS)

Petzoldt, R.W.; Moir, R.W.

1993-01-01

The use of a thin membrane to suspend an (inertial fusion energy) fuel capsule in a holder for injection into a reactor chamber is investigated. Capsule displacement and membrane deformation angle are calculated for an axisymmetric geometry for a range of membrane strain and capsule size. This information is used to calculate maximum target accelerations. Membranes must be thin (perhaps of order one micron) to minimize their effect on capsule implosion symmetry. For example, a 5 μm thick cryogenic mylar membrane is calculated to allow 1,000 m/s 2 acceleration of a 3 mm radius, 100 mg capsule. Vibration analysis (for a single membrane support) shows that if membrane vibration is not deliberately minimized, allowed acceleration may be reduced by a factor of four. A two membrane alternative geometry would allow several times greater acceleration. Therefore, alternative membrane geometry's should be used to provide greater target acceleration potential and reduce capsule displacement within the holder (for a given membrane thickness)

Capsule Development and Utilization for Material Irradiation Tests

Energy Technology Data Exchange (ETDEWEB)

Kim, Bong Goo; Kang, Y. H.; Cho, M. S. (and others)

2007-06-15

The essential technology for an irradiation test of materials and nuclear fuel has been successively developed and utilized to meet the user's requirements in Phase I(July 21, 1997 to March 31, 2000). It enables irradiation tests to be performed for a non-fissile material under a temperature control(300{+-}10 .deg. C) in a He gas environment, and most of the irradiation tests for the internal and external users are able to be conducted effectively. The basic technology was established to irradiate a nuclear fuel, and a creep capsule was also developed to measure the creep property of a material during an irradiation test in HANARO in Phase II(April 1, 2000 to March 31, 2003). The development of a specific purpose capsule, essential technology for a re-irradiation of a nuclear fuel, advanced technology for an irradiation of materials and a nuclear fuel were performed in Phase III(April 1, 2003 to February 28, 2007). Therefore, the technology for an irradiation test was established to support the irradiation of materials and a nuclear fuel which is required for the National Nuclear R and D Programs. In addition, an improvement of the existing capsule design and fabrication technology, and the development of an instrumented capsule for a nuclear fuel and a specific purpose will be able to satisfy the user's requirements. In order to support the irradiation test of materials and a nuclear fuel for developing the next generation nuclear system, it is also necessary to continuously improve the design and fabrication technology of the existing capsule and the irradiation technology.

LOFT advanced fuel rod instrumentation development

International Nuclear Information System (INIS)

Billeter, T.R.; Brown, R.L.; Chan, A.I.Y.; Day, C.K.; Meyers, S.C.; Sheen, E.M.; Stringer, J.L.

1978-01-01

Advanced fuel rod instrumentation for the Loss of Fluid Test (LOFT) reactor is being developed by the Hanford Engineering Development Laboratory for the Nuclear Regulatory Commission. This effort calls for development of sensors to measure fuel rod axial motion, fuel centerline temperature (to 2200 0 C), fuel rod plenum gas pressure (to 2500 psig), and plenum gas temperature (to 1500 0 F). A parallel test and evaluation of several modified commercial sensors was undertaken and will result in commercial availability of the final qualified sensors. Necessary test facilities were prepared for the development and evaluation effort. Tests to date indicate a three coil Linear Variable Differential Transformer (LVDT), operated from temperature compensating signal source and processing electronics, will meet the desired requirements

Irradiation test plan of the simulated DUPIC fuel

Energy Technology Data Exchange (ETDEWEB)

Bae, Ki Kwang; Yang, M. S.; Kim, B. K. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-11-01

Simulated DUPIC fuel had been irradiated from Aug. 4, 1999 to Oct. 4 1999, in order to produce the data of its in-core behavior, to verify the design of DUPIC non-instrumented capsule developed, and to ensure the irradiation requirements of DUPIC fuel at HANARO. The welding process was certified for manufacturing the mini-element, and simulated DUPIC fuel rods were manufactured with simulated DUPIC pellets through examination and test. The non-instrumented capsule for a irradiation test of DUPIC fuel has been designed and manufactured referring to the design specification of the HANARO fuel. This is to be the design basis of the instrumented capsule under consideration. The verification experiment, whether the capsule loaded in the OR4 hole meet the HANARO requirements under the normal operation condition, as well as the structural analysis was carried out. The items for this experiment were the pressure drop test, vibration test, integrity test, et. al. It was noted that each experimental result meet the HANARO operational requirements. For the safety analysis of the DUPIC non-instrumented capsule loaded in the HANARO core, the nuclear/mechanical compatibility, thermodynamic compatibility, integrity analysis of the irradiation samples according to the reactor condition as well as the safety analysis of the HANARO were performed. Besides, the core reactivity effects were discussed during the irradiation test of the DUPIC capsule. The average power of each fuel rod in the DUPIC capsule was calculated, and maximal linear power reflecting the axial peaking power factor from the MCNP results was evaluated. From these calculation results, the HANARO core safety was evaluated. At the end of this report, similar overseas cases were introduced. 9 refs., 16 figs., 10 tabs. (Author)

Biosynthesis of the polysialic acid capsule of Escherichia coli K1: factors influencing cessation of capsule expression at 150C

International Nuclear Information System (INIS)

Merker, R.I.

1987-01-01

Initial experiments were designed to determine if increases in unsaturated fatty acids (UFA) that usually occur in cells grown at 15 0 C were related to defects in membrane-associated sialyltransferase (ST) activity at 15 0 C. An E. coli K1 hybrid strain that did not increase UFA levels after growth at 15 0 C due to a mutant fabF gene was constructed. Isogenic strains with and without the fabF defect produced capsule at 33 0 C but not at 15 0 C. Membranous ST complexes isolated from both strains grown at 33 0 C transfered [ 14 C]-sialic acid (NeuNAC) from CMP-[ 14 C]-NeuNAc to endogeneous acceptors and to exogenous sialyl oligomers. Membranes from 15 0 C grown cells of the fabF + strain catalyzed incorporation of [ 14 C]NeuNAc from CMP-[ 14 C]-NeuNAc to exogenous sialyl oligomers, but required 2-4 h incubation at 33 0 C for endogenous incorporation. Membranes from the fabF mutant strain grown at 15 0 C did not incorporate [ 14 C]NeuNAc from CMP-[ 14 C]-NeuNAc under these conditions. We concluded that membrane-associated ST activity is not interrupted by low temperature increases in UFA content. Acapsular mutants derived from E. coli K1 that were defective in NeuNAc catabolism (NeuNAc aldolase) and activation or polymerization were used to examine the effects of growth at 15 0 C on NeuNAc synthesis and initiation of polysialic acid capsule synthesis. These strains accumulated high internal NeuNAc internal NeuNAc at 37 0 C, but NeuNAc was undetectable after growth at 15 0 C. Intracellular NeuNAc levels increased within 10 min. after shift from 15 0 C to 37 0 C even in the presence of rifampicin (100 g ml -1 ) or chloramphenicol (100 g ml -1 ). Extracts from these strains grown at 15 0 C and 37 0 C lacked NeuNAc synthase activity in 15 0 C assays, but were active in 37 0 C assays. We conclude that NeuNAc synthase is present but nonfunctional at 15 0 C

Performance of HTGR fuel in HFIR capsule HT-33

International Nuclear Information System (INIS)

Tiegs, T.N.; Robbins, J.M.

1979-06-01

Irradiation capsule HT-33 was a cooperative effort between General Atomic Company (GA) and Oak Ridge National Laboratory (ORNL). In this capsule ThO 2 particles (fabricated by GA), low-enriched uranium particles, inert carbon particles, and various fuel rod matrices were tested under accelerated irradiation in the High-Flux Isotope Reactor. Visual examination showed good irradiation behavior for fuel rods with slug-injected matrices (using a pitch binder) and warm-molded matrices (using a thermosetting resin binder). Rod debonding improved somewhat with fuel rods that used GLCC H-451 ground graphite shim particles rather than Speer fluid coke shim particles. Measurements of permeability (by inert gas intrusion) of the pyrocarbon on the inert particles showed that the disorder created by the neutron flux did not increase the inert gas permeability. Metallographic examination of Triso-coated particles irradiated both with and without an outer pyrocarbon coating revealed that the outer coating is necessary to suppress SiC degradation at temperatures above approximately 1375 0 C. The fission product behavior (determined by the electron microprobe) was similar in both low-enriched and high-enriched uranium particles made from weak-acid resins. Furthermore, fission product palladium caused severe SiC corrosion at time-averaged temperatures above 1400 0 C

Dissolved inorganic carbon, temperature, salinity and other variables collected from discrete sample and profile observations using CTD, bottle and other instruments from the METEOR in the North Atlantic Ocean from 1997-08-15 to 1997-09-09 (NODC Accession 0113914)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — NODC Accession 0113914 includes chemical, discrete sample, physical and profile data collected from METEOR in the North Atlantic Ocean from 1997-08-15 to 1997-09-09...

Reactor transients tests for SNR fuel elements in HFR reactor

International Nuclear Information System (INIS)

Plitz, H.

1989-01-01

In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

Irradiation behaviors of coated fuel particles, (3)

International Nuclear Information System (INIS)

Fukuda, Kousaku; Kashimura, Satoru; Iwamoto, Kazumi; Ikawa, Katsuichi

1980-07-01

This report is concerning to the irradiation experiments of the coated fuel particles, which were performed by 72F-6A and 72F-7A capsules in JMTR. The coated particles referred to the preliminary design of VHTR were prepared for the experiments in 1972 and 1973. 72F-6A capsule was irradiated at G-10 hole of JMTR fuel zone for 2 reactor cycles, and 72F-7A capsule had been planned to be irradiated at the same irradiation hole before 72F-6A. However, due to slight leak of the gaseous fission products into the vacuum system controlling irradiation temperature, irradiation of 72F-7A capsule was ceased after 85 hrs since the beginning. In the post irradiation examination, inspection to surface appearance, ceramography, X-ray microradiography and acid leaching for the irradiated particle samples were made, and crushing strength of the two particle samples was measured. (author)

New In-pile Instrumentation to Support Fuel Cycle Research and Development

Energy Technology Data Exchange (ETDEWEB)

J. Rempe; H. MacLean; R. Schley; D. Hurley; J. Daw; S. Taylor; J. Smith; J. Svoboda; D. Kotter; D. Knudson; M. Guers; S. C. Wilkins

2011-01-01

New and enhanced nuclear fuels are a key enabler for new and improved reactor technologies. For example, the goals of the next generation nuclear plant (NGNP) will not be met without irradiations successfully demonstrating the safety and reliability of new fuels. Likewise, fuel reliability has become paramount in ensuring the competitiveness of nuclear power plants. Recently, the Office of Nuclear Energy in the Department of Energy (DOE-NE) launched a new direction in fuel research and development that emphasizes an approach relying on first principle models to develop optimized fuel designs that offer significant improvements over current fuels. To facilitate this approach, high fidelity, real-time, data are essential for characterizing the performance of new fuels during irradiation testing. A three-year strategic research program is proposed for developing the required test vehicles with sensors of unprecedented accuracy and resolution for obtaining the data needed to characterize three-dimensional changes in fuel microstructure during irradiation testing. When implemented, this strategy will yield test capsule designs that are instrumented with new sensor technologies for the Advanced Test Reactor (ATR) and other irradiation locations for the Fuel Cycle Research and Development (FC R&D) program. Prior laboratory testing, and as needed, irradiation testing, of these sensors will have been completed to give sufficient confidence that the irradiation tests will yield the required data. Obtaining these sensors must draw upon the expertise of a wide-range of organizations not currently supporting nuclear fuels research. This document defines this strategic program and provides the necessary background information related to fuel irradiation testing, desired parameters for detection, and an overview of currently available in-pile instrumentation. In addition, candidate sensor technologies are identified in this document, and a list of proposed criteria for ranking

The structures of bacteriophages K1E and K1-5 explain processive degradation of polysaccharide capsules and evolution of new host specificities.

Science.gov (United States)

Leiman, Petr G; Battisti, Anthony J; Bowman, Valorie D; Stummeyer, Katharina; Mühlenhoff, Martina; Gerardy-Schahn, Rita; Scholl, Dean; Molineux, Ian J

2007-08-17

External polysaccharides of many pathogenic bacteria form capsules protecting the bacteria from the animal immune system and phage infection. However, some bacteriophages can digest these capsules using glycosidases displayed on the phage particle. We have utilized cryo-electron microscopy to determine the structures of phages K1E and K1-5 and thereby establish the mechanism by which these phages attain and switch their host specificity. Using a specific glycosidase, both phages penetrate the capsule and infect the neuroinvasive human pathogen Escherichia coli K1. In addition to the K1-specific glycosidase, each K1-5 particle carries a second enzyme that allows it to infect E. coli K5, whose capsule is chemically different from that of K1. The enzymes are organized into a multiprotein complex attached via an adapter protein to the virus portal vertex, through which the DNA is ejected during infection. The structure of the complex suggests a mechanism for the apparent processivity of degradation that occurs as the phage drills through the polysaccharide capsule. The enzymes recognize the adapter protein by a conserved N-terminal sequence, providing a mechanism for phages to acquire different enzymes and thus to evolve new host specificities.

THERMAL NEUTRON FLUX MAPPING ON A TARGET CAPSULE AT RABBIT FACILITY OF RSG-GAS REACTOR FOR USE IN k0-INAA

Directory of Open Access Journals (Sweden)

Sutisna Sutisna

2015-03-01

Full Text Available Instrumental neutron activation analysis based on the k0 method (k0-INAA requires the availability of the accurate reactor parameter data, in particular a thermal neutron flux that interact with a targets inside the target capsule. This research aims to determine and map the thermal neutron flux inside the capsule and irradiation channels used for the elemental quantification using the k0-AANI. Mapping of the thermal neutron flux (фth on two type of irradiation capsule have been done for RS01 and RS02 facilities of RSG-GAS reactor. Thermal neutron flux determined using Al-0,1%Au alloy through 197Au(n,g 198Au nuclear reaction, while the flux mapping done using statistics R. Thermal neutron flux are calculated using k0-IAEA software provided by IAEA. The results showed the average thermal neutron flux is (5.6±0.3×10+13 n.cm-2.s-1; (5.6±0.4×10+13 n.cm-2.s-1; (5.2±0.4×10+13 n.cm-2.s-1 and (5.3±0.4×10+13 n.cm-2.s-1 for Polyethylene capsule of 1st , 2nd, 3rd and 4th layer respectively. In the case of Aluminum capsule, the thermal neutron flux was lower compared to that on Polyethylene capsule. There were (3.0±0.2×10+13 n.cm-2.s-1; (2.8±0.1×10+13 n.cm-2.s-1; (3.2±0.3×10+13 n.cm-2.s-1 for 1st, 2nd and 3rd layers respectively. For each layer in the capsule, the thermal neutron flux is not uniform and it was no degradation flux in the axial direction, both for polyethylene and aluminum capsules. Contour map of eight layer on polyethylene capsule and six layers on aluminum capsule for RS01 and RS02 irradiation channels had a similar pattern with a small diversity for all type of the irradiation capsule. Keywords: thermal neutron, flux, capsule, NAA   Analisis aktivasi neutron instrumental berbasis metode k0 (k0-AANI memerlukan ketersediaan data parameter reaktor yang akurat, khususnya data fluks neutron termal yang berinteraksi dengan inti sasaran di dalam kapsul target. Penelitian ini bertujuan menentukan dan memetakan fluks neutron termal

40 CFR 600.510-08 - Calculation of average fuel economy.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Calculation of average fuel economy. 600.510-08 Section 600.510-08 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy Regulations for Model Year 1978 Passenger Automobiles...

A study on the development of instrumented capsule for the material irradiation test

Energy Technology Data Exchange (ETDEWEB)

Kang, Young Hwan; Park, J M; Choo, K N; Maeng, W Y; Park, D K; Oh, J M; Park, S J; Jung, S H; Park, J S; Kim, T R; Park, J H; Yang, S Y; Jun, Y K; Yang, S H

1997-08-01

Extensive efforts have been made to establish design and manufacturing technology for the instrumented capsule and its related system, which should be compatible with the HANARO`s characteristics. (author). 86 refs., 45 tabs., 146 figs.

Post-Irradiation Examination of Fuel Pin R54-F20A, Irradiated in a NaK Environment. RCN Report

International Nuclear Information System (INIS)

Kwast, H.

1972-12-01

Fuel pin R54-F20A has been irradiated in a NaK-environment. Temperature measurements in the NaK were carried out at average linear fission powers of 552 and 825 W/cm respectively. A maximum average canning temperature of 920°C was reached. The fuel pin was irradiated for about 50 minutes at the maximum irradiation conditions, while the total irradiation time was two hours. The irradiation had to be broken off before the end condition was reached because of malfunctioning of the fuelfailure detection system. No power peaking did occur at the upper and lower interfaces between the 50%-enriched UO 2 - and the natural UO 2 + 8 w/o UB 4 pellet. About 35% of the fuel has molten, but the fuel pin did not fail. The irradiation has been carried out in the Poolside Facility (PSF) of the High Flux Reactor (HFR) at Petten. (author)

Effect of anode firing on the performance of lanthanum and nickel co-doped SrTiO3 (La0.2Sr0.8Ti0.9Ni0.1O3-δ) anode of solid oxide fuel cell

Science.gov (United States)

Park, Byung Hyun; Choi, Gyeong Man

2015-10-01

Perovskite oxides have potential for use as alternative anode materials in solid oxide fuel cells (SOFCs) due to stability in anode atmosphere; donor-doped SrTiO3 (e.g., La0.2Sr0.8TiO3-δ) is a good candidate for this purpose. Electro-catalytic nanoparticles can be produced in oxide anodes by the ex-solution method, e.g., by incorporating Ni into a perovskite oxide in air, then reducing the oxide in H2 atmosphere. In this study, we varied the temperature (1100, 1250 °C) and atmosphere (air, H2) of La0.2Sr0.8Ti0.9Ni0.1O3-δ (LSTN) anode firing to control the degree of Ni ex-solution and microstructure. LSTN fired at 1250 °C in H2 showed the best anodic performance for scandia-stabilized zirconia (ScSZ) electrolyte-supported cells in H2 and CH4 fuels due to the favorable microstructure and Ni ex-solution.

Design, irradiation, and post-irradiation examination of the UC and (U,Pu)C fuel rods of the test groups Mol-11/K1 and Mol-11/K2

International Nuclear Information System (INIS)

Freund, D.; Elbel, H.; Steiner, H.

1976-06-01

The test groups K1 and K2 of the irradiation experiment Mol-11 are reported. Design, irradiation, and post-irradiation examination of the fuel rods irradiated are described. Mol-11/K1 consisted of one fuel rod with UC of 94% T.D. and helium bonding. This test group was intended to prove the high power irradiation capsule in pile. Mol-11/K2 consists of three fuel rods in total. One of these is presently still in the reactor. In this test group mixed carbide fuel of 83% T.D. and 15% Pu content under helium bonding is irradiated. The fuel rod K2-2 was provided with a capillary tube for the continuous measurement of fission gas pressure built up. 1.4988 stainless steel was chosen as cladding material. The final burnup lies between 35 and 70 MWd/kg M. Post-irradiation examination of the two test groups covers a theoretical analysis of the irradiation behaviour. (orig./GSCH) [de

Irradiation capsule design capable of continuously monitoring the creepdown of Zircaloy fuel cladding

International Nuclear Information System (INIS)

Thoms, K.R.; Dodd, C.V.; van der Kaa, T.; Hobson, D.O.

1978-01-01

An irradiation capsule which permits continuous monitoring of the creepdown of Zircaloy tubing has been designed and given preliminary tests. This design effort is the major element of a cooperative research program between the United States Nuclear Regulatory Commission and the Netherlands Energy Research Foundation (ECN) and is a part of the NRC-sponsored Zircaloy creepdown program. The purpose of the Zircaloy creepdown program is to provide data on the deformation characteristics of Zircaloy tubes, typical of LWR fuel element cladding, under combined axial and tangential compressive stresses. These data will be used to verify and improve the material behavior codes that are used for the description of fuel pin behavior. The first capsule of this series contains a mockup test specimen which was machined with three different diameters, nominally 10.92-mm, 10.54-mm and 11.30-mm (.430-in., .415-in., and .445-in.). This test specimen can be moved axially thereby varying the lift-off and serving as a calibration device for the eddy-current deformation monitoring probes. Fabrication of this capsule has been completed and during out-or-reactor checkout we were able to obtain a resolution of better than 0.01-mm (0.0004-in.). The capsule is scheduled for installation in the HFR on February 8, 1978, for a 26 day irradiation test. The first pressurized capsule, and therefore the first one to monitor in-reactor cladding deformation, will be installed in the HFR on May 3, 1978

Utilization of the capsule out-pile test facilities(2000-2003)

Energy Technology Data Exchange (ETDEWEB)

Cho, M. S.; Oh, J. M.; Cho, Y. G. and others

2003-06-01

Two out-pile test facilities were installed and being utilized for the non-irradiation tests outside the HANARO. The names of the facilities are the irradiation equipment design verification test facilities and the one-channel flow test device. In these facilities, the performance test of all capsules manufactured before loading in the HANARO and the design verification test for newly developed capsules were performed. The tests in these facilities include loading/unloading, pressure drop, endurance and vibration test etc. of capsules. In the period 2000{approx}2003, the performance tests for 8 material capsules of 99M-01K{approx}02M-05U were carried out, and the design verification tests of creep and fuel capsules developed newly were performed. For development of the creep capsule, pressure drop measurement, operation test of heater, T/C, LVDT and stress loading test were performed. In the design stage of the fuel capsule, the endurance and vibration test besides the above mentioned tests were carried out for verification of the safe operation during irradiation test in the HANARO. And in-chimeny bracket and the capsule supporting system were fixed and the flow tubes and the handling tools were manufactured for use at the facilities.

Investigation on shortening fabrication process of instrumented irradiation capsule of JMTR

International Nuclear Information System (INIS)

Nagata, Hiroshi; Inoue, Shuichi; Yamaura, Takayuki; Tsuchiya, Kunihiko; Nagao, Yoshiharu

2013-06-01

Refurbishment of The Japan Materials Testing Reactor (JMTR) was completed in FY2010. For damage caused by the 2011 off the Pacific coast of Tohoku Earthquake, the repair of facilities was completed in October 2012. Currently, the JMTR is in preparation for restart. Irradiation tests for LWRs safety research, science and technologies and production of RI for medical diagnosis medicine, etc. are expected after the JMTR restart. On the other hand, aiming at the attractive irradiation testing reactor, the usability improvement has been discussed. As a part of the usability improvement, shortening of turnaround time to get irradiation results from an application for irradiation use was discussed focusing on the fabrication process of irradiation capsules, where the fabrication process was analyzed and reviewed by referring a trial fabrication of the mockup capsule. As a result, it was found that the turnaround time can be shortened 2 months from fabrication period of 6 months with communize of irradiation capsule parts, application of ready-made instrumentation including the sheath heater, reconsideration of inspection process, etc. (author)

Improvement and utilization of irradiation capsule technology in HANARO

International Nuclear Information System (INIS)

Choo, Kee-Nam; Cho, Man-Soon; Kim, Bong-Goo; Lee, Cheol-Yong; Yang, Sung-Woo; Shin, Yoon-Taek; Park, Seng-Jae; Jung, Hoan-Sung

2012-01-01

Several improvements of irradiation capsule technology regarding irradiation test parameters, such as temperature and neutron flux/fluence, and regarding instrumentation have progressed at HANARO since the last KAERI-JAERI joint seminar held in 2008. The standard HANARO capsule technology that was developed for use in a commercial power plant temperature of about 300degC was improved to apply to a temperature range of 100-1000degC for the irradiation test of materials of new research reactors and future nuclear systems. Low-flux and long-term irradiation technologies have been developed at HANARO. As a beginning step of the localization of capsule instrumentation technology, the irradiation performance of a domestically produced thermocouple and LVDT will be examined at HANARO. The accuracy of an evaluation of neutron fluence and precise welding technology are also being examined at HANARO. Based on these accumulated capsule technologies, a HANARO irradiation capsule system is being actively utilized for the national R and D programme on commercial nuclear reactors and nuclear fuel cycle technology in Korea. HANARO has recently started the irradiation support of R and D relevant to future nuclear systems including SMART, VHTR, and SFR, and HANARO is preparing new support relevant to new research and Fusion reactors. (author)

Status of the material capsule irradiation and the development of the new capsule technology in HANARO

International Nuclear Information System (INIS)

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

2006-01-01

A material capsule system including a main capsule, fixing, control, cutting, and transport systems was developed for an irradiation test of non-fissile materials in HANARO. 14 irradiation capsules (12 instrumented and 2 non-instrumented capsules) have been designed, fabricated and successfully irradiated in the HANARO CT and IR test holes since 1995. The capsules were mainly designed for an irradiation of the RPV (Reactor Pressure Vessel), reactor core materials, and Zr-based alloys. Most capsules were made for KAERI material research projects, but 5 capsules were made as a part of national projects for the promotion of the HANARO utilization for universities. Based on the accumulated irradiation experience and the user's sophisticated requirements, development of new instrumented capsule technologies for a more precise control of the irradiation temperature and fluence of a specimen irrespective of the reactor operation has been performed in HANARO. (author)

A study of fuel failure behavior in high burnup HTGR fuel. Analysis by STRESS3 and STAPLE codes

International Nuclear Information System (INIS)

Martin, David G.; Sawa, Kazuhiro; Ueta, Shouhei; Sumita, Junya

2001-05-01

In current high temperature gas-cooled reactors (HTGRs), Tri-isotropic coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. This report attempts to model fuel behavior in irradiation tests using the U.K. codes STRESS3 and STAPLE. Test results in 91F-1A and HRB-22 capsules irradiation tests, which were carried out at the Japan Materials Testing Reactor of JAERI and at the High Flux Isotope Reactor of Oak Ridge National Laboratory, respectively, were employed in the calculation. The maximum burnup and fast neutron fluence were about 10%FIMA and 3 x 10 25 m -2 , respectively. The fuel for the irradiation tests was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor. The calculation results demonstrated that if only mean fracture stress values of PyC and SiC are used in the calculation it is not possible to predict any particle failures, by which is meant when all three load bearing layers have failed. By contrast, when statistical variations in the fracture stresses and particle specifications are taken into account, as is done in the STAPLE code, failures can be predicted. In the HRB-22 irradiation test, it was concluded that the first two particles which had failed were defective in some way, but that the third and fourth failures can be accounted for by the pressure vessel model. In the 91F-1A irradiation test, the result showed that 1 or 2 particles had failed towards the end of irradiation in the upper capsule and no particles failed in the lower capsule. (author)

PINEX-2: pinhole-TV imaging of fuel ejection from an internally vented capsule

International Nuclear Information System (INIS)

Berzins, G.J.; Lumpkin, A.H.

1979-01-01

The LASL pinhole-intensified TV system was used at the TREAT reactor to image an internally vented, fuel-ejection capsule designed and built by HEDL. Several improvements in the imaging system over PINEX-1 were incorporated. A sequence of 16-ms TV frames shows axial expansion, expulsion of fuel from the pin, and retention of clad integrity during the time of coverage

Refabricated and instrumented fuel rods

International Nuclear Information System (INIS)

Silberstein, K.

2005-01-01

Nuclear Fuel for power reactors capabilities evaluation is strongly based on the intimate knowledge of its behaviour under irradiation. This knowledge can be acquired from refabricated and instrumented fuel rods irradiated at different levels in commercial reactors. This paper presents the development and qualification of a new technique called RECTO related to a double-instrumented rod re-fabrication process developed by CEA/LECA hot laboratory facility at CADARACHE. The technique development includes manufacturing of the properly dimensioned cavity in the fuel pellet stack to house the thermocouple and the use of a newly designed pressure transducer. An analytic irradiation of such a double-instrumented fuel rod will be performed in OSIRIS test reactor starting October 2004. (Author)

Irradiation and performance evaluation of DUPIC fuel

International Nuclear Information System (INIS)

Bae, Ki Kwang; Yang, M. S.; Song, K. C.

2000-05-01

The objectives of the project is to establish the performance evaluation system for the experimental verification of DUPIC fuel. The scope and content for successful accomplishment of the phase 1 objectives is established as follows : irradiation test of DUPIC fuel at HANARO using a noninstrument capsule, study on the characteristics of DUPIC pellets, development of the analysis technology on the thermal behaviour of DUPIC fuel, basic design of a instrument capsule. The R and D results of the phase 1 are summarized as follows : - Performance analysis technology development of DUPIC fuel by model development for DUPIC fuel, review on the extendability of code(FEMAXI-IV, FRAPCON-3, ELESTRESS). - Study on physical properties of DUPIC fuel by design and fabrication of the equipment for measuring the thermal property. - HANARO irradiation test of simulated DUPIC fuel by the noninstrument capsule development. - PIE and result analysis

Irradiation and performance evaluation of DUPIC fuel

Energy Technology Data Exchange (ETDEWEB)

Bae, Ki Kwang; Yang, M S; Song, K C [and others

2000-05-01

The objectives of the project is to establish the performance evaluation system for the experimental verification of DUPIC fuel. The scope and content for successful accomplishment of the phase 1 objectives is established as follows : irradiation test of DUPIC fuel at HANARO using a noninstrument capsule, study on the characteristics of DUPIC pellets, development of the analysis technology on the thermal behaviour of DUPIC fuel, basic design of a instrument capsule. The R and D results of the phase 1 are summarized as follows : - Performance analysis technology development of DUPIC fuel by model development for DUPIC fuel, review on the extendability of code(FEMAXI-IV, FRAPCON-3, ELESTRESS). - Study on physical properties of DUPIC fuel by design and fabrication of the equipment for measuring the thermal property. - HANARO irradiation test of simulated DUPIC fuel by the noninstrument capsule development. - PIE and result analysis.

40 CFR 600.006-08 - Data and information requirements for fuel economy vehicles.

Science.gov (United States)

2010-07-01

... controller, battery configuration, or other components performed within 2,000 miles prior to fuel economy... fuel economy vehicles. 600.006-08 Section 600.006-08 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel...

Power measurement in the boiling capsules in R2 using delayed neutron detector

International Nuclear Information System (INIS)

Roennberg, G.

1979-03-01

LWR fuel testing is performed in the R2 reactor by irradiation in both loops and so-called boiling capsules. The loops have forced cooling, and the power can be measured calorimetrically by conventional instrumentation. The boiling capsules have convection cooling, and it has therefore been necessary to develop a special technique for power measurement, the delayed neutron detector (DND). The DND is a pneumatic rabbit system, which activates small uranium samples in the boiling capsules and counts the delayed neutrons for determination of the fission rate. This report describes the equipment used, the procedure of measurement, and the method of evaluation. (atuhor)

Nuclear instrumentation in VENUS-F

Science.gov (United States)

Wagemans, J.; Borms, L.; Kochetkov, A.; Krása, A.; Van Grieken, C.; Vittiglio, G.

2018-01-01

VENUS-F is a fast zero power reactor with 30 wt% U fuel and Pb/Bi as a coolant simulator. Depending on the experimental configuration, various neutron spectra (fast, epithermal, and thermal islands) are present. This paper gives a review of the nuclear instrumentation that is applied for reactor control and in a large variety of physics experiments. Activation foils and fission chambers are used to measure spatial neutron flux profiles, spectrum indices, reactivity effects (with positive period and compensation method or the MSM method) and kinetic parameters (with the Rossi-alpha method). Fission chamber calibrations are performed in the standard irradiation fields of the BR1 reactor (prompt fission neutron spectrum and Maxwellian thermal neutron spectrum).

The anisotropic creep behaviour of zircaloy-4 fuel cladding at 1073 K

International Nuclear Information System (INIS)

Rosinger, H.E.; Bowden, J.; Shewfelt, R.S.W.

1982-04-01

The anisotropy coefficients (F, G and H) of Hill's equation, suitably modified for creep deformation, have been determined for Zircaloy-4 fuel cladding from steady-state creep tests at an elevated temperature. Creep specimens were subjected to both uniaxial and biaxial loads (via internal pressure) at 1073 K and the strain measured concurrently in the axial and tangential directions. It has been found that Zircaloy-4 fuel cladding is almost, but not completely, isotropic at 1073 K; the values of F, G and H are 0.57, 0.48 and 0.45 respectively

Transmutation Fuels Campaign FY-09 Accomplishments Report

Energy Technology Data Exchange (ETDEWEB)

Lori Braase

2009-09-01

This report summarizes the fiscal year 2009 (FY-08) accomplishments for the Transmutation Fuels Campaign (TFC). The emphasis is on the accomplishments and relevance of the work. Detailed description of the methods used to achieve the highlighted results and the associated support tasks are not included in this report.

(La, Pr)0.8Sr0.2FeO3-δ-Sm 0.2Ce0.8O2-δ composite cathode for proton-conducting solid oxide fuel cells

KAUST Repository

Chen, Yonghong

2014-08-01

Mixed rare-earth (La, Pr)0.8Sr0.2FeO 3-δ-Sm0.2Ce0.8O2-δ (LPSF-SDC) composite cathode was investigated for proton-conducting solid oxide fuel cells based on protonic BaZr0.1Ce0.7Y 0.2O3-δ (BZCY) electrolyte. The powders of La 0.8-xPrxSr0.2FeO3-δ (x = 0, 0.2, 0.4, 0.6), Sm0.2Ce0.8O2-δ (SDC) and BaZr0.1Ce0.7Y0.2O3-δ (BZCY) were synthesized by a citric acid-nitrates self-propagating combustion method. The XRD results indicate that La0.8-xPrxSr 0.2FeO3-δ samples calcined at 950 °C exhibit perovskite structure and there are no interactions between LPSF0.2 and SDC at 1100 °C. The average thermal expansion coefficient (TEC) of LPSF0.2-SDC, BZCY and NiO-BZCY is 12.50 × 10-6 K-1, 13.51 × 10-6 K-1 and 13.47 × 10-6 K -1, respectively, which can provide good thermal compatibility between electrodes and electrolyte. An anode-supported single cell of NiO-BZCY|BZCY|LPSF0.2-SDC was successfully fabricated and operated from 700 °C to 550 °C with humidified hydrogen (∼3% H2O) as fuel and the static air as oxidant. A high maximum power density of 488 mW cm -2, an open-circuit potential of 0.95 V, and a low electrode polarization resistance of 0.071 Ω cm2 were achieved at 700 °C. Preliminary results demonstrate that LPSF0.2-SDC composite is a promising cathode material for proton-conducting solid oxide fuel cells. © 2014, Hydrogen Energy Publications, LLC. Published by Elsevier Ltd. All rights reserved.

Vented fuel experiment for gas-cooled fast reactor application

International Nuclear Information System (INIS)

Longest, A.W.; Gat, U.; Conlin, J.A.; Campana, R.J.

1976-01-01

A pressure-equalized and vented fuel rod is being irradiated in an instrumented capsule designated GB-10 to approximately 100MWd/kg-heavy metal. The fuel is a sol-gel-derived 88 at.% uranium (approximately 9% 235 U) and 12 at.% plutonium oxide, and the cladding is 20% cold-worked 316 stainless steel. The capsule is being irradiated in the Oak Ridge Research Reactor (ORR) and has exceeded a burnup of 70MWd/kg. The fuel has been operated at linear power rates of 39 and 44kW/m, and peak outer cladding temperature of 565 and 630 0 C respectively. A similar fuel rod in a previous capsule (GB-9) was subjected to 48kW/m (685 0 C). Helium gas sweeps through any portion of the three regions of the fuel rod, namely: fuel, blanket, and charcoal trap. The charcoal trap is operated at about 300 0 C. An on-line Ge(Li) detector is used to analyse release rates of several gamma-emitting noble gas isotopes. Analyses are performed primarily on sweep gas flowing through the entire fuel rod, and for sweeps over the top of the charcoal trap. Sweep gas samples are analyzed for stable noble gas isotopes. Results in the form of ratios of release rate over birth rate (R/B) and venting rate over birth rate (V/B) are derived. R/B rates range from 10 -4 % to 30% while V/B ranges from 10 -6 % to 30%. Flow conductance in the capsule was monitored by recording the flow rate and pressure drop across the fuel rod and inlet sweep line. The flow conductance has been falling with increasing burnup, currently restricting the flow to about 20ml (s.t.p.)/min at a pressure difference of about 1.5MPa. Venting rates of the gaseous fission products as a function of gas pressure in the range 6.9 to 1.4MPa have also been measured. Planned future experiments include the monitoring of tritium release, venting and cladding permeation rates, and its molecular form. First measurements have been made. A simulated leak experiment will determine the mixture of fission gases as a function of flow rate and the most

Neutron Characterization of Encapsulated ATF-1/LANL-1 Mockup Fuel Capsules

Energy Technology Data Exchange (ETDEWEB)

Vogel, Sven C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Borges, Nicholas Paul [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Losko, Adrian Simon [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mosby, Shea Morgan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Voit, Stewart Lancaster [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); White, Joshua Taylor [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Byler, Darrin David [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dunwoody, John Tyler [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nelson, Andrew Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-09-28

Twenty pellets of mock-up accident tolerant fuels UN-U3Si5 were produced at LANL and loaded in two rodlet/capsule assemblies. Tomographic imaging and diffraction measurements were performed to characterize these samples at the Flight-Path 5 and HIPPO beam lines at LANSCE/LANL between November 2016 and January 2017 as well as in August 2017. The entire ~10 cm long, ~1 cm diameter fuel volume could be characterized, however due to time constraints only 2 mm slices in 4mm increments were characterized with neutron diffraction and a 28mm subset of the entire sample was characterized with energy-resolved neutron imaging. The double encapsulation of the fuel into two steel containers does not pose a problem for the neutron analysis and the methods could be applied to enriched as well irradiated fuels.

LOFT instrumented fuel design and operating experience

International Nuclear Information System (INIS)

Russell, M.L.

1979-01-01

A summary description of the Loss-of-Fluid Test (LOFT) system instrumented core construction details and operating experience through reactor startup and loss-of-coolant experiment (LOCE) operations performed to date are discussed. The discussion includes details of the test instrumentation attachment to the fuel assembly, the structural response of the fuel modules to the forces generated by a double-ended break of a pressurized water reactor (PWR) coolant pipe at the inlet to the reactor vessel, the durability of the LOFT fuel and test instrumentation, and the plans for incorporation of improved fuel assembly test instrumentation features in the LOFT core

Spallation as a dominant source of pusher-fuel and hot-spot mix in inertial confinement fusion capsules

Science.gov (United States)

Orth, Charles D.

2016-02-01

We suggest that a potentially dominant but previously neglected source of pusher-fuel and hot-spot "mix" may have been the main degradation mechanism for fusion energy yields of modern inertial confinement fusion (ICF) capsules designed and fielded to achieve high yields—not hydrodynamic instabilities. This potentially dominant mix source is the spallation of small chunks or "grains" of pusher material into the fuel regions whenever (1) the solid material adjacent to the fuel changes its phase by nucleation and (2) this solid material spalls under shock loading and sudden decompression. We describe this mix mechanism, support it with simulations and experimental evidence, and explain how to eliminate it and thereby allow higher yields for ICF capsules and possibly ignition at the National Ignition Facility.

40 CFR 600.314-08 - Updating label values, annual fuel cost, Gas Guzzler Tax, and range of fuel economy for...

Science.gov (United States)

2010-07-01

... cost, Gas Guzzler Tax, and range of fuel economy for comparable automobiles. 600.314-08 Section 600.314-08 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy Regulations for 1977 and Later...

Data package for the Turkey Point material interaction test capsules

International Nuclear Information System (INIS)

Krogness, J.C.; Davis, R.B.

1979-01-01

Objective of the Materials Interaction Test (MIT) is to obtain interaction information on candidate package storage materials and geologies under prototypic temperatures in gamma and low level neutron fields. Compatibility, structural properties, and chemical transformations will be studied. The multiple test samples are contained within test capsules connected end-to-end to form a test train. Only passive instrumentation has been used to monitor temperatures and record neutron fluence. The test train contains seven capsules: three to test compatibility, two for structural tests, and two for chemical transformation studies. The materials tested are potential candidates for the spent fuel package canister and repository geologies

The K1 capsule is the critical determinant in the development of Escherichia coli meningitis in the rat.

OpenAIRE

Kim, K S; Itabashi, H; Gemski, P; Sadoff, J; Warren, R L; Cross, A S

1992-01-01

Although Escherichia coli strains possessing the K1 capsule are predominant among isolates from neonatal E. coli meningitis and most of these K1 isolates are associated with a limited number of 0 lipopolysaccharide (LPS) types, the basis of this association of K1 and certain 0 antigens with neonatal E. coli meningitis is not clear. The present study examined in experimental E. coli bacteremia and meningitis in newborn and adult rats whether or not the K1 capsule and/or O-LPS antigen are criti...

Capsule safety analysis of PRTF irradiation facility

International Nuclear Information System (INIS)

Suwarto

2013-01-01

Power Ramp Test Facility (PRTF) is an irradiation facility used for fuel testing of power reactor. PRTF has a capsule which is a test fuel rod container. During operation, pressurized water of 160 bars flows through in the capsule. Due to the high pressure it should be analyzed the impact of the capsule on reactor core safety. This analysis has purpose to calculate the ability of capsule pressure capacity. The analysis was carried out by calculating pressure capacity. From the calculating results it can be concluded that the capsule with pressure capacity of 438 bars will be safe to prevent the operation pressure of PRTF. (author)

Fabrication of internally instrumented reactor fuel rods

International Nuclear Information System (INIS)

Schmutz, J.D.; Meservey, R.H.

1975-01-01

Procedures are outlined for fabricating internally instrumented reactor fuel rods while maintaining the original quality assurance level of the rods. Instrumented fuel rods described contain fuel centerline thermocouples, ultrasonic thermometers, and pressure tubes for internal rod gas pressure measurements. Descriptions of the thermocouples and ultrasonic thermometers are also contained

The system K2NbF7-K2TiF6-KCl

International Nuclear Information System (INIS)

Kamenskaya, L.A.; Matveev, A.M.

1984-01-01

Using visual-polythermal and thermographical methods the ternary system K 2 NbF 7 -K 2 TiE 6 -KCl has been studied. Crystallization fields of initial components and the field of solid solutions of double compounds K 3 NbClF 7 and K 3 TiClF 6 are outlined. Ternary eutectics at 654 deg C, having the composition K 2 NbF 6 -41, K 2 TiP 6 -41, KCl-18 mol.%, is determined. Potassium fluoroniobate and fluorotitanate form continuous solid solutions unstable in the presence of the third component, potassium chloride

Surveillance instrumentation for spent-fuel safeguards

International Nuclear Information System (INIS)

McKenzie, J.M.; Holmes, J.P.; Gillman, L.K.; Schmitz, J.A.; McDaniel, P.J.

1978-01-01

The movement, in a facility, of spent reactor fuel may be tracked using simple instrumentation together with a real time unfolding algorithm. Experimental measurements, from multiple radiation monitors and crane weight and position monitors, were obtained during spent fuel movements at the G.E. Morris Spent-Fuel Storage Facility. These data and a preliminary version of an unfolding algorithm were used to estimate the position of the centroid and the magnitude of the spent fuel radiation source. Spatial location was estimated to +-1.5 m and source magnitude to +-10% of their true values. Application of this surveillance instrumentation to spent-fuel safeguards is discussed

Sources of variance in BC mass measurements from a small marine engine: Influence of the instruments, fuels and loads

Science.gov (United States)

Jiang, Yu; Yang, Jiacheng; Gagné, Stéphanie; Chan, Tak W.; Thomson, Kevin; Fofie, Emmanuel; Cary, Robert A.; Rutherford, Dan; Comer, Bryan; Swanson, Jacob; Lin, Yue; Van Rooy, Paul; Asa-Awuku, Akua; Jung, Heejung; Barsanti, Kelley; Karavalakis, Georgios; Cocker, David; Durbin, Thomas D.; Miller, J. Wayne; Johnson, Kent C.

2018-06-01

Knowledge of black carbon (BC) emission factors from ships is important from human health and environmental perspectives. A study of instruments measuring BC and fuels typically used in marine operation was carried out on a small marine engine. Six analytical methods measured the BC emissions in the exhaust of the marine engine operated at two load points (25% and 75%) while burning one of three fuels: a distillate marine (DMA), a low sulfur, residual marine (RMB-30) and a high-sulfur residual marine (RMG-380). The average emission factors with all instruments increased from 0.08 to 1.88 gBC/kg fuel in going from 25 to 75% load. An analysis of variance (ANOVA) tested BC emissions against instrument, load, and combined fuel properties and showed that both engine load and fuels had a statistically significant impact on BC emission factors. While BC emissions were impacted by the fuels used, none of the fuel properties investigated (sulfur content, viscosity, carbon residue and CCAI) was a primary driver for BC emissions. Of the two residual fuels, RMB-30 with the lower sulfur content, lower viscosity and lower residual carbon, had the highest BC emission factors. BC emission factors determined with the different instruments showed a good correlation with the PAS values with correlation coefficients R2 >0.95. A key finding of this research is the relative BC measured values were mostly independent of load and fuel, except for some instruments in certain fuel and load combinations.

Instrumentation of fuel elements and fuel plates

International Nuclear Information System (INIS)

Durand, J.P.; Fanjas, Y.

1993-01-01

When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have led to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

Instrumentation of fuel elements and fuel plates

International Nuclear Information System (INIS)

Durand, J.P.; Fanjas, Y.

1994-01-01

When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have lead to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

Measurement of $CP$ asymmetry in $B^0_s \\rightarrow D^{\\mp}_s K^{\\pm}$ decays

CERN Document Server

Aaij, Roel; Adinolfi, Marco; Affolder, Anthony; Ajaltouni, Ziad; Akar, Simon; Albrecht, Johannes; Alessio, Federico; Alexander, Michael; Ali, Suvayu; Alkhazov, Georgy; Alvarez Cartelle, Paula; Alves Jr, Antonio; Amato, Sandra; Amerio, Silvia; Amhis, Yasmine; An, Liupan; Anderlini, Lucio; Anderson, Jonathan; Andreassen, Rolf; Andreotti, Mirco; Andrews, Jason; Appleby, Robert; Aquines Gutierrez, Osvaldo; Archilli, Flavio; Artamonov, Alexander; Artuso, Marina; Aslanides, Elie; Auriemma, Giulio; Baalouch, Marouen; Bachmann, Sebastian; Back, John; Badalov, Alexey; Baldini, Wander; Barlow, Roger; Barschel, Colin; Barsuk, Sergey; Barter, William; Batozskaya, Varvara; Battista, Vincenzo; Bay, Aurelio; Beaucourt, Leo; Beddow, John; Bedeschi, Franco; Bediaga, Ignacio; Belogurov, Sergey; Belous, Konstantin; Belyaev, Ivan; Ben-Haim, Eli; Bencivenni, Giovanni; Benson, Sean; Benton, Jack; Berezhnoy, Alexander; Bernet, Roland; Bettler, Marc-Olivier; van Beuzekom, Martinus; Bien, Alexander; Bifani, Simone; Bird, Thomas; Bizzeti, Andrea; Bjørnstad, Pål Marius; Blake, Thomas; Blanc, Frédéric; Blouw, Johan; Blusk, Steven; Bocci, Valerio; Bondar, Alexander; Bondar, Nikolay; Bonivento, Walter; Borghi, Silvia; Borgia, Alessandra; Borsato, Martino; Bowcock, Themistocles; Bowen, Espen Eie; Bozzi, Concezio; Brambach, Tobias; van den Brand, Johannes; Bressieux, Joël; Brett, David; Britsch, Markward; Britton, Thomas; Brodzicka, Jolanta; Brook, Nicholas; Brown, Henry; Bursche, Albert; Busetto, Giovanni; Buytaert, Jan; Cadeddu, Sandro; Calabrese, Roberto; Calvi, Marta; Calvo Gomez, Miriam; Campana, Pierluigi; Campora Perez, Daniel; Carbone, Angelo; Carboni, Giovanni; Cardinale, Roberta; Cardini, Alessandro; Carson, Laurence; Carvalho Akiba, Kazuyoshi; Casse, Gianluigi; Cassina, Lorenzo; Castillo Garcia, Lucia; Cattaneo, Marco; Cauet, Christophe; Cenci, Riccardo; Charles, Matthew; Charpentier, Philippe; Chefdeville, Maximilien; Chen, Shanzhen; Cheung, Shu-Faye; Chiapolini, Nicola; Chrzaszcz, Marcin; Ciba, Krzystof; Cid Vidal, Xabier; Ciezarek, Gregory; Clarke, Peter; Clemencic, Marco; Cliff, Harry; Closier, Joel; Coco, Victor; Cogan, Julien; Cogneras, Eric; Collins, Paula; Comerma-Montells, Albert; Contu, Andrea; Cook, Andrew; Coombes, Matthew; Coquereau, Samuel; Corti, Gloria; Corvo, Marco; Counts, Ian; Couturier, Benjamin; Cowan, Greig; Craik, Daniel Charles; Cruz Torres, Melissa Maria; Cunliffe, Samuel; Currie, Robert; D'Ambrosio, Carmelo; Dalseno, Jeremy; David, Pascal; David, Pieter; Davis, Adam; De Bruyn, Kristof; De Capua, Stefano; De Cian, Michel; De Miranda, Jussara; De Paula, Leandro; De Silva, Weeraddana; De Simone, Patrizia; Decamp, Daniel; Deckenhoff, Mirko; Del Buono, Luigi; Déléage, Nicolas; Derkach, Denis; Deschamps, Olivier; Dettori, Francesco; Di Canto, Angelo; Dijkstra, Hans; Donleavy, Stephanie; Dordei, Francesca; Dorigo, Mirco; Dosil Suárez, Alvaro; Dossett, David; Dovbnya, Anatoliy; Dreimanis, Karlis; Dujany, Giulio; Dupertuis, Frederic; Durante, Paolo; Dzhelyadin, Rustem; Dziurda, Agnieszka; Dzyuba, Alexey; Easo, Sajan; Egede, Ulrik; Egorychev, Victor; Eidelman, Semen; Eisenhardt, Stephan; Eitschberger, Ulrich; Ekelhof, Robert; Eklund, Lars; El Rifai, Ibrahim; Elsasser, Christian; Ely, Scott; Esen, Sevda; Evans, Hannah Mary; Evans, Timothy; Falabella, Antonio; Färber, Christian; Farinelli, Chiara; Farley, Nathanael; Farry, Stephen; Fay, Robert; Ferguson, Dianne; Fernandez Albor, Victor; Ferreira Rodrigues, Fernando; Ferro-Luzzi, Massimiliano; Filippov, Sergey; Fiore, Marco; Fiorini, Massimiliano; Firlej, Miroslaw; Fitzpatrick, Conor; Fiutowski, Tomasz; Fontana, Marianna; Fontanelli, Flavio; Forty, Roger; Francisco, Oscar; Frank, Markus; Frei, Christoph; Frosini, Maddalena; Fu, Jinlin; Furfaro, Emiliano; Gallas Torreira, Abraham; Galli, Domenico; Gallorini, Stefano; Gambetta, Silvia; Gandelman, Miriam; Gandini, Paolo; Gao, Yuanning; García Pardiñas, Julián; Garofoli, Justin; Garra Tico, Jordi; Garrido, Lluis; Gaspar, Clara; Gauld, Rhorry; Gavardi, Laura; Gavrilov, Gennadii; Gersabeck, Evelina; Gersabeck, Marco; Gershon, Timothy; Ghez, Philippe; Gianelle, Alessio; Giani', Sebastiana; Gibson, Valerie; Giubega, Lavinia-Helena; Gligorov, V.V.; Göbel, Carla; Golubkov, Dmitry; Golutvin, Andrey; Gomes, Alvaro; Gotti, Claudio; Grabalosa Gándara, Marc; Graciani Diaz, Ricardo; Granado Cardoso, Luis Alberto; Graugés, Eugeni; Graziani, Giacomo; Grecu, Alexandru; Greening, Edward; Gregson, Sam; Griffith, Peter; Grillo, Lucia; Grünberg, Oliver; Gui, Bin; Gushchin, Evgeny; Guz, Yury; Gys, Thierry; Hadjivasiliou, Christos; Haefeli, Guido; Haen, Christophe; Haines, Susan; Hall, Samuel; Hamilton, Brian; Hampson, Thomas; Han, Xiaoxue; Hansmann-Menzemer, Stephanie; Harnew, Neville; Harnew, Samuel; Harrison, Jonathan; He, Jibo; Head, Timothy; Heijne, Veerle; Hennessy, Karol; Henrard, Pierre; Henry, Louis; Hernando Morata, Jose Angel; van Herwijnen, Eric; Heß, Miriam; Hicheur, Adlène; Hill, Donal; Hoballah, Mostafa; Hombach, Christoph; Hulsbergen, Wouter; Hunt, Philip; Hussain, Nazim; Hutchcroft, David; Hynds, Daniel; Idzik, Marek; Ilten, Philip; Jacobsson, Richard; Jaeger, Andreas; Jalocha, Pawel; Jans, Eddy; Jaton, Pierre; Jawahery, Abolhassan; Jing, Fanfan; John, Malcolm; Johnson, Daniel; Jones, Christopher; Joram, Christian; Jost, Beat; Jurik, Nathan; Kaballo, Michael; Kandybei, Sergii; Kanso, Walaa; Karacson, Matthias; Karbach, Moritz; Karodia, Sarah; Kelsey, Matthew; Kenyon, Ian; Ketel, Tjeerd; Khanji, Basem; Khurewathanakul, Chitsanu; Klaver, Suzanne; Klimaszewski, Konrad; Kochebina, Olga; Kolpin, Michael; Komarov, Ilya; Koopman, Rose; Koppenburg, Patrick; Korolev, Mikhail; Kozlinskiy, Alexandr; Kravchuk, Leonid; Kreplin, Katharina; Kreps, Michal; Krocker, Georg; Krokovny, Pavel; Kruse, Florian; Kucewicz, Wojciech; Kucharczyk, Marcin; Kudryavtsev, Vasily; Kurek, Krzysztof; Kvaratskheliya, Tengiz; La Thi, Viet Nga; Lacarrere, Daniel; Lafferty, George; Lai, Adriano; Lambert, Dean; Lambert, Robert W; Lanfranchi, Gaia; Langenbruch, Christoph; Langhans, Benedikt; Latham, Thomas; Lazzeroni, Cristina; Le Gac, Renaud; van Leerdam, Jeroen; Lees, Jean-Pierre; Lefèvre, Regis; Leflat, Alexander; Lefrançois, Jacques; Leo, Sabato; Leroy, Olivier; Lesiak, Tadeusz; Leverington, Blake; Li, Yiming; Likhomanenko, Tatiana; Liles, Myfanwy; Lindner, Rolf; Linn, Christian; Lionetto, Federica; Liu, Bo; Lohn, Stefan; Longstaff, Iain; Lopes, Jose; Lopez-March, Neus; Lowdon, Peter; Lu, Haiting; Lucchesi, Donatella; Luo, Haofei; Lupato, Anna; Luppi, Eleonora; Lupton, Oliver; Machefert, Frederic; Machikhiliyan, Irina V; Maciuc, Florin; Maev, Oleg; Malde, Sneha; Malinin, Alexander; Manca, Giulia; Mancinelli, Giampiero; Maratas, Jan; Marchand, Jean François; Marconi, Umberto; Marin Benito, Carla; Marino, Pietro; Märki, Raphael; Marks, Jörg; Martellotti, Giuseppe; Martens, Aurelien; Martín Sánchez, Alexandra; Martinelli, Maurizio; Martinez Santos, Diego; Martinez Vidal, Fernando; Martins Tostes, Danielle; Massafferri, André; Matev, Rosen; Mathe, Zoltan; Matteuzzi, Clara; Mazurov, Alexander; McCann, Michael; McCarthy, James; McNab, Andrew; McNulty, Ronan; McSkelly, Ben; Meadows, Brian; Meier, Frank; Meissner, Marco; Merk, Marcel; Milanes, Diego Alejandro; Minard, Marie-Noelle; Moggi, Niccolò; Molina Rodriguez, Josue; Monteil, Stephane; Morandin, Mauro; Morawski, Piotr; Mordà, Alessandro; Morello, Michael Joseph; Moron, Jakub; Morris, Adam Benjamin; Mountain, Raymond; Muheim, Franz; Müller, Katharina; Mussini, Manuel; Muster, Bastien; Naik, Paras; Nakada, Tatsuya; Nandakumar, Raja; Nasteva, Irina; Needham, Matthew; Neri, Nicola; Neubert, Sebastian; Neufeld, Niko; Neuner, Max; Nguyen, Anh Duc; Nguyen, Thi-Dung; Nguyen-Mau, Chung; Nicol, Michelle; Niess, Valentin; Niet, Ramon; Nikitin, Nikolay; Nikodem, Thomas; Novoselov, Alexey; O'Hanlon, Daniel Patrick; Oblakowska-Mucha, Agnieszka; Obraztsov, Vladimir; Oggero, Serena; Ogilvy, Stephen; Okhrimenko, Oleksandr; Oldeman, Rudolf; Onderwater, Gerco; Orlandea, Marius; Otalora Goicochea, Juan Martin; Owen, Patrick; Oyanguren, Maria Arantza; Pal, Bilas Kanti; Palano, Antimo; Palombo, Fernando; Palutan, Matteo; Panman, Jacob; Papanestis, Antonios; Pappagallo, Marco; Pappalardo, Luciano; Parkes, Christopher; Parkinson, Christopher John; Passaleva, Giovanni; Patel, Girish; Patel, Mitesh; Patrignani, Claudia; Pazos Alvarez, Antonio; Pearce, Alex; Pellegrino, Antonio; Pepe Altarelli, Monica; Perazzini, Stefano; Perez Trigo, Eliseo; Perret, Pascal; Perrin-Terrin, Mathieu; Pescatore, Luca; Pesen, Erhan; Petridis, Konstantin; Petrolini, Alessandro; Picatoste Olloqui, Eduardo; Pietrzyk, Boleslaw; Pilař, Tomas; Pinci, Davide; Pistone, Alessandro; Playfer, Stephen; Plo Casasus, Maximo; Polci, Francesco; Poluektov, Anton; Polycarpo, Erica; Popov, Alexander; Popov, Dmitry; Popovici, Bogdan; Potterat, Cédric; Price, Eugenia; Prisciandaro, Jessica; Pritchard, Adrian; Prouve, Claire; Pugatch, Valery; Puig Navarro, Albert; Punzi, Giovanni; Qian, Wenbin; Rachwal, Bartolomiej; Rademacker, Jonas; Rakotomiaramanana, Barinjaka; Rama, Matteo; Rangel, Murilo; Raniuk, Iurii; Rauschmayr, Nathalie; Raven, Gerhard; Reichert, Stefanie; Reid, Matthew; dos Reis, Alberto; Ricciardi, Stefania; Richards, Sophie; Rihl, Mariana; Rinnert, Kurt; Rives Molina, Vincente; Roa Romero, Diego; Robbe, Patrick; Rodrigues, Ana Barbara; Rodrigues, Eduardo; Rodriguez Perez, Pablo; Roiser, Stefan; Romanovsky, Vladimir; Romero Vidal, Antonio; Rotondo, Marcello; Rouvinet, Julien; Ruf, Thomas; Ruffini, Fabrizio; Ruiz, Hugo; Ruiz Valls, Pablo; Saborido Silva, Juan Jose; Sagidova, Naylya; Sail, Paul; Saitta, Biagio; Salustino Guimaraes, Valdir; Sanchez Mayordomo, Carlos; Sanmartin Sedes, Brais; Santacesaria, Roberta; Santamarina Rios, Cibran; Santovetti, Emanuele; Sarti, Alessio; Satriano, Celestina; Satta, Alessia; Saunders, Daniel Martin; Savrie, Mauro; Savrina, Darya; Schiller, Manuel; Schindler, Heinrich; Schlupp, Maximilian; Schmelling, Michael; Schmidt, Burkhard; Schneider, Olivier; Schopper, Andreas; Schune, Marie Helene; Schwemmer, Rainer; Sciascia, Barbara; Sciubba, Adalberto; Seco, Marcos; Semennikov, Alexander; Sepp, Indrek; Serra, Nicola; Serrano, Justine; Sestini, Lorenzo; Seyfert, Paul; Shapkin, Mikhail; Shapoval, Illya; Shcheglov, Yury; Shears, Tara; Shekhtman, Lev; Shevchenko, Vladimir; Shires, Alexander; Silva Coutinho, Rafael; Simi, Gabriele; Sirendi, Marek; Skidmore, Nicola; Skwarnicki, Tomasz; Smith, Anthony; Smith, Edmund; Smith, Eluned; Smith, Jackson; Smith, Mark; Snoek, Hella; Sokoloff, Michael; Soler, Paul; Soomro, Fatima; Souza, Daniel; Souza De Paula, Bruno; Spaan, Bernhard; Sparkes, Ailsa; Spradlin, Patrick; Sridharan, Srikanth; Stagni, Federico; Stahl, Marian; Stahl, Sascha; Steinkamp, Olaf; Stenyakin, Oleg; Stevenson, Scott; Stoica, Sabin; Stone, Sheldon; Storaci, Barbara; Stracka, Simone; Straticiuc, Mihai; Straumann, Ulrich; Stroili, Roberto; Subbiah, Vijay Kartik; Sun, Liang; Sutcliffe, William; Swientek, Krzysztof; Swientek, Stefan; Syropoulos, Vasileios; Szczekowski, Marek; Szczypka, Paul; Szilard, Daniela; Szumlak, Tomasz; T'Jampens, Stephane; Teklishyn, Maksym; Tellarini, Giulia; Teubert, Frederic; Thomas, Christopher; Thomas, Eric; van Tilburg, Jeroen; Tisserand, Vincent; Tobin, Mark; Tolk, Siim; Tomassetti, Luca; Tonelli, Diego; Topp-Joergensen, Stig; Torr, Nicholas; Tournefier, Edwige; Tourneur, Stephane; Tran, Minh Tâm; Tresch, Marco; Tsaregorodtsev, Andrei; Tsopelas, Panagiotis; Tuning, Niels; Ubeda Garcia, Mario; Ukleja, Artur; Ustyuzhanin, Andrey; Uwer, Ulrich; Vagnoni, Vincenzo; Valenti, Giovanni; Vallier, Alexis; Vazquez Gomez, Ricardo; Vazquez Regueiro, Pablo; Vázquez Sierra, Carlos; Vecchi, Stefania; Velthuis, Jaap; Veltri, Michele; Veneziano, Giovanni; Vesterinen, Mika; Viaud, Benoit; Vieira, Daniel; Vieites Diaz, Maria; Vilasis-Cardona, Xavier; Vollhardt, Achim; Volyanskyy, Dmytro; Voong, David; Vorobyev, Alexey; Vorobyev, Vitaly; Voß, Christian; Voss, Helge; de Vries, Jacco; Waldi, Roland; Wallace, Charlotte; Wallace, Ronan; Walsh, John; Wandernoth, Sebastian; Wang, Jianchun; Ward, David; Watson, Nigel; Websdale, David; Whitehead, Mark; Wicht, Jean; Wiedner, Dirk; Wilkinson, Guy; Williams, Matthew; Williams, Mike; Wilson, Fergus; Wimberley, Jack; Wishahi, Julian; Wislicki, Wojciech; Witek, Mariusz; Wormser, Guy; Wotton, Stephen; Wright, Simon; Wu, Suzhi; Wyllie, Kenneth; Xie, Yuehong; Xing, Zhou; Xu, Zhirui; Yang, Zhenwei; Yuan, Xuhao; Yushchenko, Oleg; Zangoli, Maria; Zavertyaev, Mikhail; Zhang, Liming; Zhang, Wen Chao; Zhang, Yanxi; Zhelezov, Alexey; Zhokhov, Anatoly; Zhong, Liang; Zvyagin, Alexander

2014-11-13

We report on measurements of the time-dependent CP violating observables in $B^0_s\\rightarrow D^{\\mp}_s K^{\\pm}$ decays using a dataset corresponding to 1.0 fb$^{-1}$ of pp collisions recorded with the LHCb detector. We find the CP violating observables $C_f=0.53\\pm0.25\\pm0.04$, $A^{\\Delta\\Gamma}_f=0.37\\pm0.42\\pm0.20$, $A^{\\Delta\\Gamma}_{\\bar{f}}=0.20\\pm0.41\\pm0.20$, $S_f=-1.09\\pm0.33\\pm0.08$, $S_{\\bar{f}}=-0.36\\pm0.34\\pm0.08$, where the uncertainties are statistical and systematic, respectively. We use these observables to make the first measurement of the CKM angle $\\gamma$ in $B^0_s\\rightarrow D^{\\mp}_s K^{\\pm}$ decays, finding $\\gamma$ = (115$_{-43}^{+28}$)$^\\circ$ modulo 180$^\\circ$ at 68% CL, where the error contains both statistical and systematic uncertainties.

Spent fuel dry storage technology development: thermal evaluation of three adjacent drywells (each containing a 0.6 kW PWR spent fuel assembly)

International Nuclear Information System (INIS)

Unterzuber, R.; Hanson, J.P.

1981-09-01

A spent fuel Adjacent Drywell Test was conducted at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site utilizing three nearly identical pressurized water reactor spent fuel assemblies each having a decay heat level of approximately 0.6 kW. Each fuel assembly was encapsulated inside the E-MAD Hot Bay and placed in an instrumented near-surface drywell storage cell for thermal testing. Each fuel assembly was sealed inside a 14-in. diam, 168-in.-long stainless steel canister and attached to a concrete-filled, 20-in.-diam, 34-in.-long, shield plug. The canister assembly was then placed in a carbon steel drywell liner which had been grouted into a hole drilled in the soil adjacent to E-MAD. The three drywells were located 25 feet apart in a linear array. Thermocouples, provided to measure canister, liner and soil temperatures, were inserted into tubes on the outside of the canister and drywell liner and were attached to plastic pipes which were grouted into holes in the soil. Temperatures from the three drywells and the adjacent soil were recorded throughout the Adjacent Drywell Test. Drywell thermal data showed virtually no thermal interaction between adjacent drywells. However, peak temperatures reached by the three drywells did show a fairly significant difference. Peak canister and drywell liner temperatures were reached in August 1981 for all three drywells. The two previously unused drywells responded similarly with peak canister and liner temperatures reaching 199 0 F and 158 0 F, respectively. Comparable peak temperatures for the third drywell which had previously contained spent fuel for nearly 21 months prior to the Adjacent Drywell Test reached 210 0 F for the canister and 169 0 F for the drywell liner. This difference is attributed to a decrease in soil thermal conductivity caused by the dryout of soil around the drywell used for previous spent fuel testing

Equilibrium ignition for ICF capsules

International Nuclear Information System (INIS)

Lackner, K.S.; Colgate, S.A.; Johnson, N.L.; Kirkpatrick, R.C.; Menikoff, R.; Petschek, A.G.

1993-01-01

There are two fundamentally different approaches to igniting DT fuel in an ICF capsule which can be described as equilibrium and hot spot ignition. In both cases, a capsule which can be thought of as a pusher containing the DT fuel is imploded until the fuel reaches ignition conditions. In comparing high-gain ICF targets using cryogenic DT for a pusher with equilibrium ignition targets using high-Z pushers which contain the radiation. The authors point to the intrinsic advantages of the latter. Equilibrium or volume ignition sacrifices high gain for lower losses, lower ignition temperature, lower implosion velocity and lower sensitivity of the more robust capsule to small fluctuations and asymmetries in the drive system. The reduction in gain is about a factor of 2.5, which is small enough to make the more robust equilibrium ignition an attractive alternative

Fission gas release modelling: developments arising from instrumented fuel assemblies, out-of-pile experiments and microstructural observations

International Nuclear Information System (INIS)

Leech, N.A.; Smith, M.R.; Pearce, J.H.; Ellis, W.E.; Beatham, N.

1990-01-01

This paper reviews the development of fission gas release modelling in thermal reactor fuel (both steady-state and transient) and in particular, illustrates the way in which experimental data have been, and continue to be, the main driving force behind model development. To illustrate this point various aspects of fuel performance are considered: temperature calculation, steady-state and transient fission gas release, grain boundary gas atom capacity and microstructural phenomena. The sources of experimental data discussed include end-of-life fission gas release measurements, instrumented fuel assemblies (e.g. rods with internal pressure transducers, fuel centre thermocouples), swept capsule experiments, out-of-pile annealing experiments and microstructural techniques applied during post-irradiation evaluation. In the case of the latter, the benefit of applying many observation and analysis techniques on the same fuel samples (the approach adopted at NRL Windscale) is emphasized. This illustrates a shift of emphasis in the modelling field from the development of large, complex thermo-mechanical computer codes to the assessment of key experimental data in order to develop and evaluate sub-models which correctly predict the observed behaviour. (author)

Hot cell examination on the surveillance capsule and HANARO capsule in IMEF

International Nuclear Information System (INIS)

Choo, Yong Sun; Oh, Wan Ho; Yoo, Byung Ok; Jung, Yang Hong; Ahn, Sang Bok; Baik, Seung Je; Song, Wung Sup; Hong, Kwon Pyo

2000-01-01

For the maintenance of integrity and safety of pressurizer of commercial power plant until its life span, it is required by US NRC 10CFR50 APP. G and H and ASTM E185-94 to periodically monitor irradiation embrittlement by neutron irradiation. In order to accomplished the requirement reactor operator has been carrying out the test by extracting the monitoring capsule embeded in reactor during the period of planned preventive maintenance. In relation to this irradiation samples are being used for prediction of reactor vessel life span and reactor vessel's adjusted reference temperature by irradiation of neutron flux enough to reach to end of life span. And also irradiation capsules with and without instrumentation are used for R and D on nuclear materials. Each capsule contains high radioactivity, therefore, post irradiation examination has to be handled by all means in the hot cell. The facility available for this purpose is Irradiated material examination facility (IMEF) to handle such works as capsule receiving, capsule cut and dismantling, sample classification, various examination, and finally development and improvement of examination equipment and instrumentation. (Hong, J. S.)

Safeguards instrument to monitor spent reactor fuel

International Nuclear Information System (INIS)

Nicholson, N.; Dowdy, E.J.; Holt, D.M.; Stump, C.

1981-01-01

A hand-held instrument for monitoring irradiated nuclear fuel inventories located in water-filled storage ponds has been developed. This instrument provides sufficient precise qualitative and quantitative information to be useful as a confirmatory technique to International Atomic Energy Agency inspectors, and is believed to be of potential use to nuclear fuel managers and to operators of spent-fuel storage facilities, both at reactor and away-from-reactor, and to operators of nuclear fuel reprocessing plants. Because the Cerenkov radiation glow can barely be seen by the unaided eye under darkened conditions, a night vision device is incorporated to aid the operator in locating the fuel assembly to be measured. Beam splitting optics placed in front of the image intensifier and a preset aperture select a predetermined portion of the observed scene for measurement of the light intensity using a photomultiplier (PM) tube and digital readout. The PM tube gain is adjusted by use of an internal optical reference source, providing long term repeatability and instrument-to-instrument cnsistency. Interchangeable lenses accommodate various viewing and measuring conditions

X-ray photoelectron spectroscopy as detection tool for coordinated or uncoordinated fluorine atoms demonstrated on fluoride systems NaF, K2TaF7, K3TaF8, K2ZrF6, Na7Zr6F31 and K3ZrF7

Science.gov (United States)

Boča, Miroslav; Barborík, Peter; Mičušík, Matej; Omastová, Mária

2012-07-01

While systems K3TaF8 and K3ZrF7 were prepared by modified molten salt method modified wet pathway was used for reproducible preparation of Na7Zr6F31. Its congruently melting character was demonstrated on simultaneous TG/DSC measurements and XRD patterns. X-ray photoelectron spectroscopy was applied for identification of differently bonded fluorine atoms in series of compounds NaF, K2TaF7, K3TaF8, K2ZrF6, Na7Zr6F31 and K3ZrF7. Three different types of fluorine atoms were described qualitatively and quantitatively. Uncoordinated fluorine atoms (F-) provide signals at lowest binding energies, followed by signals from terminally coordinated fluorine atoms (M-F) and then bridging fluorine atoms (M-F-M) at highest energy. Based on XPS F 1s signals assigned to fluorine atoms in compounds with correctly determined structure it was suggested that fluorine atoms in K3ZrF7 have partially bridging character.

Radiation research of materials using irradiation capsules

International Nuclear Information System (INIS)

Chamrad, B.

1976-01-01

The methods are briefly characterized of radiation experiments on the WWR-S research reactor. The irradiation capsule installed in the reactor including the electronic instrumentation is described. Irradiated samples temperature is stabilized by an auxiliary heat source placed in the irradiation space. The electronic control equipment of the system is automated. In irradiation experiments, experimental and operating conditions are recorded by a digital measuring centre with electric typewriter and paper tape data recording and by an analog compensating recorder. The irradiation experiment control system controls irradiated sample temperature, the supply current size and the heating element temperature of the auxiliary stabilizing source, inert and technological pressures of the capsule atmosphere and the thermostat temperature of the thermocouple junctions. (O.K.)

Irradiation of Parts of the X-Gen Nuclear Fuel Assembly made by KNF in HANARO

International Nuclear Information System (INIS)

Choo, K. N.; Cho, M. S.; Shin, Y. T.; Kim, B. G.; Lee, S. H.; Eom, K. B.

2008-01-01

An instrumented capsule has been developed at HANARO (High flux Advanced Neutron Application ReactOr) for the neutron irradiation tests of materials. The capsule system has been actively utilized for the various material irradiation tests requested by users from research institutes, universities, and the industries. As a preliminary test, some specimens made of the parts of a nuclear fuel assembly were inserted in the 05M-07U instrumented capsule and successfully irradiated at HANARO. Based on the results and experience, a new irradiation capsule of 07M-13N was designed, fabricated, and irradiated at HANARO for the evaluation of the neutron irradiation properties of the parts of the X-Gen nuclear fuel assembly made by KNF (Korea Nuclear Fuel). Specimens such as bucking and spring test specimens of spacer grid, microstructure and tensile test specimens of welded parts, tensile, irradiation growth and spring test specimens made of HANA tube, Zirlo, Zircaloy-4 and Inconel-718 were placed in the capsule. The capsule was loaded into the CT test hole of HANARO of a 30MW thermal output and the specimens were irradiated at 295 - 460 .deg. C up to a fast neutron fluence of 1.2x10 21 (n/cm 2 ) (E>1.0MeV)

Determination of melting point of mixed-oxide fuel irradiated in a fast breeder reactor

International Nuclear Information System (INIS)

Tachibana, Toshimichi

1985-01-01

The melting point of fuel is important to set its in-reactor maximum temperature in fuel design. The fuel melting point measuring methods are broadly the filament method and the capsule sealing method. The only instance of measuring the melting point of irradiated mixed oxide (U, Pu)O 2 fuel by the filament method is by GE in the United States. The capsule sealing method, while the excellent means, is difficult in weld sealing the irradiated fuel in a capsule within the cell. In the fast reactor development program, the remotely operated melting point measuring apparatus in capsule sealing the mixed (U, Pu)O 2 fuel irradiated in the experimental FBR Joyo was set in the cell and the melting point was measured, for the first time in the world. (Mori, K.)

CCT-K2.1: NRC/VNIIFTRI bilateral comparison of capsule-type standard platinum resistance thermometers from 13.8 K to 273.16 K

Energy Technology Data Exchange (ETDEWEB)

Hill, K.D.; Steele, A.G. [National Research Council of Canada, Institute for National Measurement Standards, Ottawa, ON (Canada); Dedikov, Y.A.; Shkraba, V.T. [Institute for Physical-Technical and Radiotechnical Measurements (VNIIFTRI), Moscow (Russian Federation)

2005-04-01

The Consultative Committee for Thermometry Key Comparison 2 (CCT-K2) results were published two years ago (2002 Metrologia 39 551-71). NRC served as the pilot laboratory for CCT-K2 and remains able to provide a scale and measurement system suitable for performing bilateral comparisons linked to the original key comparison results. In March 2003, measurements of two VNIIFTRI 100 {omega} capsule-style platinum resistance thermometers (CSPRTs), S/N 356 and 476, were undertaken to relate their local calibration to the results from the CCT-K2 exercise. The NRC Leeds and Northrup (L and N) CSPRT S/N 1872174 provides the link to the CCT-K2 results. The three CSPRTs were compared at the eight defining cryogenic temperatures of the International Temperature Scale of 1990 (ITS-90) in the range from 13.8033 K to 273.16 K. The reader is referred to the full text of the CCT-K2 report for a detailed explanation of the methodology employed for the comparison. Only the details unique to the measurements reported here will be addressed in this article. The NRC/VNIIFTRI bilateral comparison of capsule-style platinum resistance thermometers over the range 13.8 K to 273.16 K has revealed calibrations at VNIIFTRI to be in agreement with the KCRV of CCT-K2 within the expanded uncertainty for all temperatures of the comparison with the exception of the triple point of hydrogen at 13.8033 K. One of the two CSPRTs supplied by VNIIFTRI was found to be discrepant as revealed by differences at the triple point of water and at the lowest temperatures of the comparison, and was therefore excluded from further analysis. The linkage to the CCT-K2 data supports the evaluation of the VNIIFTRI CMCs in Appendix C of the KCDB. (authors)

A tale of tails: Sialidase is key to success in a model of phage therapy against K1-capsulated Escherichia coli

International Nuclear Information System (INIS)

Bull, J.J.; Vimr, E.R.; Molineux, I.J.

2010-01-01

Prior studies treating mice infected with Escherichia coli O18:K1:H7 observed that phages requiring the K1 capsule for infection (K1-dep) were superior to capsule-independent (K1-ind) phages. We show that three K1-ind phages all have low fitness when grown on cells in serum whereas fitnesses of four K1-dep phages were high. The difference is serum-specific, as fitnesses in broth overlapped. Sialidase activity was associated with all K1-dep virions tested but no K1-ind virions, a phenotype supported by sequence analyses. Adding endosialidase to cells infected with K1-ind phage increased fitness in serum by enhancing productive infection after adsorption. We propose that virion sialidase activity is the primary determinant of high fitness on cells grown in serum, and thus in a mammalian host. Although the benefit of sialidase is specific to K1-capsulated bacteria, this study may provide a scientific rationale for selecting phages for therapeutic use in many systemic infections.

Kimberlite Wall Rock Fragmentation: Venetia K08 Pipe Development

Science.gov (United States)

Barnett, W.; Kurszlaukis, S.; Tait, M.; Dirks, P.

2009-05-01

Volcanic systems impose powerful disrupting forces on the country rock into which they intrude. The nature of the induced brittle deformation or fragmentation can be characteristic of the volcanic processes ongoing within the volcanic system, but are most typically partially removed or obscured by repeated, overprinting volcanic activity in mature pipes. Incompletely evolved pipes may therefore provide important evidence for the types and stages of wall rock fragmentation, and mechanical processes responsible for the fragmentation. Evidence for preserved stages of fragmentation is presented from a detailed study of the K08 pipe within the Cambrian Venetia kimberlite cluster, South Africa. This paper investigates the growth history of the K08 pipe and the mechanics of pipe development based on observations in the pit, drill core and thin sections, from geochemical analyses, particle size distribution analyses, and 3D modeling. Present open pit exposures of the K08 pipe comprise greater than 90% mega-breccia of country rock clasts (gneiss and schist) with Drill core shows that below about 225 m the CRB includes increasing quantities of kimberlite. The breccia clasts are angular, clast-supported with void or carbonate cement between the clasts. Average clast sizes define sub-horizontal layers tens of metres thick across the pipe. Structural and textural observations indicate the presence of zones of re-fragmentation or zones of brittle shearing. Breccia textural studies and fractal statistics on particle size distributions (PSD) is used to quantify sheared and non- sheared breccia zones. The calculated energy required to form the non-sheared breccia PSD implies an explosive early stage of fragmentation that pre-conditions the rock mass. The pre-conditioning would have been caused by explosions that are either phreatic or phreatomagmatic in nature. The explosions are likely to have been centered on a dyke, or pulses of preceding volatile-fluid phases, which have

Developing Spent Fuel Assembly for Advanced NDA Instrument Calibration - NGSI Spent Fuel Project

Energy Technology Data Exchange (ETDEWEB)

Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Banfield, James [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Skutnik, Steven [Univ. of Tennessee, Knoxville, TN (United States)

2014-02-01

This report summarizes the work by Oak Ridge National Laboratory to investigate the application of modeling and simulation to support the performance assessment and calibration of the advanced nondestructive assay (NDA) instruments developed under the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Advanced NDA instrument calibration will likely require reference spent fuel assemblies with well-characterized nuclide compositions that can serve as working standards. Because no reference spent fuel standard currently exists, and the practical ability to obtain direct measurement of nuclide compositions using destructive assay (DA) measurements of an entire fuel assembly is prohibitive in the near term due to the complexity and cost of spent fuel experiments, modeling and simulation will be required to construct such reference fuel assemblies. These calculations will be used to support instrument field tests at the Swedish Interim Storage Facility (Clab) for Spent Nuclear Fuel.

Stability of La0.6Sr0.4Co0.2Fe0.8O3/Ce0.9Gd0.1O2 cathodes during sintering and solid oxide fuel cell operation

DEFF Research Database (Denmark)

Kiebach, Ragnar; Zhang, Weiwei; Zhang, Wei

2015-01-01

Degradation phenomena of La0.58Sr0.4Co0.2Fe0.8O3/Ce0.9Gd0.1O2 (LSCF/CGO) cathodes were investigated via post-mortem analyses of an experimental solid oxide fuel cell (SOFC) stack tested at 700 °C for 2000 h using advanced electron microscopy (SEM-EDS, HR-TEM-EDS) and time-of-flight secondary ion...... mass spectrometry (TOF-SIMS). Similar studies were carried out on non-tested reference cells for comparison. The analysis focused on the LSCF/CGO cathode and the CGO barrier layer, as the cathode degradation can be a major contributor to the overall degradation in this type of SOFC. SEM-EDS and TOF......-SIMS were used to investigate inter-diffusion across the barrier layer - electrolyte interface and the barrier layer - cathode interface. In addition, TOF-SIMS data were employed to investigate impurity distribution before and after testing. HR-TEM-EDS was used to investigate possible phase segregation...

K vp estimate intercomparison between Unfors XI, Radcal 4075 and a new CDTN multipurpose instrument

International Nuclear Information System (INIS)

Baptista N, A. T.; Oliveira, B. B.; Faria, L. O.

2014-08-01

This work compares results obtained using 3 (three) different instruments capable of non-invasively estimating the voltage applied to the electrodes of an x-ray emission equipment, namely the Unfors model Xi R/F, the Radcal Corporation model 4075 R/F and a new CDTN multipurpose instrument. Tests were carried out using the Pantak Seifert Model 320 Hs x-ray machine with equal setups for all instruments undergoing comparison. Irradiations were performed for different conditions of voltage and filtration. Although all instruments show a similar tendency to increase the k Vp estimate when aluminum filters are placed in the path of the x-ray beam, they may all be satisfactorily adopted in quality control routines of x-ray equipment by means of estimation of the applied voltage. The importance of using equally calibrated measurement instruments and according to manufacturers instructions became clear; in case it is not possible to follow these requirements, measurement-correcting methods must be applied. Using the new multipurpose instrument, the k Vp estimate is satisfactory even if the x-ray beam intensity is filtered in approximately one-tenth value layer. (author)

K vp estimate intercomparison between Unfors XI, Radcal 4075 and a new CDTN multipurpose instrument

Energy Technology Data Exchange (ETDEWEB)

Baptista N, A. T.; Oliveira, B. B.; Faria, L. O., E-mail: annibal@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear - CNEN, Av. Presidente Antonio Carlos 6627, Campus UFMG, Pampulha, CEP 31270-901 Belo Horizonte, Minas Gerais (Brazil)

2014-08-15

This work compares results obtained using 3 (three) different instruments capable of non-invasively estimating the voltage applied to the electrodes of an x-ray emission equipment, namely the Unfors model Xi R/F, the Radcal Corporation model 4075 R/F and a new CDTN multipurpose instrument. Tests were carried out using the Pantak Seifert Model 320 Hs x-ray machine with equal setups for all instruments undergoing comparison. Irradiations were performed for different conditions of voltage and filtration. Although all instruments show a similar tendency to increase the k Vp estimate when aluminum filters are placed in the path of the x-ray beam, they may all be satisfactorily adopted in quality control routines of x-ray equipment by means of estimation of the applied voltage. The importance of using equally calibrated measurement instruments and according to manufacturers instructions became clear; in case it is not possible to follow these requirements, measurement-correcting methods must be applied. Using the new multipurpose instrument, the k Vp estimate is satisfactory even if the x-ray beam intensity is filtered in approximately one-tenth value layer. (author)

2017-12-22T09:08:04Z https://www.ajol.info/index.php/all/oai oai:ojs ...

African Journals Online (AJOL)

article/61364 2017-12-22T09:08:04Z bcse:ART Pollution status of Tinishu Akaki River and its tributaries (Ethiopia) evaluated using physico-chemical parameters, major ions, and nutrients Melaku, Samuel; Laboratory of Analytical Chemistry, Gent ...

Experiments with preirradiated fuel rods in the Nuclear Safety Research Reactor

International Nuclear Information System (INIS)

Horiki, O.; Kobayashi, S.; Takariko, I.; Ishijima, K.

1992-01-01

In the Nuclear Safety Research Reactor (NSRR) owned and operated by Japan Atomic Energy Research Institute (JAERI), extensive experimental studies on the fuel behavior under reactivity initiated accident (RIA) conditions have been continued since the start of the test program in 1975. Accumulated experimental data were used as the fundamental data base of the Japanese safety evaluation guideline for reactivity initiated events in light water cooled nuclear power plants established by the nuclear safety commission in 1984. All of the data used to establish the guideline were, however, limited to those derived from the tests with fresh fuel rods as test samples because of the lack of experimental facility to handle highly radioactive materials.The guideline, therefore, introduces the peak fuel enthalpy of 85 cal/g which was adopted from the SPERT-CDC data as a provisional failure threshold of preirradiated fuel rod and, says that this value should be revised based on the NSRR experiments in the future. According to the above requirement, new NSRR experimental program with the preirradiated fuel rods as test samples was started in 1989. Test fuel rods are prepared by refabrication of the long-sized fuel rods preirradiated in commercial PWRs and BWRs into short segments and by preirradiation of short-sized test fuel rods in the Japan Material Testing Reactor(JMTR). For the tests with preirradiated fuel rods as test samples, the special experimental capsules, the automatic instrumentation fitting device, the automatic capsule assembling device and the capsule loading device were newly developed. In addition, the existing hot cave was modified to mount the capsule assembling device and the other inspection tools and, a new small iron cell was established adjacent to the cave to store the instrumentation fitting device. (author)

Some views on mechanical safety of capsules for radioactive waste

International Nuclear Information System (INIS)

Nilsson, F.

1978-02-01

Three different concepts for encapsulation of used nuclear fuel are discussed with respect to their mechanical safety. In the first concept the burnt out fuel elements are encapsulated into a copper capsule. The material properties of copper are discussed especially with reference to toughness and creep. A simple fracture mechanical analysis shows that the risk for direct fracture is negligible at the actual stress levels. The loads on the capsule are studied and are found to be normally less than 40 MPa (residual stresses). Transient loads that might arise in the handling of the capsule might however be dangerous to its integrity. The next concept is encapsulation of the fuel elements into a sintered aluminium oxide capsule. A fracture probability analysis based on Weibull's statistical fracture theory gives fracture probabilities that are acceptable. Extended studies of this concept, especially of the risk for delayed fracture, is recommended. The last concept is a Pb-Ti capsule for glassed refined fuel. An analysis of the relaxation of internal stresses is performed. The critical point of these capsules appears to be the welds on the titanium shell where the risk for a direct fracture is not neglibigle

F2α-isoprostane, Na+-K+ ATPase and membrane fluidity of placental syncytiotrophoblast cell in preeclamptic women with vitamin E supplementation

Directory of Open Access Journals (Sweden)

Franciscus D. Suyatna

2012-11-01

Full Text Available Background: The aim of our study was to analyze F2α-isoprostane level, Na+-K+ ATPase activity and placental syncytiotrophoblast cell membrane fluidity in preeclamptic women who received vitamin E supplementation.Methods: The study was conducted between September 2003 and February 2005 at Budi Kemuliaan Maternity Hospital, Central Jakarta. Samples were 6 preeclamptic women with vitamin E supplementation, 6 preeclamptic women without vitamin E supplementation and 6 normal pregnant women. The dose of vitamin E was 200 mg daily. F2α-isoprostane was measured with ELISA reader at λ of 450 nm. Cell membrane fluidity was measured by comparing the molar ratio of total cholesterol and cell membrane phospholipid concentration. The cholesterol was measured by Modular C800 using Roche reagent. Phospholipid was measured by Shimadzu RF5301PC spectrofluorometer (excitation 267 nm, emission 307 nm. Na+-K+ ATPase activity was inhibited by ouabain. Pi production was measured with Fiske and Subbarow method using spectrophotometer at λ of 660 nm. Data was analyzed using F test with one-way ANOVA.Results: Vitamin E supplementation in preeclamptic women decreased the oxidative stress, indicated by significantly lower level of F2α-isoprostane compared to those without vitamin E (26.72 ± 11.21 vs 41.85 ± 7.09 ng/mL, respectively, p = 0.017. Membrane fluidity in syncytiotrophoblast cell of preeclampsia with vitamin E group was maintained at 0.39 ± 0.08 while in those without vitamin E was 0.53 ± 0.14 (p = 0.04. Na+-K+ ATPase activity in syncytiotrophoblast cell membrane was not affected by vitamin E (p = 0.915.Conclusion: Vitamin E supplementation in preeclamptic women decreases F2α-isoprostane level and maintains cell membrane fluidity of syncytiotrophoblast cells; however, it does not increase Na+-K+ ATPase enzyme activity. (Med J Indones. 2012;21:225-9Keywords: F2α-isoprostane, membrane fluidity, Na+-K+ ATPase, preeclampsia, vitamin E

Specific heat measurements of TiB2 and 6LiF from 0.5 to 30 K

International Nuclear Information System (INIS)

Lang, Brian E.; Donaldson, Marcus H.; Woodfield, Brian F.; Burger, Arnold; Roy, Utupal N.; Lamberti, Vincent; Bell, Zane W.

2005-01-01

The specific heats of TiB 2 and 6 LiF have been measured from 0.5 to 30 K as part of a larger project in the construction of a neutron spectrometer. For this application, the measured specific heats were used to extrapolate the specific heats down to 0.1 K with lattice, electronic, and Schottky equations for the respective samples. The resultant specific heat values at 0.1 K for TiB 2 and 6 LiF are 4.08 x 10 -4 ± 0.27 x 10 -4 J/K/mol and 9.19 x 10 -9 ± 0.15 x 10 -9 J/K/mol, respectively

2018-03-20T08:36:09Z https://www.ajol.info/index.php/all/oai oai:ojs ...

African Journals Online (AJOL)

article/75816 2018-03-20T08:36:09Z ijard:ART An Overview of Benefits of Organic Agriculture as a Climate Change Adaptation and Mitigation Strategy for Nigeria Korie, OC Eze, CC Lemchi, JI Ibekwe, UC Orebiyi, JS Obasi, PC Ohajianya, DO Eze ...

Aeroheating Measurement of Apollo Shaped Capsule with Boundary Layer Trip in the Free-piston Shock Tunnel HIEST

Science.gov (United States)

Hideyuki, TANNO; Tomoyuki, KOMURO; Kazuo, SATO; Katsuhiro, ITOH; Lillard, Randolph P.; Olejniczak, Joseph

2013-01-01

An aeroheating measurement test campaign of an Apollo capsule model with laminar and turbulent boundary layer was performed in the free-piston shock tunnel HIEST at JAXA Kakuda Space Center. A 250mm-diameter 6.4%-scaled Apollo CM capsule model made of SUS-304 stainless steel was applied in this study. To measure heat flux distribution, the model was equipped with 88 miniature co-axial Chromel-Constantan thermocouples on the heat shield surface of the model. In order to promote boundary layer transition, a boundary layer trip insert with 13 "pizza-box" isolated roughness elements, which have 1.27mm square, were placed at 17mm below of the model geometric center. Three boundary layer trip inserts with roughness height of k=0.3mm, 0.6mm and 0.8mm were used to identify the appropriate height to induce transition. Heat flux records with or without roughness elements were obtained for model angles of attack 28º under stagnation enthalpy between H(sub 0)=3.5MJ/kg to 21MJ/kg and stagnation pressure between P(sub 0)=14MPa to 60MPa. Under the condition above, Reynolds number based on the model diameter was varied from 0.2 to 1.3 million. With roughness elements, boundary layer became fully turbulent less than H(sub 0)=9MJ/kg condition. However, boundary layer was still laminar over H(sub 0)=13MJ/kg condition even with the highest roughness elements. An additional experiment was also performed to correct unexpected heat flux augmentation observed over H(sub 0)=9MJ/kg condition.

Chemical, physical and profile oceanographic data collected aboard the RYAN CHOUEST in the Gulf of Mexico from 2010-09-04 to 2010-09-08 in response to the Deepwater Horizon Oil Spill event (NODC Accession 0069120)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Chemical, physical and profile oceanographic data were collected aboard the RYAN CHOUEST in the Gulf of Mexico from 2010-09-04 to 2010-09-08 in response to the...

Irradiation performance of HTGR fuel in HFIR capsule HT-31

International Nuclear Information System (INIS)

Tiegs, T.N.; Robbins, J.M.; Hamner, R.L.; Montgomery, B.H.; Kania, M.J.; Lindemer, T.B.; Morgan, C.S.

1979-05-01

The capsule was irradiated in the High Flux Isotope Reactor at ORNL to peak particle temperatures up to 1600 0 C, fast neutron fluences (0.18 MeV) up to 9 x 10 25 n/m 2 , and burnups up to 8.9% FIMA for ThO 2 particles. The oxygen release from plutonium fissions was less than calculated, possibly because of the solid solution of SrO and rare earth oxides in UO 2 . Tentative results show that pyrocarbon permeability decreases with increasing fast neutron fluence. Fission products in sol-gel UO 2 particles containing natural uranium mostly behaved similarly to those in particles containing highly enriched uranium (HEU). Thus, much of the data base collected on HEU fuel can be applied to low-enriched fuel. Fission product palladium penetrated into the SiC on Triso-coated particles. Also the SiC coating provided some retention of /sup 110m/Ag. Irradiation above about 1200 0 C without an outer pyrocarbon coating degraded the SiC coating on Triso-coated particles

2018-05-09T08:04:19Z https://www.ajol.info/index.php/all/oai oai:ojs ...

African Journals Online (AJOL)

article/24359 2018-05-09T08:04:19Z ajfm:ART Equity and VAT in Tanzania. Mugoya, Patrick KD It is normally argued that Value Added Tax (VAT) is essentially regressive in the sense that all consumers pay the same amount of the tax per unit of ...

Power ramping test in the JMTR for PCI study of water reactor fuel

International Nuclear Information System (INIS)

Nakata, H.; Kanbara, M.; Ichikawa, M.

1984-01-01

Power ramping test is essential for PCI study of water reactor fuel. Boiling water capsules have been used for the tests in the JMTR. Heat generation of fuel rod in the capsule can be changed by the He-3 power control facility during reactor operation. Four specially designed fuel rods have been ramped to about 41-43 kW/m; two of them have small gaps filled with iodine, the other two are equipped with centerline temperature thermocouple. Fuel rod elongation detector is equipped to each capsule. For the fuel rods with small gap, unique contraction followed by ordinary fuel relaxation behaviour was observed right after the fast ramping. None of them failed. Future programme includes a series of tests of fuel rods irradiated in the high-pressure water loop at the JMTR and a verification test of remedy fuel which allows daily-load-following operation of BWRs. (author)

Hydrogen permeation measurement of the reduced activation ferritic steel F82H by the vacuum thermo-balance method

International Nuclear Information System (INIS)

Yoshida, Hajime; Enoeda, Mikio; Abe, Tetsuya; Akiba, Masato

2005-03-01

Hydrogen permeation fluxes of the reduced activation ferritic steel F82H were quantitatively measured by a newly proposed method, vacuum thermo-balance method, for a precise estimation of tritium leakage in a fusion reactor. We prepared sample capsules made of F82H, which enclosed hydrogen gas. The hydrogen in the capsules permeated through the capsule wall, and subsequently desorbed from the capsule surface during isothermal heating. The vacuum thermo-balance method allows simultaneous measurement of the hydrogen permeation flux by two independent methods, namely, the net weight reduction of the sample capsule and exhaust gas analysis. Thus the simultaneous measurements by two independent methods increase the reliability of the permeability measurement. When the gas pressure of enclosed hydrogen was 0.8 atm at the sample temperature of 673 K, the hydrogen permeation flux of F82H obtained by the net weight reduction and the exhaust gas analysis was 0.75x10 18 (H 2 /m 2 s) and 2.2x10 18 (H 2 /m 2 s), respectively. The ratio of the hydrogen permeation fluxes obtained by the net weight reduction to that measured by the exhaust gas analysis was in the range from 1/4 to 1/1 in this experiment. The temperature dependence of the estimated permeation flux was similar in both methods. Taking the uncertainties of both measurements into consideration, both results are supposed to be consistent. The enhancement of hydrogen permeation flux was observed from the sample of which outer surface was mechanically polished. Through the present experiments, it has been demonstrated that the vacuum thermo-balance method is effective for the measurement of hydrogen permeation rate of F82H. (author)

Morphologically well-defined Gd0.1Ce0.9O1.95 embedded Ba0.5Sr0.5Co0.8Fe0.2O3-δ nanofiber with an enhanced triple phase boundary as cathode for low-temperature solid oxide fuel cells

Science.gov (United States)

Kim, Chanho; Park, Hyunjung; Jang, Inyoung; Kim, Sungmin; Kim, Kijung; Yoon, Heesung; Paik, Ungyu

2018-02-01

Controlling triple phase boundary (TPB), an intersection of the ionic conductor, electronic conductor and gas phase as a major reaction site, is a key to improve cell performances for low-temperature solid oxide fuel cells. We report a synthesis of morphologically well-defined Gd0.1Ce0.9O1.95 (GDC) embedded Ba0.5Sr0.5Co0.8Fe0.2O3-δ (BSCF) nanofibers and their electrochemical performances as a cathode. Electrospun fibers prepared with a polymeric solution that contains crystalline Ba0.5Sr0.5Co0.8Fe0.2O3-δ particles in ∼200 nm size and Gd(NO3)3/Ce(NO3)3 precursors in an optimized weight ratio of 3 to 2 result in one dimensional structure without severe agglomeration and morphological collapse even after a high calcination at 1000 °C. As-prepared nanofibers have fast electron pathways along the axial direction of fibers, a higher surface area of 7.5 m2 g-1, and more oxygen reaction sites at TPBs than those of GDC/BSCF composite particles and core-shell nanofibers. As a result, the Gd0.1Ce0.9O1.95 embedded Ba0.5Sr0.5Co0.8Fe0.2O3-δ nanofiber cell shows excellent performances of the maximum power density of 0.65 W cm-2 at 550 °C and 1.02 W cm-2 at 600 °C, respectively.

In situ characterization of Hanford K Basins fuel

Energy Technology Data Exchange (ETDEWEB)

Pitner, A.L.

1998-01-06

Irradiated N Reactor uranium metal fuel is stored underwater in the Hanford K East and K West Basins. In K East Basin, fuel is stored in open canisters and defected fuel is free to react with the basin water. In K West Basin, the fuel is stored in sealed canisters filled with water containing a corrosion inhibitor (potassium nitrite). To gain a better understanding of the physical condition of the fuel in these basins, visual surveys using high resolution underwater cameras were conducted. The inspections included detailed lift and look examinations of a number of fuel assemblies from selected canisters in each basin. These examinations formed the bases for selecting specific fuel elements for laboratory testing and analyses as prescribed in the characterization plan for Hanford K Basin Spent Nuclear Fuel.

40 CFR 600.209-08 - Calculation of vehicle-specific 5-cycle fuel economy values for a model type.

Science.gov (United States)

2010-07-01

...-cycle fuel economy values for a model type. 600.209-08 Section 600.209-08 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy Regulations for 1977 and Later Model Year Automobiles-Procedures for...

Polyimide capsules may hold high pressure DT fuel without cryogenic support for the National Ignition Facility indirect-drive targets

International Nuclear Information System (INIS)

Sanchez, J.J.; Letts, S.A.

1997-01-01

New target designs for the Omega upgrade laser and ignition targets in the National Ignition Facility (NIF) require thick (80 - 100 microm) cryogenic fuel layers. The Omega upgrade target will require cryogenic handling after initial fill because of the high fill pressures and the thin capsule walls. For the NIF indirectly driven targets, a larger capsule size and new materials offer hope that they can be built, filled and stored in a manner similar to the targets used in the Nova facility without requiring cryogenic handling

Meltability in system of K2TaF7-NaF-NaCl-KCl

International Nuclear Information System (INIS)

Kartsev, V.E.; Kovalev, F.V.; Korshunov, B.G.

1975-01-01

Thermographic and visual-polythermal techniques were used to study the meltability in K 2 TaF 7 -NaF-NaCl-KCl system. The tetrahedron-forming sections NaF-NaCl-K 2 TaF 7 xKCl and NaF-K 2 TaF 7 xKCl-2K 2 TaF 7 xNaCl divide the concentration tetrahedron into three particular tetrahedra: NaF-K 2 TaF 7 xKCl-2K 2 TaF 7 xNaCl-K 2 TaF 7 , NaF-NaCl-K 2 TaF 7 xKCl-2K 2 TaF 7 xaCl, and NaF-NaCl-KCl-K 2 TaF 7 xKCl. Non-variant equilibrium points in all of the particular four-component systems have been determined

Design, fabrication and irradiation test report on HANARO instrumented capsule (03M-06U) for researches of universities in 2003

International Nuclear Information System (INIS)

Choo, K. N.; Kim, B. G.; Kang, Y. H.; Choi, M. H.; Cho, M. S.; Son, J. M.; Shin, Y. T.; Park, S. J.

2005-03-01

As a part of 2003 project for active utilization of HANARO, an instrumented capsule (03M-06U) was designed, fabricated and irradiated for the irradiation test of various nuclear materials under irradiation conditions requested by external researchers from universities. The basic structure of 03M-06U capsule was based on the 00M-01U, 01M-05U and 02M-05U capsules successfully irradiated in HANARO as 2000, 2001 and 2002 projects. However, because of the limited number of specimens and budget of 4 universities, the remained space of the capsule was charged with KAERI specimens for the development of the precise temperature control technology under irradiation. The material of the specimens is mainly Fe-based alloys partially mixed with Zr, Al and Cu-Ag alloys. The capsule is composed of 5 stages having many kinds of specimens and independent electric heater in each stage. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 5 sets of Ni-Ti-Fe neutron fluence monitors installed in the capsule. Various types of specimens such as tensile, Charpy, TEM, toughness, electrical resistance specimens were inserted in the capsule. The capsule was firstly irradiated in the CT test hole of HANARO of 30MW thermal output at 275∼500±10 .deg. C up to a fast neutron fluence of 5.4 x 10 20 (n/cm 2 ) (E>1.0MeV). The obtained results will be very valuable for the related researches of the users

The Effects of NiTi Hand and Rotary Canal Master U and K-Flex Instrumentation of Root Canal Morphology

Science.gov (United States)

1992-09-01

Rotary Canal Master "U" and K-Flex Instrumentation of Root Canal Morphology 6. AUTHOR(S) Robert H. Haller 7. PERFORMING ORGANIZATION NAME(S) AND ADDRESS(ES... endodontic failure. B. STATET OF THE PREL]2U Conventional endodontic instruments have re.nained basically unchanged since the introduction of the K-type...F;hcw"e that this nO𔃾 instruneent destim tended ti- produce rounder pr rMticns with • t!-rnnVortaticn than c rnvrnticnal 4 endodontic instruments

Physics of ignition for ICF capsules

International Nuclear Information System (INIS)

Lindl, J.D.

1989-01-01

The implosion of an ICF capsule must accomplish both compression of the main fuel to several hundred grams per cubic centimeter and heating and compression of the central region of the fuel to ignition. This report discusses the physics of these conditions

Focal uptake at the rotator interval or inferior capsule of shoulder on "1"8F-FDG PET/CT is associated with adhesive capsulitis

International Nuclear Information System (INIS)

Sridharan, Radhika; Engle, Mitchell Philip; Garg, Naveen; Wei, Wei; Amini, Behrang

2017-01-01

To determine if focal increased uptake at the rotator interval (RI) and/or inferior capsule (IC) on"1"8F-FDG PET/CT (''positive PET'') predicts the presence of adhesive capsulitis (AC). Three populations were retrospectively examined. Group 1 included 1,137 consecutive "1"8F-FDG PET/CT studies and was used to determine the prevalence of focal uptake at the RI or IC. Group 2 included 361 cases from a 10-year period with "1"8F-FDG PET/CT and MRI of shoulder performed within 45 days of each other and was used to enrich the study group. Group 3 included 109 randomly selected patients from the same time frame as groups 1 and 2 and was used to generate the control group. The study group consisted of 15 cases from the three groups, which had positive PET findings. PET/CT images were assessed in consensus by two musculoskeletal radiologists. The reference standard for a diagnosis of AC was clinical and was made by review of the medical record by a pain medicine physician. The prevalence of focal activity at either the RI or IC (''positive PET'') was 0.53%. Nine patients had a clinical diagnosis of AC and 15 patients had a positive PET. The sensitivity and specificity of PET for detection of AC was 56% and 87%, respectively. PET/CT had a positive likelihood ratio for AC of 6.3 (95% CI: 2.8-14.6). Increased uptake at the RI or IC on PET/CT confers a moderate increase in the likelihood of AC. (orig.)

Postirradiation examination report of TRISO and BISO coated ThO2 particles irradiated in capsules HT-31 and HT-33

International Nuclear Information System (INIS)

Sedlak, B.J.

1980-01-01

Capsules HT-31 and HT-33 were uninstrumented capsule experiments irradiated in the target position of the High-Flux Isotope Reactor at Oak Ridge National Laboratory. The experiments were used to evaluate the irradiation performance of (1) fuel fabricated in a 240-mm-diameter coater for production scale-up, (2) TRISO ThO 2 and BISO ThO 2 particles, and (3) fuel with certain OPyC variables. A total of 16 BISO particle samples and 32 TRISO particle samples were irradiated to fast neutron fluences ranging from 4.0 to 11.7 x 10 25 n/m 2 (E > 29 fJ)/sub HTGR/ and heavy metal burnups between 3.5% and 13.2% FIMA at temperatures from 1150 0 to 1530 0 C

Dynamic design load of type 2 water-flow capsule in Nuclear Safety Research Reactor in Tokai Research Establishment of Japan Atomic Energy Research Institute, and its reuse test

International Nuclear Information System (INIS)

1981-01-01

A report by the Nuclear Safety Bureau of the Science and Technology Agency to the Nuclear Safety Commission was presented on the validity of the dynamic design load of type 2 water-flow capsule and the method of its reuse test. The safety in both aspects of the capsule was confirmed. The Nuclear Safety Research Reactor (NSRR), in which the water-flow capsule is set, is a swimming pool type reactor, fueled with enriched uranium, having heat output of 300 kW in normal operation and maximum instantaneous heat output of 23,000 MW in pulse operation. The type 2 water-flow capsule, with the initial conditions simulating a power generating LWR plant and being appropriately set, is used to acquire the data on fuel behavior and destructive power in pulse irradiation. (J.P.N.)

Dynamic design load of type 2 water-flow capsule in Nuclear Safety Research Reactor in Tokai Research Establishment of Japan Atomic Energy Research Institute, and its reuse test

Energy Technology Data Exchange (ETDEWEB)

1981-01-01

A report by the Nuclear Safety Bureau of the Science and Technology Agency to the Nuclear Safety Commission was presented on the validity of the dynamic design load of type 2 water-flow capsule and the method of its reuse test. The safety in both aspects of the capsule was confirmed. The Nuclear Safety Research Reactor (NSRR), in which the water-flow capsule is set, is a swimming pool type reactor, fueled with enriched uranium, having heat output of 300 kW in normal operation and maximum instantaneous heat output of 23,000 MW in pulse operation. The type 2 water-flow capsule, with the initial conditions simulating a power generating LWR plant and being appropriately set, is used to acquire the data on fuel behavior and destructive power in pulse irradiation.

Alternate-Fueled Combustor-Sector Performance

Science.gov (United States)

Thomas, Anna E.; Saxena, Nikita T.; Shouse, Dale T.; Neuroth, Craig; Hendricks, Robert C.; Lynch, Amy; Frayne, Charles W.; Stutrud, Jeffrey S.; Corporan, Edwin; Hankins, Terry

2013-01-01

In order to realize alternative fueling for military and commercial use, the industry has set forth guidelines that must be met by each fuel. These aviation fueling requirements are outlined in MIL-DTL-83133F(2008) or ASTM D 7566 Annex (2011) standards, and are classified as "drop-in" fuel replacements. This report provides combustor performance data for synthetic-paraffinic-kerosene- (SPK-) type (Fischer-Tropsch (FT)) fuel and blends with JP-8+100, relative to JP-8+100 as baseline fueling. Data were taken at various nominal inlet conditions: 75 psia (0.52 MPa) at 500 degF (533 K), 125 psia (0.86 MPa) at 625 degF (603 K), 175 psia (1.21 MPa) at 725 degF (658 K), and 225 psia (1.55 MPa) at 790 degF (694 K). Combustor performance analysis assessments were made for the change in flame temperatures, combustor efficiency, wall temperatures, and exhaust plane temperatures at 3, 4, and 5 percent combustor pressure drop (DP) for fuel:air ratios (F/A) ranging from 0.010 to 0.025. Significant general trends show lower liner temperatures and higher flame and combustor outlet temperatures with increases in FT fueling relative to JP-8+100 fueling. The latter affects both turbine efficiency and blade and vane lives.

The Capsule Supports Survival but Not Traversal of Escherichia coli K1 across the Blood-Brain Barrier

OpenAIRE

Hoffman, Jill A.; Wass, Carol; Stins, Monique F.; Kim, Kwang Sik

1999-01-01

The vast majority of cases of gram-negative meningitis in neonates are caused by K1-encapsulated Escherichia coli. The role of the K1 capsule in the pathogenesis of E. coli meningitis was examined with an in vivo model of experimental hematogenous E. coli K1 meningitis and an in vitro model of the blood-brain barrier. Bacteremia was induced in neonatal rats with the E. coli K1 strain C5 (O18:K1) or its K1− derivative, C5ME. Subsequently, blood and cerebrospinal fluid (CSF) were obtained for c...

Chemical, physical, profile and laboratory analysis oceanographic data collected aboard the Brooks McCall in the Gulf of Mexico from 2010-08-09 to 2010-08-12 in response to the Deepwater Horizon Oil Spill event (NODC Accession 0069056)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Chemical, physical, profile and laboratory analysis oceanographic data were collected aboard the Brooks McCall in the Gulf of Mexico from 2010-08-09 to 2010-08-12 in...

Increasing Z-pinch vacuum hohlraum capsule coupling efficiency

International Nuclear Information System (INIS)

Callahan, Debbie; Vesey, Roger Alan; Cochrane, Kyle Robert; Nikroo, A.; Bennett, Guy R.; Schroen, Diana Grace; Ruggles, Laurence E.; Porter, John L.; Streit, Jon; Mehlhorn, Thomas Alan; Cuneo, Michael Edward

2004-01-01

Symmetric capsule implosions in the double-ended vacuum hohlraum (DEH) on Z have demonstrated convergence ratios of 14-21 for 2.15-mm plastic ablator capsules absorbing 5-7 kJ of x-rays, based on backlit images of the compressed ablator remaining at peak convergence (1). Experiments with DD-filled 3.3-mm diameter capsules designed to absorb 14 kJ of x-rays have begun as an integrated test of drive temperature and symmetry, complementary to thin-shell symmetry diagnostic capsules. These capsule implosions are characterized by excellent control of symmetry (< 3% time-integrated), but low hohlraum efficiency (< 2%). Possible methods to increase the capsule absorbed energy in the DEH include mixed-component hohlraums, large diameter foam ablator capsules, transmissive shine shields between the z-pinch and capsule, higher spoke electrode x-ray transmission, a double-sided power feed, and smaller initial radius z-pinch wire arrays. Simulations will explore the potential for each of these modifications to increase the capsule coupling efficiency for near-term experiments on Z and ZR

Drying results of K-Basin fuel element 3128W (run 2)

International Nuclear Information System (INIS)

Abrefah, J.; Klinger, G.S.; Oliver, B.M.; Marshman, S.C.; MacFarlan, P.J.; Ritter, G.A.; Flament, T.A.

1998-07-01

An N-Reactor outer fuel element that had been stored underwater in the Hanford 100 Area K-East Basin was subjected to a combination of low- and high-temperature vacuum drying treatments. These studies are part of a series of tests being conducted by Pacific Northwest National Laboratory on the drying behavior of N-Reactor spent nuclear fuel elements removed from both the K-West and K-East Basins. The drying test series was designed to test fuel elements that ranged from intact to severely damaged. The fuel element discussed in this report was removed from an open K-East canister (3128W) during the first fuel selection campaign conducted in 1995, and has remained in wet storage in the Postirradiation Testing Laboratory (PTL, 327 Building) since that time. Although it was judged to be breached during in-basin (i.e., K-Basin) examinations, visual inspection of this fuel element in the hot cell indicated that it was likely intact. Some scratches on the coating covering the cladding were identified before the furnace test. The drying test was conducted in the Whole Element Furnace Testing System located in G-Cell within the PTL. This test system is composed of three basic systems: the in-cell furnace equipment, the system gas loop, and the analytical instrument package. Element 3128W was subjected to the drying processes based on those proposed under the Integrated Process Strategy, which included a hot drying step. Results of the Pressure Rise and Gas Evolution Tests suggest that most of the free water in the system was released during the extended CVD cycle (68 hr versus 8 hr for the first run). An additional ∼0.34 g of water was released during the subsequent HVD phase, characterized by multiple water release peaks, with a principle peak at ∼180 C. This additional water is attributed to decomposition of a uranium hydrate (UO 4 ·4H 2 O/UO 4 ·2H 2 O) coating that was observed to be covering the surface of the fuel element to a thickness of ∼1.6 mg/cm 2 . A

KEY COMPARISON: CCT-K2.1: NRC/VNIIFTRI bilateral comparison of capsule-type standard platinum resistance thermometers from 13.8 K to 273.16 K

Science.gov (United States)

Hill, K. D.; Steele, A. G.; Dedikov, Y. A.; Shkraba, V. T.

2005-01-01

The Consultative Committee for Thermometry Key Comparison 2 (CCT-K2) results were published two years ago (2002 Metrologia 39 551-71). NRC served as the pilot laboratory for CCT-K2 and remains able to provide a scale and measurement system suitable for performing bilateral comparisons linked to the original key comparison results. In March 2003, measurements of two VNIIFTRI 100 Ω capsule-style platinum resistance thermometers (CSPRTs), S/N 356 and 476, were undertaken to relate their local calibration to the results from the CCT-K2 exercise. The NRC Leeds and Northrup (L&N) CSPRT S/N 1872174 provides the link to the CCT-K2 results. The three CSPRTs were compared at the eight defining cryogenic temperatures of the International Temperature Scale of 1990 (ITS-90) in the range from 13.8033 K to 273.16 K. The reader is referred to the full text of the CCT-K2 report for a detailed explanation of the methodology employed for the comparison. Only the details unique to the measurements reported here will be addressed in this article. The NRC/VNIIFTRI bilateral comparison of capsule-style platinum resistance thermometers over the range 13.8 K to 273.16 K has revealed calibrations at VNIIFTRI to be in agreement with the KCRV of CCT-K2 within the expanded uncertainty for all temperatures of the comparison with the exception of the triple point of hydrogen at 13.8033 K. One of the two CSPRTs supplied by VNIIFTRI was found to be discrepant as revealed by differences at the triple point of water and at the lowest temperatures of the comparison, and was therefore excluded from further analysis. The linkage to the CCT-K2 data supports the evaluation of the VNIIFTRI CMCs in Appendix C of the KCDB. Main text. To reach the main text of this paper, click on Final Report. Note that this text is that which appears in Appendix B of the BIPM key comparison database kcdb.bipm.org/. The final report has been peer-reviewed and approved for publication by the CCT, according to the provisions

The possibility of fuel cycle design for ABC/ATW complex with molten fuel on LiF-BeF2 basis

International Nuclear Information System (INIS)

Naumov, V.S.; Bychkov, A.V.

1995-01-01

The experience gained in the field of the development of molten salt reactors (MSR) can be made a basis of chemical processing of the ABC/ATW liquid fuel. The following combination of two processing principles are proposed for the ABC/ATW fuel (LiF-BeF 2 -PuF 3,(4) - MAF n ): - continious removal of radioactive gases, volatile impurities and 'noble fission products'; - portion-by-portion electrochemical processing with removal of rare earth elements and some other fission products at an autonomous plant. After processing the fuel salt is brought back to the blanket of the ABC/ATW complex. The analysis of information previously published in different countries allows for a safe assumption that the ABC/ATW fuel cycle with liquid fuel salt is feasible and can be demonstrated experimentally

The possibility of fuel cycle design for ABC/ATW complex with molten fuel on LiF-BeF2 basis

International Nuclear Information System (INIS)

Naumov, V. S.; Bychkov, A. V.

1995-01-01

The experience gained in the field of the development of molten salt reactors (MSR) can be made a basis of chemical processing of the ABC/ATW liquid fuel. The following combination of two processing principles are proposed for the ABC/ATW fuel (LiF-BeF2-PuF3,(4)-MAFn): -continious removal of radioactive gases, volatile impurities and 'noble fission products'; -portion-by-portion electrochemical processing with removal of rare earth elements and some other fission products at an autonomous plant. After processing the fuel salt is brought back to the blanket of the ABC/ATW complex. The analysis of information previously published in different countries allows for a safe assumption that the ABC/ATW fuel cycle with liquid fuel salt is feasible and can be demonstrated experimentally

Probing cell internalisation mechanics with polymer capsules.

Science.gov (United States)

Chen, Xi; Cui, Jiwei; Ping, Yuan; Suma, Tomoya; Cavalieri, Francesca; Besford, Quinn A; Chen, George; Braunger, Julia A; Caruso, Frank

2016-10-06

We report polymer capsule-based probes for quantifying the pressure exerted by cells during capsule internalisation (P in ). Poly(methacrylic acid) (PMA) capsules with tuneable mechanical properties were fabricated through layer-by-layer assembly. The P in was quantified by correlating the cell-induced deformation with the ex situ osmotically induced deformation of the polymer capsules. Ultimately, we found that human monocyte-derived macrophage THP-1 cells exerted up to approximately 360 kPa on the capsules during internalisation.

The Hydraulic Test Procedure for Non-instrumented Irradiation Test Rig of Annular Fuel

Energy Technology Data Exchange (ETDEWEB)

Kim, Dae Ho; Lee, Kang Hee; Shin, Chang Hwan; Park, Chan Kook

2008-08-15

This report presents the procedure of pressure drop test, vibration test and endurance test for the non-instrumented rig using the irradiation test in HANARO of advanced PWR annular fuel which were designed and fabricated by KAERI. From the out-pile thermal hydraulic tests, confirm the flow rate at the 200 kPa pressure drop and measure the RMS displacement at this time. And the endurance test is confirmed the wear and the integrity of the non-instrumented rig at the 110% design flow rate. This out-pile test perform the Flow-Induced Vibration and Pressure Drop Experimental Tester(FIVPET) facility. The instruments in FIVPET facility was calibrated in KAERI and the pump and the thermocouple were certified by manufacturer.

Temperature, salinity and other variables collected from discrete sample and profile observations using CTD, bottle and other instruments from THOMAS G. THOMPSON in the North Pacific Ocean from 1985-08-04 to 1985-09-07 (NCEI Accession 0143394)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — NCEI Accession 0143394 includes discrete sample and profile data collected from THOMAS G. THOMPSON in the North Pacific Ocean from 1985-08-04 to 1985-09-07 and...

Equation of state for L.M.F.B.R. fuel (measurement of fission gas release during transients)

International Nuclear Information System (INIS)

Combette, P.; Barthelemy, P.

1979-01-01

A sample of fuel (UO 2 or UPuO 2 ) can be heated by fission in a heating transient up to energy deposition 4000 j/g, in the Silene reactor. The Kistler type capsule, the calorimeter device and the radiochemical analysis of fission products enable the pressure pulse and the fuel energy deposition to be measured. So, the relationship between the fuel vapour pressure and the fuel specific energy can be deduced. Peaks pressure (about 1 MPa) coming from fresh UO 2 vaporization, have been measured on a 7 milliseconds time scale. There is a good agreement with the E.O.S. for fresh UO 2 , which is well known for low pressure (1 MPa). Numerous tests have been done with 93% enriched UO 2 and a first test with highly active fuel containing plutonium (15 at %) has been performed. The capsule allows the released gas coming from the irradiated fuel to be retained for measurements and analysis. To investigate the mode of fuel disruption, in-pile fission-heated fuel pellets has been recorded by high speed cinematography

Focal uptake at the rotator interval or inferior capsule of shoulder on {sup 18}F-FDG PET/CT is associated with adhesive capsulitis

Energy Technology Data Exchange (ETDEWEB)

Sridharan, Radhika [Universiti Kebangsaan Malaysia Medical Centre, Department of Diagnostic Imaging, Kuala Lumpur (Malaysia); Engle, Mitchell Philip [The University of Texas M.D. Anderson Cancer Center, Department of Pain Medicine, Houston, TX (United States); Garg, Naveen [The University of Texas M.D. Anderson Cancer Center, Department of Diagnostic Imaging, Houston, TX (United States); Wei, Wei; Amini, Behrang [The University of Texas M.D. Anderson Cancer Center, Department of Biostatistics, Houston, TX (United States)

2017-04-15

To determine if focal increased uptake at the rotator interval (RI) and/or inferior capsule (IC) on{sup 18}F-FDG PET/CT (''positive PET'') predicts the presence of adhesive capsulitis (AC). Three populations were retrospectively examined. Group 1 included 1,137 consecutive {sup 18}F-FDG PET/CT studies and was used to determine the prevalence of focal uptake at the RI or IC. Group 2 included 361 cases from a 10-year period with {sup 18}F-FDG PET/CT and MRI of shoulder performed within 45 days of each other and was used to enrich the study group. Group 3 included 109 randomly selected patients from the same time frame as groups 1 and 2 and was used to generate the control group. The study group consisted of 15 cases from the three groups, which had positive PET findings. PET/CT images were assessed in consensus by two musculoskeletal radiologists. The reference standard for a diagnosis of AC was clinical and was made by review of the medical record by a pain medicine physician. The prevalence of focal activity at either the RI or IC (''positive PET'') was 0.53%. Nine patients had a clinical diagnosis of AC and 15 patients had a positive PET. The sensitivity and specificity of PET for detection of AC was 56% and 87%, respectively. PET/CT had a positive likelihood ratio for AC of 6.3 (95% CI: 2.8-14.6). Increased uptake at the RI or IC on PET/CT confers a moderate increase in the likelihood of AC. (orig.)

Development of endplug welding technology for irradiation testing capsule

Energy Technology Data Exchange (ETDEWEB)

Lee, J. W.; Shin, Y. T.; Kim, S. S.; Kim, B. K.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

2001-10-01

To evaluate the performance of newly developed nuclear fuel, it is necessary to irradiate the fuel at a research reactor and examine the irradiated fuel. For the irradiation test in a reasearch reactor, a fuel assembly which is generally called a capsule should be fabricated, considering the fuel irradiation plan and the characteristics of the reactor to be used. And also the fuel elements containing the developed fuel pellets should be made and assembled into a capsule. In this study, the welding method, welding equipment, welding conditions and parameters were developed to make fuel elements for the irradiation test at the HANARO research reactor. The TIG welding method using automatic orbital tube welding system was adopted and the welding joint design was developed for the fabrication of various kinds of irradiation fuel elements. And the optimal welding conditions and parameters were also established for the endplug welding of Zircaloy-4 cladding tube.

Summary Report for Capsule Dry Storage Project

Energy Technology Data Exchange (ETDEWEB)

JOSEPHSON, W S

2003-09-04

There are 1.936 cesium (Cs) and strontium (Sr) capsules stored in pools at the Waste Encapsulation and Storage Facility (WESF). These capsules will be moved to dry storage on the Hanford Site as an interim measure to reduce risk. The Cs/Sr Capsule Dry Storage Project (CDSP) is conducted under the assumption the capsules will eventually be moved to the repository at Yucca Mountain, and the design criteria include requirements that will facilitate acceptance at the repository. The storage system must also permit retrieval of capsules in the event vitrification of the capsule contents is pursued. A cut away drawing of a typical cesium chloride (CsCI) capsule and the capsule property and geometry information are provided in Figure 1.1. Strontium fluoride (SrF{sub 2}) capsules are similar in design to CsCl capsules. Further details of capsule design, current state, and reference information are given later in this report and its references. Capsule production and life history is covered in WMP-16938, Capsule Characterization Report for Capsule Dry Storage Project, and is briefly summarized in Section 5.2 of this report.

Turbidity, SOLAR RADIATION - ATMOSPHERIC and other data from NATHANIEL B. PALMER from 1994-11-09 to 1994-12-08 (NODC Accession 9700174)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Temperature, salinity, and other data were collected from bottle and CTD casts from the Nathaniel B. Palmer from 09 November 1994 to 08 December 1994. Data were...

Capsule HRB-15B postirradiation examination report

International Nuclear Information System (INIS)

Ketterer, J.W.; Bullock, R.E.

1981-06-01

Capsule HRB-15B design tested 184 thin graphite trays containing unbonded fuel particles to peak exposures of 6.6 x 10 25 n/m 2 (E > 29 fJ)/sub HTGR/ fast fluence, approx. 27% fissions per initial metal atom (FIMA) fissile burnup, and 6% FIMA fertile burnup at nominal time-averaged temperatures of 815 to 915 0 C. The capsule tested a variety of low-enriched uranium (approx. 19.5% U-235) fissile particle types, including UC 2 , UC/sub x/O/sub y/, UO 2 , zirconium-buffered UO 2 (referred to in this report as UO 2 /sup *), and 1:1(Th,U)O 2 with both TRISO and silicon-BISO coatings. All fertile particles were ThO 2 with BISO, silicon-BISO, or TRISO coatings. The findings indicated that all TRISO particles retained virtually all of their fission product inventories, except small quantities of silver, at these irradiation temperatures, while some of the silicon-BISO particles released significant amounts of both silver and cesium. No kernel migration, pressure vessel, or outer pyrolytic carbon (OPyC) failures were observed in the fuel particles, which had total diameters of 2 /sup */ particles exhibited no detrimental irradiation effects, but they contained pure carbon precipitates in the kernels after irradiation which were not observed in the undoped UO 2 particles. Postirradiation examination revealed no differences in the irradiation performance of three UC/sub x/O/sub y/ kernel types with varying oxygen/uranium ratios

K Basins fuel encapsulation and storage hazard categorization

International Nuclear Information System (INIS)

Porten, D.R.

1994-12-01

This document establishes the initial hazard categorization for K-Basin fuel encapsulation and storage in the 100 K Area of the Hanford site. The Hazard Categorization for K-Basins addresses the potential for release of radioactive and non-radioactive hazardous material located in the K-Basins and their supporting facilities. The Hazard Categorization covers the hazards associated with normal K-Basin fuel storage and handling operations, fuel encapsulation, sludge encapsulation, and canister clean-up and disposal. The criteria categorizes a facility based on total curies per radionuclide located in the facility. Tables 5-3 and 5-4 display the results in section 5.0. In accordance with DOE-STD-1027 and the analysis provided in section 5.0, the K East Basin fuel encapsulation and storage activity and the K West Basin storage are classified as a open-quotes Category 2close quotes Facility

Ledarskapet och de två faktorerna : En induktiv studie för att förstå hur organisationsfaktorer och könsstereotyper påverkar ledarskap

OpenAIRE

Larsson, Hampus; Tang, Vinh; Vannfält, Jessica

2018-01-01

Sammanfattning   Titel: Ledarskapet och de två faktorerna: En induktiv studie för att förstå hur organisationsfaktorer och könsstereotyper påverkar ledarskap.   Seminariedatum: 2018-01-09   Akademi: Akademin för Ekonomi, Samhälle och Teknik, Mälardalen Högskola   Ämne/kurs: FOA300, examensarbete i företagsekonomi på kandidatnivå.   Författare: Larsson, Hampus 930903, Tang, Vinh 940523, Vannfält, Jessica 911031   Handledare: Magnus Linderström   Syfte: Syftet med denna studie är att genom en i...

Scholarly productivity and professional advancement of junior researchers receiving KL2, K23, or K08 awards at a large public research institution.

Science.gov (United States)

Amory, John K; Louden, Diana K N; McKinney, Christy; Rich, Joanne; Long-Genovese, Stacy; Disis, Mary L

2017-04-01

How the productivity and careers of KL2 scholars compare with scholars receiving individual K-awards is unknown. The productivity of KL2 scholars (n=21) at our institution was compared with that of K08 (n=34) and K23 (n=26) scholars. KL2 and K23 scholars had greater productivity than K08 scholars ( p =0.01). Professional advancement was similar among groups. At our institution, scholarly productivity and professional advancement did not differ by type of K-award.

Drop-in capsule testing of plutonium-based fuels in the Advanced Test Reactor

International Nuclear Information System (INIS)

Chang, G.S.; Ryskamp, J.M.; Terry, W.K.; Ambrosek, R.G.; Palmer, A.J.; Roesener, R.A.

1996-09-01

The most attractive way to dispose of weapons-grade plutonium (WGPu) is to use it as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PuO[sub 2]) mixed with urania (UO[sub 2]). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in- reactor thermal, mechanical, and fission gas release behavior of the prototype fuel will most likely be required in a limited number of test reactor irradiations. The application to license operation with MOX fuel must be amply supported by experimental data. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) is capable of playing a key role in the irradiation, development, and licensing of these new fuel types. The ATR is a 250- MW (thermal) LWR designed to study the effects of intense radiation on reactor fuels and materials. For 25 years, the primary role of the ATR has been to serve in experimental investigations for the development of advanced nuclear fuels. Both large- and small-volume test positions in the ATR could be used for MOX fuel irradiation. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. Furthermore, these data can be obtained more quickly by using ATR instead of testing in a commercial LWR. Our previous work in this area has demonstrated that it is technically feasible to perform MOX fuel testing in the ATR. This report documents our analyses of sealed drop-in capsules containing plutonium-based test specimens placed in various ATR positions

In-pile instrumentation improvements for fuel irradiations in test reactor

International Nuclear Information System (INIS)

Blanc, J.Y.; Bernard, J.L.; Estrade, J.; Geoffroy, G.

1996-01-01

Knowledge of fuel limits and safety margins in normal and off-normal transients in nuclear power plants remains a constant preoccupation for electricity producers and fuel manufacturers. Accurate determination of such limits, through fuel irradiation testing in the OSIRIS reactor at Saclay is closely linked to the reliability of appropriate instrumentation techniques. Two paths are currently followed to obtain short experimental rods: segmented fuel coming directly from power plants, or re-fabrication of rods in hot cells with our FABRICE process. It can be associated with instrumentation such as fuel centerline thermocouple in annular pellets, pressure transducer or fission gas release measurement by gamma-spectrometry using helium sweeping, in analytic experiments. Our present development, to be implemented in 1993, is the the centerline instrumentation of a fuel column with solid pellets. Inserting the thermocouple requires a cold drilling machine, using CO 2 freezing of broken UO 2 (with liquid nitrogen). During the fuel rod irradiation itself, we try to lower the uncertainties associated to power determination, using thermal balance or neutronic calibration, or even gamma spectrometry. A description of the new test train designed for the ISABELLE water loop in OSIRIS is given, with special emphasis on instrumentation: a LVDT for measuring fuel rod elongation and eventual clad failure, and increased number and better localization of thermocouples and SPDN. The third part is devoted to the measurements by optical microdensitometry of neutron radiographs of the fuel pellet dish modification after irradiation. Dishes are generally disappearing through thermal and mechanical deformation of the pellet, and this can eventually be modelized to better understand pellet-cladding mechanical interaction. (author). 3 refs, 5 figs

Photoproduction of the f1(1285 ) meson

Science.gov (United States)

Dickson, R.; Schumacher, R. A.; Adhikari, K. P.; Akbar, Z.; Amaryan, M. J.; Anefalos Pereira, S.; Badui, R. A.; Ball, J.; Battaglieri, M.; Batourine, V.; Bedlinskiy, I.; Biselli, A.; Boiarinov, S.; Briscoe, W. J.; Burkert, V. D.; Cao, T.; Carman, D. S.; Celentano, A.; Chandavar, S.; Charles, G.; Chetry, T.; Ciullo, G.; Colaneri, L.; Cole, P. L.; Compton, N.; Contalbrigo, M.; Cortes, O.; Crede, V.; D'Angelo, A.; Dashyan, N.; De Vita, R.; De Sanctis, E.; Deur, A.; Djalali, C.; Dugger, M.; Dupre, R.; El Alaoui, A.; El Fassi, L.; Eugenio, P.; Fanchini, E.; Fedotov, G.; Filippi, A.; Fleming, J. A.; Gevorgyan, N.; Ghandilyan, Y.; Gilfoyle, G. P.; Giovanetti, K. L.; Girod, F. X.; Gothe, R. W.; Griffioen, K. A.; Guo, L.; Hafidi, K.; Hakobyan, H.; Hanretty, C.; Harrison, N.; Hattawy, M.; Holtrop, M.; Hicks, K.; Hughes, S. M.; Ilieva, Y.; Ireland, D. G.; Ishkhanov, B. S.; Isupov, E. L.; Jiang, H.; Jo, H. S.; Joosten, S.; Keller, D.; Khachatryan, G.; Khandaker, M.; Kim, A.; Kim, W.; Klein, F. J.; Kubarovsky, V.; Kuleshov, S. V.; Lanza, L.; Lenisa, P.; Livingston, K.; Lu, H. Y.; MacGregor, I. J. D.; Mattione, P.; McKinnon, B.; Meyer, C. A.; Mirazita, M.; Markov, N.; Mokeev, V.; Moriya, K.; Munevar, E.; Murdoch, G.; Nadel-Turonski, P.; Net, L. A.; Ni, A.; Osipenko, M.; Ostrovidov, A. I.; Park, K.; Pasyuk, E.; Phelps, W.; Pisano, S.; Pogorelko, O.; Price, J. W.; Prok, Y.; Puckett, A. J. R.; Raue, B. A.; Ripani, M.; Rizzo, A.; Rosner, G.; Roy, P.; Salgado, C.; Seder, E.; Sharabian, Y. G.; Skorodumina, Iu.; Smith, E. S.; Smith, G. D.; Sober, D.; Sokhan, D.; Sparveris, N.; Stepanyan, S.; Strakovsky, I. I.; Stankovic, I.; Strauch, S.; Sytnik, V.; Taiuti, M.; Ungaro, M.; Voskanyan, H.; Voutier, E.; Walford, N. K.; Watts, D. P.; Weygand, D.; Wood, M. H.; Zachariou, N.; Zana, L.; Zhang, J.; Zonta, I.; CLAS Collaboration

2016-06-01

The f1(1285 ) meson with mass 1281.0 ±0.8 MeV/c2 and width 18.4 ±1.4 MeV (full width at half maximum) was measured for the first time in photoproduction from a proton target using CLAS at Jefferson Lab. Differential cross sections were obtained via the η π+π-,K+K¯0π- , and K-K0π+ decay channels from threshold up to a center-of-mass energy of 2.8 GeV. The mass, width, and an amplitude analysis of the η π+π- final-state Dalitz distribution are consistent with the axial-vector JP=1+ f1(1285 ) identity, rather than the pseudoscalar 0- η (1295 ) . The production mechanism is more consistent with s -channel decay of a high-mass N* state and not with t -channel meson exchange. Decays to η π π go dominantly via the intermediate a0±(980 ) π∓ states, with the branching ratio Γ [a0π (noK ¯K )] /Γ [η π π (all)] =0.74 ±0.09 . The branching ratios Γ (K K ¯π ) /Γ (η π π ) =0.216 ±0.033 and Γ (γ ρ0) /Γ (η π π ) =0.047 ±0.018 were also obtained. The first is in agreement with previous data for the f1(1285 ) , while the latter is lower than the world average.

The possibility of fuel cycle design for ABC/ATW complex with molten fuel on LiF-BeF{sub 2} basis

Energy Technology Data Exchange (ETDEWEB)

Naumov, V.S.; Bychkov, A.V. [Research Institute of Atomic Reactors, Dimitrovgrad (Russian Federation)

1995-10-01

The experience gained in the field of the development of molten salt reactors (MSR) can be made a basis of chemical processing of the ABC/ATW liquid fuel. The following combination of two processing principles are proposed for the ABC/ATW fuel (LiF-BeF{sub 2}-PuF{sub 3,(4)} - MAF{sub n}): - continious removal of radioactive gases, volatile impurities and {open_quotes}noble fission products{close_quotes}; - portion-by-portion electrochemical processing with removal of rare earth elements and some other fission products at an autonomous plant. After processing the fuel salt is brought back to the blanket of the ABC/ATW complex. The analysis of information previously published in different countries allows for a safe assumption that the ABC/ATW fuel cycle with liquid fuel salt is feasible and can be demonstrated experimentally.

Ultra-clean Fischer-Tropsch (F-T) Fuels Production and Demonstration Project

Energy Technology Data Exchange (ETDEWEB)

Stephen P. Bergin

2006-06-30

The objective of the DOE-NETL Fischer-Tropsch (F-T) Production and Demonstration Program was to produce and evaluate F-T fuel derived from domestic natural gas. The project had two primary phases: (1) fuel production of ultra-clean diesel transportation fuels from domestic fossil resources; and (2) demonstration and performance testing of these fuels in engines. The project also included a well-to-wheels economic analysis and a feasibility study of small-footprint F-T plants (SFPs) for remote locations such as rural Alaska. During the fuel production phase, ICRC partnered and cost-shared with Syntroleum Corporation to complete the mechanical design, construction, and operation of a modular SFP that converts natural gas, via F-T and hydro-processing reactions, into hydrogensaturated diesel fuel. Construction of the Tulsa, Oklahoma plant started in August 2002 and culminated in the production of over 100,000 gallons of F-T diesel fuel (S-2) through 2004, specifically for this project. That fuel formed the basis of extensive demonstrations and evaluations that followed. The ultra-clean F-T fuels produced had virtually no sulfur (less than 1 ppm) and were of the highest quality in terms of ignition quality, saturation content, backend volatility, etc. Lubricity concerns were investigated to verify that commercially available lubricity additive treatment would be adequate to protect fuel injection system components. In the fuel demonstration and testing phase, two separate bus fleets were utilized. The Washington DC Metropolitan Area Transit Authority (WMATA) and Denali National Park bus fleets were used because they represented nearly opposite ends of several spectra, including: climate, topography, engine load factor, mean distance between stops, and composition of normally used conventional diesel fuel. Fuel evaluations in addition to bus fleet demonstrations included: bus fleet emission measurements; F-T fuel cold weather performance; controlled engine dynamometer

Electrochemical performance of Ni0.8Cu0.2/Ce0.8Gd0.2O1.9 cermet anodes with functionally graded structures for intermediate-temperature solid oxide fuel cell fueled with syngas

Science.gov (United States)

Miyake, Michihiro; Iwami, Makoto; Takeuchi, Mizue; Nishimoto, Shunsuke; Kameshima, Yoshikazu

2018-06-01

The electrochemical performance of layered Ni0.8Cu0.2/Ce0.8Gd0.2O1.9 (GDC) cermet anodes is investigated for intermediate-temperature solid oxide fuel cells (IT-SOFCs) at 600 °C using humidified (3% H2O) model syngas with a molar ratio of H2/CO = 3/2 as the fuel. From the results obtained, the electrochemical performance of the functionally graded multi-layered anodes is found to be superior to the mono-layered anodes. The test cell with a bi-layered anode consisting of 100 mass% Ni0.8Cu0.2/0 mass% GDC (10M/0E) and 70 mass% Ni0.8Cu0.2/30 mass% GDC (7M/3E) exhibits high power density. The test cell with a tri-layered anode consisting of 10M/0E, 7M/3E, and 50 mass% Ni0.8Cu0.2/50 mass% GDC (5M/5E) exhibits an even higher power density, suggesting that 10M/0E and 5M/5E layers contribute to the current collecting part and active part, respectively.

The use of tritium rich capsules with 25-35% deuterium to achieve ignition at the National Ignition Facility

Science.gov (United States)

Wilson, D. C.; Spears, B. K.; Hatchett, S. P., Ii; Cerjan, C. J.; Springer, P. T.; Clark, D. S.; Edwards, M. J.; Salmonson, J. D.; Weber, S. V.; Hammel, B. A.; Grim, G. P.; Herrmann, H. W.; Wilke, M. D.

2010-08-01

Diagnostics such as neutron yield, ion temperature, image size and shape, and bang time in capsules with >~25 % deuterium fuel show changes due to burn product heating. The comparison of performance between a THD(2%) and THD(35%) can help predict ignition in a TD(50%) capsule. Surrogacy of THD capsules to TD(50%) is incomplete due to variations in fuel molecular vapour pressures. TD(25-35%) capsules might be preferred to study hot spot heating, but at the risk of increased fuel/ablator mixing.

The use of tritium rich capsules with 25-35% deuterium to achieve ignition at the National Ignition Facility

International Nuclear Information System (INIS)

Wilson, D C; Grim, G P; Herrmann, H W; Wilke, M D; Spears, B K; Ii, S P Hatchett; Cerjan, C J; Springer, P T; Clark, D S; Edwards, M J; Salmonson, J D; Weber, S V; Hammel, B A

2010-01-01

Diagnostics such as neutron yield, ion temperature, image size and shape, and bang time in capsules with >∼25 % deuterium fuel show changes due to burn product heating. The comparison of performance between a THD(2%) and THD(35%) can help predict ignition in a TD(50%) capsule. Surrogacy of THD capsules to TD(50%) is incomplete due to variations in fuel molecular vapour pressures. TD(25-35%) capsules might be preferred to study hot spot heating, but at the risk of increased fuel/ablator mixing.

The use of tritium rich capsules with 25-35% deuterium to achieve ignition at the National Ignition Facility

Energy Technology Data Exchange (ETDEWEB)

Wilson, D C; Grim, G P; Herrmann, H W; Wilke, M D [Los Alamos National Laboratory, Los Alamos, NM, 87545 (United States); Spears, B K; Ii, S P Hatchett; Cerjan, C J; Springer, P T; Clark, D S; Edwards, M J; Salmonson, J D; Weber, S V; Hammel, B A, E-mail: dcw@lanl.go [Lawrence Livermore National Laboratory, Livermore, CA, 94550 (United States)

2010-08-01

Diagnostics such as neutron yield, ion temperature, image size and shape, and bang time in capsules with >{approx}25 % deuterium fuel show changes due to burn product heating. The comparison of performance between a THD(2%) and THD(35%) can help predict ignition in a TD(50%) capsule. Surrogacy of THD capsules to TD(50%) is incomplete due to variations in fuel molecular vapour pressures. TD(25-35%) capsules might be preferred to study hot spot heating, but at the risk of increased fuel/ablator mixing.

Development of remote welding techniques for in-pile IASCC capsules and evaluation of material integrity on capsules for long irradiation period

Science.gov (United States)

Shibata, A.; Nakano, J.; Ohmi, M.; Kawamata, K.; Nakagawa, T.; Tsukada, T.

2012-03-01

To simulate irradiation assisted stress corrosion cracking (IASCC) behavior by in-pile experiments, it is necessary to irradiate specimens up to a neutron fluence that is higher than the IASCC threshold fluence. Pre-irradiated specimens must be relocated from pre-irradiation capsules to in-pile capsules. Hence, a remote welding machine has been developed. And the integrity of capsule housing for a long term irradiation was evaluated by tensile tests in air and slow strain rate tests in water. Two type specimens were prepared. Specimens were obtained from the outer tubes of capsule irradiated to 1.0-3.9 × 1026 n/m2 (E > 1 MeV). And specimens were irradiated in a leaky capsule to 0.03-1.0 × 1026 n/m2. Elongation more than 15% in tensile test at 423 K was confirmed and no IGSCC fraction was shown in SSRT at 423 K which was estimated as temperature at the outer tubes of the capsule under irradiation.

Development of remote welding techniques for in-pile IASCC capsules and evaluation of material integrity on capsules for long irradiation period

International Nuclear Information System (INIS)

Shibata, A.; Nakano, J.; Ohmi, M.; Kawamata, K.; Nakagawa, T.; Tsukada, T.

2012-01-01

To simulate irradiation assisted stress corrosion cracking (IASCC) behavior by in-pile experiments, it is necessary to irradiate specimens up to a neutron fluence that is higher than the IASCC threshold fluence. Pre-irradiated specimens must be relocated from pre-irradiation capsules to in-pile capsules. Hence, a remote welding machine has been developed. And the integrity of capsule housing for a long term irradiation was evaluated by tensile tests in air and slow strain rate tests in water. Two type specimens were prepared. Specimens were obtained from the outer tubes of capsule irradiated to 1.0–3.9 × 10 26 n/m 2 (E > 1 MeV). And specimens were irradiated in a leaky capsule to 0.03–1.0 × 10 26 n/m 2 . Elongation more than 15% in tensile test at 423 K was confirmed and no IGSCC fraction was shown in SSRT at 423 K which was estimated as temperature at the outer tubes of the capsule under irradiation.

40 CFR 600.207-08 - Calculation and use of vehicle-specific 5-cycle-based fuel economy values for vehicle...

Science.gov (United States)

2010-07-01

...-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy Regulations for 1977 and Later Model Year... for each vehicle under § 600.114-08 and as approved in § 600.008-08 (c), are used to determine vehicle... fuel economy value exists for an electric vehicle configuration, all values for that vehicle...

The fuel to clad heat transfer coefficient in advanced MX-type fuel pins

International Nuclear Information System (INIS)

Caligara, F.; Campana, M.; Mandler, R.; Blank, H.

1979-01-01

Advanced fuels (mixed carbides, nitrides and carbonitrides) are characterised by a high thermal conductivity compared to that of oxide fuels (5 times greater) and their behaviour under irradiation (amount of swelling, fracture behaviour, restructuring) is far more sensitive to the design parameters and to the operating temperature than that of oxide fuels. The use of advanced fuels is therefore conditioned by the possibility of mastering the above phenomena, and the full exploitation of their favorable neutron characteristics depends upon a good understanding of the mutual relationships of the various parameters, which eventually affect the mechanical stability of the pin. By far the most important parameter is the radial temperature profile which controls the swelling of the fuel and the build-up of stress fields within the pin. Since the rate of fission gas swelling of these fuels is relatively large, a sufficient amount of free space has to be provided within the pin. This space originally appears as fabrication porosity and as fuel-to-clad clearance. Due to the large initial gap width and to the high fuel thermal conductivity, the range of the fuel operating temperatures is mainly determined by the fuel-to-clad heat transfer coefficient h, whose correct determination becomes one of the central points in modelling. During the many years of modelling activity in the field of oxide fuels, several theoretical models have been developed to calculate h, and a large amount of experimental data has been produced for the empirical adjustment of the parameters involved, so that the situation may be regarded as rather satisfactory. The analysis lead to the following conclusions. A quantitative comparison of experimental h-values with existing models for h requires rather sophisticated instrumented irradiation capsules, which permit the measurement of mechanical data (concerning fuel and clad) together with heat rating and temperatures. More and better well-instrumented

Upgrade of Dhruva fuel channel flow instrumentation

International Nuclear Information System (INIS)

Gadgil, Kaustubh; Awale, P.K.; Sengupta, C.; Sumanth, P.; Roy, Kallol

2014-01-01

Dhruva, a 100 MW Heavy Water moderated and cooled, vertical tank-type Research Reactor, using metallic natural Uranium fuel has flow instrumentation for all the 144 fuel channels, consisting of venturi and triplicate DP gauges for each fuel channel. These gauges provide contacts for generation of reactor trip on low flow through fuel channel. These DP gauges were facing numerous generic and ageing related failures over the years and was also difficult to maintain owing to obsolescence. While considering an upgrade for these DP gauges, it was also planned to replace the existing Coolant Low Flow Trip (CLFT) system with a computer based Reactor Trip Logic System (RTLS). Being a retrofit job, the existing panels for mounting the gauges, cable layout, impulse tubing layout, etc. were retained, thereby simplifying the site execution, reducing reactor down time and also reducing person-milli-Sievert consumption. A customized Electronic DP Indicating Switch (EDPIS) was conceptualized for achieving these objectives. Such a design, utilizing a standard DP transmitter with customized electronic circuitry, was developed, evaluated and finalized after a series of factory trials, field trials and prototyping. The instrument design included contact input for existing CLFT system and also provision for 4-20 mA current output for the proposed computer based RTLS. The display and form factor of the instrument remained identical to older one and ensures familiarity of O and M personnel. Since EDPIS is classified as Safety Class IA, stringent type tests, hardware FMEA and V and V of the micro-controller software were carried out as per the requirements laid down by relevant standards for qualification of these instruments. Being a customized instrument, the manufacturing process was closely monitored and was followed by stringent QA plan and acceptance tests. A total of 396 gauges were replaced in a phased manner during scheduled fuelling outages and thereby did not affect reactor

Postirradiation examinations of fuel pins from the GCFR F-1 series of mixed-oxide fuel pins at 5.5 at. % burnup

International Nuclear Information System (INIS)

Strain, R.V.; Johnson, C.E.

1978-05-01

Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760 0 C. The maximum diametral change that occurred during irradiation was 0.2% ΔD/D 0 . The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. Teh postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred

Postirradiation examinations of fuel pins from the GCFR F-1 series of mixed-oxide fuel pins at 5. 5 at. % burnup

Energy Technology Data Exchange (ETDEWEB)

Strain, R V; Johnson, C E

1978-05-01

Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760/sup 0/C. The maximum diametral change that occurred during irradiation was 0.2% ..delta..D/D/sub 0/. The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. Teh postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred.

NMR relaxation dispersion of Miglyol molecules confined inside polymeric micro-capsules.

Science.gov (United States)

Nechifor, Ruben; Ardelean, Ioan; Mattea, Carlos; Stapf, Siegfried; Bogdan, Mircea

2011-11-01

Frequency dependent NMR relaxation studies have been carried out on Miglyol molecules confined inside core shell polymeric capsules to obtain a correlation between capsule dimension and the measurable parameters. The polymeric capsules were prepared using an interfacial polymerization technique for three different concentrations of Miglyol. It was shown that the variation of Miglyol concentration influences the capsule dimension. Their average size was estimated using the pulsed field gradient diffusometry technique. The relaxation dispersion curves were obtained at room temperature by a combined use of a fast field cycling instrument and a high-field instrument. The frequency dependence of relaxation rate shows a transition from a diffusion-limited to a surface-limited relaxation regime. Copyright © 2011 John Wiley & Sons, Ltd.

Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

International Nuclear Information System (INIS)

Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye

2013-01-01

It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses

Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

Energy Technology Data Exchange (ETDEWEB)

Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses.

Development of thermocouple re-instrumentation technique for irradiated fuel rod. Techniques for making center hole into UO2 pellets and thermocouple re-instrumentation to fuel rod

International Nuclear Information System (INIS)

Shimizu, Michio; Saito, Junichi; Oshima, Kunio

1995-07-01

The information on FP gas pressure and centerline temperature of fuel pellets during power transient is important to study the pellet clad interaction (PCI) mechanism of high burnup LWR fuel rods. At the Department of JMTR, a re-instrumentation technique of FP gas pressure gage for an irradiated fuel rod was developed in 1990. Furthermore, a thermocouple re-instrumentation technique was successfully developed in 1994. Two steps were taken to carry out the development program of the thermocouple re-instrumentation technique. In the first step, a drilling technique was developed for making a center hole of the irradiated fuel pellets. Various drilling tests were carried out using dummy of fuel rods consisted of Ba 2 FeO 3 pellets and Zry-2 cladding. On this work it is important to keep the pellets just the state cracked at a power reactor. In these tests, the technique to fix the pellets by frozen CO 2 was used during the drilling work. Also, diamond drills were used to make the center hole. These tests were completed successfully. A center hole, 54mm depth and 2.5mm diameter, was realized by these methods. The second step of this program is the in-pile demonstration test on an irradiated fuel rod instrumented dually a thermocouple and FP gas pressure gage. The demonstration test was carried out at the JMTR in 1995. (author)

Research reactor fuel transport in the U.K

Energy Technology Data Exchange (ETDEWEB)

Panter, R [U.K. Atomic Energy Authority, Harwell (United Kingdom)

1983-09-01

This paper describes the containers currently used for transport of fresh or spent fuel elements for Research and Materials Test Reactors in the U.K., their status, operating procedures and some of the practical difficulties. In the U.K., MTR fuel cycle work is almost entirely the responsibility of the U.K. Atomic Energy Authority.

High Performance Capsule Implosions on the Omega Laser Facility with Rugby Hohlraums

Science.gov (United States)

Robey, Harry F.

2009-11-01

Rugby-shaped hohlraums have been proposed as a method for x-ray drive enhancement for indirectly-driven capsule implosions [1]. This concept has recently been tested in a series of shots on the OMEGA laser facility at the Laboratory for Laser Energetics at the University of Rochester. In this talk, experimental results are presented comparing the performance of D2-filled capsules between standard cylindrical Au hohlraums and rugby-shaped hohlraums. Not only did the rugby hohlraums demonstrate 18% more x-ray drive energy as compared with the cylinders, but the high-performance design of these implosions (both cylinder and rugby) also provided 20X more DD neutrons than any previous indirectly-driven campaign on Omega (and 3X more than ever achieved on Nova implosions driven with nearly twice the laser energy). This increase in performance enables, for the first time, a measurement of the neutron burn history of an indirectly-driven implosion. Previous DD neutron yields had been too low to register this key measurement of capsule performance and the effects of dynamic mix. A wealth of additional data on the fuel areal density from the suite of charged particle diagnostics was obtained on a subset of the shots that used D^3He rather than D2 fuel. Comparisons of the experimental results with numerical simulations are shown to be in excellent agreement. The design techniques employed in this campaign, e.g., smaller NIF-like laser entrance holes and hohlraum case-to-capsule ratios, provide added confidence in the pursuit of ignition on the National Ignition Facility. [4pt] [1] P. Amendt, C. Cerjan, D. E. Hinkel, J. L. Milovich, H.-S. Park, and H. F. Robey, ``Rugby-like hohlraum experimental designs for demonstrating x-ray drive enhancement'', Phys. Plasmas 15, 012702 (2008).

40 CFR 600.206-08 - Calculation and use of FTP-based and HFET-based fuel economy values for vehicle configurations.

Science.gov (United States)

2010-07-01

... EMISSIONS OF MOTOR VEHICLES Fuel Economy Regulations for 1977 and Later Model Year Automobiles-Procedures... economy value exists for an electric vehicle configuration, all values for that vehicle configuration are... HFET-based fuel economy values for vehicle configurations. 600.206-08 Section 600.206-08 Protection of...

Dielectric and magnetic properties of Ba-, La- and Pb-doped Bi0.8Gd0.1M0.1Fe0.9Ti0.1O3 perovskite ceramics

Directory of Open Access Journals (Sweden)

Radheshyam Rai

2014-04-01

Full Text Available The multiferroic Bi0.8Gd0.1M0.1Fe0.9Ti0.1O3, (where M = Ba (DB, La (DL and Pb (DP has been synthesized by using solid-state reaction technique. Effects of Ba, La and Pb substitution on the structure, electrical and ferroelectric properties of Bi0.8Gd0.1M0.1Fe0.9Ti0.1O3 samples have been studied by performing X-ray diffraction, dielectric and magnetic measurements. The crystal structures of the ceramic samples have a tetragonal phase. The vibrating sample magnetometer (VSM measurement shows a significant change in the magnetic properties of Ba-doped Bi0.8Gd0.1M0.1Fe0.9Ti0.1O3 as compared to La- and Pb-doped ceramics. It is seen that coercive field (HC and remanent magnetization (MR increases with Ba-doped ceramics but decreases for La- and Pb-doped ceramics.

fK /f{pi} in Full QCD with Domain Wall Valence Quarks

Energy Technology Data Exchange (ETDEWEB)

Silas Beane; Paulo Bedaque; Konstantinos Orginos; Martin Savage

2007-05-01

We compute the ratio of pseudoscalar decay constants f{sub K}/f{sub {pi}} using domain-wall valence quarks and rooted improved Kogut-Susskind sea quarks. By employing continuum chiral perturbation theory, we extract the Gasser-Leutwyler low-energy constant L{sub 5}, and extrapolate f{sub K}/f{sub {pi}} to the physical point. We find: f{sub K}/f{sub {pi}} = 1.218 {+-} 0.002{sub -0.024}{sup +0.011} where the first error is statistical and the second error is an estimate of the systematic due to chiral extrapolation and fitting procedures. This value agrees within the uncertainties with the determination by the MILC collaboration, calculated using Kogut-Susskind valence quarks, indicating that systematic errors arising from the choice of lattice valence quark are small.

Monitoring instrumentation spent fuel management program. Final report

International Nuclear Information System (INIS)

1979-01-01

Preliminary monitoring system methodologies are identified as an input to the risk assessment of spent fuel management. Conceptual approaches to instrumentation for surveillance of canister position and orientation, vault deformation, spent fuel dissolution, temperature, and health physics conditions are presented. In future studies, the resolution, reliability, and uncertainty associated with these monitoring system methodologies will be evaluated

Production of 131I gelatin capsules

International Nuclear Information System (INIS)

Freud, A.; Hirshfeld, N.; Canfi, A.; Melamud, Y.

1997-01-01

Radioiodine ( 131 I) hard-gelatin capsules are widely used for the diagnosis and treatment of various thyroid disorders. Until 1980 radioiodine was supplied by us as a liquid dosage. This proved to be a rather inconvenient form since it resulted in inaccurate dosing by the physicians and caused frequent contamination of the patients and the hospital personnel. In an attempt to overcome these problems we have designed and constructed a production facility for capsules in which 1311 is packaged. Because of the extreme precautions necessary in handling radioactive compounds, encapsulation of radioactive materials requires specifically designed production techniques, special instrumentation and unique quality control procedures that are not encountered in the standard capsule production processes in the pharmaceutical industry

FUEL CONSUMPTION EFFECT OF COMMERCIAL TURBOFANS ON GLOBAL WARMING

Energy Technology Data Exchange (ETDEWEB)

Onder Turan; T. Hikmet Karakoc [School of Civil Aviation, Anadolu University, Eskisehir (Turkey)

2008-09-30

The main objective pursued in this study is to parametrically investigate the fuel consumption effect of commercial turbofans on global warming. In this regard, Of the important parameters, specific fuel consumption of a commercial turbofans is taken into consideration. In order to minimize the effect of fuel consumption on global warming, the values of engine design parameters are optimized for maintaining minimum specific fuel consumption of high bypass turbofan engine under different flight conditions and design criteria. The backbones of optimization approach consisted of elitism-based genetic algorithm coupled with real parametric cycle analysis of a turbofan engine. For solving optimization problem a new software program is developed in MATLAB, while objective function is determined for minimizing the specific fuel consumption by considering the following parameters such as the fan pressure ratio ({pi}{sub f}), bypass ratio ({alpha}) and the fuel heating value [h{sub PR}-(kJ/kg)]. Accordingly, it may be concluded that the software program developed can successfully solve optimization problems at 1.2{le}{pi}{sub f}{le}2, 2{le}{alpha}{le}10 and 23000{le}h{sub PR}{le}120000 with aircraft flight Mach number {le}0.8. Fuel types used in preliminary engine cycle analysis were JP-4, JP-5, JP-8 and hydrogen in this paper.

UO2-PuO2 fuel pin capsule-irradiations of the test series FR 2-5a

International Nuclear Information System (INIS)

Dienst, W.; Goetzmann, O.; Schulz, B.

1975-06-01

In the capsule-irradiation test series FR 2-5a, short UO 2 -PuO 2 fuel pins (80 mm fuel length) of 7 mm diameter were irradiated in a thermal neutron flux at mean rod powers of 400 - 450 W/cm and mean cladding surface temperatures of 500 - 550 0 C to burnups of 0.6, 1.8 and 5.0 at% (U + Pu). Void volume redistribution in the fuel pins was examined in micrographs of cross-sections by measuring crack widths, central void diameters, and fuel porosity. The width of the radial cracks at the outer fuel rim was taken as a basis for measuring the irradiation-induced densification of the UO 2 -PuO 2 fuel. The result was that the final fuel density after irradiation-induced densification amounted to 92 - 94% TD and had already been reached after 0.6 at% burnup. The porosity measurement on fuel cross-sections was to show a possible dependence of the radial porosity redistribution on the initial sintered density. Examining the fuel pin diameters after irradiation showed permanent cladding strains after 5 at% burnup, which must be due to mechanical interaction with the fuel. To judge if the chemical compatibility between the fuel and the cladding of Cr-Ni-stainless steel 1.4988, the depths of chemical attack on the cladding inside was measured by micrographs of fuel pin cross-sections. (orig./GSC) [de

High temperature radioisotope capsule

International Nuclear Information System (INIS)

Bradshaw, G.B.

1976-01-01

A high temperature radioisotope capsule made up of three concentric cylinders, with the isotope fuel located within the innermost cylinder is described. The innermost cylinder has hemispherical ends and is constructed of a tantalum alloy. The intermediate cylinder is made of a molybdenum alloy and is capable of withstanding the pressure generated by the alpha particle decay of the fuel. The outer cylinder is made of a platinum alloy of high resistance to corrosion. A gas separates the innermost cylinder from the intermediate cylinder and the intermediate cylinder from the outer cylinder

K Basin spent nuclear fuel characterization

Energy Technology Data Exchange (ETDEWEB)

LAWRENCE, L.A.

1999-02-10

The results of the characterization efforts completed for the N Reactor fuel stored in the Hanford K Basins were Collected and summarized in this single referencable document. This summary provides a ''road map'' for what was done and the results obtained for the fuel characterization program initiated in 1994 and scheduled for completion in 1999 with the fuel oxidation rate measurement under moist inert atmospheres.

Compact Design of 10 kW Proton Exchange Membrane Fuel Cell Stack Systems with Microcontroller Units

Directory of Open Access Journals (Sweden)

Hsiaokang Ma

2014-04-01

Full Text Available In this study, fuel, oxidant supply and cooling systems with microcontroller units (MCU are developed in a compact design to fit two 5 kW proton exchange membrane fuel cell (PEMFC stacks. At the initial stage, the testing facility of the system has a large volume (2.0 m × 2.0 m × 1.5 m with a longer pipeline and excessive control sensors for safe testing. After recognizing the performance and stability of stack, the system is redesigned to fit in a limited space (0.4 m × 0.5 m × 0.8 m. Furthermore, the stack performance is studied under different hydrogen recycling modes. Then, two similar 5 kW stacks are directly coupled with diodes to obtain a higher power output and safe operation. The result shows that the efficiency of the 5 kW stack is 43.46% with a purge period of 2 min with hydrogen recycling and that the hydrogen utilization rate µf is 66.31%. In addition, the maximum power output of the twin-coupled module (a power module with two stacks in electrical cascade/parallel arrangement is 9.52 kW.

Dissolved inorganic carbon, alkalinity, temperature, salinity and other variables collected from discrete sample and profile observations using Alkalinity titrator, CTD and other instruments from the PELAGIA in the North Atlantic Ocean from 2007-08-30 to 2007-09-27 (NODC Accession 0110258)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — NODC Accession 0110258 includes chemical, discrete sample, physical and profile data collected from PELAGIA in the North Atlantic Ocean from 2007-08-30 to 2007-09-27...

Developments in the LASL Fuel Pin Imaging System: PINEX-3A

International Nuclear Information System (INIS)

Lumpkin, A.H.; Berzins, G.J.; Cosimi, R.A.; O'Hare, T.E.; Davidson, J.R.

1979-01-01

The LASL Fuel Pin Imaging System was evaluated using a series of 10 TREAT transients, each of approx. 240-MW peak power. HEDL provided the fuel-ejection type capsule with annular fuel pellets. The pin visibility threshold was determined to be approx. 20-MW of TREAT power (approx. 130 W/g), almost an order of magnitude improvement over our PINEX-2 threshold. The impact of changes in instrumentation, imaging apertures, and fluors that produced the improved sensitivity are reported. Results of a time-integrated imaging technique are also presented

Isolation and mutagenesis of a capsule-like complex (CLC from Francisella tularensis, and contribution of the CLC to F. tularensis virulence in mice.

Directory of Open Access Journals (Sweden)

Aloka B Bandara

Full Text Available BACKGROUND: Francisella tularensis is a category-A select agent and is responsible for tularemia in humans and animals. The surface components of F. tularensis that contribute to virulence are not well characterized. An electron-dense capsule has been postulated to be present around F. tularensis based primarily on electron microscopy, but this specific antigen has not been isolated or characterized. METHODS AND FINDINGS: A capsule-like complex (CLC was effectively extracted from the cell surface of an F. tularensis live vaccine strain (LVS lacking O-antigen with 0.5% phenol after 10 passages in defined medium broth and growth on defined medium agar for 5 days at 32°C in 7% CO₂. The large molecular size CLC was extracted by enzyme digestion, ethanol precipitation, and ultracentrifugation, and consisted of glucose, galactose, mannose, and Proteinase K-resistant protein. Quantitative reverse transcriptase PCR showed that expression of genes in a putative polysaccharide locus in the LVS genome (FTL_1432 through FTL_1421 was upregulated when CLC expression was enhanced. Open reading frames FTL_1423 and FLT_1422, which have homology to genes encoding for glycosyl transferases, were deleted by allelic exchange, and the resulting mutant after passage in broth (LVSΔ1423/1422_P10 lacked most or all of the CLC, as determined by electron microscopy, and CLC isolation and analysis. Complementation of LVSΔ1423/1422 and subsequent passage in broth restored CLC expression. LVSΔ1423/1422_P10 was attenuated in BALB/c mice inoculated intranasally (IN and intraperitoneally with greater than 80 times and 270 times the LVS LD₅₀, respectively. Following immunization, mice challenged IN with over 700 times the LD₅₀ of LVS remained healthy and asymptomatic. CONCLUSIONS: Our results indicated that the CLC may be a glycoprotein, FTL_1422 and -FTL_1423 were involved in CLC biosynthesis, the CLC contributed to the virulence of F. tularensis LVS, and a CLC

Postirradiation examination of capsule GF-4

International Nuclear Information System (INIS)

Kovacs, W.J.; Sedlak, B.J.

1980-10-01

The GF-4 capsule test was irradiated in the SILOE reactor at Grenoble, France between April 8, 1975 and July 26, 1976. High-enriched uranium (HEU) UC 2 and weak acid resin (WAR) UC/sub x/O/sub y/ fissile and ThO 2 fertile particles were tested. Postirradiation examination of cured-in-place fuel rods showed no fuel rod/graphite element interaction. In addition, all rods exhibited adequate structural integrity. Irradiation-induced dimensional changes for rods containing all TRISO-coated fuel were consistent with model predictions; however, rods containing BISO-coated fuel exhibited greater volumetric contractions than predicted

Design and fabrication report on capsule (11M 19K for out of pile test) for irradiation testing of research reactor materials at HANARO

Energy Technology Data Exchange (ETDEWEB)

Kim, B.G.; Yang, S.W.; Park, S.J.; Shim, K.T.; Choo, K.N.; Oh, J.M.; Lee, B.C.; Choi, M.H.; Kim, D.J.; Kim, J.M.; Kang, S.H.; Chun, Y.B.; Kim, T.K.; Jeong, Y.H.

2012-05-15

As a part of the research reactor development project with a plate type fuel, the irradiation tests of graphite (Gr), beryllium (Be), and zircaloy 4 materials using the capsule have been investigating to obtain the mechanical characteristics such as an irradiation growth, hardness, swelling and tensile strength at the temperature below 100 .deg. C and the 30 MW reactor power. Then, A capsule to be able to irradiate materials(graphite, Be, zircaloy 4) under 100 .deg. C at the HANARO was designed and fabricated. After performing out of pile testing in single channel test loop by using the capsule, the final design of the capsules to be irradiated in CT and IR2 test hole of HANARO was approved, and 2 sets of capsule were fabricated. These capsules will be loaded in CT and IR2 test hole of HANARO, and be started the irradiation from the end of June, 2012. After performing the irradiation testing of 2 sets of capsule, PIE (Post Irradiation Examination) on irradiated specimens (Gr, Be, and zircaloy 4) will be carry out in IMEF (Irradiated Material Examination Facility). So, the irradiation testing will be contributed to obtain the characteristic data induced neutron irradiation on Gr, Be, and zircaloy 4. And then, it is convinced that these data will be also contributed to obtain the license for JRTR (Jordan Research and Training Reactor) and new research reactor in Korea, and export research reactors.

Test plan for surface and subsurface examinations of K-east and K-west fuel elements

International Nuclear Information System (INIS)

Pitner, A.L.

1997-01-01

The test plan for subsurface examinations on damaged K East and K West Basin fuel elements is presented. The purpose of these examinations is to inspect damaged areas on the fuel elements for the presence of voids, sludge, or broken fuel, and to obtain samples from the damaged areas for subsequent characterization tests

Uncertainty Quantification of Calculated Temperatures for the U.S. Capsules in the AGR-2 Experiment

Energy Technology Data Exchange (ETDEWEB)

Lybeck, Nancy [Idaho National Lab. (INL), Idaho Falls, ID (United States); Einerson, Jeffrey J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pham, Binh T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hawkes, Grant L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-03-01

A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN-3636). The AGR-2 test was inserted in the B-12 position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in June 2010 and successfully completed irradiation in October 2013, resulting in irradiation of the TRISO fuel for 559.2 effective full power days (EFPDs) during approximately 3.3 calendar years. The AGR-2 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS) (Pham and Einerson 2014). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as-run thermal analysis has been performed separately on each of four AGR-2 U.S. capsules for the entire irradiation as discussed in (Hawkes 2014). The ABAQUS code’s finite element-based thermal model predicts the daily average volume-average fuel temperature and peak fuel temperature in each capsule. This thermal model involves complex physical mechanisms (e.g., graphite holder and fuel compact shrinkage) and properties (e.g., conductivity and density). Therefore, the thermal model predictions are affected by uncertainty in input parameters and by incomplete knowledge of the underlying physics leading to modeling assumptions. Therefore, alongside with the deterministic predictions from a set of input thermal conditions, information about prediction uncertainty is instrumental for the ART

Compact 4-kHz XeF-laser with multisectional discharge gap

Science.gov (United States)

Andramanov, A. V.; Kabaev, S. A.; Lazhintsev, Boris V.; Nor-Arevyan, Vladimir A.; Selemir, V. D.

2005-03-01

An electric-discharge XeF-laser with a pulse repetition rate up to 4 kHz was developed. The laser electrode unit was made on the basis of plate-like electrodes with inductive-capacity discharge stabilization. The narrow discharge width laser energy was 3 mJ by using He/Xe/NF3 and Ne/Xe/NF3 mixtures at the total pressure of 0.8 atm and 1.2 atm, respectively. The maximum laser efficiency was ~ 0.73% The gas flow was formed with the help of a diametrical fan rotated by the direct-current motor with 80 W power. The gas velocity of 20 m/s in the interelectrode gap was achieved. The laser pulse energy for a pulse repetition rate up to 3.5...4 kHz was virtually equal to the laser pulse energy in the infrequently-repeating-pulse regime. The average output power of 12 W at the pulse repetition rate of 4 kHz was achieved. The relative root-mean-square pulse-to-pulse variation of the output energy σ = 2.5% was reached.

Interrelation of transport properties, defect structure and spin state of Ni3+ in La1.2Sr0.8Ni0.9Fe0.1O4+δ

Science.gov (United States)

Gilev, A. R.; Kiselev, E. A.; Zakharov, D. M.; Cherepanov, V. A.

2017-10-01

The total conductivity, Seebeck coefficient and oxygen non-stoichiometry for La1.2Sr0.8Ni0.9Fe0.1O4+δ have been measured vs temperature and oxygen partial pressure P(O2). The measurements were carried out at 800, 850, 900 and 950 °C within the P(O2) range of 10-5-0.21 atm. La1.2Sr0.8Ni0.9Fe0.1O4+δ was shown to be oxygen deficient in all temperature and P(O2) ranges studied. The calculated values of the partial molar enthalpy of oxygen depend very slightly on oxygen content (δ), indicating that La1.2Sr0.8Ni0.9Fe0.1O4+δ with the oxygen deficiency can be considered an ideal solution. The model of point defect equilibria in La1.2Sr0.8Ni0.9Fe0.1O4+δ has been proposed and fitted to experimental dependencies. Subsequent joint analysis of the defect structure and transport properties revealed that electron holes can coexist in both localized and quasi-delocalized states in the oxide: the former corresponded to high-spin state Ni3+ and the latter - to low-spin state Ni3+. The mobilities of localized electron holes were shown to be significantly lower in comparison to quasi-delocalized ones. The behavior of localized electron holes was explained in terms of a small polaron conduction mechanism; in contrast, quasi-delocalized electron holes were described in terms of a band conduction approach. The small polaron conduction mechanism was shown to be predominant in the Sr- and Fe-co-doped lanthanum nickelate.

Peak effect and superconducting properties of SmFeAsO{sub 0.8}F{sub 0.2} wires

Energy Technology Data Exchange (ETDEWEB)

Chen, Y L; Cui, Y J; Yang, Y; Zhang, Y; Wang, L; Zhao, Y [Key Laboratory of Magnetic Levitation Technologies and Maglev Trains, Ministry of Education of China, and Superconductivity R and D Center (SRDC), Southwest Jiaotong University, Chengdu, Sichuan 610031 (China); Cheng, C H; Sorrell, C [School of Materials Science and Engineering, University of New South Wales, Sydney, NSW 2052 (Australia)], E-mail: yzhao@swjtu.edu.cn

2008-11-15

Ta-sheathed SmFeAsO{sub 0.8}F{sub 0.2} superconducting wires with T{sub c} = 52.5 K have been fabricated using the powder-in-tube (PIT) method and the superconducting properties of the wires have been investigated. The wires exhibit a very large intragrain critical current density at a temperature below 30 K. A peak effect with maximal J{sub c} = 0.6 MA cm{sup -2} at 10 K under 6 T field was observed. The peak field H{sub pear} is strongly temperature-dependent. A severe weak-link effect depresses the development of global supercurrent owing to a very short coherence length. The wires also show a power law temperature dependence for the irreversibility line with H{sub irr}{approx_equal}(1-T/T{sub c}){sup 1.5}. The H-T phase diagram was found to be similar to that of other superconducting cuprates.

Hanford K basins spent nuclear fuel project update

International Nuclear Information System (INIS)

Williams, N.H.; Hudson, F.G.

1997-07-01

Twenty one hundred metric tons of spent nuclear fuel (SNF) are currently stored in the Hanford Site K Basins near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported to the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building

Estimation of the development possibility of the ABC/ATW fuel cycle based on LiF-BeF2 fuel salt. Part 2

International Nuclear Information System (INIS)

Bychkov, A.V.; Naumov, V.S.

1994-01-01

The aim of the first chapter was generalization of data on solubility and equilibrium states of fission product and actinide fluorides in fluoride salt melts-solvents and fuel composition melts based on LiF-BeF 2 mixture which was proposed as fuel basis for ABC/ATW facility. The second chapter is devoted to description of processes proposed for the chemical-technological complex of the ABC/ATW facility and their physico-chemical peculiarities. The complex is responsible for the removal of fission products and actinides from irradiated fuel salt

Capsule Typing of Haemophilus influenzae by Matrix-Assisted Laser Desorption/Ionization Time-of-Flight Mass Spectrometry.

Science.gov (United States)

Månsson, Viktor; Gilsdorf, Janet R; Kahlmeter, Gunnar; Kilian, Mogens; Kroll, J Simon; Riesbeck, Kristian; Resman, Fredrik

2018-03-01

Encapsulated Haemophilus influenzae strains belong to type-specific genetic lineages. Reliable capsule typing requires PCR, but a more efficient method would be useful. We evaluated capsule typing by using matrix-assisted laser desorption/ionization time-of-flight (MALDI-TOF) mass spectrometry. Isolates of all capsule types (a-f and nontypeable; n = 258) and isogenic capsule transformants (types a-d) were investigated. Principal component and biomarker analyses of mass spectra showed clustering, and mass peaks correlated with capsule type-specific genetic lineages. We used 31 selected isolates to construct a capsule typing database. Validation with the remaining isolates (n = 227) showed 100% sensitivity and 92.2% specificity for encapsulated strains (a-f; n = 61). Blinded validation of a supplemented database (n = 50) using clinical isolates (n = 126) showed 100% sensitivity and 100% specificity for encapsulated strains (b, e, and f; n = 28). MALDI-TOF mass spectrometry is an accurate method for capsule typing of H. influenzae.

Preparation and performance of intermediate-temperature fuel cells based on Gd-doped ceria electrolytes with different compositions

International Nuclear Information System (INIS)

Li, Zhimin; Mori, Toshiyuki; Yan, Pengfei; Wu, Yuanyuan; Li, ZhiPeng

2012-01-01

Highlights: ► Gd 0.1 Ce 0.9 O 1.95 electrolyte had less density of oxygen vacancies ordering. ► Gd 0.2 Ce 0.8 O 1.9 fuel cell showed better performance than Gd 0.1 Ce 0.9 O 1.95 . ► The relationship between microstructures and performance for cells were discussed. ► Gd 0.2 Ce 0.8 O 1.9 electrolyte with higher grain boundary conductivity was concluded. - Abstract: In this work, the effect of two frequently used Gd x Ce 1−x O 2−x/2 electrolytes (x = 0.1 and x = 0.2) on the performance of fuel cells operated at intermediate temperature was studied. The microstructures of ceria electrolytes responsible for the performance were discussed. Electrochemical measurements of as-prepared cells showed that the cell with Gd 0.2 Ce 0.8 O 1.9 electrolyte had a better performance than that of Gd 0.1 Ce 0.9 O 1.95 . It can be concluded that the increase of grain boundary conductivity of Gd 0.2 Ce 0.8 O 1.9 electrolyte contributes to its better cell performance.

Fuel taxes: An important instrument for climate policy

International Nuclear Information System (INIS)

Sterner, Thomas

2007-01-01

This article shows that fuel taxes serve a very important role for the environment and that we risk a backlash of increased emissions if they are abolished. Fuel taxes have restrained growth in fuel demand and associated carbon emissions. Although fuel demand is large and growing, our analysis shows that it would have been much higher in the absence of domestic fuel taxes. People often assert that fuel demand is inelastic but there is strong research evidence showing the opposite. The price elasticity is in fact quite high but only in the long-run: in the short run it may be quite inelastic which has important implications for policy makers. Had Europe not followed a policy of high fuel taxation but had low US taxes, then fuel demand would have been twice as large. Hypothetical transport demand in the whole OECD area is calculated for various tax scenarios and the results show that fuel taxes are the single most powerful climate policy instrument implemented to date-yet this fact is not usually given due attention in the debate

Postirradiation examination of capsules GF-1, GF-2, and GF-3

International Nuclear Information System (INIS)

Kovacs, W.J.; Blanchard, R.; Pointud, M.L.

1980-09-01

The GF-1, GF-2, and GF-3 capsule tests were irradiated in the Siloe reactor at Grenoble, France, between October 31, 1973, and July 25, 1975. High-enriched uranium (HEU) mixed oxide (8Th,U)O 2 fissile and ThO 2 fertile particles were tested over the following exposure conditions: 3.8 to 11.0 x 10 25 n/m 2 (E > 29 fJ)/sub HTGR/; 960 0 to 1120 0 C time volume average temperature; and mixed oxide (8Th,U)O 2 burnup between 5.3 and 11.4% FIMA and ThO 2 burnup between 1.1 and 3.6% FIMA. Postirradiation examination of HTGR fuel rods in capsules GF-1, GF-2, and GF-3 showed acceptable structural integrity and irradiation-induced dimensional changes that were consistent with model predictions. Pressure vessel failure levels between different TRISO-coated (8Th,U)O 2 fissile particle types showed that the 400-μm-diameter kernel design was more conservative than the 500-μm-diameter designs

Postirradiation thermal analysis of capsule P13T

International Nuclear Information System (INIS)

Ketterer, J.W.

1978-12-01

In determining fuel rod temperature histories for the P13T capsule, a technique which combined measured temperature and dimensional data, TAC-2D computer modeling, and a calculational procedure was employed. TAC-2D models were constructed for each of the capsule's four fuel bodies and temperature matching runs were made at five time points of the irradiation history. The agreement between TAC-calculated and measured temperatures was good; at all times the TAC-calculated temperatures were within 20 0 C of the Chromel-Alumel (C/A) measurements and 40 0 C of the corrected tungsten-rhenium (W/Re) temperatures. Thermocouple decalibration was treated in detail and corrected temperatures for all W/Re thermocouples were calculated over the irradiation period

Status for development of a capsule and instruments for high-temperature irradiation in HANARO

Energy Technology Data Exchange (ETDEWEB)

Cho, Man Soon; Choo, Kee Nam; Lee, Chul Yong; Yang, Seong Woo; Shim, Kyue Taek; Chung, Hwan-Sung [Korea Atomic Energy Research Institute, Taejeon (Korea, Republic of)

2012-03-15

As the reactors planned in the Gen-IV program will be operated at high temperature and under high neutron flux, the requirements for irradiation of materials at high temperature are recently being gradually increased. The irradiation tests of materials in HANARO up to the present have been performed usually at temperatures below 300degC at which the RPV materials of the commercial reactors are being operated. To overcome the restriction for high-temperature use of Al thermal media of the existing standard capsule, a new capsule with double thermal media composed of two kinds of materials such as Al-Ti and Al-graphite was designed and fabricated as a more advanced capsule than the single thermal media capsule. (author)

Introduction to the physics of ICF capsules

International Nuclear Information System (INIS)

Lindl, J.D.

1989-01-01

Inertial Confinement Fusion is an approach to fusion which relies on the inertia of the fuel mass to provide confinement. To achieve conditions under which this confinement is sufficient for efficient thermonuclear burn, high gain ICF targets designed to be imploded directly by laser light. These capsules are generally a spherical shell which is filled with low density DT gas. The shell is composed of an outer region which forms the ablator and an inner region of frozen or liquid DT which forms the main fuel. Energy from the driver is delivered to the ablator which heats up and expands. As the ablator expands and blows outward, the rest of the shell is forced inward to conserve momentum. In this implosion process, several features are important. We define the in-flight-aspect-ratio (IFAR) as the ratio of the shell radius R as it implodes to its thickness ΔR. Hydrodynamic instabilities during the implosion impose limits on this ratio which results in a minimum pressure requirement of about 100 Mbar. The convergence ratio is defined as the ratio of the initial outer radius of the ablator to the final compressed radius of the hot spot. This hot spot is the central region of the compressed fuel which is required to ignite the main fuel in high gain designs. Typical convergence ratios are 30--40. To maintain a nearly spherical shape during the implosion, when convergence ratios are this large, the flux delivered to the capsule must be uniform to a few percent. The remainder of this paper discusses the conditions necessary to achieve thermonuclear ignition in these ICF capsules

ORR irradiation experiment OF-1: accelerated testing of HTGR fuel

International Nuclear Information System (INIS)

Tiegs, T.N.; Long, E.L. Jr.; Kania, M.J.; Thoms, K.R.; Allen, E.J.

1977-08-01

The OF-1 capsule, the first in a series of High-Temperature Gas-Cooled Reactor fuel irradiations in the Oak Ridge Research Reactor, was irradiated for more than 9300 hr at full reactor power (30 MW). Peak fluences of 1.08 x 10 22 neutrons/cm 2 (> 0.18 MeV) were achieved. General Atomic Company's magazine P13Q occupied the upper two-thirds of the test space and the ORNL magazine OF-1 the lower one-third. The ORNL portion tested various HTGR recycle particles and fuel bonding matrices at accelerated flux levels under reference HTGR irradiation conditions of temperature, temperature gradient, and fast fluence exposure

X-ray drive of beryllium capsule implosions at the National Ignition Facility

International Nuclear Information System (INIS)

Wilson, D C; Yi, S A; Simakov, A N; Kline, J L; Kyrala, G A; Olson, R E; Zylstra, A B; Dewald, E L; Tommasini, R; Ralph, J E; Strozzi, D J; Celliers, P M; Schneider, M B; MacPhee, A G; Callahan, D A; Hurricane, O A; Milovich, J L; Hinkel, D E; Rygg, J R; Rinderknecht, H G

2016-01-01

National Ignition Facility experiments with beryllium capsules have followed a path begun with “high-foot” plastic capsule implosions. Three shock timing keyhole targets, one symmetry capsule, a streaked backlit capsule, and a 2D backlit capsule were fielded before the DT layered shot. After backscatter subtraction, laser drive degradation is needed to match observed X-ray drives. VISAR measurements determined drive degradation for the picket, trough, and second pulse. Time dependence of the total Dante flux reflects degradation of the of the third laser pulse. The same drive degradation that matches Dante data for three beryllium shots matches Dante and bangtimes for plastic shots N130501 and N130812. In the picket of both Be and CH hohlraums, calculations over-estimate the x-ray flux > 1.8 keV by ∼100X, while calculating the total flux correctly. In beryllium calculations these X-rays cause an early expansion of the beryllium/fuel interface at ∼3 km/s. VISAR measurements gave only ∼0.3 km/s. The X-ray drive on the Be DT capsule was further degraded by an unplanned decrease of 9% in the total picket flux. This small change caused the fuel adiabat to rise from 1.8 to 2.3. The first NIF beryllium DT implosion achieved 29% of calculated yield, compared to CH capsules with 68% and 21%. (paper)

Plan for characterization of K Basin spent nuclear fuel and sludge

International Nuclear Information System (INIS)

Lawrence, L.A.; Marschman, S.C.

1995-06-01

This plan outlines a characterization program that supports the accelerated Path Forward scope and schedules for the Spent Nuclear Fuel stored in the Hanford K Basins. This plan is driven by the schedule to begin fuel transfer by December 1997. The program is structured for 4 years and is limited to in-situ and laboratory examinations of the spent nuclear fuel and sludge in the K East and K West Basins. The program provides bounding behavior of the fuel, and verification and acceptability for three different sludge disposal pathways. Fuel examinations are based on two shipping campaigns for the K West Basin and one from the K East Basin. Laboratory examinations include physical condition, hydride and oxide content, conditioning testing, and dry storage behavior

CERCA's fuel elements instrumentation manufacturing

International Nuclear Information System (INIS)

Harbonnier, G.; Jarousse, C.; Pin, T.; Febvre, M.; Colomb, P.

2005-01-01

When research and test reactors wish to further understand the Fuel Elements behavior when operating as well as mastering their irradiation conditions, operators carry out neutron and thermo hydraulic analysis. For thermal calculation, the codes used have to be preliminary validated, at least in the range of the reactor safety operational limits. When some further investigations are requested either by safety authorities or for its own reactor needs, instrumented tools are the ultimate solution for providing representative measurements. Such measurements can be conducted for validating thermal calculation codes, at nominal operating condition as well as during transients ones, or for providing numerous and useful data in the frame of a new products qualification program. CERCA, with many years of experience for implanting thermocouples in various products design, states in this poster his manufacturing background on instrumented elements, plates or targets. (author)

Are PSR 0656+14, PSR 0950+08, and PSR 1822-09 gamma ray pulsars?

Science.gov (United States)

Brown, Lawrence E.; Hartmann, Dieter H.

1993-01-01

The possible discovery of three new gamma-ray pulsars PSR 0656+14, PSR 0950+08, and PSR 1822-09 (Ma, Lu, Yu, and Young, 1993) in data obtained with the COS-B experiment is reinvestigated using a refined technique for pulsar light curve analysis. The results of this study do not confirm the previously claimed gamma-ray pulsar nature of any of these pulsars. Even when using the standard epoch folding technique in conjunction with energy-dependent acceptance cones, we do not detect pulsed gamma-ray emission from these sources. We suspect that insufficient position accuracy is the cause for the discrepancy between our results and those of Ma et al. (1993). We do not rule out that any one of the three candidates, or all of them, is in fact a gamma-ray pulsar, but their spin properties must differ from those derived by Ma et al. (1993). More work is needed to determine the correct high-energy properties of these three sources.

Kinetics of the reaction of F atoms with O₂ and UV spectrum of FO₂ radicals in the gas phase at 295 K

DEFF Research Database (Denmark)

Ellermann, T.; Sehested, J.; Nielsen, O.J.

1994-01-01

The ultraviolet absorption spectrum of FO2 radicals and the kinetics of the reaction of F atoms with O2 have been studied in the gas phase at 295 K using pulse radiolysis combined with kinetic UV spectroscopy. At 230 nm, sigma(FO2) = (5.08 +/- 0.70) X 10(-18) cm2 molecule-1. The kinetics of the r......The ultraviolet absorption spectrum of FO2 radicals and the kinetics of the reaction of F atoms with O2 have been studied in the gas phase at 295 K using pulse radiolysis combined with kinetic UV spectroscopy. At 230 nm, sigma(FO2) = (5.08 +/- 0.70) X 10(-18) cm2 molecule-1. The kinetics...

Applications and experience with a new instrumented fuel element

International Nuclear Information System (INIS)

Morris, F.M.

1972-01-01

Previously reported information to TRIGA Reactor Conference I concerning the development of a new concept in an instrumented fuel element is updated and expanded. The evaluation of these new instrumented elements is discussed and some areas of application to reactor behavior are described. Experiments concerning temperature and flux mapping under varying conditions are investigated and conclusions are given. (author)

Capsule Performance Optimization for the National Ignition Facility

Science.gov (United States)

Landen, Otto

2009-11-01

The overall goal of the capsule performance optimization campaign is to maximize the probability of ignition by experimentally correcting for likely residual uncertainties in the implosion and hohlraum physics used in our radiation-hydrodynamic computational models before proceeding to cryogenic-layered implosions and ignition attempts. This will be accomplished using a variety of targets that will set key laser, hohlraum and capsule parameters to maximize ignition capsule implosion velocity, while minimizing fuel adiabat, core shape asymmetry and ablator-fuel mix. The targets include high Z re-emission spheres setting foot symmetry through foot cone power balance [1], liquid Deuterium-filled ``keyhole'' targets setting shock speed and timing through the laser power profile [2], symmetry capsules setting peak cone power balance and hohlraum length [3], and streaked x-ray backlit imploding capsules setting ablator thickness [4]. We will show how results from successful tuning technique demonstration shots performed at the Omega facility under scaled hohlraum and capsule conditions relevant to the ignition design meet the required sensitivity and accuracy. We will also present estimates of all expected random and systematic uncertainties in setting the key ignition laser and target parameters due to residual measurement, calibration, cross-coupling, surrogacy, and scale-up errors, and show that these get reduced after a number of shots and iterations to meet an acceptable level of residual uncertainty. Finally, we will present results from upcoming tuning technique validation shots performed at NIF at near full-scale. Prepared by LLNL under Contract DE-AC52-07NA27344. [4pt] [1] E. Dewald, et. al. Rev. Sci. Instrum. 79 (2008) 10E903. [0pt] [2] T.R. Boehly, et. al., Phys. Plasmas 16 (2009) 056302. [0pt] [3] G. Kyrala, et. al., BAPS 53 (2008) 247. [0pt] [4] D. Hicks, et. al., BAPS 53 (2008) 2.

Instrumentation of cars for fuel economy. Final report

Energy Technology Data Exchange (ETDEWEB)

Morris, J E

1982-04-01

The development of an electronic system to control the air-fuel ratio (A/F) and ignition timing of an internal combustion engine to optimize fuel economy is described. Dynamometer and drive cycle testing of the system was performed. The results showed that a significant improvement in fuel economy can be achieved by a control system of the type developed. It is clear, however, that considerably more work needs to be done. One area mentioned is the need for more systematic fuel economy testing against speed and load as control parameters are varied for optimization, and a more economic air bypass system must be developed. (LCL)

Instruments for reducing the specific fuel consumption of cars

International Nuclear Information System (INIS)

Hammer, S.; Maibach, M.; Marti, P.

2001-01-01

This report for the Swiss Federal Office of Energy (SFOE) presents three possible courses of action that are to be taken to reduce the specific fuel consumption of private cars. The report first examines existing targets and the degree to which they have been met up to now, whereby the situation both in Switzerland and in the European Union is looked at. The report makes a suggestion for a future target scenario and elaborates three possible ways to met these targets: regulations on fuel-consumption, a bonus/malus system and tradable certificates. For each of the proposed instruments, the report examines implementation variants and discusses the means for their implementation. The report presents the best models for each of the implementation-variants on the basis of comparisons and the results of evaluations of their effects. For these chosen variants, the authors present comparisons of their effect on fuel consumption in graphical form and recommend tradable certificates as the best instrument

Fuel cracking in relation to fuel oxidation in support of an out-reactor instrumented defected fuel experiment

Energy Technology Data Exchange (ETDEWEB)

Quastel, A.; Thiriet, C. [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Lewis, B., E-mail: brent.lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada); Corcoran, E., E-mail: emily.corcoran@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

2014-07-01

An experimental program funded by the CANDU Owners Group (COG) is studying an out-reactor instrumented defected fuel experiment in Stern Laboratories (Hamilton, Ontario) with guidance from Atomic Energy of Canada Limited (AECL). The objective of this test is to provide experimental data for validation of a mechanistic fuel oxidation model. In this experiment a defected fuel element with UO{sub 2} pellets will be internally heated with an electrical heater element, causing the fuel to crack. By defecting the sheath in-situ the fuel will be exposed to light water coolant near normal reactor operating conditions (pressure 10 MPa and temperature 265-310{sup o}C) causing fuel oxidation, especially near the hotter regions of the fuel in the cracks. The fuel thermal conductivity will change, resulting in a change in the temperature distribution of the fuel element. This paper provides 2D r-θ plane strain solid mechanics models to simulate fuel thermal expansion, where conditions for fuel crack propagation are investigated with the thermal J integral to predict fuel crack stress intensity factors. Finally since fuel crack geometry can affect fuel oxidation this paper shows that the solid mechanics model with pre-set radial cracks can be coupled to a 2D r-θ fuel oxidation model. (author)

The evaluation of kefir pure culture starter: Liquid-core capsule entrapping microorganisms isolated from kefir grains.

Science.gov (United States)

Wang, Liang; Zhong, Hao; Liu, Keying; Guo, Aizhen; Qi, Xianghui; Cai, Meihong

2016-10-01

The main purpose of this study was to develop a pure culture starter for producing kefir. In order to accomplish starter recycling, yeasts (Kluyveromyces marxianus strain, Pichia kudriavzevii clone), lactic acid bacteria (Lactobacillus kefiri strain F4Aa, Lactobacillus kefiri strain NM131-7, Lactobacillus kefiri strain NM132-3, Lactobacillus kefiri strain NM180-3, respectively), and acetic acid bacteria (Acetobacter lovaniensis strain) were entrapped in liquid core capsules based on the distribution ratio in kefir grains. The microbiological, antimicrobial, and chemical properties of kefir made with capsules (M) and kefir grains (K) were measured and compared. According to the results of plate counts in different selective medium, the number of yeasts and bacteria in the liquid core capsules gradually increased and stabilized after eight fermentation cycles. The results of gas chromatography-mass spectrometry showed that almost all the aroma components existed in the two type of kefir, except the ethyl lactate. There was no significant difference in alcohol content, protein content, and fat content, except the acidity and sugar content. Water holding capacity of kefir K was higher than kefir M. There were 14 same free amino acids in kefir M and kefir K, and the content of most free amino acids was similar. In antimicrobial test, there was no significant difference in both kefirs. © The Author(s) 2016.

Hyperthermal K--TeF6 molecular beam scattering

International Nuclear Information System (INIS)

Wagner, A.F.; Young, C.E.; Pobo, L.G.; Wexler, S.

1982-01-01

Angular distributions of K + product ions from collisions of a beam of hyperthermal K atoms with a cross beam of thermal TeF 6 molecules were determined at 13.7 and 23.7 eV (lab). The angular yields of K atom products from the same system were too low to permit measurement of angular distributions. From the integrated yields, the K + ion/K atom branching ratio was determined to be greater than 10 3 . In addition to the extremely large branching ratio, the differential cross sections exhibited several other unusual characteristics: (a) the lack of small angle scattering, corresponding to virtual absence of covalent scattering, (b) two peaks in the differential cross section with an outer rainbow feature at very large scattering angles (approx.275 eV deg). The observations are unexpected from previous experimental and theoretical studies of electron transfer reactions and from the electronic and structural properties of TeF 6 and TeF - 6 . A simplified dynamics model based on formation of electronically excited TeF - 6 in the initial electron transfer, followed by inner crossings leading to formation of electronically and vibrationally unexcited TeF - 6 or dissociation to TeF - 5 and other ionic products, has been developed which accounts for the experimental results. The model suggests that the observed two peaks in the differential cross section are due to the production of TeF - 6 (inner peak) or TeF - 5 and other ionic dissociation products (outer peak). The model also suggests that the observed branching ratio requires a vertical electron affinity of < or =1.9 eV, much lower than its adiabatic electron affinity of 3.3 eV

President Ilves kõneles Saint Gallenis

Index Scriptorium Estoniae

2009-01-01

President Toomas Hendrik Ilves esines 9. mail 2009 Šveitsis, Saint Galleni 39. majandus- ja poliitikasümpoosionil kõnega teemal "20 aastat pärast Berliini müüri langemist - Euroopa uued väljakutsed". Vabariigi President töövisiidil Šveitsis 08.-09.05.2009

Analysis of in-core coolant temperatures of FFTF instrumented fuels tests at full power

International Nuclear Information System (INIS)

Hoth, C.W.

1981-01-01

Two full size highly instrumented fuel assemblies were inserted into the core of the Fast Flux Test Facility in December of 1979. The major objectives of these instrumented tests are to provide verification of the FFTF core conditions and to characterize temperature patterns within FFTF driver fuel assemblies. A review is presented of the results obtained during the power ascents and during irradiation at a constant reactor power of 400 MWt. The results obtained from these instrumented tests verify the conservative nature of the design methods used to establish core conditions in FFTF. The success of these tests also demonstrates the ability to design, fabricate, install and irradiate complex, instrumented fuel tests in FFTF using commercially procured components

Irradiation Test Plan and Safety Analysis of the Fatigue Capsule(05S-05K)

Energy Technology Data Exchange (ETDEWEB)

Cho, Man Soon; Kim, B. G.; Kang, Y. H.; Choo, K. N.; Sohn, J. M.; Park, S. J.; Shin, Y. T.; Seo, C. K

2007-01-15

In this report, the design, fabrication, the out-pile test and the irradiation test plan of the fatigue capsule 05S-05K were described and the safety aspect during the design, fabrication and irradiation test was reviewed. A cyclic load device necessary for the fatigue test was newly designed and manufactured. By using the cyclic load device the performance test and the preliminary fatigue test were performed with STS316L specimen of {phi}1.8 mm x 12.5 mm gage length under the same condition(550 .deg. C) as the temperature of the specimen during the irradiation test. As a result of the test, the fracture of the specimen occurs at a total of 70,120 cycles, at which the displacement was 2.02 mm. The reactivity effect was reviewed and an analysis for the structural and thermal integrity was performed to review the safety of the capsule, which will be irradiated at a temperature higher than 550 .deg. C And the thermal analysis shows that the temperatures of the parts are less than the melting temperatures of the corresponding materials. The structural analysis considering this temperature shows that the combined stress on the outer tube is less than the allowable stress limits and so the structural integrity is maintained.

Drying results of K-Basin fuel element 0309M (Run 3)

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-07-01

An N-Reactor outer fuel element that had been stored underwater in the Hanford 100 Area K-West Basin was subjected to a combination of low- and high-temperature vacuum drying treatments. These studies are part of a series of tests being conducted by Pacific Northwest National Laboratory on the drying behavior of spent nuclear fuel elements removed from both the K-West and K-East Basins. The drying test series was designed to test fuel elements that ranged from intact to severely damaged. The fuel element discussed in this report was removed from K-West canister 0309M during the second fuel selection campaign, conducted in 1996, and has remained in wet storage in the Postirradiation Testing Laboratory (PTL, 327 Building) since that time. The fuel element was broken in two pieces, with a relatively clean fracture, and the larger piece was tested. A gray/white coating was observed. This was the first test of a damaged fuel element in the furnace. K-West canisters can hold up to seven complete fuel assemblies, but, for purposes of this report, the element tested here is designated as Element 0309M. Element 0309M was subjected to drying processes based on those proposed under the Integrated Process Strategy, which included a hot drying step

Structural characterization and electron density distribution studies of (La{sub 0.8}Ca{sub 0.2})(Cr{sub 0.9−x}Co{sub 0.1}Mn{sub x})O{sub 3}

Energy Technology Data Exchange (ETDEWEB)

Saravanan, R. [Research Centre and Post Graduate Department of physics, The Madura College, Madurai 625011 (India); Thenmozhi, N., E-mail: thenmozhi.n6@gmail.com [PG and Research Department of Physics, NMSSVN College, Nagamalai, Madurai 625019 (India); Fu, Yen-Pei [Department of materials Science and Engineering, National Dong-Hwa University, Shou-Feng, Hualien 974, Taiwan (China)

2016-07-15

The doped lanthanum chromite (La{sub 0.8}Ca{sub 0.2})(Cr{sub 0.9−x}Co{sub 0.1}Mn{sub x})O{sub 3} (x=0.03, 0.06, 0.09 and 0.12) were synthesized by solid state reaction technique. The samples have been characterized by X-ray diffraction for structural and charge density analysis. XRD data show that the grown samples are orthorhombic in structure with single phase. The spatial charge density distribution in the unit cell for the synthesized samples has been studied using maximum entropy method. Further, the samples were analyzed by UV–visible spectrometry for optical properties and scanning electron microscopy for surface morphology. From the optical data, it was found that the direct band gap of the samples range from 2.27 to 2.46 eV. The samples were also investigated by vibrating sample magnetometry for magnetic properties. From VSM data, it is inferred that all the samples in this series are found to be predominantly antiferromagnetic in nature. Since the doped lanthanum chromites have good mechanical properties and electrical conductivity at high temperature, these materials are used in solid oxide fuel cells (SOFC).

Failure of the capsule for coated particles irradiation

International Nuclear Information System (INIS)

Yamaki, Jikei; Nomura, Yasushi; Nagamatsuya, Takaaki; Yamahara, Takeshi; Sakai, Haruyuki

1975-10-01

During operation cycle No. 27 of the JMTR (Japan Material Testing Reactor) on May 20, 1974, leakage of the fission product gas occurred from the capsule 72F-7A, which contained coated particles for the irradiation; the coated particles are for the development of a multi-purpose high temperature gas cooled reactor. The capsule was designed for heat 1600 0 C. Three nickel plates as the heat reflector were sandwiched in between the plates of titanium and zirconium, which were adsorbents for the impurity gases in the cladding tube (Nb-1%Zr). Temperatures of the plates were about 1000 0 C under the irradiation, so one metal diffused into the other metal through interfaces, resulting in the formation of an alloy. Its melting point was lower than those of metals in the capsule. The cladding material Nb-1%Zr was melted by the alloy and finally a pin hole developed through the cladding. The process of failure, design of the capsule, post-irradiation test of the capsule and the failure-reproducing experiment with a mock-up capsule are described. (auth.)

Asymmetric-shell ignition capsule design to tune the low-mode asymmetry during the peak drive

International Nuclear Information System (INIS)

Gu, Jianfa; Dai, Zhensheng; Song, Peng; Zou, Shiyang; Ye, Wenhua; Zheng, Wudi; Gu, Peijun; Wang, Jianguo; Zhu, Shaoping

2016-01-01

The low-mode radiation flux asymmetry in the hohlraum is a main source of performance degradation in the National Ignition Facility (NIF) implosion experiments. To counteract the deleterious effects of the large positive P2 flux asymmetry during the peak drive, this paper develops a new tuning method called asymmetric-shell ignition capsule design which adopts the intentionally asymmetric CH ablator layer or deuterium-tritium (DT) ice layer. A series of two-dimensional implosion simulations have been performed, and the results show that the intentionally asymmetric DT ice layer can significantly improve the fuel ρR symmetry, hot spot shape, hot spot internal energy, and the final neutron yield compared to the spherical capsule. This indicates that the DT asymmetric-shell capsule design is an effective tuning method, while the CH ablator asymmetric-shell capsule could not correct the fuel ρR asymmetry, and it is not as effective as the DT asymmetric-shell capsule design.

Asymmetric-shell ignition capsule design to tune the low-mode asymmetry during the peak drive

Science.gov (United States)

Gu, Jianfa; Dai, Zhensheng; Song, Peng; Zou, Shiyang; Ye, Wenhua; Zheng, Wudi; Gu, Peijun; Wang, Jianguo; Zhu, Shaoping

2016-08-01

The low-mode radiation flux asymmetry in the hohlraum is a main source of performance degradation in the National Ignition Facility (NIF) implosion experiments. To counteract the deleterious effects of the large positive P2 flux asymmetry during the peak drive, this paper develops a new tuning method called asymmetric-shell ignition capsule design which adopts the intentionally asymmetric CH ablator layer or deuterium-tritium (DT) ice layer. A series of two-dimensional implosion simulations have been performed, and the results show that the intentionally asymmetric DT ice layer can significantly improve the fuel ρR symmetry, hot spot shape, hot spot internal energy, and the final neutron yield compared to the spherical capsule. This indicates that the DT asymmetric-shell capsule design is an effective tuning method, while the CH ablator asymmetric-shell capsule could not correct the fuel ρR asymmetry, and it is not as effective as the DT asymmetric-shell capsule design.

Asymmetric-shell ignition capsule design to tune the low-mode asymmetry during the peak drive

Energy Technology Data Exchange (ETDEWEB)

Gu, Jianfa, E-mail: gu-jianfa@iapcm.ac.cn; Dai, Zhensheng, E-mail: dai-zhensheng@iapcm.ac.cn; Song, Peng; Zou, Shiyang; Ye, Wenhua; Zheng, Wudi; Gu, Peijun; Wang, Jianguo; Zhu, Shaoping [Institute of Applied Physics and Computational Mathematics, Beijing 100088 (China)

2016-08-15

The low-mode radiation flux asymmetry in the hohlraum is a main source of performance degradation in the National Ignition Facility (NIF) implosion experiments. To counteract the deleterious effects of the large positive P2 flux asymmetry during the peak drive, this paper develops a new tuning method called asymmetric-shell ignition capsule design which adopts the intentionally asymmetric CH ablator layer or deuterium-tritium (DT) ice layer. A series of two-dimensional implosion simulations have been performed, and the results show that the intentionally asymmetric DT ice layer can significantly improve the fuel ρR symmetry, hot spot shape, hot spot internal energy, and the final neutron yield compared to the spherical capsule. This indicates that the DT asymmetric-shell capsule design is an effective tuning method, while the CH ablator asymmetric-shell capsule could not correct the fuel ρR asymmetry, and it is not as effective as the DT asymmetric-shell capsule design.

The Thermal-hydraulic Performance Test Report for the Non-instrumented Irradiation Test Rig of Annular Fuel

Energy Technology Data Exchange (ETDEWEB)

Kim, Dae Ho; Lee, Kang Hee; Shin, Chang Hwan

2008-09-15

This report presents the results of pressure drop test, vibration test and endurance test for the non-instrumented rig using the irradiation test in HANARO of the double cooled annular fuel which were designed and fabricated by KAERI. From the out-pile thermal hydraulic tests, corresponding to the pressure drop of 200 kPa is measured to be about 9.72 kg/sec. Vibration frequency for the non-instrumented rig ranges from 5.0 to 10.7 kg/s. RMS(Root Mean Square) displacement for non-instrumented rig is less than 11.73 m, and the maximum displacement is less than 54.87m. The flow rate for endurance test were 10.5 kg/s, which was 110% of 9.72 kg/s. And the endurance test was carried out for 3 days. The test results found not to the wear and satisfied to the limits of pressure drop, flow rate, vibration and wear in the non-instrumented rig. This test was performed at the FIVPET facility.

Analysis of Ignition Testing on K-West Basin Fuel

Energy Technology Data Exchange (ETDEWEB)

J. Abrefah; F.H. Huang; W.M. Gerry; W.J. Gray; S.C. Marschman; T.A. Thornton

1999-08-10

Approximately 2100 metric tons of spent nuclear fuel (SNF) discharged from the N-Reactor have been stored underwater at the K-Basins in the 100 Area of the Hanford Site. The spent fuel has been stored in the K-East Basin since 1975 and in the K-West Basin since 1981. Some of the SNF elements in these basins have corroded because of various breaches in the Zircaloy cladding that occurred during fuel discharge operations and/or subsequent handling and storage in the basins. Consequently, radioactive material in the fuel has been released into the basin water, and water has leaked from the K-East Basin into the soil below. To protect the Columbia River, which is only 380 m from the basins, the SNF is scheduled to be removed and transported for interim dry storage in the 200 East Area, in the central portion of the Site. However, before being shipped, the corroded fuel elements will be loaded into Multi-Canister OverPacks and conditioned. The conditioning process will be selected based on the Integrated Process Strategy (IPS) (WHC 1995), which was prepared on the basis of the dry storage concept developed by the Independent Technical Assessment (ITA) team (ITA 1994).

Analysis of Ignition Testing on K-West Basin Fuel

International Nuclear Information System (INIS)

Abrefah, J.; Huang, F.H.; Gerry, W.M.; Gray, W.J.; Marschman, S.C.; Thornton, T.A.

1999-01-01

Approximately 2100 metric tons of spent nuclear fuel (SNF) discharged from the N-Reactor have been stored underwater at the K-Basins in the 100 Area of the Hanford Site. The spent fuel has been stored in the K-East Basin since 1975 and in the K-West Basin since 1981. Some of the SNF elements in these basins have corroded because of various breaches in the Zircaloy cladding that occurred during fuel discharge operations and/or subsequent handling and storage in the basins. Consequently, radioactive material in the fuel has been released into the basin water, and water has leaked from the K-East Basin into the soil below. To protect the Columbia River, which is only 380 m from the basins, the SNF is scheduled to be removed and transported for interim dry storage in the 200 East Area, in the central portion of the Site. However, before being shipped, the corroded fuel elements will be loaded into Multi-Canister OverPacks and conditioned. The conditioning process will be selected based on the Integrated Process Strategy (IPS) (WHC 1995), which was prepared on the basis of the dry storage concept developed by the Independent Technical Assessment (ITA) team (ITA 1994)

Growth of solid solutions with colquiriite structure LiCa0,2Sr0,8AlF6: Ce3+

International Nuclear Information System (INIS)

Shavelev, A A; Nizamutdinov, A S; Semashko, V V; Marisov, M A

2014-01-01

Aim of this work were experiments on growing new materials based on fluoride crystals with the colquiriite structure LiSr 0,8 Ca 0,2 F 6 , as well as the study of their phase composition. It is shown that for a series of crystals LiSr 0,8 Ca 0,2 F 6 distribution of reflections observed corresponds to the colquiriite structure, and the dependence of the lattice constant in the transition from LiCaAlF 6 crystal to LiSrAlF 6 crystal is linear. Also it found that absorption coefficient in mixed samples is much larger than in not mixed

Nuclear reactor instrumentation

International Nuclear Information System (INIS)

Duncombe, E.; McGonigal, G.

1976-01-01

Reference is made to the instrumentation of liquid metal cooled fast reactors. In order to ensure the safe operation of such reactors it is necessary to constantly monitor the coolant flowing through the fuel assemblies for temperature and rate of flow, requiring a large number of sensors. An improved and simplified arrangement is claimed in which the fuel assemblies feed a fraction of coolant to three instrument units arranged to sense the temperature and rate of flow of samples of coolant. Each instrument unit comprises a sleeve housing a sensing unit and has a number of inlet ducts arranged for receiving coolant from a fuel assembly together with a single outlet. The sensing unit has three thermocouple hot junctions connected in series, the hot junctions and inlet ducts being arranged in pairs. Electromagnetic windings around an inductive core are arranged to sense variation in flow of liquid metal by flux distortion. Fission product sensing means may also be provided. Full constructional details are given. (U.K.)

Behavioral and motoric testing of transgenic minipigs - focus on F0, F1, and F2 generations

Czech Academy of Sciences Publication Activity Database

Bohuslavová, Božena; Kučerová, S.; Mačáková, Monika; Ellederová, Zdeňka; Motlík, Jan

2015-01-01

Roč. 78, Suppl 2 (2015), s. 16-16 ISSN 1210-7859. [Conference on Animal Models for neurodegenerative Diseases /3./. 08.11.2015-10.11.2015, Liblice] R&D Projects: GA MŠk ED2.1.00/03.0124; GA MŠk(CZ) 7F14308 Institutional support: RVO:67985904 Keywords : minipig Subject RIV: FH - Neurology

20% Efficient Zn0.9Mg0.1O:Al/Zn0.8Mg0.2O/Cu(In,Ga)(S,Se)2 Solar Cell Prepared by All-Dry Process through a Combination of Heat-Light-Soaking and Light-Soaking Processes.

Science.gov (United States)

Chantana, Jakapan; Kato, Takuya; Sugimoto, Hiroki; Minemoto, Takashi

2018-04-04

Development of Cd-free Cu(In,Ga)(S,Se) 2 (CIGSSe)-based thin-film solar cells fabricated by an all-dry process is intriguing to minimize optical loss at a wavelength shorter than 520 nm owing to absorption of the CdS buffer layer and to be easily integrated into an in-line process for cost reduction. Cd-free CIGSSe solar cells are therefore prepared by the all-dry process with a structure of Zn 0.9 Mg 0.1 O:Al/Zn 0.8 Mg 0.2 O/CIGSSe/Mo/glass. It is demonstrated that Zn 0.8 Mg 0.2 O and Zn 0.9 Mg 0.1 O:Al are appropriate as buffer and transparent conductive oxide layers with large optical band gap energy values of 3.75 and 3.80 eV, respectively. The conversion efficiency (η) of the Cd-free CIGSSe solar cell without K-treatment is consequently increased to 18.1%. To further increase the η, the Cd-free CIGSSe solar cell with K-treatment is next fabricated and followed by posttreatment called the heat-light-soaking (HLS) + light-soaking (LS) process, including HLS at 110 °C followed by LS under AM 1.5G illumination. It is disclosed that the HLS + LS process gives rise to not only the enhancement of carrier density but also the decrease in the carrier recombination rate at the buffer/absorber interface. Ultimately, the η of the Cd-free CIGSSe solar cell with K-treatment prepared by the all-dry process is enhanced to the level of 20.0%.

Tunable exchange bias effect in magnetic Bi0.9Gd0.1Fe0.9Ti0.1O3 nanoparticles at temperatures up to 250K

DEFF Research Database (Denmark)

Basith, M. A.; Khan, F. A.; Ahmmad, Bashir

2015-01-01

that the strength of the exchange bias effect is tunable by the field cooling. The HEB values are also found to be dependent on the temperature. This magnetically tunable exchange bias obtained at temperatures up to 250K in Bi0.9Gd0.1Fe0.9Ti0.1O3 nanoparticles may be worthwhile for potential applications.......The exchange bias (EB) effect has been observed in magnetic Bi0.9Gd0.1Fe0.9Ti0.1O3 nanoparticles.The influence of magnetic field cooling on the exchange bias effect has also been investigated. The magnitude of the exchange bias field (HEB) increases with the cooling magnetic field, showing...

First principles simulation on the K0.8Fe2Se2 high-temperature structural superconductor

International Nuclear Information System (INIS)

Guo, Rui; Yang, Shizhong; Khosravi, Ebrahim; Zhao, Guang-Lin; Bagayoko, Diola

2013-01-01

Highlights: • The superconductor K 0.8 Fe 2 Se 2 super cell size, shape, and atomic positions are fully optimized using first principles density functional theory method. • Each K atom donates 0.8 |e| with K vacancies in the supercell, each Fe atom donates 0.4 |e|, while each Se atom gains 0.7 |e| ∼ 0.8 |e|. • Fe atoms show magnetic moment fluctuation and possible strong spin-orbital coupling. -- Abstract: Since the synthesis of the first ones in 2008, iron-based high temperature superconductors have been the subject of many studies. This great interest is partly due to their higher, upper magnetic field, smaller Fermi surface around the Γ point, and a larger coherence length. This work is focused on A x Fe 2 Se 2 structural superconductor (FeSe, 11 hierarchy; A = K, Cs) as recently observed. ARPES data show novel, electronic structure and a hole-free Fermi surface which is different from previously observed Fermi surface images. We use ab initio density functional theory method to simulate the electronic structure of the novel superconductor A x Fe 2 Se 2 . We compare this electronic structure with those of other Fe-based superconductors

Out-pile Test of Double Cladding Fuel Rod Mockups for a Nuclear Fuel Irradiation Test

Energy Technology Data Exchange (ETDEWEB)

Sohn, Jaemin; Park, Sungjae; Kang, Younghwan; Kim, Harkrho; Kim, Bonggoo; Kim, Youngki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-05-15

An instrumented capsule for a nuclear fuel irradiation test has been developed to measure fuel characteristics, such as a fuel temperature, internal pressure of a fuel rod, a fuel pellet elongation and a neutron flux during an irradiation test at HANARO. In the future, nuclear fuel irradiation tests under a high temperature condition are expected from users. To prepare for this request, we have continued developing the technology for a high temperature nuclear fuel irradiation test at HANARO. The purpose of this paper is to verify the possibility that the temperature of a nuclear fuel can be controlled at a high temperature during an irradiation test. Therefore we designed and fabricated double cladding fuel rod mockups. And we performed out-pile tests using these mockups. The purposes of a out-pile test is to analyze an effect of a gap size, which is between an outer cladding and an inner cladding, on the temperature and the effect of a mixture ratio of helium gas and neon gas on the temperature. This paper presents the design and fabrication of double cladding fuel rod mockups and the results of the out-pile test.

Cobalt-free cathode material SrFe{sub 0.9}Nb{sub 0.1}O{sub 3-{delta}} for intermediate-temperature solid oxide fuel cells

Energy Technology Data Exchange (ETDEWEB)

Zhou, Qingjun [State Key Laboratory of Superhard Materials and College of Physics, Jilin University, Changchun 130012 (China); College of Science, Civil Aviation University of China, Tianjin 300300 (China); Zhang, Leilei; He, Tianmin [State Key Laboratory of Superhard Materials and College of Physics, Jilin University, Changchun 130012 (China)

2010-02-15

A cobalt-free cubic perovskite oxide, SrFe{sub 0.9}Nb{sub 0.1}O{sub 3-{delta}} (SFN) was investigated as a cathode for intermediate-temperature solid oxide fuel cells (IT-SOFCs). XRD results showed that SFN cathode was chemically compatible with the electrolyte Sm{sub 0.2}Ce{sub 0.8}O{sub 1.9} (SDC) for temperatures up to 1050 C. The electrical conductivity of SFN sample reached 34-70 S cm{sup -1} in the commonly operated temperatures of IT-SOFCs (600-800 C). The area specific resistance was 0.138 {omega} cm{sup 2} for SFN cathode on SDC electrolyte at 750 C. A maximum power density of 407 mW cm{sup -2} was obtained at 800 C for single-cell with 300 {mu}m thick SDC electrolyte and SFN cathode. (author)

Progress on the Hanford K basins spent nuclear fuel project

International Nuclear Information System (INIS)

Culley, G.E.; Fulton, J.C.; Gerber, E.W.

1996-01-01

This paper highlights progress made during the last year toward removing the Department of Energy's (DOE) approximately, 2,100 metric tons of metallic spent nuclear fuel from the two outdated K Basins at the Hanford Site and placing it in safe, economical interim dry storage. In the past year, the Spent Nuclear Fuel (SNF) Project has engaged in an evolutionary process involving the customer, regulatory bodies, and the public that has resulted in a quicker, cheaper, and safer strategy for accomplishing that goal. Development and implementation of the Integrated Process Strategy for K Basins Fuel is as much a case study of modern project and business management within the regulatory system as it is a technical achievement. A year ago, the SNF Project developed the K Basins Path Forward that, beginning in December 1998, would move the spent nuclear fuel currently stored in the K Basins to a new Staging and Storage Facility by December 2000. The second stage of this $960 million two-stage plan would complete the project by conditioning the metallic fuel and placing it in interim dry storage by 2006. In accepting this plan, the DOE established goals that the fuel removal schedule be accelerated by a year, that fuel conditioning be closely coupled with fuel removal, and that the cost be reduced by at least $300 million. The SNF Project conducted coordinated engineering and technology studies over a three-month period that established the technical framework needed to design and construct facilities, and implement processes compatible with these goals. The result was the Integrated Process Strategy for K Basins Fuel. This strategy accomplishes the goals set forth by the DOE by beginning fuel removal a year earlier in December 1997, completing it by December 1999, beginning conditioning within six months of starting fuel removal, and accomplishes it for $340 million less than the previous Path Forward plan

Drying results of K-Basin fuel element 1990 (Run 1)

International Nuclear Information System (INIS)

Marschman, S.C.; Abrefah, J.; Klinger, G.S.; Oliver, B.M.; MacFarlan, P.J.; Ritter, G.A.

1998-06-01

The water-filled K-Basins in the Hanford 100-Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first of those tests (Run 1), which was conducted on an N-Reactor inner fuel element (1990) that had been stored underwater in the K-West Basin (see Section 2.0). This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The testing was conducted in the Whole Element Furnace Testing System, described in Section 3.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in Section 4.0, and the experimental results provided in Section 5.0. These results are further discussed in Section 6.0

Res Sep 2014 Cover Tp 08.09.14.cdr

Indian Academy of Sciences (India)

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2010). ISSN 0971-8044. Regn No.KRNAVBGE-340/2012-2014. Licenced to Post without prepayment No.6. Posted at MBC, GPO, Bangalore 560 001, 12.09.14. Registered with Registrar of Newspapers in India vide Regn. No. 66273/96.

Repurposing an irradiated instrumented TRIGA fuel element for regular use

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Paulo F.; Souza, Luiz C.A., E-mail: pfo@cdtn.br, E-mail: lcas@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2015-07-01

TRIGA IPR-R1 is a research reactor also used for training and radioisotope production, located at the Centro de Desenvolvimento da Tecnologia Nuclear da Comissao Nacional de Energia Nuclear (Nuclear Technology Development Centre, Brazilian National Nuclear Energy Commission - CDTN/CNEN). Its first criticality occurred in November 1960. All original fuel elements were aluminum-clad. In 1971 nine new fuel elements, stainless steel-clad were acquired. One of them was an instrumented fuel element (IFE), equipped with 3 thermocouples. The IFE was introduced into the core only on August 2004, and remained there until July 2007. It was removed from the core after the severing of contacts between the thermocouples and their extension cables. After an unsuccessful attempt to recover electrical access to the thermocouples the IFE was transferred from the reactor pool to an auxiliary spent fuel storage well, with water, in the reactor room. In December 2011 the IFE was transferred to an identical well, dry, where it remains so far. This work is a proposal for recovery of this instrumented fuel element, by removing the cable guide rod and adaptation of a superior terminal plug similar to conventional fuel elements. This will enable its handling through the same tool used for regular fuel elements and its return to the reactor core. This is a delicate intervention in terms of radiological protection, and will require special care to minimize the exposure of operators. (author)

Reversible operation of microtubular solid oxide cells using La0.6Sr0.4Co0.2Fe0.8O3-δ-Ce0.9Gd0.1O2-δ oxygen electrodes

Science.gov (United States)

López-Robledo, M. J.; Laguna-Bercero, M. A.; Larrea, A.; Orera, V. M.

2018-02-01

Yttria stabilized zirconia (YSZ) based microtubular solid oxide fuel cells (mT-SOFCs) using La0.6Sr0.4Co0.2Fe0.8O3-δ (LSCF) and Ce0.9Gd0.1O2-δ (GDC) as the oxygen electrode, along with a porous GDC electrolyte-electrode barrier layer, were fabricated and characterized in both fuel cell (SOFC) and electrolysis (SOEC) operation modes. The cells were anode-supported, the NiO-YSZ microtubular supports being made by Powder Extrusion Moulding (PEM). The cells showed power densities of 695 mW cm-2 at 800 °C and 0.7 V in SOFC mode, and of 845 mA cm-2 at 800 °C and 1.3 V in SOEC mode. AC impedance experiments performed under different potential loads demonstrated the reversibility of the cells. These results showed that these cells, prepared with a method suitable for using on an industrial scale, are highly reproducible and reliable, as well as very competitive as reversible SOFC-SOEC devices operating at intermediate temperatures.

EST Table: FS910254 [KAIKOcDNA[Archive

Lifescience Database Archive (English)

Full Text Available FS910254 E_FL_fufe_19K18_F_0 10/09/28 87 %/274 aa ref|NP_001098702.1| nanos-like pr...otein [Bombyx mori] gb|ABS17681.1| nanos-like protein [Bombyx mori] 10/09/12 low homology 10/08/29 n.h 10/09

K-Basin spent nuclear fuel characterization data report 2

International Nuclear Information System (INIS)

Abrefah, J.; Gray, W.J.; Ketner, G.L.; Marschman, S.C.; Pyecha, T.D.; Thornton, T.A.

1996-03-01

An Integrated Process Strategy has been developed to package, condition, transport, and store in an interim storage facility the spent nuclear fuel (SNF) currently residing in the K-Basins at Hanford. Information required to support the development of the condition process and to support the safety analyses must be obtained from characterization testing activities conducted on fuel samples from the Basins. Some of the information obtained in the testing was reported in PNL-10778, K-Basin Spent Nuclear Fuel Characterization Data Report (Abrefah et al. 1995). That report focused on the physical, dimensional, metallographic examinations of the first K-West (KW) Basin SNF element to be examined in the Postirradiation Testing Laboratory (PTL) hot cells; it also described some of the initial SNF conditioning tests. This second of the series of data reports covers the subsequent series of SNF tests on the first fuel element. These tests included optical microscopy analyses, conditioning (drying and oxidation) tests, ignition tests, and hydrogen content tests

K-Basin spent nuclear fuel characterization data report 2

Energy Technology Data Exchange (ETDEWEB)

Abrefah, J.; Gray, W.J.; Ketner, G.L.; Marschman, S.C.; Pyecha, T.D.; Thornton, T.A.

1996-03-01

An Integrated Process Strategy has been developed to package, condition, transport, and store in an interim storage facility the spent nuclear fuel (SNF) currently residing in the K-Basins at Hanford. Information required to support the development of the condition process and to support the safety analyses must be obtained from characterization testing activities conducted on fuel samples from the Basins. Some of the information obtained in the testing was reported in PNL-10778, K-Basin Spent Nuclear Fuel Characterization Data Report (Abrefah et al. 1995). That report focused on the physical, dimensional, metallographic examinations of the first K-West (KW) Basin SNF element to be examined in the Postirradiation Testing Laboratory (PTL) hot cells; it also described some of the initial SNF conditioning tests. This second of the series of data reports covers the subsequent series of SNF tests on the first fuel element. These tests included optical microscopy analyses, conditioning (drying and oxidation) tests, ignition tests, and hydrogen content tests.

System design description for sampling fuel in K basins

International Nuclear Information System (INIS)

Baker, R.B.

1996-01-01

This System Design Description provides: (1) statements of the Spent Nuclear Fuel Projects (SNFP) needs requiring sampling of fuel in the K East and K West Basins, (2) the sampling equipment functions and requirements, (3) a general work plan and the design logic being followed to develop the equipment, and (4) a summary description of the design for the sampling equipment. The report summarizes the integrated application of both the subject equipment and the canister sludge sampler in near-term characterization campaigns at K Basins

Instrumentation needs in LWR severe fuel damage experiments

International Nuclear Information System (INIS)

McCormick, R.D.

1980-01-01

The Class 9 type nuclear accident is defined and the Three Mile Island type accident and proposed Idaho National Engineering Laboratory experiment series are described in some detail. Different types of severe fuel damage experiments are briefly discussed in order to show typical measurement requirements. General instrumentation needs and problems encountered in Class 9 accident research are outlined. It is concluded that the extremely high temperatures, high nuclear radiation fields, and oxidizing atmosphere will necessitate instrument development programs. Noncontact type sensing will be necessary in most of the molten core experiments

SMART instruments for radiological surveillance at the back end of nuclear fuel cycle

International Nuclear Information System (INIS)

Pendharkar, K.A.; Jauhri, G.S.; Ganesh, G.; Kulkarni, V.V.

2001-01-01

The back end of Nuclear Fuel Cycle mainly consists of Fuel Reprocessing Plant and Waste Management Plant for treatment of different types of wastes generated during processing of spent fuel. A fuel reprocessing plant handles annually several million curies of fission product activity and few hundred kg of plutonium. A Waste Management Facility associated with a reprocessing plant also handles several million curies of fission product activity. In both the plants several types of radiological measurements have to be carried out to ensure that the individual doses are well below regulatory limits and release of radioactivity to environment (through stack and through liquid effluent) is below the limit stipulated in technical specifications of the plant. The measurements comprise individual external dose, measurement of radiation level in different areas of the plant, assessment of air-borne activity due to plutonium and fission products in different areas of the plant, radioactivity release to environment through liquid effluents and through stack. In order to carry out the above mentioned measurements large number of different types of instruments are required. The existing instruments are analog instruments. These instruments have served well. However they have certain limitations with respect to flexibility and extra functionality. In this respect the 'SMART' instruments have distinct advantages. The advantages, that are offered by the 'SMART' instrument in making the radiological surveillance programme more effective, are brought out in the paper. (author)

Strategy for thermo-gravimetric analysis of K East fuel samples

International Nuclear Information System (INIS)

Lawrence, L.A.

1997-01-01

A strategy was developed for the Thermo-Gravimetric Analysis (TGA) testing of K East fuel samples for oxidation rate determinations. Tests will first establish if there are any differences for dry air oxidation between the K West and K East fuel. These tests will be followed by moist inert gas oxidation rate measurements. The final series of tests will consider pure water vapor i.e., steam

Simulation of a 250 kW diesel fuel processor/PEM fuel cell system

Science.gov (United States)

Amphlett, J. C.; Mann, R. F.; Peppley, B. A.; Roberge, P. R.; Rodrigues, A.; Salvador, J. P.

Polymer-electrolyte membrane (PEM) fuel cell systems offer a potential power source for utility and mobile applications. Practical fuel cell systems use fuel processors for the production of hydrogen-rich gas. Liquid fuels, such as diesel or other related fuels, are attractive options as feeds to a fuel processor. The generation of hydrogen gas for fuel cells, in most cases, becomes the crucial design issue with respect to weight and volume in these applications. Furthermore, these systems will require a gas clean-up system to insure that the fuel quality meets the demands of the cell anode. The endothermic nature of the reformer will have a significant affect on the overall system efficiency. The gas clean-up system may also significantly effect the overall heat balance. To optimize the performance of this integrated system, therefore, waste heat must be used effectively. Previously, we have concentrated on catalytic methanol-steam reforming. A model of a methanol steam reformer has been previously developed and has been used as the basis for a new, higher temperature model for liquid hydrocarbon fuels. Similarly, our fuel cell evaluation program previously led to the development of a steady-state electrochemical fuel cell model (SSEM). The hydrocarbon fuel processor model and the SSEM have now been incorporated in the development of a process simulation of a 250 kW diesel-fueled reformer/fuel cell system using a process simulator. The performance of this system has been investigated for a variety of operating conditions and a preliminary assessment of thermal integration issues has been carried out. This study demonstrates the application of a process simulation model as a design analysis tool for the development of a 250 kW fuel cell system.

FY09 Advanced Instrumentation and Active Interrogation Research for Safeguards

International Nuclear Information System (INIS)

Chichester, D.L.; Pozzi, S.A.; Seabury, E.H.; Dolan, J.L.; Flaska, M.; Johnson, J.T.; Watson, S.M.; Wharton, J.

2009-01-01

Multiple small-scale projects have been undertaken to investigate advanced instrumentation solutions for safeguard measurement challenges associated with advanced fuel cycle facilities and next-generation fuel reprocessing installations. These activities are in support of the U.S. Department of Energy's Fuel Cycle Research and Development program and its Materials Protection, Accounting, and Control for Transmutation (MPACT) campaign. (1) Work was performed in a collaboration with the University of Michigan (Prof. Sara Pozzi, co-PI) to investigate the use of liquid-scintillator radiation detectors for assaying mixed-oxide (MOX) fuel, to characterize its composition and to develop advanced digital pulse-shape discrimination algorithms for performing time-correlation measurements in the MOX fuel environment. This work included both simulations and experiments and has shown that these techniques may provide a valuable approach for use within advanced safeguard measurement scenarios. (2) Work was conducted in a collaboration with Oak Ridge National Laboratory (Dr. Paul Hausladen, co-PI) to evaluate the strengths and weaknesses of the fast-neutron coded-aperture imaging technique for locating and characterizing fissile material, and as a tool for performing hold-up measurements in fissile material handling facilities. This work involved experiments at Idaho National Laboratory, using MOX fuel and uranium metal, in both passive and active interrogation configurations. A complete analysis has not yet been completed but preliminary results suggest several potential uses for the fast neutron imaging technique. (3) Work was carried out to identify measurement approaches for determining nitric acid concentration in the range of 1-4 M and beyond. This work included laboratory measurements to investigate the suitability of prompt-gamma neutron activation analysis for this measurement and product reviews of other commercial solutions. Ultrasonic density analysis appears to be

Demonstration of high coupling efficiency to Al capsule in rugby hohlraum on NIF

Science.gov (United States)

Ping, Y.; Smalyuk, V.; Amendt, P.; Bennett, D.; Chen, H.; Dewald, E.; Goyon, C.; Graziani, F.; Johnson, S.; Khan, S.; Landen, O.; Nikroo, A.; Pino, J.; Ralph, J.; Seugling, R.; Strozzi, D.; Tipton, R.; Tommasini, R.; Wang, M.; Loomis, E.; Merritt, E.; Montgomery, D.

2017-10-01

A new design of the double-shell approach predicts a high coupling efficiency from the hohlraum to the capsule, with 700 kJ in the capsule instead of 200kJ in the conventional low-Z single-shell scheme, improving prospects of double-shell performance. A recent experiment on NIF has evaluated a first step toward this goal of energy coupling using 0.7x subscale Al capsule, Au rugby hohlraum and 1MJ drive. A shell velocity of 150 μm/ns was measured, DANTE peak temperature of 255 eV was measured, and shell kinetic energy of 36 kJ was inferred using a rocket model, all close to predictions and consistent with 330kJ of total energy coupled to the capsule. Data analysis and more results from subsequent experiments will be presented. In the next step, an additional 2x increase of total coupled energy up to 700 kJ is projected for full-scale 2-MJ drive in U Rugby hohlraum. This work was performed under DOE contract DE-AC52-07NA27344.

Estimation of gastric residence time of the Heidelberg capsule in humans: effect of varying food composition

International Nuclear Information System (INIS)

Mojaverian, P.; Ferguson, R.K.; Vlasses, P.H.; Rocci, M.L. Jr.; Oren, A.; Fix, J.A.; Caldwell, L.J.; Gardner, C.

1985-01-01

In animal and human studies, the gastric emptying of large (greater than 1 mm) indigestible solids is due to the activity of the interdigestive migrating myoelectric complex. The gastric residence time (GRT) of an orally administered, nondigestible, pH-sensitive, radiotelemetric device (Heidelberg capsule) was evaluated in three studies in healthy volunteers. In 6 subjects, the GRT of the Heidelberg capsule was compared with the half-emptying time (t1/2) of diethylenetriaminepentaacetic acid labeled with technetium 99m after a 4-ml/kg liquid fatty meal. The mean (+/-SD) GRT (4.3 +/- 1.4 h) was significantly (p less than 0.001) longer than the mean t1/2 (1.1 +/- 0.3 h); the GRT was prolonged compared with the t1/2 in each subject. In a randomized, crossover trial in 10 subjects, frequent feeding caused a dramatic prolongation in mean GRT of the capsule compared with the fasting state (greater than 14.5 vs. 0.5 h, p less than 0.005). In another crossover study in 6 subjects, the GRT of the capsule was evaluated after an overnight fast, a standard breakfast including solid food, and a liquid meal (i.e., 200 ml of diluted light cream). The mean GRT was 2.6 +/- 0.9 h after the liquid meal vs. 1.2 +/- 0.8 h after fasting (p less than 0.025). The mean GRT after the breakfast was 4.8 +/- 1.5 h, which was significantly greater than that after fasting (p less than 0.001) and after the liquid meal (p less than 0.01). These data suggest that the GRT of the Heidelberg capsule is a marker of the interdigestive migrating myoelectric complex in humans, the interdigestive migrating myoelectric complex can be markedly delayed by frequent feedings with solids, and the interdigestive migrating myoelectric complex is delayed by both liquid and solid meals

Estimation of gastric residence time of the Heidelberg capsule in humans: effect of varying food composition

Energy Technology Data Exchange (ETDEWEB)

Mojaverian, P.; Ferguson, R.K.; Vlasses, P.H.; Rocci, M.L. Jr.; Oren, A.; Fix, J.A.; Caldwell, L.J.; Gardner, C.

1985-08-01

In animal and human studies, the gastric emptying of large (greater than 1 mm) indigestible solids is due to the activity of the interdigestive migrating myoelectric complex. The gastric residence time (GRT) of an orally administered, nondigestible, pH-sensitive, radiotelemetric device (Heidelberg capsule) was evaluated in three studies in healthy volunteers. In 6 subjects, the GRT of the Heidelberg capsule was compared with the half-emptying time (t1/2) of diethylenetriaminepentaacetic acid labeled with technetium 99m after a 4-ml/kg liquid fatty meal. The mean (+/-SD) GRT (4.3 +/- 1.4 h) was significantly (p less than 0.001) longer than the mean t1/2 (1.1 +/- 0.3 h); the GRT was prolonged compared with the t1/2 in each subject. In a randomized, crossover trial in 10 subjects, frequent feeding caused a dramatic prolongation in mean GRT of the capsule compared with the fasting state (greater than 14.5 vs. 0.5 h, p less than 0.005). In another crossover study in 6 subjects, the GRT of the capsule was evaluated after an overnight fast, a standard breakfast including solid food, and a liquid meal (i.e., 200 ml of diluted light cream). The mean GRT was 2.6 +/- 0.9 h after the liquid meal vs. 1.2 +/- 0.8 h after fasting (p less than 0.025). The mean GRT after the breakfast was 4.8 +/- 1.5 h, which was significantly greater than that after fasting (p less than 0.001) and after the liquid meal (p less than 0.01). These data suggest that the GRT of the Heidelberg capsule is a marker of the interdigestive migrating myoelectric complex in humans, the interdigestive migrating myoelectric complex can be markedly delayed by frequent feedings with solids, and the interdigestive migrating myoelectric complex is delayed by both liquid and solid meals.

Re-irradiation tests of spent fuel at JMTR by means of re-instrumentation technique

International Nuclear Information System (INIS)

Nakamura, Jinichi; Shimizu, Michio; Endo, Yasuichi; Nabeya, Hideaki; Ichise, Kenichi; Saito, Junichi; Oshima, Kunio; Uetsuka, Hiroshi

1999-01-01

JAERI has developed re-irradiation test procedures of spent fuel irradiated at commercial reactor by means of re-instrumentation technique. Full length rods irradiated at commercial LWRs were re-fabricated to short length rods, and rod inner pressure gauges and fuel center thermocouples were re-instrumented to the rods. Re-irradiation tests to study the fuel behavior during power change were carried out by means of BOCA/OSF-1 facility at the JMTR. In the tests to study the fission gas release during power change, the rod inner pressure increase was observed during power change, especially during power reduction. The fission gas release during power reduction is estimated to be the release from fission gas bubbles on the grain boundary caused by the thermal stress in the pellet during power reduction. Re-irradiation test of gadolinia added fuel was performed by means of dual re-instrumentation technique (fuel center thermocouples and rod inner pressure gauge). A stepwise fission gas release during power change, and the following fuel center temperature change due to gap conductance change were observed. (author)

Hanford K Basins spent nuclear fuels project update

International Nuclear Information System (INIS)

Hudson, F.G.

1997-01-01

Twenty one hundred metric tons of spent nuclear fuel are stored in two concrete pools on the Hanford Site, known as the K Basins, near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current wet pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in the K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported into the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building (CSB) in the 200 Area for staging prior to hot conditioning. The conditioning step to remove chemically bound water is performed by holding the MCO at 300 C under vacuum. This step is necessary to prevent excessive pressure buildup during interim storage that could be caused by corrosion. After conditioning, MCOs will remain in the CSB for interim storage until a national repository is completed

Hanford K Basins spent nuclear fuels project update

Energy Technology Data Exchange (ETDEWEB)

Hudson, F.G.

1997-10-17

Twenty one hundred metric tons of spent nuclear fuel are stored in two concrete pools on the Hanford Site, known as the K Basins, near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current wet pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in the K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported into the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building (CSB) in the 200 Area for staging prior to hot conditioning. The conditioning step to remove chemically bound water is performed by holding the MCO at 300 C under vacuum. This step is necessary to prevent excessive pressure buildup during interim storage that could be caused by corrosion. After conditioning, MCOs will remain in the CSB for interim storage until a national repository is completed.

Drying results of K-Basin fuel element 5744U (Run 4)

International Nuclear Information System (INIS)

Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-07-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site. Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the fourth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 5744U. This element (referred to as Element 5744U) was stored underwater in the K-West Basin from 1983 until 1996. Element 5744U was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0

Drying Results of K-Basin Fuel Element 2660M (Run 7)

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1999-01-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the seventh of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 2660M. This element (referred to as Element 2660M) was stored underwater in the K-West Basin from 1983 until 1996. Element 2660M was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0

Drying Results of K-Basin Fuel Element 6513U (Run 8)

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1999-01-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL)on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the eighth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 6513U. This element (referred to as Element 6513U) was stored underwater in the K-West Basin from 1983 until 1996. Element 6513U was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0 and discussed in Section 6.0

Drying results of K-Basin fuel element 1164M (run 6)

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-08-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the sixth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 1164 M. This element (referred to as Element 1164M) was stored underwater in the K-West Basin from 1983 until 1996. Element 1164M was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0

La(0.4)Ba(0.6)Fe(0.8)Zn(0.2)O(3-delta) as cathode in solid oxide fuel cells for simultaneous NO reduction and electricity generation.

Science.gov (United States)

Zhou, Renjie; Bu, Yunfei; Xu, Dandan; Zhong, Qin

2014-01-01

A perovskite-type oxide La(0.4)Ba(0.6)Fe(0.8)Zn(0.2)O(3-delta) (LBFZ) was investigated as the cathode material for simultaneous NO reduction and electricity generation in solid oxide fuel cells (SOFCs). The microstructure of LBFZ was demonstrated by X-ray diffraction and scanning electron microscopy. The results showed that a single cubic perovskite LBFZ was formed after calcined at 1100 degrees C. Meanwhile, the solid-state reaction between LBFZ and Ce(0.8)Sm(0.2)O(1.9) (SDC) at 900 degrees C was negligible. To measure the electrochemical properties, SOFC units were constructed with Sm(0.9)Sr(0.1)Cr(0.5)Fe(0.5)O3 as the anode, SDC as the electrolyte and LBFZ as the cathode. The maximum power density increased with the increasing NO concentration and temperature. The cell resistance is mainly due to the cathodic polarization resistance.

On the Potential Role of Species Separation in DT Fuels on Implosion Performance

Science.gov (United States)

Amendt, Peter; Bellei, Claudio; Wilks, Scott; Haines, Malcolm; Casey, Dan; Li, C. K.; Petrasso, Richard

2012-10-01

The measurement of strong, self-generated electric fields (1-10 GVolts/m) in imploding capsules [1], their attribution to polarized (plasma) shock fronts [2], and the identification of plasma-enhanced binary species diffusion from barodiffusion and electrodiffusion [3] have led to a growing interest in the potential role of species separation in inertial-confinement-fusion (ICF) thermonuclear fuels. The potential for anomalous heating from transient frictional or resistive drag between D and T across a finite thickness shock front will be assessed and applied towards ignition thresholds and understanding some outstanding anomalies in the Omega implosion database.[4pt] [1] J.R. Rygg et al., Science 319, 1223 (2008); C.K. Li et al., Phys. Rev. Lett. 100, 225001 (2008).[0pt] [2] P.A. Amendt, J.L. Milovich, S.C. Wilks, C.K. Li, R.D. Petrasso and F.H. S'eguin, Plasma Phys. Control. Fusion 51, 124048 (2009).[0pt] [3] P. Amendt, C. Bellei and S.C. Wilks, Phys. Rev. Lett. (to appear).

Preparation of the capsule for the experiment with enriched fuel, task E-14-16-00-29; Priprema kapsula za eksperiment sa obogacenim gorivom, zad. E-14-16-00-29

Energy Technology Data Exchange (ETDEWEB)

Anastasijevic, P; Pavlovic, A; Zivkovic, S; Nikolic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1968-12-15

Preparation of the experiment with highly enriched fuel was based on the experience gained in constructing the capsule for natural uranium fuel. The possibility of using capsules with enriched UO{sub 2} in the RA reactor was analyzed. Special attention was devoted to the analysis of heat removal from the fuel rods. Capsule for irradiation of enriched UO{sub 2} was designed. The most important change appeared in sealing zircaloy-2 with stainless steel part. This must be done by diffusion welding. Since there is no device for diffusion welding in our country the elements will be sent to Saclay, France. All the parts made of domestic material were fabricated in the Institute. Purchase of elements that must be imported is delayed due to procedures for obtaining foreign currency. The fabrication of the capsule can be completed within 25 days upon obtaining the mentioned material from abroad. This report contains the safety report for the experiment with the enriched fuel. Na osnovu dosadasnjeg rada i iskustva na izgradnji kapsula za neobogaceno gorivo izvrsena je priprema za eksperimenat sa obogacenim gorivom. Analizirana je mogucnost koriscenja kapsula sa obogadenim UO{sub 2} u reaktoru RA. Narocito je posvecena paznja analizi odvodjenja generisane toplote iz sipki gorivnih elemenata. Projektovana je kapsula za ozracivanje obogacenog UO{sub 2}. Najznacajnija izmena je kod spajanja dela od cirkaloja-2 sa delom od nerdjajuceg celika. To spajanje morace se izvrsiti na masini za difuziono zavarivanje. Kako takvu masinu nemamo u zemlji elemente cemo dati da se zavare u Saclay-u. Svi delovi za koje nije bio potreban uvozni materijal izradjeni su u nasem Institutu. Kapsula nije montirana jer materijal iz uvoza koji se ugradjuje u kapsulu nije jos stigao u zemlju. Nabavka uvoznlh elemenata nije izvrsena na vreme zbog zakasnjenja kod odobravanja deviznih sredstava. Potpuna montaza kapsule moze se obaviti 25 dana po dobijanju pomenutog materijala. U radu je dat i izvestaj o

Bowman Capsulitis Predicts Poor Kidney Allograft Outcome in T Cell-Mediated Rejection.

Science.gov (United States)

Gallan, Alexander J; Chon, W James; Josephson, Michelle A; Cunningham, Patrick N; Henriksen, Kammi J; Chang, Anthony

2018-02-28

Acute T cell-mediated rejection (TCMR) is an important cause of renal allograft loss. The Banff classification for tubulointerstitial (type I) rejection is based on the extent of both interstitial inflammation and tubulitis. Lymphocytes may also be present between parietal epithelial cells and Bowman capsules in this setting, which we have termed "capsulitis." We conducted this study to determine the clinical significance of capsulitis. We identified 42 patients from the pathology archives at the University of Chicago with isolated Banff type I TCMR from 2010-2015. Patient demographic data, Banff classification, and graft outcome measurements were compared between capsulitis and non-capsulitis groups using Mann-Whitney U test. Capsulitis was present in 26 (62%), and was more frequently seen in Banff IB than IA TCMR (88% vs 44%, P=.01). Patients with capsulitis had a higher serum creatinine at biopsy (4.6 vs 2.9mg/dL, P=.04) and were more likely to progress to dialysis (42% vs 13%, P=.06) with fewer recovering their baseline serum creatinine (12% vs 38%, P=.08). Patients with both Banff IA TCMR and capsulitis have clinical outcomes similar or possibly worse than Banff IB TCMR compared to those with Banff IA and an absence of capsulitis. Capsulitis is an important pathologic parameter in the evaluation of kidney transplant biopsies with potential diagnostic, prognostic, and therapeutic implications in the setting of TCMR. Copyright © 2018. Published by Elsevier Inc.

Design, fabrication and installation of irradiation facilities -Advanced nuclear material development-

International Nuclear Information System (INIS)

Kim, Yong Seong; Lee, Jeong Yeong; Lee, Seong Ho; Ji, Dae Yeong; Kim, Seok Hoon; An, Seong Ho; Kim, Dong Hoon; Seok, Ho Cheon; Kim, Joon Yeon; Yang, Seong Hong

1994-07-01

The objective of this study is to design and construct the steady state fuel test loop and non-instrumented capsules to be installed in KMRR. The principle contents of this project are to design, fabricate the steady-state fuel test loop and non-instrumented capsule to be installed in KMRR for nuclear technology development. This project will be completed in 1996, so preparation of design criteria for fuel test loop have been performed in 1993 as the first year of the first phase in implementing this project. Also design and pressure drop test of non-instrumented capsule have been performed in 1993

Simulations of indirectly driven gas-filled capsules at the National Ignition Facility

Energy Technology Data Exchange (ETDEWEB)

Weber, S. V.; Casey, D. T.; Eder, D. C.; Pino, J. E.; Smalyuk, V. A.; Remington, B. A.; Rowley, D. P.; Yeamans, C. B.; Tipton, R. E.; Barrios, M.; Benedetti, R.; Berzak Hopkins, L.; Bleuel, D. L.; Bond, E. J.; Bradley, D. K.; Caggiano, J. A.; Callahan, D. A.; Cerjan, C. J.; Clark, D. S.; Divol, L. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); and others

2014-11-15

Gas-filled capsules imploded with indirect drive on the National Ignition Facility have been employed as symmetry surrogates for cryogenic-layered ignition capsules and to explore interfacial mix. Plastic capsules containing deuterated layers and filled with tritium gas provide a direct measure of mix of ablator into the gas fuel. Other plastic capsules have employed DT or D{sup 3}He gas fill. We present the results of two-dimensional simulations of gas-filled capsule implosions with known degradation sources represented as in modeling of inertial confinement fusion ignition designs; these are time-dependent drive asymmetry, the capsule support tent, roughness at material interfaces, and prescribed gas-ablator interface mix. Unlike the case of cryogenic-layered implosions, many observables of gas-filled implosions are in reasonable agreement with predictions of these simulations. Yields of TT and DT neutrons as well as other x-ray and nuclear diagnostics are matched for CD-layered implosions. Yields of DT-filled capsules are over-predicted by factors of 1.4–2, while D{sup 3}He capsule yields are matched, as well as other metrics for both capsule types.

Crystallization and preliminary X-ray diffraction analysis of ybfF, a new esterase from Escherichia coli K12

Energy Technology Data Exchange (ETDEWEB)

Park, Suk-Youl; Lee, Sang-Hak; Lee, Jieun; Jung, Che-Hun; Kim, Jeong-Sun, E-mail: jsunkim@chonnam.ac.kr [Department of Chemistry and Institute of Basic Sciences, Chonnam National University, Gwangju 500-757 (Korea, Republic of)

2007-12-01

The crystallization of ybfF, a new esterase from E. coli, and the collection of diffraction data to 1.1 Å resolution are reported. The product of the recently discovered ybfF gene, which belongs to the esterase family, does not show high sequence similarity to other esterases. To provide the molecular background to the enzymatic mechanism of the ybfF esterase, the ybfF protein from Escherichia coli K12 (Ec-ybfF) was cloned, expressed and purified. The Ec-ybfF protein was crystallized from 60% Tacsimate and 0.1 M bis-Tris propane buffer pH 7.0. Diffraction data were collected to 1.10 Å resolution using synchrotron radiation. The crystal belongs to the orthorhombic space group P2{sub 1}2{sub 1}2{sub 1}, with unit-cell parameters a = 66.09, b = 90.71, c = 92.88 Å. With two Ec-ybfF molecules in the asymmetric unit, the crystal volume per unit protein weight is 2.17 Å{sup 3} Da{sup −1}, corresponding to a solvent content of 42%.

Experimental Investigation of the Spiral Structure of a Magnetic Capsule Endoscope

Directory of Open Access Journals (Sweden)

Wanan Yang

2016-06-01

Full Text Available Fitting a wireless capsule endoscope (WCE with a navigation feature can maximize its functional benefits. The rotation of a spiral-type capsule can be converted to translational motion. The study investigated how the spiral structure and rotational speed affected the capsule's translation speed. A hand-held instrument, including two permanent magnets, a stepper motor, a controller and a power supplier, were designed to generate rotational magnetic fields. The surfaces of custom-built permanent magnet rings magnetized radially were mounted in spiral lines with different lead angles and diameters, acting as mock-up capsules. The experimental results demonstrate that the rotational speed of the magnetic field and the spiral have significant effects on the translational speed of a capsule. The spiral line with a larger lead angle and the rotating magnetic field with a higher speed can change the capsule's rotation into a translational motion more efficiently in the intestine.

Postirradiation gamma scans of GCFR capsule GB-10 at ORNL

International Nuclear Information System (INIS)

Tiegs, T.N.

1977-11-01

The Gas-Cooled Fast-Breeder Reactor capsule GB-10 was examined by gamma spectroscopy at Oak Ridge National Laboratory after fuel rod irradiation tests. The short-lived iodine fission products concentrated at the upper fuel-blanket interface, and cesium fission products concentrated at the fuel-blanket interfaces and in the charcoal trap. High concentrations of ruthenium isotopes were observed in the same positions at which neutron radiographs showed inclusions in the central void

Reducing the effects of X-ray pre-heat in double shell NIF capsules by over-coating the high Z shell

Science.gov (United States)

Wilson, Douglas; Milovich, J. L.; Daughton, W. S.; Loomis, E. N.; Sauppe, J. P.; Dodd, E. S.; Merritt, E. C.; Montgomery, D. S.; Renner, D. B.; Haines, B. M.; Cardenas, T.; Desjardins, T.; Palaniyappan, S.; Batha, S. H.

2017-10-01

Hohlraum generated X-rays will penetrate the ablator of a double shell capsule and be absorbed in the outer surface of the inner capsule. The ablative pressure this generates drives a shock into the central fuel, and a reflected shock that reaches the inner high-Z shell surface before the main shock even enters the fuel. With a beryllium over-coat preheat X-rays deposit just inside the beryllium/high z interface. The beryllium tamps the preheat expansion, eliminating ablation, and dramatically reducing pressure. The slow shock or pressure wave it generates is then overtaken by the main shock, avoiding an early shock in the fuel and increasing capsule yield.

Test reports for K Basins vertical fuel handling tools

Energy Technology Data Exchange (ETDEWEB)

Meling, T.A.

1995-02-01

The vertical fuel handling tools, for moving N Reactor fuel elements, were tested in the 305 Building Cold Test Facility (CTF) in the 300 Area. After fabrication was complete, the tools were functionally tested in the CTF using simulated N Reactor fuel rods (inner and outer elements). The tools were successful in picking up the simulated N Reactor fuel rods. These tools were also load tested using a 62 pound dummy to test the structural integrity of each assembly. The tools passed each of these tests, based on the performance objectives. Finally, the tools were subjected to an operations acceptance test where K Basins Operations personnel operated the tool to determine its durability and usefulness. Operations personnel were satisfied with the tools. Identified open items included the absence of a float during testing, and documentation required prior to actual use of the tools in the 100 K fuel storage basin.

Drying Results of K-Basin Fuel Element 6603M (Rune 5)

International Nuclear Information System (INIS)

Oliver, B.M.; Ritter, G.A.; Klinger, G.S.; Abrefah, J.; Greenwood, L.R.; MacFarlan, P.J.; Marschman, S.C.

1999-01-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium spent nuclear fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the fifth of those tests conducted on an N-Reactor outer fuel element (6603M) which had been stored underwater in the Hanford 100 Area K-West basin from 1983 until 1996. This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments which were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0. The test conditions and methodologies are given in Section 3.0. Inspections on the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0. Discussion of the results is given in Section 6.0

Drying Results of K-Basin Fuel Element 6603M (Rune 5)

Energy Technology Data Exchange (ETDEWEB)

B.M. Oliver; G.A. Ritter; G.S. Klinger; J. Abrefah; L.R. Greenwood; P.J. MacFarlan; S.C. Marschman

1999-09-24

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium spent nuclear fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the fifth of those tests conducted on an N-Reactor outer fuel element (6603M) which had been stored underwater in the Hanford 100 Area K-West basin from 1983 until 1996. This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments which were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0. The test conditions and methodologies are given in Section 3.0. Inspections on the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0. Discussion of the results is given in Section 6.0.

ROTARY FUEL INJECTION PUMP WEAR TESTING USING A 30 %/ 70% ATJ/F-24 FUEL BLEND

Science.gov (United States)

2017-09-30

DD-MM-YYYY) 30-09-2017 2. REPORT TYPE Interim Report 3. DATES COVERED (From - To) September 2013 – September 2017 4. TITLE AND SUBTITLE Rotary... Corrosion Inhibitor/Lubricity Improver cm...fuels, to full scale equipment and fleet testing to determine resulting component and vehicle performance. This report covers investigation into the

Fuel temperature prediction during high burnup HTGR fuel irradiation test. US-JAERI irradiation test for HTGR fuel

International Nuclear Information System (INIS)

Sawa, Kazuhiro; Fukuda, Kousaku; Acharya, R.

1995-01-01

This report describes the preirradiation thermal analysis of the HRB-22 capsule designed for an irradiation test in a removable beryllium position of the High Flux Isotope Reactor(HFIR) at Oak Ridge National Laboratory. This test is being carried out under Annex 2 of the Arrangement between the U.S. Department of Energy and the Japan Atomic Energy Research Institute on Cooperation in Research and Development regarding High-Temperature Gas-cooled Reactors. The fuel used in the test is an advanced type. The advanced fuel was designed aiming at burnup of about 10%FIMA(% fissions per initial metallic atom) which was higher than that of the first charge fuel for the High Temperature Engineering Test Reactor(HTTR) and was produced in Japan. CACA-2, a heavy isotope and fission product concentration calculational code for experimental irradiation capsules, was used to determine time-dependent fission power for the fuel compacts. The Heat Engineering and Transfer in Nine Geometries(HEATING) code was used to solve the steady-state heat conduction problem. The diameters of the graphite fuel body, which contains the fuel compacts, and of the primary pressure vessel were determined such that the requirements of running the fuel compacts at an average temperature less than 1250degC and of not exceeding a maximum fuel temperature of 1350degC were met throughout the four cycles of irradiation. The detail design of the capsule was carried out based on this analysis. (author)

Status of IVO-FR2-Vg7 experiment for irradiation of fast reactor fuel rods

International Nuclear Information System (INIS)

Elbel, H.; Kummerer, K.; Bojarsky, K.; Lopez Jimenez, J.; Otero de la Gandara, J.L.

1979-01-01

Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid) concerning a Joint Irradiation Program of Fast Reactor Fuel Rods. The design of fuel rods in general is defined, and, in particular of those with a density 94% DT and diameter 7.6 mm up to a burn-up of 7% FIMA, to be irradiated in the FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsules used in this irradiation, other possibilities of irradiation in the reactor will also be described. (auth.)

Instruments for reducing the specific fuel consumption of cars; Instrumente zur Absenkung des spezifischen Treibstoffverbrauchs von Personenwagen

Energy Technology Data Exchange (ETDEWEB)

Hammer, S.; Maibach, M. [Infras, Zuerich (Switzerland); Marti, P. [Metron, Brugg (Switzerland)

2001-07-01

This report for the Swiss Federal Office of Energy (SFOE) presents three possible courses of action that are to be taken to reduce the specific fuel consumption of private cars. The report first examines existing targets and the degree to which they have been met up to now, whereby the situation both in Switzerland and in the European Union is looked at. The report makes a suggestion for a future target scenario and elaborates three possible ways to met these targets: regulations on fuel-consumption, a bonus/malus system and tradable certificates. For each of the proposed instruments, the report examines implementation variants and discusses the means for their implementation. The report presents the best models for each of the implementation-variants on the basis of comparisons and the results of evaluations of their effects. For these chosen variants, the authors present comparisons of their effect on fuel consumption in graphical form and recommend tradable certificates as the best instrument.

EST Table: FS924814 [KAIKOcDNA[Archive

Lifescience Database Archive (English)

Full Text Available FS924814 E_FL_fwgP_11K20_F_0 10/09/28 39 %/114 aa ref|NP_001129360.1| osiris 9 [Bom...byx mori] gb|ACI23620.1| osiris 9 [Bombyx mori] 10/09/13 n.h 10/08/29 n.h 10/09/10 n.h 10/09/10 n.h 10/09/10 n.h FS935058 fwgP ...

EST Table: FS934649 [KAIKOcDNA[Archive

Lifescience Database Archive (English)

Full Text Available FS934649 E_FL_fwgP_40K10_F_0 10/09/28 31 %/210 aa ref|NP_001129360.1| osiris 9 [Bom...byx mori] gb|ACI23620.1| osiris 9 [Bombyx mori] 10/09/13 n.h 10/08/29 n.h 10/09/10 low homology 10/09/10 n.h 10/09/10 n.h FS923180 fwgP ...

EST Table: FS912585 [KAIKOcDNA[Archive

Lifescience Database Archive (English)

Full Text Available FS912585 E_FL_fufe_26K20_F_0 10/09/28 100 %/191 aa ref|NP_001098700.1| nanos-M [Bom...byx mori] dbj|BAF73619.1| nanos-M [Bombyx mori] 10/09/12 low homology 10/08/29 n.h 10/09/10 low homology 10/09/10 low homology 10/09/10 n.h FS918893 fufe ...

Impacts of Implosion Asymmetry And Hot Spot Shape On Ignition Capsules

Science.gov (United States)

Cheng, Baolian; Kwan, Thomas J. T.; Wang, Yi-Ming; Yi, S. Austin; Batha, Steve

2017-10-01

Implosion symmetry plays a critical role in achieving high areal density and internal energy at stagnation during hot spot formation in ICF capsules. Asymmetry causes hot spot irregularity and stagnation de-synchronization that results in lower temperatures and areal densities of the hot fuel. These degradations significantly affect the alpha heating process in the DT fuel as well as on the thermonuclear performance of the capsules. In this work, we explore the physical factors determining the shape of the hot spot late in the implosion and the effects of shape on Î+/-particle transport. We extend our ignition theory [1-4] to include the hot spot shape and quantify the effects of the implosion asymmetry on both the ignition criterion and capsule performance. We validate our theory with the NIF existing experimental data Our theory shows that the ignition criterion becomes more restrictive with the deformation of the hot spot. Through comparison with the NIF data, we demonstrate that the shape effects on the capsules' performance become more explicit as the self-heating and yield of the capsules increases. The degradation of the thermonuclear burn by the hot spot shape for high yield shots to date can be as high as 20%. Our theory is in good agreement with the NIF data. This work was performed under the auspices of the U.S. Department of Energy by the Los Alamos National Laboratory under Contract No. W-7405-ENG-36.

Irradiation Test in HANARO of the Parts of an X-Gen Nuclear Fuel Assembly

Energy Technology Data Exchange (ETDEWEB)

Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

2008-08-15

An instrumented capsule of 07M-13N was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of the parts of an X-Gen nuclear fuel assembly for PWR requested by KNF. Some specimens requested by Westinghouse Co. and Hanyang university were also inserted. 389 KNF specimens such as bucking and spring test specimens of 1x1 cell spacer grid, tensile, microstructure and tensile of welded parts, irradiation growth, spring test specimens made of HANA tube, Zirlo, Zircaloy-4, Inconel-718 were placed in the capsule. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 7 sets of Ni-Ti-Fe (2 sets contain additional Nb-Ag) neutron fluence monitors installed in the capsule. The capsule was irradiated for 59.19days (4 cycles) in the CT test hole of HANARO of a 30MW thermal output at 300 {approx} 420 .deg. C(for KNF specimens) up to a fast neutron fluence of 1.27x10{sup 21}(n/cm{sup 2}) (E>1MeV). After an irradiation test, the main body of the capsule was cut off at the bottom of the protection tube with a cutting system and it was transported to the IMEF (Irradiated Materials Examination Facility). The irradiated specimens were tested to evaluate the irradiation performance of the parts of an X-Gen fuel assembly in the IMEF hot cell.

Challenges and achievements of instrumentation for failed fuel identification in PFBR

International Nuclear Information System (INIS)

Upadhyay, Chandra Kant; Sivaramakrishna, M.; Nageswaran, A.; Nagaraj, C.P.; Madhusoodanan, K.; Krishnakumar, B.

2010-01-01

Failed fuel identification system is provided to locate and remove the failed fuel sub assembly. It comprises of selector valve mechanism to sample the flow from each assembly and associated instrumentation to detect the activity in the sample, indicating the clad failure of the respective subassembly. The development includes sampling pump, its power supply, motor to rotate the selector valve, encoder, brake, gear box, neutron detectors and Instrumentation and Control for all these with interlocks and related logic. These systems are expected to work under many physical constraints and in harsh environmental conditions, such as high temperature. This paper discusses the various challenges and achievements towards this system design. (author)

Electrochemical performance of Nd1.8Ce0.2CuO4+δ:Ce0.9Gd0.1O2 composite cathode for intermediate temperature solid oxide fuel cells

International Nuclear Information System (INIS)

Khandale, A.P.; Bhoga, S.S.

2012-01-01

Intermediate temperature solid oxide fuel cells (IT-SOFCs) are viewed as a promising power generation systems with high efficiency and low pollution. Recently, mixed ionic-electronic conductors (MIECs), with K 2 NiF 4 - type structure, attracted much attention as cathode for IT-SOFC

Development and testing of metallic fuels with high minor actinide content

International Nuclear Information System (INIS)

Meyer, M.K.; Hayes, S.L.; Kennedy, J.R.; Keiser, D.D.; Hilton, B.A.; Frank, S.M.; Kim, Y.-S.; Chang, G.; Ambrosek, R.G.

2003-01-01

Metallic alloys are promising candidates for use as fuels for transmutation and in advanced closed nuclear cycles. Metallic alloys have high heavy metal atom density, relatively high thermal conductivity, favorable gas release behavior, and lend themselves to remote recycle processes. Both non-fertile and uranium-bearing metal fuels containing minor actinide are under consideration for use as transmutation fuels by the U.S. Advanced Fuel Cycle (AFC) program, however, little irradiation performance data exists for fuel forms containing significant fractions of minor actinides. The first irradiation tests of non-fertile high-actinide-content fuels are scheduled to begin in early 2003 in the Advanced Test Reactor (ATR). The irradiation test matrix was designed to provide basic information on the irradiation behavior of binary Pu-Zr alloy fuel and the effect of the minor actinides americium and neptunium on alloy fuel behavior, together and separately. Five variants of transuranic containing zirconium-based alloy fuels are included in the AFC-1 irradiation test matrix. These are (in wt.%) Pu-40Zr, Pu-60Zr, Pu-12Am-40Zr, Pu-10Np-40Zr and Pu-10Np-10Am-40Zr. PuN-ZrN based fuels containing Am and Np are also included. All five of the fuel alloys have been fabricated in the form of cylindrical fuel slugs by arc-casting. Short melt times, on the order or 5-20 seconds, prevent the volatilization of significant quantities of americium metal, despite the high melt temperatures characteristic of the arc-melting process. Alloy microstructure have been characterized by x-ray diffraction and scanning electron microscopy. Thermal analysis has also been performed. The AFC-1 irradiation experiment configuration consists of twenty-four sodium bonded fuel specimens sealed in helium filled secondary capsules. The first capsule has a design burnup to 7 at.% 239 Pu; goal peak burnup of the second capsule is ∼18 at%. Capsule assemblies are placed within an aluminum flow-through basket

Seismic analysis for shroud facility in-pile tube and saturated temperature capsules

International Nuclear Information System (INIS)

Iimura, Koichi; Yamaura, Takayuki; Ogawa, Mitsuhiro

2009-07-01

At Oarai Research and Development Center, Japan Atomic Energy Agency (JAEA), the plan of repairing and refurbishing Japan Materials Testing Reactor (JMTR) has progressed in order to restart JMTR operation in the fiscal 2011. As a part of effective use of JMTR, the neutron irradiation tests of LWR fuels and materials has been planned in order to study their soundness. By using Oarai Shroud Facility (OSF-1) and Fuel Irradiation Facility with the He-3 gas control system for power lamping test using Boiling Water Capsules (BOCA Irradiation Facility), the irradiation tests with power ramping will be carried out to study the soundness of fuel under LWR Transient condition. OSF-1 is the irradiation facility of shroud type that can insert and eject the capsule under reactor operation, and is composed of 'In-pile Tube', 'Cooling system' and 'Capsule exchange system'. BOCA Irradiation Facility is the facility which simulates irradiation environment of LWR, and is composed of 'Boiling water Capsule', 'Capsule control system' and 'Power control system by He-3'. By using Saturated temperature Capsules and the water environment control system, the material irradiation tests under the water chemistry condition of LWR will be carried out to clarify the mechanism of IASCC. In JMTR, these facilities are in service at the present. However, the detailed design for renewal or remodeling was carried out based on the new design condition in order to be correspondent to the irradiation test plan after restart JMTR operation. In this seismic analysis of the detailed design, each equipment classification and operating state were arranged with 'Japanese technical standards of the structure on nuclear facility for test research' and 'Technical guidelines for seismic design of nuclear power plants on current, and then, stress calculation and evaluation were carried out by FEM piping analysis code 'SAP' and structure analysis code 'ABAQUS'. About the stress of the seismic force, it was proven

Status Report on Irradiation Capsules Designed to Evaluate FeCrAl-UO₂ Interactions

Energy Technology Data Exchange (ETDEWEB)

Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-06-24

This status report provides the background and current status of a series of irradiation capsules that were designed and are being built to test the interactions between candidate FeCrAl cladding for enhanced accident tolerant applications and prototypical enriched commercial UO₂ fuel in a neutron radiation environment. These capsules will test the degree, if any, of fuel cladding chemical interactions (FCCI) between FeCrAl and UO₂. The capsules are to be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory to burn-ups of 10, 30, and 50 GWd/MT with a nominal target temperature at the interfaces between the pellets and clad of 350°C.

Development of a Low Temperature Irradiation Capsule for Research Reactor Materials

International Nuclear Information System (INIS)

Choo, Kee Nam; Cho, Man Soon; Lee, Cheol Yong; Yang, Sung Woo; Shin, Yoon Taek; Park, Seng Jae; Kang, Suk Hoon; Kang, Young Hwan; Park, Sang Jun

2013-01-01

A new capsule design was prepared and tested at HANARO for a neutron irradiation of core materials of research reactors as a part of the research reactor development project. Irradiation testing of the materials including graphite, beryllium, and zircaloy-4 that are supposed to be used as core materials in research reactors was required for irradiation at up to 8 reactor operation cycles at low temperature (<100 .deg. C). Therefore, three instrumented capsules were designed and fabricated for an evaluation of the neutron irradiation properties of the core materials (Graphite, Be, Zircaloy-4) of research reactors. The capsules were first designed and fabricated to irradiate materials at low temperature (<100 .deg. C) for a long cycle of 8 irradiation cycles at HANARO. Therefore, the safety of the new designed capsule should be fully checked before irradiation testing. Out-pile performance and endurance testing before HANARO irradiation testing was performed using a capsule under a 110% condition of a reactor coolant flow amount. The structural integrity of the capsule was analyzed in terms of a vibration-induced fatigue cracking of a rod tip of the capsule that is suspected to be the most vulnerable part of a capsule. Another two capsules were irradiated at HANARO for 4 cycles, and one capsule was transferred to a hot cell to examine the integrity of the rod tip of the capsule. After confirming the soundness of the 4 cycle-irradiated capsule, the remaining capsule was irradiated at up to 8 cycles at HANARO. Based on the structural integrity analysis of the capsule, an improved capsule design will be suggested for a longer irradiation test at HANARO

AGR-2 Irradiated Test Train Preliminary Inspection and Disassembly First Look

Energy Technology Data Exchange (ETDEWEB)

Ploger, Scott [Idaho National Lab. (INL), Idaho Falls, ID (United States); Demkowciz, Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-05-01

The AGR 2 irradiation experiment began in June 2010 and was completed in October 2013. The test train was shipped to the Materials and Fuels Complex in July 2014 for post-irradiation examination (PIE). The first PIE activities included nondestructive examination of the test train, followed by disassembly of the test train and individual capsules and detailed inspection of the capsule contents, including the fuel compacts and their graphite fuel holders. Dimensional metrology was then performed on the compacts, graphite holders, and steel capsule shells. AGR 2 disassembly and metrology were performed with the same equipment used successfully on AGR 1 test train components. Gamma spectrometry of the intact test train gave a preliminary look at the condition of the interior components. No evidence of damage to compacts or graphite components was evident from the isotopic and gross gamma scans. Disassembly of the AGR 2 test train and its capsules was conducted rapidly and efficiently by employing techniques refined during the AGR 1 disassembly campaign. Only one major difficulty was encountered while separating the test train into capsules when thermocouples (of larger diameter than used in AGR 1) and gas lines jammed inside the through tubes of the upper capsules, which required new tooling for extraction. Disassembly of individual capsules was straightforward with only a few minor complications. On the whole, AGR 2 capsule structural components appeared less embrittled than their AGR 1 counterparts. Compacts from AGR 2 Capsules 2, 3, 5, and 6 were in very good condition upon removal. Only relatively minor damage or markings were visible using high resolution photographic inspection. Compact dimensional measurements indicated radial shrinkage between 0.8 to 1.7%, with the greatest shrinkage observed on Capsule 2 compacts that were irradiated at higher temperature. Length shrinkage ranged from 0.1 to 0.9%, with by far the lowest axial shrinkage on Capsule 3 compacts�

EdF speaks about economic advantages of fuel reprocessing as compared with interim storage

International Nuclear Information System (INIS)

Anon.

1997-01-01

The French company Electricite de France (EdF) will prefer nuclear fuel reprocessing and plutonium recycling to spent fuel storage also in the years after 2000. This option is economically advantageous if the proportional cost of reprocessing does not exceed 1900 FRF/kg heavy metal. Economic analysis shows that this is feasible. EdF will soon have to reprocess annually about 1000 Mt spent fuel to supply enough plutonium for MOX fuel fabrication to feed as many as 28 PWR units and the Superphenix reactor. Spent fuel reprocessing is seen as promising as long as the efficiency of the MOX fuel approaches that of natural uranium based fuel. The French national industrial, political and legal context of EdF operations is also considered. (P.A.)

Nondestructive inspection of General Purpose Heat Source (GPHS) fueled clad girth welds

International Nuclear Information System (INIS)

Reimus, M. A. H.; George, T. G.; Lynch, C.; Padilla, M.; Moniz, P.; Guerrero, A.; Moyer, M. W.; Placr, A.

1998-01-01

The General-Purpose Heat Source (GPHS) provides power for space missions by transmitting the heat of 238 Pu decay to an array of thermoelectric elements. The GPHS is fabricated using an iridium-alloy to contain the 238 PuO 2 fuel pellet. GPHS capsules will be utilized in the upcoming Cassini mission to explore Saturn and its moons. The physical integrity of the girth weld is important to mission safety and performance. Because past experience had revealed a potential for initiation of small cracks in the girth weld overlap zone, a nondestructive inspection of each capsule weld is required. An ultrasonic method was used to inspect the welds of capsules fabricated for the Galileo mission. The instrument, transducer, and method used were state of the art at the time (early 1980s). The ultrasonic instrumentation and methods used to inspect the Cassini GPHSs was significantly upgraded from those used for the Galileo mission. GPHSs that had ultrasonic reflectors in excess of the reject specification level were subsequently inspected with radiography to provide additional engineering data used to accept/reject the heat source. This paper describes the Galileo-era ultrasonic instrumentation and methods and the subsequent upgrades made to support testing of Cassini GPHSs. Also discussed is the data obtained from radiographic examination and correlation to ultrasonic examination results

Nondestructive inspection of General Purpose Heat Source (GPHS) fueled clad girth welds

International Nuclear Information System (INIS)

Reimus, M.A.; George, T.G.; Lynch, C.; Padilla, M.; Moniz, P.; Guerrero, A.; Moyer, M.W.; Placr, A.

1998-01-01

The General-Purpose Heat Source (GPHS) provides power for space missions by transmitting the heat of 238 Pu decay to an array of thermoelectric elements. The GPHS is fabricated using an iridium-alloy to contain the 238 PuO 2 fuel pellet. GPHS capsules will be utilized in the upcoming Cassini mission to explore Saturn and its moons. The physical integrity of the girth weld is important to mission safety and performance. Because past experience had revealed a potential for initiation of small cracks in the girth weld overlap zone, a nondestructive inspection of each capsule weld is required. An ultrasonic method was used to inspect the welds of capsules fabricated for the Galileo mission. The instrument, transducer, and method used were state of the art at the time (early 1980s). The ultrasonic instrumentation and methods used to inspect the Cassini GPHSs was significantly upgraded from those used for the Galileo mission. GPHSs that had ultrasonic reflectors in excess of the reject specification level were subsequently inspected with radiography to provide additional engineering data used to accept/reject the heat source. This paper describes the Galileo-era ultrasonic instrumentation and methods and the subsequent upgrades made to support testing of Cassini GPHSs. Also discussed is the data obtained from radiographic examination and correlation to ultrasonic examination results. copyright 1998 American Institute of Physics

On the knowledge of the (NH4)3SiF7 type of structure

International Nuclear Information System (INIS)

Hofmann, B.; Hoppe, R.

1979-01-01

New obtained are Rb 3 SiF 7 (a = 7.95 9 ; c = 5.82 3 A), Cs 3 SiF 7 (a = 8.30 6 ; c = 6.17 0 A), Rb 3 TiF 7 (a = 8.20 2 ; c = 5.97 9 A), Cs 3 TiF 7 (a = 8.47 3 ; c = 6.31 3 A). and Cs 2 RbSiF 7 (a = 8.19 8 ; c = 6.01 9 A); Cs 2 KSiF 7 (a = 8.11 5 , c = 5.97 2 A), Rb 2 CsSiF 7 (a = 8.09 9 ; c = 5.89 9 A), Rb 2 KSiF 7 (a = 7.88 3 ; c = 5.72 4 A), all colourless, Rb 3 CrF 7 (a = 8.08 4 ; c = 5.90 2 A), Cs 3 CrF 7 (a 8.39 0 ; c = 6.24 7 A), both pink, K 3 MnF 7 (a = 11.14 6 ; b = 11.00 5 ; c = 5.63 1 A), Rb 3 MnF 7 (a = 8.05 0 ; c = 5.89 0 A), Cs 3 MnF 7 (a 8.36 9 ; c = 6.23 3 A), each lemon yellow, and Rb 3 NiF 7 (a = 7.97 8 ; c = 5.85 7 A) and Cs 3 NiF 7 (a = 8.30 7 ; c = 6.19 2 A), both bright carmine red. Proposals for structure, due to Guinier-Simon powder datas (Cu-Kα), are based on the assumption that these fluorides (K 3 MnF 7 exception) are isotypic with K 3 SiF 7 , P4/mbm. Calculations of the Madelung Part of Lattice Energy, MAPLE, confirm the adopted parameters of position. Effective Coordination Numbers, ECoN, calculated by use of Mean Fictive Ionic Radii, MEFIR, are calculated and discussed. Raman spectra indicate the presence of octahedral groups [MF 6 ] in case of Rb 3 SiF 7 , and Cs 3 TiF 7 . The magnetic behaviour of A 3 MnF 7 is measured (70-293 K). (author)

Nuclear-powered pacemaker fuel cladding study

International Nuclear Information System (INIS)

Shoup, R.L.

1976-01-01

The composite of metals and alloys used in the fabrication of 238 Pu cardiac pacemaker fuel capsules resists the effects of high temperatures, high mechanical forces, and chemical corrosives and provides more than adequate protection to the fuel pellet even from deliberate attempts to dissolve the cladding in inorganic acids. This does not imply that opening a pacemaker fuel capsule by inorganic acids is impossible but that it would not be a wise choice

A screen-printed Ce 0.8Sm 0.2O 1.9 film solid oxide fuel cell with a Ba 0.5Sr 0.5Co 0.8Fe 0.2O 3- δ cathode

Science.gov (United States)

Zhang, Yaohui; Huang, Xiqiang; Lu, Zhe; Liu, Zhiguo; Ge, Xiaodong; Xu, Jiahuan; Xin, Xianshuang; Sha, Xueqing; Su, Wenhui

Screen-printing technology was developed to fabricate Ce 0.8Sm 0.2O 1.9 (SDC) electrolyte films onto porous NiO-SDC green anode substrates. After sintering at 1400 °C for 4 h, a gas-tight SDC film with a thickness of 12 μm was obtained. A novel cathode material of Ba 0.5Sr 0.5Co 0.8Fe 0.2O 3- δ was subsequently applied onto the sintered SDC electrolyte film also by screen-printing and sintered at 970 °C for 3 h to get a single cell. A fuel cell of Ni-SDC/SDC (12 μm)/Ba 0.5Sr 0.5Co 0.8Fe 0.2O 3- δ provides the maximum power densities of 1280, 1080, 670, 370, 180 and 73 mW cm -2 at 650, 600, 555, 505, 455 and 405 °C, respectively, using hydrogen as fuel and stationary air as oxidant. When dry methane was used as fuel, the maximum power densities are 876, 568, 346 and 114 mW cm -2 at 650, 600, 555 and 505 °C, respectively. The present fuel cell shows excellent performance at lowered temperatures.

Preliminary Hazards Analysis of K-Basin Fuel Encapsulation and Storage

International Nuclear Information System (INIS)

Strickland, G.C.

1994-01-01

This Preliminary Hazards Analysis (PHA) systematically examines the K-Basin facilities and their supporting systems for hazards created by abnormal operating conditions and external events (e.g., earthquakes) which have the potential for causing undesirable consequences to the facility worker, the onsite individual, or the public. The operational activities examined are fuel encapsulation, fuel storage and cooling. Encapsulation of sludges in the basins is not examined. A team of individuals from Westinghouse produced a set of Hazards and Operability (HAZOP) tables documenting their examination of abnormal process conditions in the systems and activities examined in K-Basins. The purpose of this report is to reevaluate and update the HAZOP in the original Preliminary Hazard Analysis of K-Basin Fuel Encapsulation and Storage originally developed in 1991

K-Basin spent nuclear fuel characterization data report

International Nuclear Information System (INIS)

Abrefah, J.; Gray, W.J.; Ketner, G.L.; Marschman, S.C.; Pyecha, T.D.; Thornton, T.A.

1995-11-01

The spent nuclear fuel (SNF) project characterization activities will be furnishing technical data on SNF stored at the K Basins in support of a pathway for placement of a ''stabilized'' form of SNF into an interim storage facility. This report summarizes the results so far of visual inspection of the fuel samples, physical characterization (e.g., weight and immersion density measurements), metallographic examinations, and controlled atmosphere furnace testing of three fuel samples shipped from the KW Basin to the Postirradiation Testing Laboratory (PTL). Data on sludge material collected by filtering the single fuel element canister (SFEC) water are also discussed in this report

Structure and selectivity trends in crystalline urea-functionalized anion-binding capsules

Energy Technology Data Exchange (ETDEWEB)

Rajbanshi, Arbin [Oak Ridge National Laboratory (ORNL); Custelcean, Radu [ORNL

2012-01-01

A tripodal trisurea receptor (L1) persistently self-assembles with various divalent oxoanion salts M{sub n}X (M = Na, K, Mg, Ca, Cd; X = SO{sub 4}{sup 2-}, SO{sub 3}{sup 2-}, SeO{sub 4}{sup 2-}, CrO{sub 4}{sup 2-}) into isomorphous series of crystalline frameworks in three different compositions: MX(L1){sub 2}(H{sub 2}O){sub 6} (M = Mg, Ca, Cd) (1), Na{sub 2}X(L1){sub 2}(H{sub 2}O){sub 4} (2) and K{sub 2}X(L1){sub 2}(H{sub 2}O){sub 2} (3). Single-crystal X-ray structural analysis revealed that all three series of structures adopt a NaCl-type topology, consisting of alternating anionic X(L1){sub 2}{sup 2-} capsules and M(H{sub 2}O){sub 6}{sup 2+}, Na{sub 2}(H{sub 2}O){sub 4}{sup 2+} or K{sub 2}(H{sub 2}O){sub 2}{sup 2+} hydrated cations. The capsules provide a complementary environment to tetrahedral oxoanions via 12 hydrogen bonds from six urea groups lining the cavities of the capsules. The persistent formation of the capsules facilitated the investigation of structural trends and structure-selectivity relationships across series 1-3. First, it was found that the size of the capsules is relatively unresponsive to the change in the encapsulated anion, resulting in good shape and size recognition in the separation of anions by competitive crystallizations. Second, it was found that the size of the capsules varies linearly with the size of the external cation, which provides a way for tuning the anion encapsulation selectivity. However, no straightforward dependence was found between the size of the capsules and the relative selectivity for different-sized tetrahedral oxoanions in competitive crystallizations.

Thermal conductivity of fresh and irradiated U-Mo fuels

Science.gov (United States)

Huber, Tanja K.; Breitkreutz, Harald; Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.; Elgeti, Stefan; Reiter, Christian; Robinson, Adam. B.; Smith, Frances. N.; Wachs, Daniel. M.; Petry, Winfried

2018-05-01

The thermal conductivity of fresh and irradiated U-Mo dispersion and monolithic fuel has been investigated experimentally and compared to theoretical models. During in-pile irradiation, thermal conductivity of fresh dispersion fuel at a temperature of 150 °C decreased from 59 W/m·K to 18 W/m·K at a burn-up of 4.9·1021 f/cc and further to 9 W/m·K at a burn-up of 6.1·1021 f/cc. Fresh monolithic fuel has a considerably lower thermal conductivity of 15 W/m·K at a temperature of 150 °C and consequently its decrease during in-pile irradiation is less steep than for dispersion fuel. For a burn-up of 3.5·1021 f/cc of monolithic fuel, a thermal conductivity of 11 W/m·K at a temperature of 150 °C has been measured by Burkes et al. (2015). The difference of decrease for both fuels originates from effects in the matrix that occur during irradiation, like for dispersion fuel the gradual disappearance of the Al matrix with increased burn-up and the subsequent growth of an interaction layer (IDL) between the U-Mo fuel particle and Al matrix and subsequent matrix hardening. The growth of fission gas bubbles and the decomposition of the U-Mo crystal lattice also affect both dispersion and monolithic fuel.

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

Review of consequences of uranium hydride formation in N-Reactor fuel elements stored in the K-Basins

Energy Technology Data Exchange (ETDEWEB)

Weber, J.W.

1994-09-28

The 105-K Basins on the Hanford site are used to store uranium fuel elements and assemblies irradiated in and discharged from N Reactor. The storage cylinders in KW Basin are known to have some broken N reactor fuel elements in which the exposed uranium is slowly reacting chemically with water in the cylinder. The products of these reactions are uranium oxide, hydrogen, and potentially some uranium hydride. The purpose of this report is to document the results f the latest review of potential, but highly unlikely accidents postulated to occur as closed cylinders containing N reactor fuel assemblies are opened under water in the KW basin and as a fuel assembly is raised from the basin in a shipping cask for transportation to the 327 Building for examination as part of the SNF Characterization Program. The postulated accidents reviews in this report are considered to bound all potential releases of radioactivity and hydrogen. These postulated accidents are: (1) opening and refill of a cylinder containing significant amounts of hydrogen and uranium hydride; and (2) draining of the single element can be used to keep the fuel element submerged in water after the cask containing the can and element is lifted from the KW Basin. Analysis shows the release of radioactivity to the site boundary is significantly less than that allowed by the K Basin Safety Evaluation. Analysis further shows there would be no damage to the K Basin structure nor would there be injury to personnel for credible events.

40 CFR 600.007-08 - Vehicle acceptability.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Vehicle acceptability. 600.007-08... FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy Regulations for 1977 and Later Model Year Automobiles-General Provisions § 600.007-08 Vehicle acceptability. (a) All...

kVp estimate intercomparison between Unfors XI, Radcal 4075 and a new CDTN multipurpose instrument

International Nuclear Information System (INIS)

Baptista Neto, A.T.; Oliveira, B.B.; Faria, L.O.

2015-01-01

In this work we compare the kVp estimate between CDTN multipurpose instrument, UnforsXI and Radcal 4075 meters under different combinations of voltage and filtration. The non-invasively measurements made using x-ray diagnostic and interventional radiology devices show similar tendencies to increase the kVp estimate when aluminum filters are placed in the path of the x-ray beam. The results reveal that the kVp estimate made by the CDTN multipurpose instrument is always satisfactory for highly filtered beam intensities. - Highlights: • We compare the kVp estimate between CDTN instrument and 2 different kVp meters. • The new CDTN multipurpose instrument performance was found to be satisfactory. • All instruments increase kVp estimative for increasing additional filtration. • They are suitable for quality control routines in x-ray diagnostic radiology

Highly nitrogen-doped carbon capsules: scalable preparation and high-performance applications in fuel cells and lithium ion batteries.

Science.gov (United States)

Hu, Chuangang; Xiao, Ying; Zhao, Yang; Chen, Nan; Zhang, Zhipan; Cao, Minhua; Qu, Liangti

2013-04-07

Highly nitrogen-doped carbon capsules (hN-CCs) have been successfully prepared by using inexpensive melamine and glyoxal as precursors via solvothermal reaction and carbonization. With a great promise for large scale production, the hN-CCs, having large surface area and high-level nitrogen content (N/C atomic ration of ca. 13%), possess superior crossover resistance, selective activity and catalytic stability towards oxygen reduction reaction for fuel cells in alkaline medium. As a new anode material in lithium-ion battery, hN-CCs also exhibit excellent cycle performance and high rate capacity with a reversible capacity of as high as 1046 mA h g(-1) at a current density of 50 mA g(-1) after 50 cycles. These features make the hN-CCs developed in this study promising as suitable substitutes for the expensive noble metal catalysts in the next generation alkaline fuel cells, and as advanced electrode materials in lithium-ion batteries.

HANARO fuel irradiation test(II)

Energy Technology Data Exchange (ETDEWEB)

Sohn, D. S.; Kim, H. R.; Chae, H. T.; Lee, B. C.; Lee, C. S.; Kim, B. G.; Lee, C. B.; Hwang, W

2001-04-01

In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiatied at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%.

Fuel fabrication instrumentation and control system overview

International Nuclear Information System (INIS)

Bennett, D.W.; Fritz, R.L.

1980-10-01

A process instrumentation and control system is being developed for automated fabrication of breeder reactor fuel at the Hanford Engineering Development Laboratory (HEDL) in Richland, Washington. The basic elements of the control system are a centralized computer system linked to distributed local computers, which direct individual process applications. The control philosophy developed for the equipment automation program stresses system flexibility and inherent levels of redundant control capabilities. Four different control points have been developed for each unit process operation

Test plan for K-Basin fuel handling tools

International Nuclear Information System (INIS)

Bridges, A.E.

1995-01-01

The purpose of this document is to provide the test plan and procedures for the acceptance testing of the handling tools enveloped for the removal of an N-Reactor fuel element from its storage canister in the K-Basins storage pool and insertion into the Single fuel Element Can for subsequent shipment to a Hot Cell for examination. Examination of these N-Reactor fuel elements is part of the overall characterization effort. New hand tools were required since previous fuel movement has involved grasping the fuel in a horizontal position. The 305 Building Cold Test Facility will be used to conduct the acceptance testing of the Fuel Handling Tools. Upon completion of this acceptance testing and any subsequent training of operators, the tools will be transferred to the 105 KW Basin for installation and use

The upgraded control and instrumentation system of C9 irradiation device

International Nuclear Information System (INIS)

Pulpa, A.; Parvu, C.; Valeca, S. C.; Ancuta, M.; Ana, E.

2016-01-01

C9 Capsule is an irradiation device of TRIGA SSR, which was designed for nuclear fuel cycling testing. It simulates the load follow-up by the power reactor, i.e. CANDU 700MW operating with variable load. The irradiation tests in the power cycling conditions were conceived for complete characterization of the fuel behaviour. The irradiation conditions are similar to those found in nuclear power plant when it is operated in load following mode. The power cycling tests are performed by using an under flux moving system. The paper presents the upgraded control and instrumentation system of the ''9C irradiation device, designed and manufactured in order to enhance the performance of this system for better surveillance and processing of the acquired experimental data. (authors)

Postirradiation examination of capsules P13R and P13S

International Nuclear Information System (INIS)

Scott, C.B.; Harmon, D.P.; Holzgraf, J.F.

1976-01-01

Capsules P13R and P13S were the seventh and eighth in a series of irradiation tests conducted under the ERDA-sponsored HTGR Fuels and Core Development Program. Reference type LHTGR fuel fabricated with a broad spectrum of property and process variables was irradiated to extreme temperature and fluence conditions. Postirradiation examination revealed that the bonded fuel rods exhibited good stability after irradiation to fast neutron fluences of 12.4 x 10 21 n/cm 2 (E greater than 0.18 MeV), which is 55 percent beyond the LHTGR peak design fast neutron fluence of 8.0 x 10 21 n/cm 2 . Thermal cycling to high temperatures did not adversely affect fuel rod integrity. Particle batches with coating designs representative of the design requirements envisioned for the LHTGR exhibited excellent irradiation performance. Ten batches of fissile and fertile particles were irradiated without coating failure to fast neutron exposures which exceeded the LHTGR peak design exposure by 35 to 52 percent. Capsules P13R and P13S were considered to be very successful qualification tests of LHTGR fuel components. These results provided a substantial data base for the LHTGR Fuel Product Specification and Performance Models used in HTGR core design studies, and demonstrated the excellent irradiation performance of reference LHTGR fuel to well beyond peak design exposures

40 kW Stirling Engine for Solid Fuel

DEFF Research Database (Denmark)

Carlsen, Henrik; Trærup, Jens

1996-01-01

The external combustion in a Stirling engine makes it very attractive for utilisation of solid fuels in decentralised combined heat and power (CHP) plants. Only a few projects have concentrated on the development of Stirling engines specifically for biomass. In this project, a Stirling engine has...... been designed primarily for utilisation of wood chips. Maximum shaft power is 40 kW corresponding to an electric output of 36 kW. Biomass needs more space in the combustion chamber compared to gas and liquid fuels, and a large heat transfer area is necessary. The design of the new Stirling engine has...... been adapted to the special demands of combustion of wood chips, resulting in a large engine compared to engines for gas or liquid fuels. The engine has four-cylinders arranged in a square. The design is made as a hermetic unit, where the alternator is built into the pressurised crankcase so...

Muonium chemistry: kinetics of the gas phase reaction Mu + F/sub 2/. -->. MuF + F from 300 to 400 K

Energy Technology Data Exchange (ETDEWEB)

Garner, D M; Fleming, D G; Brewer, J H [British Columbia Univ., Vancouver (Canada). Faculty of Medicine

1978-04-01

The MSR (muonium spin rotation) technique was used to measure the chemical reaction rate for ..mu.. + F/sub 2/ ..-->.. ..mu..F + F in N/sub 2/ moderator at approximately 1 atm from 295 to 383 K giving the Arrhenius expression: log/sub 10/k (l/mole s) = (10.83+-0.20) - (200+-50)/T, with k = (1.46+-0.11) X 10/sup 10/ l/mole s at 300 K. This is a least 6.8 times the room temperature rate constant for the analogous H atom reaction. The measured activation energy and enhancement over the H reaction rate are indicative of significant tunnelling in the Mu reaction, in agreement with the recent collinear quantum mechanical calculations of Connor et al.

Dynamic modeling and experimental investigation of a high temperature PEM fuel cell stack

DEFF Research Database (Denmark)

Nguyen, Gia; Sahlin, Simon Lennart; Andreasen, Søren Juhl

2016-01-01

High temperature polymer fuel cells operating at 100 to 200◦C require simple fuel processing and produce high quality heat that can integrate well with domestic heating systems. Because the transportation of hydrogen is challenging, an alternative option is to reform natural gas on site....... This article presents the development of a dynamic model and the comparison with experimental data from a high temperature proton exchange membrane fuel cell stack operating on hydrogen with carbon monoxide concentrations up to 0.8%, and temperatures from 155 to 175◦C. The dynamic response of the fuel cell...... is investigated with simulated reformate gas. The dynamic response of the fuel cell stack was compared with a step change in current from 0.09 to 0.18 and back to 0.09 A/cm2 . This article shows that the dynamic model calculates the voltage at steady state well. The dynamic response for a change in current shows...

327 SNF fuel return to K-Basin quality process plan

International Nuclear Information System (INIS)

Ham, J.E.

1998-01-01

The B and W Hanford Company's (BWHC) 327 Facility, in the 300 Area of the Hanford Site, contains Spent Nuclear Fuel (SNF) single fuel element canisters (SFEC) and fuel remnant canisters (FRC) which are to be returned to K-Basin. Seven shipments of up to six fuel canisters will be loaded into the CNS 1-13G Cask and transported to 105-KE

K basins sludge removal sludge pretreatment system

International Nuclear Information System (INIS)

Chang, H.L.

1997-01-01

The Spent Nuclear Fuels Program is in the process of planning activities to remove spent nuclear fuel and other materials from the 100-K Basins as a remediation effort for clean closure. The 105 K- East and K-West Basins store spent fuel, sludge, and debris. Sludge has accumulated in the 1 00 K Basins as a result of fuel oxidation and a slight amount of general debris being deposited, by settling, in the basin water. The ultimate intent in removing the sludge and fuel is to eliminate the environmental risk posed by storing fuel at the K Basins. The task for this project is to disposition specific constituents of sludge (metallic fuel) to produce a product stream through a pretreatment process that will meet the requirements, including a final particle size acceptable to the Tank Waste Remediation System (TWRS). The purpose of this task is to develop a preconceptual design package for the K Basin sludge pretreatment system. The process equipment/system is at a preconceptual stage, as shown in sketch ES-SNF-01 , while a more refined process system and material/energy balances are ongoing (all sketches are shown in Appendix C). Thus, the overall process and 0535 associated equipment have been conservatively selected and sized, respectively, to establish the cost basis and equipment layout as shown in sketches ES- SNF-02 through 08

Plasma barodiffusion in inertial-confinement-fusion implosions: application to observed yield anomalies in thermonuclear fuel mixtures.

Science.gov (United States)

Amendt, Peter; Landen, O L; Robey, H F; Li, C K; Petrasso, R D

2010-09-10

The observation of large, self-generated electric fields (≥10(9)  V/m) in imploding capsules using proton radiography has been reported [C. K. Li, Phys. Rev. Lett. 100, 225001 (2008)]. A model of pressure gradient-driven diffusion in a plasma with self-generated electric fields is developed and applied to reported neutron yield deficits for equimolar D3He [J. R. Rygg, Phys. Plasmas 13, 052702 (2006)] and (DT)3He [H. W. Herrmann, Phys. Plasmas 16, 056312 (2009)] fuel mixtures and Ar-doped deuterium fuels [J. D. Lindl, Phys. Plasmas 11, 339 (2004)]. The observed anomalies are explained as a mild loss of deuterium nuclei near capsule center arising from shock-driven diffusion in the high-field limit.

Definition of multi-state weighted k-out-of-n: F systems

DEFF Research Database (Denmark)

Ding, Yi; Wu, Qiuwei; Zio, Enrico

2012-01-01

of the Multi-state Weighted k-out-of-n: G System- the Multi-state Weighted k-out-of-n: F System has not been clearly defined and discussed. In this short communication, the basic definition of the Multi-state Weighted k-out-of-n: F System model is proposed. The relationship between the Multi-state Weighted k...

Temperature measurement of the reactor materials samples irradiated in the fuel channels of the RA reactor - Annex 16

International Nuclear Information System (INIS)

Nikolic, M.; Djalovic, M.

1964-01-01

Reactor materials as graphite, stainless steel, magnox, zirconium alloys, etc. were exposed to fast neutron flux inside the fuel elements specially adapted for this purpose. Samples in the form ampoules were placed in capsules inside the fuel channels and cooled by heavy water which cools the fuel elements. In order to monitor the samples temperature 42 thermocouples were placed in the samples. That was necessary for reactor safety reasons and for further interpretation of measured results. Temperature monitoring was done continuously by multichannel milivoltmeters. This paper describes the technique of introducing the thermocouples, compensation instruments, control of the cold ends and adaptation of the instruments for precision (0.5%) temperature measurement in the range 30 deg - 130 deg C; 30 deg - 280 deg C and 30 deg - 80 deg C [sr

Musik, sprog og integration. Evalueringsrapport 08-09

DEFF Research Database (Denmark)

Christensen, Finn Holst

2009-01-01

Projektet er rettet mod at styrke børnenes musikalske udvikling samt at bidrage til udviklingen af deres personlige, sociale og sproglige kompetencer i rammen af en kompetent musikfaglig undervisning. Krop og bevægelse spiller en central rolle, især i overgangen mellem førskole og skole, med særl...

Strategy for phase 2 whole element furnace testing K West fuel

International Nuclear Information System (INIS)

Lawrence, L.A.

1998-01-01

A strategy was developed for the second phase of the whole element furnace testing of damaged fuel removed from the K West Basin. The Phase 2 testing can be divided into three groups covering oxidation of whole element in moist inert atmospheres, drying elements for post Cold Vacuum Drying staging tests, and drying additional K West elements to provide confirmation of the results from the first series of damaged K West fuel drying studies

Instrumentation of fuel safety test rods of the PWR system in the Phebus reactor

International Nuclear Information System (INIS)

Schley, Robert; Leveque, J.P.; Aujollet, J.M.; Dutraive, Pierre; Colome, Jean; Bouly, J.C.

1979-01-01

The tests were performed in an experimental cell centred in the core of the PHEBUS water reactor of 50 MW. The CEA make two types of apparatus for testing the safety of PWR fuel. One is for testing a single fuel stick and the other a bunch of 25 sticks. The instrumentation described enables the main parameters of the test to be known: temperatures of the fuel - central temperature of the UO 2 - cladding surface temperatures; temperature of the cooling circuits - thermal balance - temperatures of the structures, etc.; coolant pressure; internal pressure of the fuel sticks; direction and flow rate of the fluid. This instrumentation and the technological problems to be overcome are described and the results of the first tests carried out are given [fr

Operation and postirradiation examination of ORR capsule OF-2: accelerated testing of HTGR fuel

International Nuclear Information System (INIS)

Tiegs, T.N.; Thoms, K.R.

1979-03-01

Irradiation capsule OF-2 was a test of High-Temperature Gas-Cooled Reactor fuel types under accelerated irradiation conditions in the Oak Ridge Research Reactor. The results showed good irradiation performance of Triso-coated weak-acid-resin fissile particles and Biso-coated fertile particles. These particles had been coated by a fritted gas distributor in the 0.13-m-diam furnace. Fast-neutron damage (E > 0.18 MeV) and matrix-particle interaction caused the outer pyrocarbon coating on the Triso-coated particles to fail. Such failure depended on the optical anisotropy, density, and open porosity of the outer pyrocarbon coating, as well as on the coke yield of the matrix. Irradiation of specimens with values outside prescribed limits for these properties increased the failure rate of their outer pyrocarbon coating. Good irradiation performance was observed for weak-acid-resin particles with conversions in the range from 15 to 75% UC 2

Avaliação de cápsulas de cerâmica e instrumentos de medida de tensão usados em tensiômetros Evaluation of ceramic capsules and instruments of tension measurement used in tensiometers

Directory of Open Access Journals (Sweden)

Neuzo B. de Moraes

2006-03-01

provided same tension values, although the capsules from Ceara state presented higher conductance than the ones from USA. The tension values obtained by the instruments of tension measurements differed statistically, although the difference was within the precision (1.0 kPa. A high correlation exists among the tensions obtained by the two instruments of tension measurement.

Electric conductivity of salt melts containing KCL, KF and K2TaF7

International Nuclear Information System (INIS)

Agulyanskij, A.I.; Stangrit, P.T.; Konstantinov, V.I.

1978-01-01

Given are electric conductivity measurement results depending on the temperature and composition of the molten KF-K 2 TaF 7 , KCl-K 2 TaF 7 systems and also melts close in their composition to industrial electrolytes, KCl-KF (in mass ratio of 2:1) with addition of K 2 TaF 7 up to 25 mass%. Presented are electric conductivity molecular isotherms of the KF-K 2 TaF 7 , KCl-K 2 TaF 7 systems at 800 deg C and specific electric conductivity dependence of KCl-KF-K 2 TaF 7 melts on K 2 TaF 7 composition at 800 deg C and 900 deg C. Proceeding from the shape of molecular and specific electric conductivity isotherms a conclusion is made about existence of the following tantalum-containing ions: TaF 7 2- , TaF 6 - and TaF 6 Cl 2- in the investigated melts

Risk Characterization for Future Training Scenarios at the Massachusetts Military Reservation (MMR), Final Results

Science.gov (United States)

2005-08-01

1979) 0.63d ocK f= owK (1) where Kow = octanol to water partitioning coefficient, L/kg foc = fraction of organic carbon of the soil Barber...Software 59 2-Methylthiophene 554-14-3 0.135 EPI Software 60 2-Nitrophenol 88-75-5 0.039 EPI Software 61 2- Octanone 111-13-7 0.148 EPI Software 62 2...Nitrophenol 88-75-5 6.09E-09 1.72E-12 2- Octanone 111-13-7 2.56E-11 7.23E-15 2-Pentanone 107-87-9 1.83E-08 5.18E-12 2-Propanol 67-63-0 4.08E-09 1.15E-12 2

Laboratory comparison of the mechanical properties of TRUShape with several nickel-titanium rotary instruments.

Science.gov (United States)

Elnaghy, A M; Elsaka, S E

2017-08-01

To assess and compare the mechanical properties of TRUShape (TRS) with several nickel-titanium rotary instruments. Cyclic fatigue, torsional resistance, flexibility and surface microhardness of TRS (size 25, 0.06v taper), ProTaper Next X2 (PTN X2, size 25, 0.06 taper), ProTaper Gold (PTG F2; size 25, 0.08 taper) and ProTaper Universal (PTU F2; size 25, 0.08 taper) instruments were evaluated. The topographical structures of the fracture surfaces of instruments were assessed using a scanning electron microscope. The cyclic fatigue resistance, torsional resistance and microhardness data were analysed using one-way analysis of variance (anova) and Tukey's post hoc tests. The fragment length and bending resistance data were analysed statistically with the Kruskal-Wallis H-test and Mann-Whitney U-tests. The statistical significance level was set at P instruments revealed significantly higher resistance to cyclic fatigue than TRS and PTU instruments (P instruments revealed significantly higher torsional resistance compared with the other instruments (P instrument had significantly higher flexibility than the other tested brands (P instruments had lower resistance to cyclic fatigue and lower flexibility compared with PTG and PTN instruments. TRS, PTG and PTU instruments had lower resistance to torsional stress than PTN instruments. TRS and PTG instruments had comparable surface microhardness. © 2016 International Endodontic Journal. Published by John Wiley & Sons Ltd.

Saturation behavior of irradiation hardening in F82H irradiated in the HFIR

Energy Technology Data Exchange (ETDEWEB)

Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Shiba, K.; Tanigawa, H.; Ando, M. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Klueh, R.L. [Oak Ridge National Laboratory, TN (United States); Stoller, R. [ORNL - Oak Ridge National Laboratory, Materials Science and Technology Div., Oak Ridge, AK TN (United States)

2007-07-01

Full text of publication follows: Post irradiation tensile tests on reduced activation ferritic/martensitic steel, F82H have been conducted over the past two decades using Japan Materials Testing Reactor (JMTR) of JAEA, and Fast Flux Testing Facility (FFTF) of PNNL and High Flux Isotope Reactor (HFIR) of ORNL, USA, under Japan/US collaboration programs. According to these results, F82H does not demonstrate irradiation hardening above 673 K up to 60 dpa. The current study has been concentrated on hardening behavior at temperature around 573 K. A series of low temperature irradiation experiment has been conducted at the HFIR under the international collaborative research between JAEA/US-DOE. In this collaboration, the irradiation condition is precisely controlled by the well matured capsule designing and instrumentation. This paper summarizes recent results of the irradiation experiments focused on F82H and its modified steels compared with the irradiation properties database on F82H. Post irradiation tensile tests have been conducted on the F82H and its modified steels irradiated at 573 K and the dose level was up to 25 dpa. According to these results, irradiation hardening of F82H is saturated by 9 dpa and the as-irradiated 0.2 % proof stress is less than 1 GPa at ambient temperature. The deterioration of total elongation was also saturated by 9 dpa irradiation. The ductility of some modified steels which showed larger total elongation than that of F82H before irradiation become the same level as that of standard F82H steel after irradiation, even though its magnitude of irradiation hardening is smaller than that of F82H. This suggests that the more ductile steel demonstrates the more ductility loss at this temperature, regardless to the hardening level. The difference in ductility loss behavior between various tensile specimens will be discussed as the ductility could depend on the specimen dimension. (authors)

Description of fuel element brush assembly's fabrication for 105-K west

International Nuclear Information System (INIS)

Maassen, D.P.

1997-01-01

This report is a description of the process to redesign and fabricate, as well as, describe the features of the Fuel Element Brush Assembly used in the 105-K West Basin. This narrative description will identify problems that occurred during the redesigning and fabrication of the 105-K West Basin Fuel Element Brush Assembly and specifically address their solutions

Dry storage of Magnox fuel

International Nuclear Information System (INIS)

1986-09-01

This work, commissioned by the CEGB, studies the feasibility of a combination of short-term pond storage and long-term dry storage of Magnox spent fuel as a cheaper alternative to reprocessing. Storage would be either at the reactor site or a central site. Two designs are considered, based on existing design work done by GEC-ESL and NNC; the capsule design developed by NNC and with storage in passive vaults for up to 100 yrs and the GEC-ESL tube design developed at Wylfa for the interim storage of LWR. For the long-term storage of Magnox spent fuel the GEC-ESL tubed vault all-dry storage method is recommended and specifications for this method are given. (U.K.)

Characterization program management plan for Hanford K Basin Spent Nuclear Fuel

International Nuclear Information System (INIS)

Lawrence, L.A.

1995-01-01

A management plan was developed for Westinghouse Hanford Company (WHC) and Pacific Northwest Laboratories (PNL) to work together on a program to provide characterization data to support removal, conditioning and subsequent dry storage of the spent nuclear fuels stored at the Hanford K Basins. The Program initially supports gathering data to establish the current state of the fuel in the two basins. Data Collected during this initial effort will apply to all SNF Project objectives. N Reactor fuel has been degrading with extended storage resulting in release of material to the basin water in K East and to the closed conisters in K West. Characterization of the condition of these materials and their responses to various conditioning processes and dry storage environments are necessary to support disposition decisions. Characterization will utilize the expertise and capabilities of WHC and PNL organizations to support the Spent Nuclear Fuels Project goals and objectives. This Management Plan defines the structure and establishes the roles for the participants providing the framework for WHC and PNL to support the Spent Nuclear Fuels Project at Hanford

Enteric coated HPMC capsules plugged with 5-FU loaded microsponges: a potential approach for treatment of colon cancer

Directory of Open Access Journals (Sweden)

Ankita Gupta

2015-09-01

Full Text Available The work was aimed at developing novel enteric coated HPMC capsules (ECHC plugged with 5 Florouracil (5-FU loaded Microsponges in combination with calcium pectinate beads. Modified quasi-emulsion solvent diffusion method was used to formulate microsponges based on 32 factorial design and the effects of independent variables (volume of organic solvent and Eudragit RS100 content on the dependent variables (Particle size, %EE & % CDR were determined. The optimized microsponges (F4 were characterized by SEM, PXRD, TGA and were plugged along with calcium pectinate beads in HPMC capsules and the HPMC capsules were further coated with enteric polymer Eudragit L 100 (Ed-L100 and/ or Eudrgit S 100 (Ed-S 100 in different proportions. In vitro release study of ECHC was performed in various release media sequentially SGF for 2 h, followed by SIF for the next 6 h and then in SCF (in the presence and absence of pectinase enzyme for further 16 h. Drug release was retarded on coating with EdS-100 in comparison to blend of EdS-100: EdL-100 coating. The percentage of 5-FU released at the end of 24 h from ECHC 3 was 97.83 ± 0.12% in the presence of pectinase whereas in control study it was 40.08 ± 0.02% drug. The optimized formulation was subjected to in vivo Roentgenographic studies in New Zealand white rabbits to analyze the in vivo behavior of the developed colon targeted capsules. Pharmacokinetic studies in New Zealand white rabbits were conducted to determine the extent of systemic exposure provided by the developed formulation in comparison to 5-FU aqueous solutions. Thus, enteric coated HPMC capsules plugged with 5-FU loaded microsponges and calcium pectinate beads proved to be promising dosage form for colon targeted drug delivery to treat colorectal cancer.

Improvement of the Ca determination accuracy with k (0)-INAA using an HPGe coaxial detector with extended energy range efficiency calibration

Czech Academy of Sciences Publication Activity Database

Kučera, Jan; Kubešová, Marie; Lebeda, Ondřej

2018-01-01

Roč. 315, č. 3 (2018), s. 671-675 ISSN 0236-5731. [7th International K0-Users Workshop. Montreal, 03.09.2017-08.09.2017] R&D Projects: GA ČR(CZ) GBP108/12/G108; GA MŠk LM2015056 Institutional support: RVO:61389005 Keywords : k(0)-INAA * Ca determination * HPGe detector * High-energy efficiency calibration * Co-56 activity standard Subject RIV: CB - Analytical Chemistry, Separation OBOR OECD: Analytical chemistry Impact factor: 1.282, year: 2016

The high temperature out-of-pile test of LVDT for elongation measurement of fuel pellet

Energy Technology Data Exchange (ETDEWEB)

Son, J. M.; Kim, B. K.; Jo, M. S.; Joo, K. N.; Park, S. J.; Gang, Y. H.; Kim, Y. J. [KAERI, Taejon (Korea, Republic of)

2003-10-01

As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), the elongation measurement technique of the fuel pellet is being developed using LVDT(Linear Variable Differential Transformer). The well qualified out-of-pile test were needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation instrumented capsule, because LVDT is very sensitive to variation of temperature. Therefore, the high temperature out-of-pile test system for fuel pellet elongation was developed, and this test was performed under the temperature condition between room temperature and 300 .deg. C with increasing the elongation from 0 to 5 mm. The LVDT's high temperature characteristics and temperature sensitivity of LVDT were analyzed through this experiment. Based on the result of this test, the method for the application of LVDT and elongation detector at high temperature was introduced. It is known that the results will be used to predict accurately the elongation of fuel pellet during irradiation test.

Characteristics of X ray calibration fields for performance test of radiation measuring instruments

International Nuclear Information System (INIS)

Shimizu, Shigeru; Takahashi, Fumiaki; Sawahata, Tadahiro; Tohnami, Kohichi; Kikuchi, Hiroshi; Murayama, Takashi

1999-02-01

Performance test and calibration of the radiation measuring instruments for low energy photons are made using the X ray calibration fields which are monochromatically characterized by filtration of continuous X ray spectrum. The X ray calibration field needs to be characterized by some quality conditions such as quality index and homogeneity coefficient. The present report describes quality conditions, spectrum and some characteristics of X ray irradiation fields in the Facility of Radiation Standard of the Japan Atomic Energy Research Institute (FRS-JAERI). Fifty nine X ray qualities with the quality index of 0.6, 0.7, 0.8 and 0.9 were set for the tube voltages between 10 kV and 350 kV. Estimation of X ray spectrum measured with a Ge detector was made in terms of exposure, ambient dose equivalent and fluence for all the obtained qualities. Practical irradiation field was determined as the dose distribution uniformity is within ±3%. The obtained results improve the quality of X ray calibration fields and calibration accuracy. (author)

Prompt $K_{S}^{0}$ production in $pp$ collisions at $\\sqrt{s}$ = 0.9 TeV

CERN Document Server

Aaij, R; Adeva, B; Adinolfi, M; Adrover, C; Affolder, A; Agari, M; Ajaltouni, Z; Albrecht, J; Alessio, F; Alexander, M; Alfonsi, M; Alvarez Cartelle, P; Alves Jr, A.A; Amato, S; Amhis, Y; Amoraal, J; Anderson, J; Antunes Nobrega, R; Appleby, R; Aquines Gutierrez, O; Arefyev, A; Arrabito, L; Artuso, M; Aslanides, E; Auriemma, G; Bachmann, S; Bagaturia, Y; Bailey, D.S; Balagura, V; Baldini, W; Barber, G; Barham, C; Barlow, R.J; Barsuk, S; Basiladze, S; Bates, A; Bauer, C; Bauer, Th; Bay, A; Bediaga, I; Bellunato, T; Belous, K; Belyaev, I; Benayoun, M; Bencivenni, G; Bernet, R; Bernhard, R.P; Bettler, M-O; van Beuzekom, M; Bibby, J.H; Bifani, S; Bizzeti, A; Bjrnstad, P.M; Blake, T; Blanc, F; Blanks, C; Blouw, J; Blusk, S; Bobrov, A; Bocci, V; Bochin, B; Bonaccorsi, E; Bondar, A; Bondar, N; Bonivento, W; Borghi, S; Borgia, A; Bos, E; Bowcock, T.J.V; Bozzi, C; Brambach, T; van den Brand, J; Brarda, L; Bressieux, J; Brisbane, S; Britsch, M; Brook, N.H; Brown, H; Brusa, S; Buchler-Germann, A; Bursche, A; Buytaert, J; Cadeddu, S; Caicedo Carvajal, J.M; Callot, O; Calvi, M; Calvo Gomez, M; Camboni, A; Cameron, W; Camilleri, L; Campana, P; Carbone, A; Carboni, G; Cardinale, R; Cardini, A; Carroll, J; Carson, L; Carvalho Akiba, K; Casse, G; Cattaneo, M; Chadaj, B; Charles, M; Charpentier, Ph; Cheng, J; Chiapolini, N; Chlopik, A; Christiansen, J; Ciambrone, P; Cid Vidal, X; Clark, P.J; Clarke, P.E.L; Clemencic, M; Cliff, H.V; Closier, J; Coca, C; Coco, V; Cogan, J; Collins, P; Comerma-Montells, A; Constantin, F; Conti, G; Contu, A; Cooke, P; Coombes, M; Corajod, B; Corti, G; Cowan, G.A; Currie, R; DAlmagne, B; DAmbrosio, C; DAntone, I; Da Silva, W; Dane, E; David, P; De Bonis, I; De Capua, S; De Cian, M; De Lorenzi, F; De Miranda, J.M; De Paula, L; De Simone, P; Decamp, D; Decreuse, G; Degaudenzi, H; Deissenroth, M; Del Buono, L; Densham, C.J; Deplano, C; Deschamps, O; Dettori, F; Dickens, J; Dijkstra, H; Dima, M; Donleavy, S; Dornan, P; Dossett, D; Dovbnya, A; Dumps, R; Dupertuis, F; Dwyer, L; Dzhelyadin, R; Eames, C; Easo, S; Egede, U; Egorychev, V; Eidelman, S; van Eijk, D; Eisele, F; Eisenhardt, S; Eklund, L; d'Enterria, D; Esperante Pereira, D; Est`eve, L; Fanchini, E; Farber, C; Fardell, G; Farinelli, C; Farry, S; Fave, V; Felici, G; Fernandez Albor, V; Ferro-Luzzi, M; Filippov, S; Fitzpatrick, C; Flegel, W; Fontanelli, F; Forti, C; Forty, R; Fournier, C; Franek, B; Frank, M; Frei, C; Frosini, M; Fungueirino Pazos, J.L; Furcas, S; Gallas Torreira, A; Galli, D; Gandelman, M; Gandini, P; Gao, Y; Garnier, J-C; Garrido, L; Gascon, D; Gaspar, C; Gaspar De Valenzuela Cue, A; Gassner, J; Gauvin, N; Gavillet, P; Gersabeck, M; Gershon, T; Ghez, Ph; Gibson, V; Gilitsky, Yu; Gligorov, V.V; Gobel, C; Golubkov, D; Golutvin, A; Gomes, A; Gong, G; Gong, H; Gordon, H; Grabalosa Gandara, M; Gracco, V; Graciani Diaz, R; Granado Cardoso, L.A; Grauges, E; Graziani, G; Grecu, A; Gregson, S; Guerrer, G; Gui, B; Gushchin, E; Guz, Yu; Guzik, Z; Gys, T; Haefeli, G; Haines, S.C; Hampson, T; Hansmann-Menzemer, S; Harji, R; Harnew, N; Harrison, P.F; He, J; Hennessy, K; Henrard, P; Hernando Morata, J.A; van Herwijnen, E; Hicheur, A; Hicks, E; Hilke, H.J; Hofmann, W; Holubyev, K; Hopchev, P; Hulsbergen, W; Hunt, P; Huse, T; Huston, R.S; Hutchcroft, D; Iacoangeli, F; Iakovenko, V; Iglesias Escudero, C; Ilgner, C; Imong, J; Jacobsson, R; Jahjah Hussein, M; Jamet, O; Jans, E; Jansen, F; Jaton, P; Jean-Marie, B; John, M; Johnson, D; Jones, C.R; Jost, B; Kapusta, F; Karbach, T.M; Kashchuk, A; Katvars, S; Keaveney, J; Kerzel, U; Ketel, T; Keune, A; Khalil, S; Khanji, B; Kim, Y.M; Knecht, M; Koblitz, S; Konoplyannikov, A; Koppenburg, P; Korolev, M; Kozlinskiy, A; Kravchuk, L; Kristic, R; Krocker, G; Krokovny, P; Kruse, F; Kruzelecki, K; Kucharczyk, M; Kudryashov, I; Kukulak, S; Kumar, R; Kvaratskheliya, T; La Thi, V.N; Lacarrere, D; Lai, A; Lambert, R.W; Lanfranchi, G; Langenbruch, C; Latham, T; Le Gac, R; Lees, J-P; Lef`evre, R; Leflat, A; Lefrancois, J; Lehner, F; Lenzi, M; Leroy, O; Lesiak, T; Li, L; Li, Y.Y; Li Gioi, L; Libby, J; Lieng, M; Lindner, R; Lindsey, S; Linn, C; Liu, B; Liu, G; Lochner, S; Lopes, J.H; Lopez Asamar, E; Lopez-March, N; Loveridge, P; Luisier, J; Mcharek, B; Machefert, F; Machikhiliyan, I.V; Maciuc, F; Maev, O; Magnin, J; Maier, A; Malde, S; Mamunur, R.M.D; Manca, G; Mancinelli, G; Mangiafave, N; Marconi, U; Marki, R; Marks, J; Martellotti, G; Martens, A; Martin, L; Martinez Santos, D; Massaferri, A; Mathe, Z; Matteuzzi, C; Matveev, V; Maurice, E; Maynard, B; Mazurov, A; McGregor, G; McNulty, R; Mclean, C; Merk, M; Merkel, J; Merkin, M; Messi, R; Metlica, F.C.D; Miglioranzi, S; Minard, M-N; Moine, G; Monteil, S; Moran, D; Morant, J; Morris, J.V; Moscicki, J; Mountain, R; Mous, I; Muheim, F; Muresan, R; Murtas, F; Muryn, B; Musy, M; Mylroie-Smith, J; Naik, P; Nakada, T; Nandakumar, R; Nardulli, J; Nawrot, A; Nedos, M; Needham, M; Neufeld, N; Neustroev, P; Nicol, M; Nicolas, L; Nies, S; Niess, V; Nikitin, N; Noor, A; Oblakowska-Mucha, A; Obraztsov, V; Oggero, S; Okhrimenko, O; Oldeman, R; Orlandea, M; Ostankov, A; Palacios, J; Palutan, M; Panman, J; Papadelis, A; Papanestis, A; Pappagallo, M; Parkes, C; Parkinson, C.J; Passaleva, G; Patel, G.D; Patel, M; Paterson, S.K; Patrick, G.N; Patrignani, C; Pauna, E; Pauna (Chiojdeanu), C; Pavel (Nicorescu), C; Pazos Alvarez, A; Pellegrino, A; Penso, G; Pepe Altarelli, M; Perazzini, S; Perego, D.L; Perez Trigo, E; Perez-Calero Yzquierdo, A; Perret, P; Pessina, G; Petrella, A; Petrolini, A; Picatoste Olloqui, E; Pie Valls, B; Piedigrossi, D; Pietrzyk, B; Pinci, D; Playfer, S; Plo Casasus, M; Poli-Lener, M; Polok, G; Poluektov, A; Polycarpo, E; Popov, D; Popovici, B; Poss, S; Potterat, C; Powell, A; Pozzi, S; du Pree, T; Pugatch, V; Puig Navarro, A; Qian, W; Rademacker, J.H; Rakotomiaramanana, B; Raniuk, I; Raven, G; Redford, S; Reece, W; dos Reis, A.C; Ricciardi, S; Riera, J; Rinnert, K; Roa Romero, D.A; Robbe, P; Rodrigues, E; Rodrigues, F; Rodriguez Cobo, C; Rodriguez Perez, P; Rogers, G.J; Romanovsky, V; Rondan Sanabria, E; Rosello, M; Rospabe, G; Rouvinet, J; Roy, L; Ruf, T; Ruiz, H; Rummel, C; Rusinov, V; Sabatino, G; Saborido Silva, J.J; Sagidova, N; Sail, P; Saitta, B; Sakhelashvili, T; Salzmann, C; Sambade Varela, A; Sannino, M; Santacesaria, R; Santinelli, R; Santovetti, E; Sapunov, M; Sarti, A; Satriano, C; Satta, A; Savidge, T; Savrie, M; Savrina, D; Schaack, P; Schiller, M; Schleich, S; Schmelling, M; Schmidt, B; Schneider, O; Schneider, T; Schopper, A; Schune, M-H; Schwemmer, R; Sciubba, A; Seco, M; Semennikov, A; Senderowska, K; Serra, N; Serrano, J; Shao, B; Shapkin, M; Shapoval, I; Shatalov, P; Shcheglov, Y; Shears, T; Shekhtman, L; Shevchenko, V; Shires, A; Sigurdsson, S; Simioni, E; Skottowe, H.P; Skwarnicki, T; Smale, N; Smith, A; Smith, A.C; Smith, N.A; Sobczak, K; Soler, F.J.P; Solomin, A; Somogy, P; Soomro, F; Souza De Paula, B; Spaan, B; Sparkes, A; Spiridenkov, E; Spradlin, P; Srednicki, A; Stagni, F; Stahl, S; Steiner, S; Steinkamp, O; Stenyakin, O; Stoica, S; Stone, S; Storaci, B; Straumann, U; Styles, N; Szczekowski, M; Szczypka, P; Szumlak, T; TJampens, S; Tarkovskiy, E; Teodorescu, E; Terrier, H; Teubert, F; Thomas, C; Thomas, E; van Tilburg, J; Tisserand, V; Tobin, M; Topp-Joergensen, S; Tran, M.T; Traynor, S; Trunk, U; Tsaregorodtsev, A; Tuning, N; Ukleja, A; Ullaland, O; Uwer, U; Vagnoni, V; Valenti, G; Van Lysebetten, A; Vazquez Gomez, R; Vazquez Regueiro, P; Vecchi, S; Velthuis, J.J; Veltri, M; Vervink, K; Viaud, B; Videau, I; Vieira, D; Vilasis-Cardona, X; Visniakov, J; Vollhardt, A; Volyanskyy, D; Voong, D; Vorobyev, A; Vorobyev, An; Voss, H; Wacker, K; Wandernoth, S; Wang, J; Ward, D.R; Webber, A.D; Websdale, D; Whitehead, M; Wiedner, D; Wiggers, L; Wilkinson, G; Williams, M.P; Williams, M; Wilson, F.F; Wishahi, J; Witek, M; Witzeling, W; Woodward, M.L; Wotton, S.A; Wyllie, K; Xie, Y; Xing, F; Yang, Z; Ybeles Smit, G; Young, R; Yushchenko, O; Zeng, M; Zhang, L; Zhang, Y; Zhelezov, A; Zverev, E

2010-01-01

The production of K_short mesons in pp collisions at a centre-of-mass energy of 0.9 TeV is studied with the LHCb detector at the Large Hadron Collider. The luminosity of the analysed sample is determined using a novel technique, involving measurements of the beam currents, sizes and positions, and is found to be 6.8 +/- 1.0 microbarn^-1. The differential prompt K_short production cross-section is measured as a function of the K_short transverse momentum and rapidity in the region 0 < pT < 1.6 GeV/c and 2.5 < y < 4.0. The data are found to be in reasonable agreement with previous measurements and generator expectations.

Klebsiella pneumoniae capsule expression is necessary for colonization of large intestines of streptomycin-treated mice

DEFF Research Database (Denmark)

Favre-Bonte, S.; Licht, Tine Rask; Forestier, C.

1999-01-01

The role of the Klebsiella pneumoniae capsular polysaccharide (K antigen) during colonization of the mouse large intestine was assessed with mild-type K. pneumoniae LM21 and its isogenic capsule-defective mutant. When bacterial strains were fed alone to mice, the capsulated bacteria persisted...... in the intestinal tract at levels of 10(8) CFU/g of feces while the capsule-defective strain colonized at low levels, 10(4) CFU/g of feces. In mixed-infection experiments, the mutant was rapidly outcompeted by the wild type. In situ hybridization on colonic sections revealed that bacterial cells of both strains...... were evenly distributed in the mucus layer at day 1 after infection, while at day 20 the wild type remained dispersed and the capsule-defective strain was seen in clusters in the mucus layer. These results suggest that capsular polysaccharide plays an important role in the gut colonization ability of K...

Exergy and Energy Analysis of Combustion of Blended Levels of Biodiesel, Ethanol and Diesel Fuel in a DI Diesel Engine

International Nuclear Information System (INIS)

Khoobbakht, Golmohammad; Akram, A.; Karimi, Mahmoud; Najafi, G.

2016-01-01

Highlights: • Exergy analysis showed that thermal efficiency of diesel engine was 36.61%. • Energy loss and work output rates were 71.36 kW and 41.22 kW, respectively. • Exergy efficiency increased with increasing engine load and speed. • Exergy efficiency increased with increasing biodiesel and bioethanol. • 0.17 L of biodiesel, 0.08 L of ethanol in 1 L of diesel at 1900 rpm and 94% load had maximum exergy efficiency. - Abstract: In this study, the first and second laws of thermodynamics are employed to analyze the energy and energy in a four-cylinder, direct injection diesel engine using blended levels of biodiesel and ethanol in diesel fuel. Also investigated the effect of operating factors of engine load and speed as well as blended levels of biodiesel and ethanol in diesel fuel on the exergy efficiency. The experiments were designed using a statistical tool known as Design of Experiments (DoE) based on central composite rotatable design (CCRD) of response surface methodology (RSM). The resultant quadratic models of the response surface methodology were helpful to predict the response parameter (exergy efficiency) further to identify the significant interactions between the input factors on the responses. The results depicted that the exergy efficiency decreased with increasing percent by volume biodiesel and ethanol fuel. The fuel blend of 0.17 L biodiesel and 0.08 L of ethanol added to 1 L of diesel (equivalent with D80B14E6) at 1900 rpm and 94% load was realized have the most exergy efficiency. The results of energy and exergy analyses showed that 43.09% of fuel exergy was destructed and the average thermal efficiency was approximately 36.61%, and the exergetic efficiency was approximately 33.81%.

High-throughput identification of chemical inhibitors of E. coli Group 2 capsule biogenesis as anti-virulence agents.

Directory of Open Access Journals (Sweden)

Carlos C Goller

Full Text Available Rising antibiotic resistance among Escherichia coli, the leading cause of urinary tract infections (UTIs, has placed a new focus on molecular pathogenesis studies, aiming to identify new therapeutic targets. Anti-virulence agents are attractive as chemotherapeutics to attenuate an organism during disease but not necessarily during benign commensalism, thus decreasing the stress on beneficial microbial communities and lessening the emergence of resistance. We and others have demonstrated that the K antigen capsule of E. coli is a preeminent virulence determinant during UTI and more invasive diseases. Components of assembly and export are highly conserved among the major K antigen capsular types associated with UTI-causing E. coli and are distinct from the capsule biogenesis machinery of many commensal E. coli, making these attractive therapeutic targets. We conducted a screen for anti-capsular small molecules and identified an agent designated "C7" that blocks the production of K1 and K5 capsules, unrelated polysaccharide types among the Group 2-3 capsules. Herein lies proof-of-concept that this screen may be implemented with larger chemical libraries to identify second-generation small-molecule inhibitors of capsule biogenesis. These inhibitors will lead to a better understanding of capsule biogenesis and may represent a new class of therapeutics.

Advances in pediatric gastroenterology: introducing video camera capsule endoscopy.

Science.gov (United States)

Siaw, Emmanuel O

2006-04-01

The video camera capsule endoscope is a gastrointestinal endoscope approved by the U.S. Food and Drug Administration in 2001 for use in diagnosing gastrointestinal disorders in adults. In 2003, the agency approved the device for use in children ages 10 and older, and the endoscope is currently in use at Arkansas Children's Hospital. A capsule camera, lens, battery, transmitter and antenna together record images of the small intestine as the endoscope makes its way through the bowel. The instrument is used with minimal risk to the patient while offering a high degree of accuracy in diagnosing small intestine disorders.

2. Sino-German workshop on fuel cells. Book of abstracts

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-07-01

The CD-ROM contains the content of 13 lectures and 19 proposals for joint projects, which were presented on the second Sino-German Workshop on fuel cells. The topics of the 13 lectures are: Ab-initio calculations of oxygen species on low-index platinum surfaces (Pachenko, M.T.M. Koper, T.E. Shubina, S.J. Mitchell, E. Roduner). Cross-Linked (Composite) Polyaryl Blend Membranes for Membrane Fuel Cells. (J.A. Kerres, A. Ullrich, W. Zhang, M. Hein, V. Gogel, L. Joerissen, Th. Frey, A. Friedrich). Performance and Methanol Permeation of Direct Methanol Fuel Cells: Dependence on Operating Conditions and on Electrode Structure(V. Gogel, Th. Frey, K.A. Friedrich, L. Joerissen, J. Garche, Z. Yongsheng). Experimental Investigation of Flow Bed Configuration Effect on Performance of Liquid Feed Direct Methanol Fuel Cells. (H. Guo, C.F. Ma, M.H. Wang, F. Ye, J. Yu, Y. Wang, C.Y. Wang). Improvement of MEAs for DMFC by a tuned production sequence assisted by mathematical modelling (Lindermeir, G. Rosenthal, U. Kunz, U. Hoffmann). Performance of the self-breathing air DMFC with solution grafted PVDF-g-PSSA membranes (X. Qiu, G. Guo, W. Li, W. Zhu, L. Chen). Modeling the Effects of Methanol Crossover on DMFC (J. Zhang, Y. Wang). The characteristics of 40 kW PEM fuel cell engine for vehicle(M. Hou, P. Ming, H. Zhang). A New and Simple Method for Preparing Biocathode in Biofuel Cells (D. Sun, C. Cai, X. Li, W. Xing, T. Lu). Nonlinear Model Reduction of a Dynamic Two-dimensional Molten Carbonate Fuel Cell Model (M. Mangold, M. Sheng). Recent Advances in Design and Fabrication of Low-Temperature Solid Oxide Fuel Cells (C. Xia, G. Meng). Novel CVD Techniques for Micro- and IT-SOFC Fabrication (G. Meng, H. Song, Q. Dong, D. Peng). Fundamental properties of La{sub 0.6}Sr{sub 0.4}Co{sub 0.8}Fe{sub 0.2}O{sub 3-{delta}} at high temperatures (S. Wang, T.-L. Wen).

Instrumentation Technologies for Improving an Irradiation Testing of Nuclear Fuels and Materials at the HANARO

International Nuclear Information System (INIS)

Kim, Bong Goo; Park, Sung Jae; Choo, Ki Nam

2011-01-01

Over 50 years of nuclear fuels and materials irradiation testing has led to many countries developing significant improvements in instrumentation to monitor physical parameters and to control the test conditions in Materials Test Reactors (MTRs) or research reactors. Recent effort to deploy new fuels and materials in existing and advanced reactors has increased the demand for well-instrumented irradiation tests. Specifically, demand has increased for tests with sensors capable of providing real-time measurement of key parameters, such as temperature, geometry changes, thermal conductivity, fission gas release, cracking, coating buildup, thermal and fast flux, etc. This review paper documents the current state of instrumentation technologies in MTRs in the world and summarizes on-going research efforts to deploy new sensors. There is increased interest to irradiate new materials and reactor fuels for advanced PWRs and the Gen-IV reactor systems, such as SFRs (Sodium-cooled Fast Reactors), VHTRs (Very-High-Temperature Reactors), SCWRs (Supercritical-Water-cooled Reactors) and GFRs (Gas-cooled Fast Reactor). This review documents the current state of instrumentation technologies in MTRs in the world, identifies challenges faced by previous testing methods and how these challenges were overcome. A wide range of sensors are available to measure key parameters of interest during fuels and materials irradiations in MTRs. Such sensors must be reliable, small size, highly accurate, and able to withstand harsh conditions. On-going development efforts are focusing on providing MTR users a wider range of parameter measurements with increased accuracy. In addition, development efforts are focusing on reducing the impact of sensor on measurements by reducing sensor size. This report includes not only status of instrumentation using research reactors in the world to irradiate nuclear fuels and materials but also future directions relating to instrumentation technologies for

Coupling of a 2.5 kW steam reformer with a 1 kW el PEM fuel cell

Science.gov (United States)

Mathiak, J.; Heinzel, A.; Roes, J.; Kalk, Th.; Kraus, H.; Brandt, H.

The University of Duisburg-Essen has developed a compact multi-fuel steam reformer suitable for natural gas, propane and butane. This steam reformer was combined with a polymer electrolyte membrane fuel cell (PEM FC) and a system test of the process chain was performed. The fuel processor comprises a prereformer step, a primary reformer, water gas shift reactors, a steam generator, internal heat exchangers in order to achieve an optimised heat integration and an external burner for heat supply as well as a preferential oxidation step (PROX) as CO purification. The fuel processor is designed to deliver a thermal hydrogen power output from 500 W to 2.5 kW. The PEM fuel cell stack provides about 1 kW electrical power. In the following paper experimental results of measurements of the single components PEM fuel cell and fuel processor as well as results of the coupling of both to form a process chain are presented.

K(3)TaF(8) from laboratory X-ray powder data.

Science.gov (United States)

Smrcok, Lubomír; Cerný, Radovan; Boca, Miroslav; Macková, Iveta; Kubíková, Blanka

2010-02-01

The crystal structure of tripotassium octafluoridotantalate, K(3)TaF(8), determined from laboratory powder diffraction data by the simulated annealing method and refined by total energy minimization in the solid state, is built from discrete potassium cations, fluoride anions and monocapped trigonal-prismatic [TaF(7)](2-) ions. All six atoms in the asymmetric unit are in special positions of the P6(3)mc space group: the Ta and one F atom in the 2b (3m) sites, the K and two F atoms in the 6c (m) sites, and one F atom in the 2a (3m) site. The structure consists of face-sharing K(6) octahedra with a fluoride anion at the center of each octahedron, forming chains of composition [FK(3)](2+) running along [001] with isolated [TaF(7)](2-) trigonal prisms in between. The structure of the title compound is different from the reported structure of Na(3)TaF(8) and represents a new structure type.

Study of behavior of cermet fuel elements on IGR reactor under RIA type accident condition

International Nuclear Information System (INIS)

Vasil'ev, Yu.S.; Vurim, A.D.; Koltyshev, S.M.; Pakhnits, V.A.; Tukhvatulin, Sh.T.; Popov, V.V.; Ryzhkov, A.N.

1996-01-01

In 1993 December in IGR reactor of Inst. of Atomic Energy of National Nuclear Center of Republic of Kazakstan the second batch of in-pile testing of perspective cermet fuel elements under the condition, simulating RIA type accident was conducted. In the second batch of testing during eight start-ups 10 cermet fuel elements were examined. Among which 8 of monolith type and 2 fuel elements with false jacket beside cladding (FJF), as well as, 6 standard fuel elements of WWER-1000 type reactor with dioxide fuel were tested. 2 fuel elements - cermet and standard were placed into capsule filled with water. To measure energy release for the each start-up two fission monitor and inside core control gauge were placed. In all the start-ups operation mode of IGR was neutron pulse. Power of fuel element kept changing from 151 to 336 k W; energy release was 38-93 kJ/gr m 235 U; maximum temperature of cermet fuel was 1943-2173 K, of dioxide fuel - 1923-2843 K. The testing has demonstrated that operability of cermet fuel elements under reactivity accident condition with pulse width of 0,2 s is, at least, not less that operability of dioxide fuel elements, through advantages of cermet fuel under these conditions are revealed to the least extent

Power ramp tests of MOX fuel rods. HBWR irradiation with the instrument rig, IFA-591

International Nuclear Information System (INIS)

Ozawa, Takayuki; Abe, Tomoyuki

2006-03-01

Plutonium-uranium mixed oxide (MOX) fuel rods of instrumental rig IFA-591 were ramped in HBWR to study the Advanced Thermal Reactor (ATR) MOX fuel behavior during transient operation and to determine a failure threshold of the MOX fuel rods. Eleven segments were base-irradiated in ATR 'FUGEN' up to 18.4 GWd/t. Zirconium liner claddings were adopted for four segments of them. As the results of non-destructive post irradiation examinations (PIEs) after the base-irradiation and before the ramp tests, no remarkable behavior affecting the integrity of fuel assembly and fuel rod was confirmed. All segments to be used for the ramp tests, which consisted of the multi-step ramp tests and the single-step ramp tests, had instrumentations for in-pile measurements of cladding elongation or plenum pressure, and heated up to the maximum linear power of 58.3-68.4 kW/m without failure. The major results of ramp tests are as follows: There is no difference in PCMI behaviors between two type rods of Zry-2 and Zirconium liner claddings from the in-pile measurements of cladding elongation and plenum pressure. The computations of cladding elongation and inner pressure gave slightly lower elongation and pressure than the in-pile measurements during the ramp-test. However, the cladding relaxation during the power hold was in good agreement, and the fission gas release behavior during cooling down could be evaluated by taking into account the relaxation of contact pressure between pellet and cladding. Although the final power during IFA-591 ramp tests reached the higher linear power than the failure threshold power of UO 2 fuel rods, no indication of fuel failure was observed during the ramp tests. The cladding relaxation due to the creep deformation of the MOX pellets at high temperature could be confirmed at the power steps during the multi-ramp test. The fission gas release due to the emancipation from PCMI stress was observed during the power decreasing. The burn-up dependence could be

Study of the Effect of (U0.8Pu0.2O2 Uranium–Plutonium Mixed Fuel Fission Products on a Living Organism

Directory of Open Access Journals (Sweden)

Ayagoz Baimukhanova

2016-08-01

Full Text Available The article describes the results of experiments conducted on pigs to determine the effect of plutonium, which is the most radiotoxic and highly active element in the range of mixed fuel (U0.8Pu0.2O2 fission products, on living organisms. The results will allow empirical prediction of the emergency plutonium radiation dose for various organs and tissues of humans in case of an accident in a reactor running on mixed fuel (U0.8Pu0.2O2.

An Analysis of the Thermal and Structure Behaviour of the UO2-PuO2-Fuel in the Irradiation Experiment of the UO2-PuO2-Fuel in the Irradiation Experiment FR2 Capsule Test Series 5a

International Nuclear Information System (INIS)

Lopez Jimenez, J.; Helmut, E.

1981-01-01

In the Karlsruhe research reactor FR2 nine fuel pins were irradiated within three irradiation capsules in the course of the test series 5a. The pins contained UO 2 -PuO 2 fuel pellets. They reached bump values of about 6, 17 and 47 Mwd/Kg Me with linear rod powers of 400 to 600 W/cm and clad surface temperature between 500 and 700 degree centigree. A detailed analysis of the fuel structuration data (columnar-grain and equiaxed- -grain growth regions) have allowed to determine, with the help of physic-mathematical models, the radii of these regions and the heat transfer through the contact zone between fuel and clad depending on the bump. The results of the analysis showed that the fuel surface temperature rose with increasing burnup. (Author) 16 refs

Controllable solvothermal synthesis and photocatalytic properties of complex (oxy)fluorides K{sub 2}TiOF{sub 4}, K{sub 3}TiOF{sub 5}, K{sub 7}Ti{sub 4}O{sub 4}F{sub 7} and K{sub 2}TiF{sub 6}

Energy Technology Data Exchange (ETDEWEB)

Sheng Jie [Division of Nanomaterials and Nanochemistry, Hefei National Laboratory for Physical Sciences at Microscale, Hefei, Anhui 230026 (China); Department of Chemistry, University of Science and Technology of China, Hefei, Anhui 230026 (China); Tang Kaibin, E-mail: kbtang@ustc.edu.cn [Division of Nanomaterials and Nanochemistry, Hefei National Laboratory for Physical Sciences at Microscale, Hefei, Anhui 230026 (China); Department of Chemistry, University of Science and Technology of China, Hefei, Anhui 230026 (China); Cheng Wei; Wang Junli; Nie Yanxiang; Yang Qing [Division of Nanomaterials and Nanochemistry, Hefei National Laboratory for Physical Sciences at Microscale, Hefei, Anhui 230026 (China); Department of Chemistry, University of Science and Technology of China, Hefei, Anhui 230026 (China)

2009-11-15

Complex (oxy)fluorides K{sub 2}TiF{sub 6}, K{sub 2}TiOF{sub 4}, K{sub 3}TiOF{sub 5} and K{sub 7}Ti{sub 4}O{sub 4}F{sub 7} have been successfully synthesized for the first time through a controllable solvothermal route involving different solvents, for example, methanol, methanol-H{sub 2}O and methanol-H{sub 2}O{sub 2}. The as-prepared products were characterized by X-ray powder diffraction, N{sub 2} surface area adsorption, scanning electron microscope, Fourier transform infrared spectroscopy, UV-vis absorption spectra and X-ray fluorescence. The influences of reaction conditions such as the ratio of methanol to H{sub 2}O{sub 2} or methanol to H{sub 2}O, reaction temperature on the phase, crystallizability and purity of the (oxy)fluorides products were discussed in detail. Meanwhile, the photocatalytic behaviors of the as-prepared K{sub 2}TiF{sub 6}, K{sub 2}TiOF{sub 4}, K{sub 3}TiOF{sub 5} and K{sub 7}Ti{sub 4}O{sub 4}F{sub 7} were evaluated by degradation of rhodamine B molecules, and the results showed that all of the products possessed photocatalytic activities in the order of K{sub 2}TiOF{sub 4} > K{sub 2}TiF{sub 6} > K{sub 7}Ti{sub 4}O{sub 4}F{sub 7} > K{sub 3}TiOF{sub 5} at room temperature under the UV light.

Preparation of functions of computer code GENGTC and improvement for two-dimensional heat transfer calculations for irradiation capsules

International Nuclear Information System (INIS)

Nomura, Yasushi; Someya, Hiroyuki; Ito, Haruhiko.

1992-11-01

Capsules for irradiation tests in the JMTR (Japan Materials Testing Reactor), consist of irradiation specimens surrounded by a cladding tube, holders, an inner tube and a container tube (from 30mm to 65mm in diameter). And the annular gaps between these structural materials in the capsule are filled with liquids or gases. Cooling of the capsule is done by reactor primary coolant flowing down outside the capsule. Most of the heat generated by fission in fuel specimens and gamma absorption in structural materials is directed radially to the capsule container outer surface. In thermal performance calculations for capsule design, an one(r)-dimensional heat transfer computer code entitled (Generalyzed Gap Temperature Calculation), GENGTC, originally developed in Oak Ridge National Laboratory, U.S.A., has been frequently used. In designing a capsule, are needed many cases of parametric calculations with respect to changes materials and gap sizes. And in some cases, two(r,z)-dimensional heat transfer calculations are needed for irradiation test capsules with short length fuel rods. Recently the authors improved the original one-dimensional code GENGTC, (1) to simplify preparation of input data, (2) to perform automatic calculations for parametric survey based on design temperatures, ect. Moreover, the computer code has been improved to perform r-z two-dimensional heat transfer calculation. This report describes contents of the preparation of the one-dimensional code GENGTC and the improvement for the two-dimensional code GENGTC-2, together with their code manuals. (author)

Viscosity of melts of the system KCl-KBF4-K2TiF6

International Nuclear Information System (INIS)

Nguyen, D.K.; Danek, V.

1997-01-01

The viscosity of melts of the system KCl-KBF 4 -K 2 TiF 6 has been measured by means of the computerized torsional pendulum method. The viscosity of KCl is higher that of KBF 4 at the same temperature, most probably due to the substantial overheating of KBF 4 . In the ternary system the viscosity increases with increasing with increasing content of K 2 TiF 6 . Additivity of algorithms of viscosity was adopted as the ideal behaviour of the mixture. Negative deviations from such additive behaviour were found in the binary system KCl-KBF 4 probably due to the breaks of the weak B-Cl-B bridges caused by the excess of Cl - ions. Positive deviations from the ideal behaviour were found in the binaries KCl-K 2 TiF 6 and KBF 4 -K 2 TiF 6 due to the formation of larger anions TiF 6 Cl 3- and TiF 7 3- caused by the reactions K 2 TiF 6 (l) + KCl(l) = K 3 TiF 6 Cl(l) and KBF 4 (l) + K 2 TiF 6 (l) = K 3 TiF 7 (l) + BF 3 (g). Statistically significant ternary interaction confirmed that the above chemical reactions take place also in the ternary system. (authors)

Metal membrane-type 25-kW methanol fuel processor for fuel-cell hybrid vehicle

Science.gov (United States)

Han, Jaesung; Lee, Seok-Min; Chang, Hyuksang

A 25-kW on-board methanol fuel processor has been developed. It consists of a methanol steam reformer, which converts methanol to hydrogen-rich gas mixture, and two metal membrane modules, which clean-up the gas mixture to high-purity hydrogen. It produces hydrogen at rates up to 25 N m 3/h and the purity of the product hydrogen is over 99.9995% with a CO content of less than 1 ppm. In this fuel processor, the operating condition of the reformer and the metal membrane modules is nearly the same, so that operation is simple and the overall system construction is compact by eliminating the extensive temperature control of the intermediate gas streams. The recovery of hydrogen in the metal membrane units is maintained at 70-75% by the control of the pressure in the system, and the remaining 25-30% hydrogen is recycled to a catalytic combustion zone to supply heat for the methanol steam-reforming reaction. The thermal efficiency of the fuel processor is about 75% and the inlet air pressure is as low as 4 psi. The fuel processor is currently being integrated with 25-kW polymer electrolyte membrane fuel-cell (PEMFC) stack developed by the Hyundai Motor Company. The stack exhibits the same performance as those with pure hydrogen, which proves that the maximum power output as well as the minimum stack degradation is possible with this fuel processor. This fuel-cell 'engine' is to be installed in a hybrid passenger vehicle for road testing.

Effect of Heat Flux on the Specimen Temperature of an LBE Capsule

International Nuclear Information System (INIS)

Kang, Y. H.; Park, S. J.; Cho, M. S.; Choo, K. N.; Lee, Y. S.

2011-01-01

For application of high-temperature irradiation tests in the HANARO reactor for Gen IV reactor material development, a number of newly designed LBE capsules have been investigated at KAERI since 2008. Recent study on heat transfer experiment of an LBE capsule with a single heater has shown that the specimen temperature of the mock-up increased linearly with an increase of heat input. The work highlighted only the heat transfer capability of an LBE capsule with a single heater as a simulated specimen in a liquid metal medium. Hence, a new LBE capsule with multi specimen sets has been designed and fabricated for the heat transfer experiment of an LBE capsule of 11M-01K. In this paper, a series of thermal analyses and heat transfer experiments for a newly designed LBE capsule was implemented to study the effect of an increase in the value of heat input and its influence on temperature distribution in the capsule mock-up

A generalized scaling law for the ignition energy of inertial confinement fusion capsules

International Nuclear Information System (INIS)

Herrmann, M.C.

2001-01-01

The minimum energy needed to ignite an inertial confinement fusion capsule is of considerable interest in the optimization of an inertial fusion driver. Recent computational work investigating this minimum energy has found that it depends on the capsule implosion history, in particular, on the capsule drive pressure. This dependence is examined using a series of LASNEX simulations to find ignited capsules which have different values of the implosion velocity, fuel adiabat and drive pressure. It is found that the main effect of varying the drive pressure is to alter the stagnation of the capsule, changing its stagnation adiabat, which, in turn, affects the energy required for ignition. To account for this effect a generalized scaling law has been devised for the ignition energy, E ign ∝α if 1.88±0.05 υ -5.89±0.12 P -0.77±0.03 . This generalized scaling law agrees with the results of previous work in the appropriate limits. (author)

Unreviewed safety question evaluation of 100K East and 100K West in-basin fuel characterization program activities

International Nuclear Information System (INIS)

Alwardt, L.D.

1995-01-01

The purpose of this report is to provide the basis for answers to an Unreviewed Safety Question (USQ) safety evaluation of the 105K East (KE) and 105K West (KW) in-basin activities associated with the fuel characterization program as described in the characterization shipping plan. The significant activities that are common to both 105 KE and 105 KW basins are the movement of canisters from their main basin storage locations (or potentially from the 105 KE Tech View Pit if a dump table is available) to the south loadout pit transfer channel, hydrogen generation testing in the single element fuel container, loading the single element fuel container into the shipping cask, loading of the shipping cask onto a flat-bed trailer, return of the test fuel elements or element pieces from the 327 facility, placement of the fuel elements back into Mark 2 canisters, and placement of the canisters in the main storage basin. Decapping of canisters in the south loadout pit transfer channel and re-encapsulation of canisters are activities specific to the 105 KW basin. The scope of this safety evaluation includes only those characterization fuel shipment activities performed in the 105 KE and 105 KW fuel storage basin structures up to installation of the overpack. The packaging safety evaluation report governs the shipment of the fuel elements. The K Basins Plant Review Committee has determined that the in-basin activities associated with the fuel characterization program fuel shipments are bounded by the current safety envelop and do not constitute an unreviewed safety question. This determination is documented on Attachment 1

Relation of zonal plasma drift and wind in the equatorial F region as derived from CHAMP observations

Directory of Open Access Journals (Sweden)

J. Park

2013-06-01

Full Text Available In this paper we estimate zonal plasma drift in the equatorial ionospheric F region without counting on ion drift meters. From June 2001 to June 2004 zonal plasma drift velocity is estimated from electron, neutral, and magnetic field observations of Challenging Mini-satellite Payload (CHAMP in the 09:00–20:00 LT sector. The estimated velocities are validated against ion drift measurements by the Republic of China Satellite-1/Ionospheric Plasma and Electrodynamics Instrument (ROCSAT-1/IPEI during the same period. The correlation between the CHAMP (altitude ~ 400 km estimates and ROCSAT-1 (altitude ~ 600 km observations is reasonably high (R ≈ 0.8. The slope of the linear regression is close to unity. However, the maximum westward drift and the westward-to-eastward reversal occur earlier for CHAMP estimates than for ROCSAT-1 measurements. In the equatorial F region both zonal wind and plasma drift have the same direction. Both generate vertical currents but with opposite signs. The wind effect (F region wind dynamo is generally larger in magnitude than the plasma drift effect (Pedersen current generated by vertical E field, thus determining the direction of the F region vertical current.

Magnetron-sputtered La{sub 0.6}Sr{sub 0.4}Co{sub 0.2}Fe{sub 0.8}O{sub 3} nanocomposite interlayer for solid oxide fuel cells

Energy Technology Data Exchange (ETDEWEB)

Solovyev, A. A., E-mail: andrewsol@mail.ru; Ionov, I. V.; Shipilova, A. V.; Kovalchuk, A. N.; Syrtanov, M. S. [Tomsk Polytechnic University (Russian Federation)

2017-03-15

A thin layer of a La{sub 0.6}Sr{sub 0.4}Co{sub 0.2}Fe{sub 0.8}O{sub 3} (LSCF) is deposited between the electrolyte and the La{sub 0.6}Sr{sub 0.4}Co{sub 0.2}Fe{sub 0.8}O{sub 3}/Ce{sub 0.9}Gd{sub 0.1}O{sub 2} (LSCF/CGO) cathode layer of a solid oxide fuel cell (SOFC) by pulsed magnetron sputtering using an oxide target of LSCF. The films were completely dense and well adherent to the substrate. The effects of annealing in temperature range from 200 to 1000 °C on the crystalline structure of the LSCF films have been studied. The films of nominal thickness, 250–500 nm, are crystalline when annealed at temperatures above 600 °C. The crystalline structure, surface topology, and morphology of the films were determined using X-ray diffraction (XRD), atomic force microscopy (AFM), and scanning electron microscopy (SEM), respectively. To study the electrochemical characteristics of the deposited-film, solid oxide fuel cells using 325-nm LSCF films as interlayer between the electrolyte and the cathode have been fabricated. The LSCF interlayer improves the overall performance of the SOFC by increasing the interfacial area between the electrolyte and cathode. The electrolyte-supported cells with the interlayer have 30% greater, overall power output compared to that achieved with the cells without interlayer. The LSCF interlayer could also act as a transition layer that improves adhesion and relieves both thermal stress and lattice strain between the cathode and the electrolyte. Our results demonstrate that pulsed magnetron sputtering provides a low-temperature synthesis route for realizing ultrathin nanocrystalline LSCF film layers for intermediate- or low-temperature solid oxide fuel cells.

HANARO fuel irradiation test (II): revision

Energy Technology Data Exchange (ETDEWEB)

Sohn, D. S.; Kim, H.; Chae, H. T.; Lee, C. S.; Kim, B. G.; Lee, C. B

2001-04-01

In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiated at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%. This report is the revision of KAERI/TR-1816/2001 on the irradiation test for HANARO fuel.