WorldWideScience

Sample records for in-pile thermocycling testing

  1. In-pile thermocycling testing and post-test analysis of beryllium divertor mockups

    Energy Technology Data Exchange (ETDEWEB)

    Giniatulin, R.; Mazul, I. [Efremov Inst., St. Petersburg (Russian Federation); Melder, R.; Pokrovsky, A.; Sandakov, V.; Shiuchkin, A.

    1998-01-01

    The main damaging factors which impact the ITER divertor components are neutron irradiation, cyclic surface heat loads and hydrogen environment. One of the important questions in divertor mockups development is the reliability of beryllium/copper joints and the beryllium resistance under neutron irradiation and thermal cycling. This work presents the experiment, where neutron irradiation and thermocyclic heat loads were applied simultaneously for two beryllium/copper divertor mockups in a nuclear reactor channel to simulate divertor operational conditions. Two mockups with different beryllium grades were mounted facing each other with the tantalum heater placed between them. This device was installed in the active zone of the nuclear reactor SM-2 (Dimitrovgrad, Russia) and the tantalum block was heated by neutron irradiation up to a high temperature. The main part of the heat flux from the tantalum surface was transported to the beryllium surface through hydrogen, as a result the heat flux loaded two mockups simultaneously. The mockups were cooled by reactor water. The device was lowered to the active zone so as to obtain the heating regime and to provide cooling lifted. This experiment was performed under the following conditions: tantalum heater temperature - 1950degC; hydrogen environment -1000 Pa; surface heat flux density -3.2 MW/m{sup 2}; number of thermal cycles (lowering and lifting) -101; load time in each cycle - 200-5000 s; dwell time (no heat flux, no neutrons) - 300-2000 s; cooling water parameters: v - 1 m/s, Tin - 50degC, Pin - 5 MPa; neutron fluence -2.5 x 10{sup 20} cm{sup -2} ({approx}8 years of ITER divertor operation from the start up). The metallographic analysis was performed after experiment to investigate the beryllium and beryllium/copper joint structures, the results are presented in the paper. (author)

  2. Gas reactor in-pile safety test project (GRIST-2)

    International Nuclear Information System (INIS)

    Kelley, A.P. Jr.; Arbtin, E.; St Pierre, R.

    1979-01-01

    Although out-of-pile tests may be expected to confirm individual phenomena models in core disruptive accident analysis codes, only in-pile tests are capable of verifying the extremely complex integrated model effects within the appropriate time phase for these accidents. For this reason, the GRIST-2 project, the purpose of which is to design and construct an in-pile helium loop capable of transient safety testing in the TREAT facility in Idaho, forms a cornerstone of the US GCFR safety program. The project organization, experiment program, facility, helium system design, and schedule which have been selected to meet the objectives are described

  3. Development of in-pile test and evaluation technology

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Yung Hwan; Park, Jong Man; Joo, Kee Nam; Park, Duk Keun; Park, Se Jin; Oh, Jong Myung; Kim, Tae Ryong; Park Jin Suk; Lee, Jae Han [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-08-01

    To develop the in-pile test and evaluation technologies using KMRR, basic design of instrumented capsule and auxiliary system for material irradiation test and the related studies are performed. First, reactor and test hole characteristics are summarized, and conceptual design requirements of capsule to KMRR are reviewed. And fundamental principles and criteria for the instrumented capsule design are summarized. Basic design and analysis of instrumented capsule are performed, and design of capsule supporting system are also performed and structural integrity of the system is analyzed. Based on the prior studies, test mock-ups are designed and manufactured, and thermohydraulic and vibration tests are prepared. And, as in-pile test evaluation technologies, KMRR neutron dosimetry and mechanical tests related to material irradiation are investigated. 67 figs, 30 tabs, 41 refs. (Author).

  4. Enhanced in-pile instrumentation at the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2011-07-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and realtime flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted. (authors)

  5. Enhanced In-Pile Instrumentation at the Advanced Test Reactor

    Science.gov (United States)

    Rempe, Joy L.; Knudson, Darrell L.; Daw, Joshua E.; Unruh, Troy; Chase, Benjamin M.; Palmer, Joe; Condie, Keith G.; Davis, Kurt L.

    2012-08-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

  6. MCNP calculations for the HCPB submodules in-pile test

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J. [Section Nuclear and Reactor Physics, ECN Nuclear Research, Petten (Netherlands)

    1998-11-01

    This report describes the MCNP calculations that have been performed for the Helium Cooled Pebble Bed (HCPB) Submodules In-pile Test that has been planned for irradiation in the materials testing High Flux Reactor (HFR) at Petten. In this test, four HSM-8 submodules will be placed at core position H4. The report presents the neutron flux and power density profiles to be expected in the submodules. For the gamma induced heating only a rough estimation could be made. In the HCPB submodules the total specific heating does not exceed (36.7 {+-} 2.9)[W/cc]. 8 refs.

  7. Design criteria and fabrication in-pile test section of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1997-10-01

    Safety state fuel test loop will be equipped in HANARO to obtain the development and betterments of advanced fuel and materials through the irradiation tests. The objective of this study is to determine the design criteria and technical specification of in-pile test section and to specify the manufacturing requirements of in-pile test section. HANARO fuel test loop was designed to meet the CANDU and PWR fuel testing and in-pile section will be manufactured and installed in HANARO. The design criteria and technical specification of in-pile test section could be used the fuel and materials design with for irradiation testing IPS of HANARO fuel test loop. This results will become guidances for the planning and programming of irradiation testing. (author). 12 refs., tabs., figs.

  8. The Phebus fission products in pile test programme

    International Nuclear Information System (INIS)

    Bussac, J.; Holtbecker, H.

    1988-01-01

    The need for quantifying the radioactive materials escaping from an LWR Nuclear Power Plant following a melt-down accident has arisen relatively late in the nuclear reactor technology development process. The TMI-2 accident in 1979 and the Chernobyl accident in 1986 have confirmed the importance of a good knowledge of phenomena which take place in a plant undergoing extreme accident conditions. After an extensive resarch effort which has involved the major nuclear countries for several years, we are now at the stage where a selective and converging attitude should be taken towards the wide range of problems underlying severe accidents. Selective, because we must understand what is important and what could be neglected. Converging, because we must arrive at a consensus at international level on the methods to treat these problems and a common understanding of the main scientific phenomena and the models to correctly represent them. After a large amount of separate effects tests and semi-integral in-pile and out-of-pile experiments, the Phebus FP project is being started as an experimental effort to quantify the relative importance of complicated processes and to give an insight into the interconnection of various mechanisms. The overall objective of this programme is to provide a qualified data base of integral in-pile experiments to validate codes dealing with FP transport in reactor core, primary cooling system and containment. This paper describes mainly the motivations and objectives of the Phebus PF programme

  9. Program of in-pile IASCC testing under the simulated actual plant condition. Development of technique for in-pile IASCC initiation test in JMTR

    International Nuclear Information System (INIS)

    Ugachi, Hirokazu; Tsukada, Takashi; Kaji, Yoshiyuki; Nagata, Nobuaki; Dozaki, Koji; Takiguchi, Hideki

    2003-01-01

    Irradiation assisted stress corrosion cracking (IASCC) is caused by the synergistic effects of neutron irradiation, stress and corrosion by high temperature water. It is, therefore, essential to perform in-pile SCC tests, which are material tests under the conditions simulating those of actual LWR operation, in order to clarify the precise mechanism of the phenomenon, though mainly out-of-pile SCC tests for irradiated materials have been carried out in this research field. There are, however, many difficulties to perform in-pile SCC tests. Performing in-pile SCC tests, essential key techniques must be developed. Hence as a part of development of the key techniques for in-pile SCC tests, we have embarked on development of the test technique which enables us to obtain the information concerning the effect of such parameters as applied stress level, water chemistry, irradiation conditions, etc. on the crack initiation behavior. Although it is difficult to detect the crack initiation in in-pile SCC tests, the crack initiation can be evaluated by the detection of specimen rupture if the cross section area of the specimen is small enough. Therefore, we adopted the uniaxial constant loading (UCL) test with small tensile specimens. This paper will describe the current status of the development of several techniques for in-pile SCC initiation tests in JMTR and the results of the performance tests of the designed testing unit using the out-of-pile loop facility. (author)

  10. A comprehensive in-pile test of PWR fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang Rixin; Zhang Shucheng; Chen Dianshan (Academia Sinica, Beijing (China). Inst. of Atomic Energy)

    1991-02-01

    An in-pile test of PWR fuel bundle has been conducted in HWRR at IAE of China. This paper describes the structure of the test bundle (3x3-2), fabrication process and quality control of the fuel rod, irradiation conditions and the main Post Irradiation Examination (PIE) results. The test fuel bundle was irradiated under the PWR operation and water chemistry conditions with an average linear power of 381 W/cm and reached an average burnup of 25010 MWd/tU of the fuel bundle. After the test, destructive and non-destructive examination of the fuel rods was conducted at hot laboratories. The fission gas release was 10.4-23%. The ridge height of cladding was 3 to 8 {mu}m. The hydrogen content of the cladding was 80 to 140 ppm. The fuel stack height was increased by 2.9 to 3.3 mm. The relative irradiation growth was about 0.11 to 0.17% of the fuel rod length. During the irradiation test, no fuel rod failure or other abnormal phenomena had been found by the on-line fuel failure monitoring system of the test loop and water sampling analysis. The structure of the test fuel assembly was left undamaged without twist and detectable deformation. (orig.).

  11. In-pile IASCC growth tests of irradiated stainless steels in JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko; Nakano, Junichi; Nishiyama, Yutaka [Japan Atomic Energy Agency, Nuclear Safety Research Center, Tokai, Ibaraki (Japan); Shibata, Akira; Ohmi, Masao [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation-assisted stress corrosion cracking (IASCC) test plan to evaluate in-situ effects of neutron/{gamma}-ray irradiation on crack growth of irradiated stainless steels under high-temperature water conditions for commercial boiling water reactors (BWRs) using the Japan Materials Testing Reactor (JMTR). Crack growth rate and its electrochemical corrosion potential (ECP) dependence are different between in-pile test and post irradiation examination (PIE), but these differences are not fully understood. The objectives of the present study are to understand the difference between in-pile and out-of-pile IASCC growth and to confirm the effectiveness of mitigation due to lowering ECP on in-pile crack growth rates. For in-pile crack growth tests, we have selected a large compact tension specimen such as 0.5T-CT because of validity of SCC growth test at a high stress intensity factor (K-value). For loading a 0.5T-CT specimen up to K - 30 MPa {radical}m, we have adopted a lever type loading unit for in-pile crack growth tests in the JMTR. In this report, an in-pile test plan for crack growth of irradiated SUS316L stainless steels under simulated BWR conditions in the JMTR and current status of development of in-pile crack growth test techniques are presented. (author)

  12. Advanced In-pile Instrumentation for Material and Test Reactors

    International Nuclear Information System (INIS)

    Rempe, J.L.; Knudson, D.L.; Daw, J.E.; Unruh, T.C.; Chase, B.M.; Davis, K.L.; Palmer, A.J.; Schley, R.S.

    2013-06-01

    The US Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified; and the progress of other development efforts is summarized. As reported in this paper, INL staff is currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating 'advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors. (authors)

  13. Advanced In-Pile Instrumentation for Materials Testing Reactors

    Science.gov (United States)

    Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T. C.; Chase, B. M.; Davis, K. L.; Palmer, A. J.; Schley, R. S.

    2014-08-01

    The U.S. Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified, and the progress of other development efforts is summarized. As reported in this paper, INL researchers are currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating `advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers, are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors.

  14. Seismic analysis of the in-pile test section

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. M.; Park, K. N.; Chi, D. Y.; Park, S. K.; Sim, B. S.; Ahn, S. H.; Lee, C. Y.; Kim, Y. J. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    This study gives the results of the seismic analysis of the IPS (In Pile Section) with lower bracket support. The results cover the natural frequency and seismic response of the IPS for the SSE and OBE events. An FE (Finite Element) model which includes the two vessels of the IPS and its support structure were analyzed by ABAQUS.

  15. Development of enclosure technique of tag gas for in-pile creep test

    International Nuclear Information System (INIS)

    Izaki, Toru; Ichikawa, Shoichi; Soroi, Masatoshi; Ito, Chikara

    2004-01-01

    Outline of the enclosure technique of tag gas for in-pile creep test is stated. In order to carry out in-pile creep test, the sample can enclose tag gas before the test and then the sample is inserted into MARICO-2 (Material Testing Rig with Temperature Control) in FBR 'JOYO' MK-III for the irradiation test. Outline of in-pile creep test using tag gas, enclosure system of tag gas, detection of a part of broken sample and identification of sample are explained. 126-, 128-, 129-, 131-, 132-, and 134-Xe are used as tag gases. The samples are identified by RIMS (Laser Resonance Ionization Mass Spectroscopy) in ppt order. ODS ferritic steel will be tested by the method in the next step. (S.Y.)

  16. Portable nucleic acid thermocyclers.

    Science.gov (United States)

    Almassian, David R; Cockrell, Lisa M; Nelson, William M

    2013-11-21

    A nucleic acid thermal cycler is considered to be portable if it is under ten pounds, easily carried by one individual, and battery powered. Nucleic acid amplification includes both polymerase chain reaction (e.g. PCR, RT-PCR) and isothermal amplification (e.g. RPA, HDA, LAMP, NASBA, RCA, ICAN, SMART, SDA). There are valuable applications for portable nucleic acid thermocyclers in fields that include clinical diagnostics, biothreat detection, and veterinary testing. A system that is portable allows for the distributed detection of targets at the point of care and a reduction of the time from sample to answer. The designer of a portable nucleic acid thermocycler must carefully consider both thermal control and the detection of amplification. In addition to thermal control and detection, the designer may consider the integration of a sample preparation subsystem with the nucleic acid thermocycler. There are a variety of technologies that can achieve accurate thermal control and the detection of nucleic acid amplification. Important evaluation criteria for each technology include maturity, power requirements, cost, sensitivity, speed, and manufacturability. Ultimately the needs of a particular market will lead to user requirements that drive the decision between available technologies.

  17. Minutes of the workshop on bases of in-pile irradiation tests

    International Nuclear Information System (INIS)

    1997-03-01

    The Workshop on Bases of In-pile Irradiation Tests was held on January 29th and 30th, 1997 at the Ibarakiken Sangyo Kaikan in Mito, Ibaraki. The purpose is to discuss upgrading an in-pile irradiation test, promoting the utilization of the research and testing reactors and also activating the research potential of JAERI transversely. Main topics are the role and future plan of the research and testing reactors, a challenge to an advanced irradiation test, development of peripheral techniques for irradiation tests and future trends of the in-pile irradiation test in the 21st century. It was mainly pointed out that the in-pile irradiation test based on an analytical method using interpolation and extrapolation procedures met a turning point and that the upgrading of the irradiation and testing method should be indispensable for regaining the latest frontiers of an irradiation study using the research and testing reactors. The new concepts were also proposed on the irradiation correlation and modeling for the design of innovative materials. It was also recognized the key issues of the irradiation study in future should be an advanced irradiation testing method which can combine various types of irradiation field and control the irradiation conditions freely. In the next century in which large accelerator or new neutron source competes with research and testing reactors for neutron irradiation tests, themes of research using in-pile irradiation tests will be upgrading of the light water reactor, development of fusion reactor, basic research, biological and medical research, radioisotope production and semiconductors manufacturing, etc. It was also concluded the research and testing reactors will keep their main role in neutron irradiation research in future. This report briefly summarizes the content of 16 presentations and the discussion. The result of the questionnaires on the utilization of research and testing reactors to the participants is also attached. (J.P.N.)

  18. In-Pile Tests for IASCC Growth Behavior of Irradiated 316L Stainless Steel under Simulated BWR Condition in JMTR

    Science.gov (United States)

    Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko; Nakano, Junichi; Nishiyama, Yutaka

    The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation test plan to evaluate in-situ effects of neutron/γ-ray irradiation on stress corrosion crack (SCC) growth of irradiated stainless steels using the Japan Materials Testing Reactor (JMTR). SCC growth rate and its dependence on electrochemical corrosion potential (ECP) are different between in-pile test and post irradiation examination (PIE). These differences are not fully understood because of a lack of in-pile data. This paper presents a systematic review on SCC growth data of irradiated stainless steels, an in-pile test plan for crack growth of irradiated SUS316L stainless steel under simulated BWR conditions in the JMTR, and the development of the in-pile test techniques.

  19. Thermal and mechanical analyses for the HCPB Submodules in-pile test

    Energy Technology Data Exchange (ETDEWEB)

    Bakker, K. [Fuels, Actinides and Isotopes, Nuclear Research and Consultancy Group NRG, Netherlands Energy Research Foundation ECN, Petten (Netherlands)

    1998-12-01

    A description is given of the Finite Element Method (FEM) and thermal and mechanical computations that have been performed for the Helium Cooled Pebble Bed (HCPB) Submodules in-pile tests, which have been planned for irradiation in the High Flux Reactor (HFR) at Petten. In this test, four submodules will be placed at core position H4. The report presents the temperature and stress distribution for the highest powered submodule of these four submodules. 9 refs

  20. Design of in-pile section monitoring system in fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Sim, B. S.; Park, K. N.; Park, S. K.; Chi, D. Y.; Lee, J. M.; Ahn, S. H.; Lee, C. Y.; Kim, Y. J. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    In this paper a brief summary of the monitoring system of In-Pile Section(IPS) is described. To meet the user requirements on the test fuel and irradiation conditions, various instruments are installed on the test fuel pin itself and the appropriate locations in IPS. The requirements and descriptions for instruments, gas supply system, and the data acquisition system to sample and record parameters are described.

  1. Manufacturing and performance tests of in-pile creep measuring machine of zirconium alloys

    International Nuclear Information System (INIS)

    Choi, Y.; Kim, B. G.; Kang, Y. H.

    2000-01-01

    A mock-up of the in-pile creep test machine of zirconium alloys for HANARO was designed and manufactured, which performance tests were carried. The dimension of the in-pile creep machine is 55 mm in diameter and 700 mm in length for HANARO, respectively. Load is transferred to specimen by through the working mechanisms in which the contraction of bellows by gas pressure moves a yoke and an upper grip connected to a specimen, simultaneously. It was observed that the extension of the specimen mounted in grips was transferred to a linear voltage differential transformer perfectly by a yoke and a push rod in a bearing. The displacement of specimen with applied pressure was determined with the LVDT and a pressure gauge, respectively. Resultant stress-strain behaviors of the specimen was determined by the displacement-applied gas pressure curve, which showed similar values obtained with a standard tensile test machine

  2. In-pile irradiation test program and safety analysis report of the KAERI fuel for HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Wan; Ryu, Woo Suck; Byun, Taek Sang; Park, Jong Man; Lee, Byung Chul; Kim, Hack No; Park, Hee Tae; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-05-01

    Localization of HANARO fuel has been carried out successfully, and design and fabrication technologies of the fuel are recently arrived the final stage of development. The performance of the fuel which has been fabricated in KAERI is confirmed through out-of-pile characterization, and the quality assurance procedure and assessment criteria are described. In order to verify the KAERI fuel, thus, in-pile irradiation test program of the KAERI fuel is scheduled in HANARO. This report summarizes the in-pile testing schedule, design documents of test rods and assemblies, fabrication history and out-of-pile characteristics of test rods, irradiation test condition and power history, post-irradiation examination scheme, linear power generation distribution, and safety analysis results. The design code for HANARO fuel is used to analyze the centerline temperature and swelling of the KAERI fuels. The results show that at 120 kW/m of linear power the maximum centerline temperature is 267 deg C which is much lower than the limitation temperature of 350 deg C, and that the swelling is 9.3 % at 95 at% lower than criterion of 20 %. Therefore, the KAERI fuels of this in-pile irradiation test is assessed to show good performance of integrity and safety in HANARO. 10 tabs., 7 figs., 3 refs. (Author).

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

    International Nuclear Information System (INIS)

    Zmitko, M.

    2002-01-01

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

  4. A probabilistic safety assessment of in-pile test loop in HWRR

    International Nuclear Information System (INIS)

    Cao Xuewu; Li Zhaohuan

    1991-07-01

    The PSA methodology has been applied to the in-pile test loop which is installed in the Heavy Water Research Reactor (HWRR). This loop is designed and operated for fuel assembly testing of the Qinshan PWR plant. This analysis is to assess the safety and to evaluate the design of this operating loop. The procedure and models are similar to a PSA on nuclear power plant. The major contents in the analysis consist of the familiarization of the object, the investigation and selection of accident initiators, setting events and fault trees, data collections, quantitative calculations, qualitative and result analyses and final conclusion. This analysis is only limited to the initiators of in-pile loop itself and possible errors made by operators during normal operation. The accident occurence is less than 10 -4 a -1 which may be recommended as an acceptance risk for safety operation of an in-pile test loop. Finally, suggestions have been raised to improve the design of test loop, especially in reducing operation errors by local operators

  5. In-pile test of tritium recovery from lithium oxide

    International Nuclear Information System (INIS)

    Kurasawa, Toshimasa; Yoshida, Hiroshi; Watanabe, Hitoshi; Takeshita, Hidefumi; Miyauchi, Takejiro; Matsui, Tomoaki

    1984-05-01

    In-situ tritium recovery experiment with sintered lithium oxide pellets was performed under a high neutron fluence in the JRR-2. The irradiation hole VT-10 is the vertical one in the fuel rods region of the reactor, and the neutron flux is as follows: the thermal neutron flux with the epithermal neutron; 1.12 x 10 14 n/cm 2 . sec, the fast neutron flux; 1.0 x 10 12 n/cm 2 . sec. Irradiation material is the four pellets of cylindrical Li 2 O with the size of 11mm-OD, 1.8mm-ID, 10mm-H, and their total weight is 6.67g(the apparent bulk density 86%TD). A sweep gas capsule with a inner heater was constructed for the present study. Irradiation temperatures were regulated in the high temperature range, 470 -- 760 0 C. Four cycles of irradiation tests were carried out from May to August in 1983, and the effective thermal neutron fluence and the burnup of 6 Li were 5.9 x 10 19 nvt and 0.24% of total lithium(natural abundance of Li), respectively. The amount of generated tritium was calculated to be 31.2Ci by using a value of the depression factor of the thermal neutron flux(0.148) and the effective neutron cross section(543b) for the 6 Li(n, α) 3 H reaction. Present report describes the tritium release behavior in the in-situ tritium recovery apparatus and discuss the effects of the moisture, the hydrogen spiking, the irradiation temperature, etc.. Problems relative to a real time measurement of a comparatively high tritium concentration(10 -1 -- 10 2 μCi/cm 3 ) in the helium gas stream were also investigated. (author)

  6. Evaluation of results from an in-pile creep test in the Studsvik R2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pettersson, Kjell [Entropy Materials, Stockholm (Sweden)

    2002-01-01

    An in-pile creep test with bowing of cladding tubes has been performed in a hot water loop in the Studsvik R2 reactor . One test was performed in the core and one outside the core. The out-of-pile sample showed some minor primary creep strain while the in-pile specimen deformed at a steady rate of 5x10{sup -7}/h . However, when the results were compared to a broader data base of Zircaloy in-pile creep it became clear that the creep deformation observed is a primary creep which occurs before the irradiation creep in Zircaloy reaches a constant steady state creep rate. This primary stage is interpreted as a consequence of the development of an irradiation induced microstructure in Zircaloy which does not reach a steady state until a dose of about 10{sup 21} n/cm{sup 2} . At this stage the steady state irradiation creep starts. From this interpretation it is concluded that it is quite feasible to use the test method on pre-irradiated material in which it can be expected that the steady state will be reached already after short irradiation times.

  7. Vanadium—lithium in-pile loop for comprehensive tests of vanadium alloys and multipurpose coatings

    Science.gov (United States)

    Lyublinski, I. E.; Evtikhin, V. A.; Ivanov, V. B.; Kazakov, V. A.; Korjavin, V. M.; Markovchev, V. K.; Melder, R. R.; Revyakin, Y. L.; Shpolyanskiy, V. N.

    1996-10-01

    The reliable information on design and material properties of self-cooled Li sbnd Li blanket and liquid metal divertor under neutron radiation conditions can be obtained using the concept of combined technological and material in-pile tests in a vanadium—lithium loop. The method of in-pile loop tests includes studies of vanadium—base alloys resistance, weld resistance under mechanical stress, multipurpose coating formation processes and coatings' resistance under the following conditions: high temperature (600-700°C), lithium velocities up to 10 m/s, lithium with controlled concentration of impurities and technological additions, a neutron load of 0.4-0.5 MW/m 2 and level of irradiation doses up to 5 dpa. The design of such an in-pile loop is considered. The experimental data on corrosion and compatibility with lithium, mechanical properties and welding technology of the vanadium alloys, methods of coatings formation and its radiation tests in lithium environment in the BOR-60 reactor (fast neutron fluence up to 10 26 m -2, irradiation temperature range of 500-523°C) are presented and analyzed as a basis for such loop development.

  8. The design of in-pile test section for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, K. N.; Lee, J. M.; Shim, B. S.; Zee, D. Y.; Park, S. H.; Ahn, S. H.; Lee, J. Y.; Kim, Y. J. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    As an equipment for nuclear fuel's general performance irradiation test in HANARO, Fuel Test Loop(FTL) has been developed that can irradiate the pin to the maximum number of 3 at the core irradiation hole(IR1 hole) by considering for it's utility and user's irradiation requirement. 3-Pin FTL consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). IPS consists for IPS Vessel assembly, In-Pool Piping, IPS Support, In-Pool Piping Support etc. Design that such IPS considers interference item consisted to do not bear in existing facilities by one. IVA that is connected to the OPS are controlled and regulated by means of system pressure, system temperature and the water quality. IPS Vessel assembly is consisted of outer pressure vessel, inner pressure vessel, IPS head, inner assembly and test fuel carrier. After 3-Pin FTL development which is expected to be finished by the 2006, FTL will be used for the irradiation test of the new PWR-type fuel and can maximize the usage of HANARO.

  9. In-pile creep test technique for zirconium alloys examination in BR-10 reactor channels

    International Nuclear Information System (INIS)

    Pevchikh, Yu.M.; Kruglov, A.S.; Troyanov, V.M.

    2002-01-01

    The irradiation enhanced creep phenomenon was discovered in stainless steels as a specific physical process accompanying high-intensity neutron flux irradiation in fast reactors. IPPE is also experienced in irradiation creep test activities, studying different types of materials under irradiation in BR-10 fast reactor. Series of in-channel type test facilities were constructed and tested in BR-10 reactor's 'dry' channels in order to carry out full-scale instrumented examination regarded to in-pile creep behaviour of different reactor materials. As a result, a specific test technique, named 'Tensometric method', has been developed and experimentally proved to be power enough in order to investigate irradiation creep of materials right in situ under neutron irradiation. The main peculiarity of test facility, which is constructed to apply the tensometric method, consists in absence of any special deformation-measurement cell at all. The in-pile creep strain measurement technique developed at IPPE is based on the non-direct measurement of specimen's deformation (either linear tensile strain or angular twisting one), which directly affects the loaded draws' tension parameters. Starting from 1993, in-pile creep experiments to investigate in-reactor creep behaviour of E110 and E635 zirconium alloys were carried out in BR-10. Experimental results and data collected during more than 20-year of BR-10 in-reactor creep test experience can be assumed as a strong evidence that the tensometric technique is a powerful instrument, which can give a chance to study different irradiation effects on reactor materials directly under irradiation. (author)

  10. In-pile tests of HTGR fuel particles and fuel elements

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Kolesov, V.S.; Deryugin, A.I.

    1985-01-01

    Main types of in-pile tests for specimen tightness control at the initial step, research of fuel particle radiation stability and also study of fission product release from fuel elements during irradiation are described in this paper. Schemes and main characteristics of devices used for these tests are also given. Principal results of fission gas product release measurements satisfying HTGR demands are illustrated on the example of fuel elements, manufactured by powder metallurgy methods and having TRISO fuel particles on high temperature pyrocarbon and silicon carbide base. (author)

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

    International Nuclear Information System (INIS)

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

    1989-10-01

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

  12. Data collection for the methods of in-pile materials testing

    International Nuclear Information System (INIS)

    Markina, N.V.; Rudkevich, A.V.; Lebedeva, E.E.

    1985-01-01

    Problems, relating to the creation of automated data banks intended for accumulatioon and systematization of data according to the methods of in-pile materials testing, taking into account their specific nature are discussed. The architecture of the bank, realized on the BESM-6 computer is described. The bank contains data on about 300 methods. The library of requests, which may be used by a user comprises about 20 descriptions. A new request description preparation and input take 5-20 min. The system answering time is from 30 sec to 3-4 min

  13. Integrated, digital experiment transient control and safety protection of an in-pile test

    International Nuclear Information System (INIS)

    Thomas, R.W.; Whitacre, R.F.; Klingler, W.B.

    1982-01-01

    The Sodium Loop Safety Facility experimental program has demonstrated that in-pile loop fuel failure transient tests can be digitally controlled and protected with reliability and precision. This was done in four nuclear experiments conducted in the Engineering Test Reactor operated by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory. Loop sodium flow and reactor power transients can be programmed to sponsor requirements and verified prior to the test. Each controller has redundancy, which reduces the effect of single failures occurring during test transients. Feedback and reject criteria are included in the reactor power control. Timed sequencing integrates the initiation of the controllers, programmed safety set-points, and other experiment actions (e.g., planned scram). Off-line and on-line testing is included. Loss-of-flow, loss-of-piping-integrity, boiling-window, transient-overpower, and local fault tests have been successfully run using this system

  14. Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

    Energy Technology Data Exchange (ETDEWEB)

    Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-09-01

    An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)

  15. Evaluation on sweep gas pressure drop in fusion blanket mock-up for in-pile test

    International Nuclear Information System (INIS)

    Ishitsuka, Etsuo; Kawamura, Hiroshi; Sagawa, Hisashi; Nagakura, Masaaki; Kanzawa, Toru.

    1993-03-01

    In the ITER/CDA (Conceptual Design Activity) of a tritium breeding blanket, Japan have proposed the pebble-typed blanket. The in-pile mock-up test will be preparing in JMTR (Japan Materials Testing Reactor) for Japanese engineering design with the pebble-typed blanket. Therefore, the He sweep gas pressure drop in the pebble bed was measured for the design of the mock-up used on in-pile test. From the results of this test, it was clear that the pressure drop was predicted on Kozeny- Carman's equation within +25 ∼ -60 %, and that the pressure drop was not affected by moisture concentration (< 100 ppm). (author)

  16. Review Report on the Design of In-Pile Test Section(IPS)

    International Nuclear Information System (INIS)

    Lee, Jong Min; Park, Kook Nam; Shim, Bong Sik; Lee, Chung Young; Chi, Dae Young; Park, Su Ki; Ahn, Sung Ho; Kim, Young Ki; Lee, Kye Hong; Kim, Kwan Hyun

    2009-01-01

    The In-Pile Test Section(IPS) accommodating fuel pins has loaded IR-1 hole in HANARO has double pressure vessel for the design conditions of 350 deg. C, 17.5 MPa and is composed of outer assembly and inner assembly. Dummy fuel, dummy fuel supports and Top flange are the main components in inner assembly and inner pressure vessel, outer pressure vessel and head are the components in outer assembly. The IPS at current status has dummy fuels and confirm the requirements for the IPS design improvements during the design, manufacturing and installation process. Head, Top Flange, Instrumentation Feed through, Lifting Eye, Fuel Carrier Leg, Retainer and Nozzle cover are the main parts that the design needs to be changed. This report suggest the needs for the IPS design modification and it would be reflected to the new IPS design which would accommodating test fuel pins

  17. Probabilistic Modeling of Updating Epistemic Uncertainty In Pile Capacity Prediction With a Single Failure Test Result

    Directory of Open Access Journals (Sweden)

    Indra Djati Sidi

    2017-12-01

    Full Text Available The model error N has been introduced to denote the discrepancy between measured and predicted capacity of pile foundation. This model error is recognized as epistemic uncertainty in pile capacity prediction. The statistics of N have been evaluated based on data gathered from various sites and may be considered only as a eneral-error trend in capacity prediction, providing crude estimates of the model error in the absence of more specific data from the site. The results of even a single load test to failure, should provide direct evidence of the pile capacity at a given site. Bayes theorem has been used as a rational basis for combining new data with previous data to revise assessment of uncertainty and reliability. This study is devoted to the development of procedures for updating model error (N, and subsequently the predicted pile capacity with a results of single failure test.

  18. Review Report on the Design of In-Pile Test Section(IPS)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jong Min; Park, Kook Nam; Shim, Bong Sik; Lee, Chung Young; Chi, Dae Young; Park, Su Ki; Ahn, Sung Ho; Kim, Young Ki; Lee, Kye Hong; Kim, Kwan Hyun

    2009-01-15

    The In-Pile Test Section(IPS) accommodating fuel pins has loaded IR-1 hole in HANARO has double pressure vessel for the design conditions of 350 deg. C, 17.5 MPa and is composed of outer assembly and inner assembly. Dummy fuel, dummy fuel supports and Top flange are the main components in inner assembly and inner pressure vessel, outer pressure vessel and head are the components in outer assembly. The IPS at current status has dummy fuels and confirm the requirements for the IPS design improvements during the design, manufacturing and installation process. Head, Top Flange, Instrumentation Feed through, Lifting Eye, Fuel Carrier Leg, Retainer and Nozzle cover are the main parts that the design needs to be changed. This report suggest the needs for the IPS design modification and it would be reflected to the new IPS design which would accommodating test fuel pins.

  19. Capsule development and utilization for material irradiation tests; study on the in-pile creep measuring method of zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong; Lee, Byung Kee; Lee, Jong Jea; Kim, Chang Sik; Kim, B. Hun; Cho, I. Sik [Sunmoon University, Asan (Korea)

    2002-02-01

    The final objective of this project is to obtain a design and fabrication technology of an in-pile creep test machine of zirconium alloys. First, design concepts of the in-pile creep test machines of various foreign countries were reviewed and a preliminary design of the equipment was carried. Second, the mock-up of the in-pile creep test machine was fabricated based on the preliminary design. The mock-up consisted of upper and lower grips, a yoke, a pressure chamber including a bellows, a push rod and LVDT. Each part was made of 304 L stainless steel. The average surface roughness of the parts was 1.0-14.7 {mu}m. The mock-up precisely determined an extension of a specimen by gas pressure. Finally, in-pile creep capsule was designed, fabricated and modified. High pure aluminum blocks were put in the capsule. Considering heat transfer coefficients of helium and nitrogen gases, the cooling efficiency is about 4 .deg. C at the condition of 300 .deg. C creep test. Yield strength, ultimate tensile strength and elongation at 300 .deg. C were 335 MPa, 591 MPa, 19.8%, respectively. which were lower than the values at room temperature, 353 MPa, 740 MPa, 12.5%. This study gave an important technology related to design, fabrication and performance tests of the in-pile creep test machine, which is applied to the fabrication of a special capsule and also used for the fundamental data for the fabrication of various in-pile creep capsules. 6 refs., 45 figs., 5 tabs. (Author)

  20. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    International Nuclear Information System (INIS)

    Agarwal, Vivek; Smith, James A.; Jewell, James Keith

    2015-01-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  1. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, Vivek [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Human Factors, Controls, and Statistics; Smith, James A. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design; Jewell, James Keith [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design

    2015-02-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  2. Design Improvement of Double Pressure Vessel in the In-pile Test Section

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

    To carry out an irradiation test of nuclear fuels, a nuclear fuel test rig should be fabricated and installed in the in-pile test section (IPS), which is installed in the reactor hall. While carrying out an irradiation test, sealing out coolant which passes through the test rig is one of the most important issues. In particular, although the double pressure vessel is assembled with the IPS head by two o-rings and six bolts, 15.5 MPa of highly pressurized coolant leaks through the gap between the vessel and IPS head. Because the temperature of the coolant in the test loop is 300 .deg. C , and the pool of HANARO is 40 .deg. C, the double pressure vessel is necessary to insulate them. Therefore, a new design to prevent the leakage of coolant needs to be developed. In this study, EB welding technique is considered to assemble the double pressure vessel and the IPS head, and their mechanical design is modified to enable the welding process. In this study, an improved design for sealing out the coolant at the pressure boundary between the double pressure vessel and the IPS head has been developed. An EB weld is applied to seal out the pressure boundary, and its sealing performance is verified by NDE, a cross section test, and a hydraulic pressure test. From the verification test results, the improved design can be used in fabricating the IPS for a nuclear fuel irradiation test.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-02-15

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

  4. In situ monitored in-pile creep testing of zirconium alloys

    Science.gov (United States)

    Kozar, R. W.; Jaworski, A. W.; Webb, T. W.; Smith, R. W.

    2014-01-01

    The experiments described herein were designed to investigate the detailed irradiation creep behavior of zirconium based alloys in the HALDEN Reactor spectrum. The HALDEN Test Reactor has the unique capability to control both applied stress and temperature independently and externally for each specimen while the specimen is in-reactor and under fast neutron flux. The ability to monitor in situ the creep rates following a stress and temperature change made possible the characterization of creep behavior over a wide stress-strain-rate-temperature design space for two model experimental heats, Zircaloy-2 and Zircaloy-2 + 1 wt%Nb, with only 12 test specimens in a 100-day in-pile creep test program. Zircaloy-2 specimens with and without 1 wt% Nb additions were tested at irradiation temperatures of 561 K and 616 K and stresses ranging from 69 MPa to 455 MPa. Various steady state creep models were evaluated against the experimental results. The irradiation creep model proposed by Nichols that separates creep behavior into low, intermediate, and high stress regimes was the best model for predicting steady-state creep rates. Dislocation-based primary creep, rather than diffusion-based transient irradiation creep, was identified as the mechanism controlling deformation during the transitional period of evolving creep rate following a step change to different test conditions.

  5. In-pile test results of HANA claddings in Halden research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Choi, Byoung Kwon; Jeong, Yong Hwan; Jung, Yun Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    It is a kind of facing tasks in the nuclear industry to develop advanced claddings for high burn-up fuel which is safer and more economical than the existing conventional ones. Since 1997, taking an initiative in KAERI, the Zr cladding development team has carried out the R and D activities for the development of the advanced claddings to be used in the high burn-up fuel (>70,000 MWD.MTU). The team had produced the advanced claddings (HANA, High-performance Alloy for Nuclear Application) from the patented composition and manufacturing process in the international collaboration with U.S. and Japan. Now, the HANA claddings have being demonstrated their good performances from the out-of-pile tests including the corrosion, creep, burst, tensile, microstructures LOCA, RIA, wear, and so on. In parallel to the out-of-pile performance tests, the HANA claddings are being undertaken to evaluate their in-pile properties in Halden research reactor. In this study, it is included the test overviews, conditions, and results of the HANA claddings in the Halden reactor.

  6. In-pile Creep Tests of Zircaloy Tubing in the Studsvik R2 Reactor. Final Report

    International Nuclear Information System (INIS)

    Tomani, Hans; Lindeloew, Ulf

    2000-12-01

    In this report are presented the findings of a prototype creep test on Zr4 guide tube specimens exposed in-pile and out-of-pile and stressed by constant bending moments. The calculated initial deflection curvature caused by the applied bending moment agrees very well with the measured initial values. Furthermore, the measurement results show excellent consistency. The dominant impact of neutron irradiation is clearly demonstrated. After 3 cycles (∼1300 hours) the irradiation creep is 4 times as large as the thermal creep. This is the case at least when fresh tube material is used. Irradiation creep progresses steadily, but the creep rate is not quite constant during the 3 irradiation cycles. The thermal creep, on the other hand, quickly saturates and there is hardly any further deflection after the second cycle for the specimen situated above the core. A limitation with the rig has been that the tube deflection became limited by the rig carrier body of the rig in the neutron flux (core) that disqualified the results of a fourth irradiation cycle actually performed in the fall of 1998. The test method appears to be suitable for testing the bending creep of different guide tube materials or designs under PWR conditions

  7. Design modification of the in-pile test section for increase of sealing capability

    Energy Technology Data Exchange (ETDEWEB)

    Hong, J T; Ahn, S H; Joung, C Y; Jeong, H Y; Lee, J M; Sim, B S [Department of Research Reactor Utilization and Development, Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2012-03-15

    Since KAERI established the fuel test loop (FTL) at HANARO in 2009, KAERI has carried out several experiments to verify the performances of the equipment. Based on the experiments, the design modification of the In-Pile test Section (IPS) has been processed to improve some difficulties such as difficulty in ejecting the inner assembly of the IPS from the pressure vessel, difficulty of the sealing process of the cooling water, etc. At first, because the cooling water of HANARO in KAERI consists of an open-pool type, if a certain shock is generated during the disassembly process, the cooling water can be spattered out of the pool. Therefore, two jacking bolts will be added on the top flange part of the inner assembly to decrease the shock. Second, at the pressure boundary of the IPS where MI-cables go through, the brazing process has been used to seal out the cooling water. However, because the length of the IPS is up to 5.5 meters, it is too difficult and time consuming to carry out the brazing process at the end part of the IPS. Therefore, the brazing process will be replaced with the mechanical sealing structure to simplify the assembly process. (author)

  8. In-pile experiments and test facilities proposed for fast reactor safety

    International Nuclear Information System (INIS)

    Grolmes, M.A.; Avery, R.; Goldman, A.J.; Fauske, H.K.; Marchaterre, J.F.; Rose, D.; Wright, A.E.

    1976-01-01

    The role of in-pile experiments in support of the resolution of fast breeder reactor safety and licensing issues has been re-examined, with emphasis on key safety issues. Experiment needs have been related to the specific characteristics of these safety issues and to realistic requirements for additional test facility capabilities which can be achieved and utilized within the next ten years. It is found that those safety issues related to the energetics of core disruptive accidents have the largest impact on new facility requirements. However, utilization of existing facilities with modifications can provide for a continuing increase in experiment capability and experiment results on a timely bases. Emphasis has been placed upon maximum utilization of existing facilities and minimum requirements for new facilities. This evaluation has concluded that a new Safety Test Facility, STF, along with major modifications to the EBR II facility, improvement in TREAT capabilities, the existing Sodium Loop Safety Facility and corresponding Support Facilities provide the essential elements of the Safety Research Experiment Facilities (SAREF) required for resolution of key issues

  9. Evaluation on sweep gas pressure drop in fusion blanket mock-up for in-pile test

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, Etsuo; Kawamura, Hiroshi; Sagawa, Hisashi (Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment); Nagakura, Masaaki; Kanzawa, Toru.

    1993-03-01

    In the ITER/CDA (Conceptual Design Activity) of a tritium breeding blanket, Japan have proposed the pebble-typed blanket. The in-pile mock-up test will be preparing in JMTR (Japan Materials Testing Reactor) for Japanese engineering design with the pebble-typed blanket. Therefore, the He sweep gas pressure drop in the pebble bed was measured for the design of the mock-up used on in-pile test. From the results of this test, it was clear that the pressure drop was predicted on Kozeny- Carman's equation within +25 [approx] -60 %, and that the pressure drop was not affected by moisture concentration (< 100 ppm). (author).

  10. Evaluation on sweep gas pressure drop in fusion blanket mock-up for in-pile test

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, Etsuo; Kawamura, Hiroshi; Sagawa, Hisashi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Nagakura, Masaaki; Kanzawa, Toru

    1993-03-01

    In the ITER/CDA (Conceptual Design Activity) of a tritium breeding blanket, Japan have proposed the pebble-typed blanket. The in-pile mock-up test will be preparing in JMTR (Japan Materials Testing Reactor) for Japanese engineering design with the pebble-typed blanket. Therefore, the He sweep gas pressure drop in the pebble bed was measured for the design of the mock-up used on in-pile test. From the results of this test, it was clear that the pressure drop was predicted on Kozeny- Carman`s equation within +25 {approx} -60 %, and that the pressure drop was not affected by moisture concentration (< 100 ppm). (author).

  11. [Wear intensity and surface roughness of microhybrid composite and ceramic occlusal veneers on premolars after the thermocycling and cyclic mechanical loading tests].

    Science.gov (United States)

    Zhang, H Y; Jiang, T; Cheng, M X; Zhang, Y W

    2018-02-18

    To evaluate the wear intensity and surface roughness of occlusal veneers on premolars made of microhybrid composite resin or two kinds of ceramics in vitro after the thermocycling and cyclic mechanical loading tests. In the study,24 fresh extracted human premolars without root canal treatment were prepared (cusps reduction of 1.5 mm in thickness to simulate middle to severe tooth wear, the inclinations of cusps were 20°). The prepared teeth were restored with occlusal veneers made of three different materials: microhybrid composite, heat-pressed lithium disilicate ceramic and computer-aided design/computer-aided manufacturing (CAD/CAM) lithium disilicate ceramic in the thickness of 1.5 mm. The occlusal veneers were cemented with resin cement. The specimens were fatigued using the thermocycling and cyclic mechanical loading tests after being stored in water for 72 h. The wear of specimens was measured using gypsum replicas and 3D laser scanner before and after the thermocycling and cyclic mechanical loading tests and the mean lost distance (mm) was used to indicate the level of wear. The surfaces of occlusal contact area were observed and the surface roughness was recorded using 3D laser scanning confocal microscope before and after the fatigue test. Differences between the groups were compared using ONE-way ANOVA(Pcomposite group, heat-pressed lithium disilicate ceramic group, and CAD/CAM lithium disilicate ceramic group was (-0.13±0.03) mm, (-0.05±0.01) mm and (-0.05±0.01) mm, the wear of microhybrid composite was significantly higher than the two ceramic groups(Pcomposite was significantly higher than the two ceramic groups(Pcomposite(P=0.005) and CAD/CAM lithium disilicate ceramic (P=0.010). From the view of wear speed, microhybrid composite was significantly higher than the two kinds of ceramics, but it was similar to enamel when the opposing tooth was natural. The surface roughness before the themocycling and cyclic mechanical loading test of microhybrid

  12. New Sensors for In-Pile Temperature Detection at the Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    Rempe, J.L.; Knudson, D.L.; Daw, J.E.; Condie, K.G.; Wilkins, S. Curtis

    2009-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. As a user facility, the ATR is supporting new users from universities, laboratories, and industry, as they conduct basic and applied nuclear research and development to advance the nation's energy security needs. A key component of the ATR NSUF effort is to develop and evaluate new in-pile instrumentation techniques that are capable of providing measurements of key parameters during irradiation. This paper describes the strategy for determining what instrumentation is needed and the program for developing new or enhanced sensors that can address these needs. Accomplishments from this program are illustrated by describing new sensors now available and under development for in-pile detection of temperature at various irradiation locations in the ATR.

  13. Heat transfer in the in-pile test section and penetration region of 3-pin fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Dae Young; Lee, Chung Young; Sim, Bong Shick; Park, Kook Nam; Park, Su Ki; Lee, Jong Min; Kim, Young Jin

    2003-12-01

    This report studies two types of normal heat transfer. One is the heat loss from the pressure vessel of In-Pile Test Section to HANARO pool water via IPS insulation gas gap. The other is the heat transfer of the Penetration Cooling Water System including the effect of the Foamglas insulator at the penetration region. The heat transfer from IPS insulation gas gap has been performed according to the detail design results from NUKEM. The heat loss also occurs at the concrete penetration region between the HANARO pool water and the FTL pipe gallery. The Foamglas insulator has been already installed at the MCW piping of the penetration region. This insulation effect has been reviewed. The Penetration Cooling Water System has been designed to fulfill the design requirement not to exceed the allowable temperature at the penetration concrete wall. The cooling ability and heat loss of PCW system has been reviewed with the insulation effect.

  14. 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.

  15. Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

    International Nuclear Information System (INIS)

    Daw, J.E.; Knudson, D.L.; Villard, J.F.; Liothin, J.; Destouches, C.; Rempe, J.L.; Matheron, P.; Lambert, T.

    2015-01-01

    Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physical property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were

  16. Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

    Energy Technology Data Exchange (ETDEWEB)

    Daw, J.E.; Knudson, D.L. [Idaho National Laboratory, Idaho Falls, ID 83415, (United States); Villard, J.F.; Liothin, J.; Destouches, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Rempe, J.L. [Rempe and Associates, LLC, Idaho Falls, ID, 83404 (United States); Matheron, P. [CEA, DEN, DEC, Uranium Fuels Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Lambert, T. [CEA, DEN, DEC, Innovative Fuel Design and Irradiation Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France)

    2015-07-01

    Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physical property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were

  17. In-pile test of Li 2TiO 3 pebble bed with neutron pulse operation

    Science.gov (United States)

    Tsuchiya, K.; Nakamichi, M.; Kikukawa, A.; Nagao, Y.; Enoeda, M.; Osaki, T.; Ioki, K.; Kawamura, H.

    2002-12-01

    Lithium titanate (Li 2TiO 3) is one of the candidate materials as tritium breeder in the breeding blanket of fusion reactors, and it is necessary to show the tritium release behavior of Li 2TiO 3 pebble beds. Therefore, a blanket in-pile mockup was developed and in situ tritium release experiments with the Li 2TiO 3 pebble bed were carried out in the Japan Materials Testing Reactor. In this study, the relationship between tritium release behavior from Li 2TiO 3 pebble beds and effects of various parameters were evaluated. The ( R/ G) ratio of tritium release ( R) and tritium generation ( G) was saturated when the temperature at the outside edge of the Li 2TiO 3 pebble bed became 300 °C. The tritium release amount increased cycle by cycle and saturated after about 20 pulse operations.

  18. Annealing tests of in-pile irradiated oxide coated U-Mo/Al-Si dispersed nuclear fuel

    Science.gov (United States)

    Zweifel, T.; Valot, Ch.; Pontillon, Y.; Lamontagne, J.; Vermersch, A.; Barrallier, L.; Blay, T.; Petry, W.; Palancher, H.

    2014-09-01

    U-Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800 °C of in-pile irradiated U-Mo/Al-Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500 °C and 670 °C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500 °C FGs are released from IDL/matrix interfaces. The second peak at 670 °C representing the main release of FGs originates from the interaction between U-Mo and matrix in the vicinity of the cladding.

  19. In-Pile Testing and Instrumentation for Development of Generation-IV Fuels and Materials. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2013-12-01

    For many years, the increase in efficiency in the production of nuclear electricity has been an economic challenge in many countries which have developed this kind of energy. The increase in fuel burnup and fuel residence time leads to a reduction in the volume of fresh fuel loaded and spent fuel discharged, respectively. More demanding nuclear fuel cycle parameters are combined with a need to operate nuclear power plants with maximal availability and load factors, in load-follow mode and with longer fuel cycles. In meeting these requirements, fuel has to operate in a demanding environment of high radiation fields, high temperatures, high mechanical stresses and high coolant flow. Requirements of increased fuel reliability and minimal fuel failures also remain in force. Under such circumstances, continuous development of more radiation resistant fuel materials, especially advanced cladding, careful and incremental examinations, and improved understanding and modelling of high burnup fuel behaviour are required. Following a recommendation of the IAEA Technical Working Group on Fuel Performance and Technology, the Technical Meeting on In-pile Testing and Instrumentation for Development of Generation-IV Fuels and Materials was held in Halden, Norway, on 21-24 August 2012. The purpose of the meeting was to review the current status and the progress in methods and technologies used for the in-pile testing of nuclear fuel achieved since the previous IAEA meeting on In-core Instrumentation and Reactor Core Assessment, also held in Halden in 2007. Emphasis was placed on advanced techniques applied for the understanding of high burnup fuel behaviour of water cooled power reactors that represent the vast majority of the current nuclear reactor fleet. However, the meeting also included papers and discussion on testing techniques applied or developed specifically for new fuel and structural materials considered for Generation-IV systems. The meeting was attended by 43

  20. Thermal-hydraulic analyses for in-pile SCWR fuel qualification test loops and SCWR material loop

    Energy Technology Data Exchange (ETDEWEB)

    Vojacek, A.; Mazzini, G.; Zmitkova, J.; Ruzickova, M. [Research Centre Rez (Czech Republic)

    2014-07-01

    One of the R&D directions of Research Centre Rez is dedicated to the supercritical water-cooled reactor concept (SCWR). Among the developed experimental facilities and infrastructure in the framework of the SUSEN project (SUStainable ENergy) is construction and experimental operation of the supercritical water loop SCWL focusing on material tests. At the first phase, this SCWL loop is assembled and operated out-of-pile in the dedicated loop facilities hall. At this out-of-pile operation various operational conditions are tested and verified. After that, in the second phase, the SCWL loop will be situated in-pile, in the core of the research reactor LVR-15, operated at CVR. Furthermore, it is planned to carry out a test of a small scale fuel assembly within the SuperCritical Water Reactor Fuel Qualification Test (SCWR-FQT) loop, which is now being designed. This paper presents the results of the thermal-hydraulic analyses of SCWL loop out-of-pile operation using the RELAP5/MOD3.3. The thermal-hydraulic modeling and the performed analyses are focused on the SCWL loop model validation through a comparison of the calculation results with the experimental results obtained at various operation conditions. Further, the present paper focuses on the transient analyses for start-up and shut-down of the FQT loop, particularly to explore the ability of system codes ATHLET 3.0A to simulate the transient between subcritical conditions and supercritical conditions. (author)

  1. In-Pile Experiment of a New Hafnium Aluminide Composite Material to Enable Fast Neutron Testing in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Douglas L. Porter; James R. Parry; Heng Ban

    2010-06-01

    A new hafnium aluminide composite material is being developed as a key component in a Boosted Fast Flux Loop (BFFL) system designed to provide fast neutron flux test capability in the Advanced Test Reactor. An absorber block comprised of hafnium aluminide (Al3Hf) particles (~23% by volume) dispersed in an aluminum matrix can absorb thermal neutrons and transfer heat from the experiment to pressurized water cooling channels. However, the thermophysical properties, such as thermal conductivity, of this material and the effect of irradiation are not known. This paper describes the design of an in-pile experiment to obtain such data to enable design and optimization of the BFFL neutron filter.

  2. In-pile data analysis of the comparative WWER/PWR test IFA-503.1. Final report.

    Energy Technology Data Exchange (ETDEWEB)

    Volkov, B.; Devold, H.; Ryazantzev, E.; Yakovlev, V.

    1999-04-15

    The comparative WWER/PWR test in IFA-503.1 was commenced in July 1995 and successfully finished at the end of November 1998. The main objective of the test was generation of representative and comparative data of standard WWER-440 fuel fabricated at the 'MSZ' Electrostal (Russia) and PWR type fuel manufactured at IFE Kjeller (Norway). The test assembly comprised two clusters, each with 3 WWER rods and 3 PWR type rods. Eight rods with two types of fuel were instrumented with expansion thermometers, four rods were equipped with both fuel stack elongation detectors and pressure transducers. All sensors worked satisfactorily during the test. The average burnups achieved in the lower and upper clusters were around 25 and 20 MWd/kgUO{sub 2}, respectively. Some difference in densification of the two types of fuel was revealed during the first irradiation period. However, the fuel temperatures and commencement of fuel stack swelling were similar despite this fact. At the end of the test the rig was moved to a higher flux position in the HBWR core with the aim of promoting FGR and to compare the behaviour of the two types of fuel under higher power. Pressure measurements indicated a comparable low FGR (around 1 percent) in both types of rods. The centreline temperatures measured in the PWR rods were very close to the Halden FGR threshold whilst the WWER fuel temperatures were slightly lower. Despite the differences found in the behaviour of the two types of fuel during the test, the analysis of the in-pile data showed that these differences would not affect the fuel efficiency, at least, up to the burnup achieved in the test. It is supposed that these differences can be related to the fuel microstructure, in particular to the fuel grain and pore sizes (author) (ml)

  3. In-pile data analysis of the comparative WWER/PWR test IFA-503.1. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Volkov, B.; Devold, H.; Ryazantzev, E.; Yakovlev, V

    1999-04-15

    The comparative WWER/PWR test in IFA-503.1 was commenced in July 1995 and successfully finished at the end of November 1998. The main objective of the test was generation of representative and comparative data of standard WWER-440 fuel fabricated at the 'MSZ' Electrostal (Russia) and PWR type fuel manufactured at IFE Kjeller (Norway). The test assembly comprised two clusters, each with 3 WWER rods and 3 PWR type rods. Eight rods with two types of fuel were instrumented with expansion thermometers, four rods were equipped with both fuel stack elongation detectors and pressure transducers. All sensors worked satisfactorily during the test. The average burnups achieved in the lower and upper clusters were around 25 and 20 MWd/kgUO{sub 2}, respectively. Some difference in densification of the two types of fuel was revealed during the first irradiation period. However, the fuel temperatures and commencement of fuel stack swelling were similar despite this fact. At the end of the test the rig was moved to a higher flux position in the HBWR core with the aim of promoting FGR and to compare the behaviour of the two types of fuel under higher power. Pressure measurements indicated a comparable low FGR (around 1 percent) in both types of rods. The centreline temperatures measured in the PWR rods were very close to the Halden FGR threshold whilst the WWER fuel temperatures were slightly lower. Despite the differences found in the behaviour of the two types of fuel during the test, the analysis of the in-pile data showed that these differences would not affect the fuel efficiency, at least, up to the burnup achieved in the test. It is supposed that these differences can be related to the fuel microstructure, in particular to the fuel grain and pore sizes (author) (ml)

  4. In-pile test of Li{sub 2}TiO{sub 3} pebble bed with neutron pulse operation

    Energy Technology Data Exchange (ETDEWEB)

    Tsuchiya, K. E-mail: tsuchiya@oarai.jaeri.go.jp; Nakamichi, M.; Kikukawa, A.; Nagao, Y.; Enoeda, M.; Osaki, T.; Ioki, K.; Kawamura, H

    2002-12-01

    Lithium titanate (Li{sub 2}TiO{sub 3}) is one of the candidate materials as tritium breeder in the breeding blanket of fusion reactors, and it is necessary to show the tritium release behavior of Li{sub 2}TiO{sub 3} pebble beds. Therefore, a blanket in-pile mockup was developed and in situ tritium release experiments with the Li{sub 2}TiO{sub 3} pebble bed were carried out in the Japan Materials Testing Reactor. In this study, the relationship between tritium release behavior from Li{sub 2}TiO{sub 3} pebble beds and effects of various parameters were evaluated. The (R/G) ratio of tritium release (R) and tritium generation (G) was saturated when the temperature at the outside edge of the Li{sub 2}TiO{sub 3} pebble bed became 300 deg. C. The tritium release amount increased cycle by cycle and saturated after about 20 pulse operations.

  5. Physics design of fast reactor safety test facilities for in-pile experiments

    International Nuclear Information System (INIS)

    Travelli, A.; Matos, J.E.; Snelgrove, J.L.; Shaftman, D.H.; Tzanos, C.P.; Lam, S.K.; Pennington, E.M.; Woodruff, W.L.

    1976-01-01

    A determined effort to identify and resolve current Fast Breeder Reactor safety testing needs has recently resulted in a number of conceptual designs for FBR safety test facilities which are very complex and diverse both in their features and in their purpose. The paper discusses the physics foundations common to most fast reactor safety test facilities and the constraints which they impose on the design. The logical evolution, features, and capabilities of several major conceptual designs are discussed on the basis of this common background

  6. Conceptual Design for the In-Pile Test Section(IPS) Improvement

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jong Min; Shim, Bong Sik; Lee, Chung Young; Chi, Dae Young; Park, Su Ki; Ahn, Sung Ho; Kim, Young Ki; Lee, Kye Hong; Kim, Kwan Hyun

    2009-01-15

    Conceptual design on the IPS, instrumentation and 1/4' Tubing, test fuel supports, test fuel inspection station and inner assembly O-ring replacement was suggested for their improvement. There is a need Jacking bolt on the Top flange for the inner assembly disassemble from the outer assembly and the replacement about the top flange O-ring to metal gasket to secure pressure boundary. Mechanical sealing was suggested instead of brazing. Instrumentation and tubing route should be modified for the reduction of bending and protection from unexpected occasion. Concept on the test fuel inspection station under the consideration of canal dimension. Top flange Bolt handling tool, O-ring replacement platform and O-ring replacement tool was designed for the O-ring replacement at the inner assembly.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  8. In-pile loop OWL-2 and irradiation tests done with it

    International Nuclear Information System (INIS)

    Suzuki, Shinobu; Ikeshima, Yoshiaki; Kawano, Masakatsu; Watanabe, Hiroyuki; Sato, Hitoshi; Tanaka, Isao

    1990-11-01

    The OWL-2 which was built in the JMTR as the biggest water loop in Japan has been operating for irradiation service since February 1972. The desired objective of the OWL-2, contributing to the development of various nuclear fuels and materials for the light water power reactor and to reactor engineering, has been so fully achieved that the OWL-2 is planned to be dismantled. After the dismantling, a loop, needed for the research and development of the breeding blanket for the fusion reactor, is going to be installed in place of the OWL-2 as a part of the JMTR Modification Program. This paper deals with the history of the OWL-2 with an emphasis on the technical affairs taken into consideration when designing the OWL-2, the irradiation tests, development of the turbine flowmeter, results of the surveillance test of the material of the in-reactor tube, the knowledge gained in the course of the investigation into the cause of transgranular stress corrosion cracking (TGSCC) which developed in the wall of the in-reactor tube, and countermeasures taken to prevent TGSCC from recurring. (author)

  9. In-Pile Sub-Miniature Fission Chambers Testing in BR2

    Science.gov (United States)

    Vermeeren, L.; Wéber, M.; Blandin, Ch.; Breaud, S.

    2003-06-01

    Three innovative sub-miniature fission chambers (SMFC), designed and manufactured at the Nuclear Measurement Systems Laboratory (LSMN) of CEA/Cadarache, were extensively tested in the BR2 research reactor at SCK•CEN, Mol. We present the experimental results for the (thermal) neutron sensitivity, the gamma-induced signal, the signal due to activation, the current picked up by the signal cable, the global current/voltage characteristics and the long term behaviour up to a thermal neutron fluence of 2.7·1021 n/cm2. We also compare the data with results from calculations with our FCD computer code. The onset of the saturation domain is well predicted by FCD; the neutron sensitivities can be accounted for perfectly after a refinement of the FCD model.

  10. In-pile testing of ITER first wall mock-ups at relevant thermal loading conditions

    Science.gov (United States)

    Litunovsky, N.; Gervash, A.; Lorenzetto, P.; Mazul, I.; Melder, R.

    2009-04-01

    The paper describes the experimental technique and preliminary results of thermal fatigue testing of ITER first wall (FW) water-cooled mock-ups inside the core of the RBT-6 experimental fission reactor (RIAR, Dimitrovgrad, Russia). This experiment has provided simultaneous effect of neutron fluence and thermal cycling damages on the mock-ups. A PC-controlled high-temperature graphite ohmic heater was applied to provide cyclic thermal load onto the mock-ups surface. This experiment lasted for 309 effective irradiation days with a final damage level (CuCrZr) of 1 dpa in the mock-ups. About 3700 thermal cycles with a heat flux of 0.4-0.5 MW/m 2 onto the mock-ups were realized before the heater fails. Then, irradiation was continued in a non-cycling mode.

  11. 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

  12. Development of in-pile instruments for fuel and material irradiation tests

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Akira; Kitagishi, Shigeru; Kimura, Nobuaki; Saito, Takashi; Nakamura, Jinichi; Ohmi, Masao; Izumo, Hironobu; Tsuchiya, Kunihiko [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    To get measurement data with high accuracy for fuel and material behavior studies in irradiation tests, two kinds of measuring equipments have been developed; these are the Electrochemical Corrosion Potential (ECP) sensor and the Linear Voltage Differential Transformer (LVDT) type gas pressure gauge. The ECP sensor has been developed to determine the corrosive potential under high temperature and high pressure water conditions. The structure of the joining parts was optimized to avoid stress concentration. The ECP sensor showed enough performance at 288degC and at 9MPa conditions. The LVDT type rod inner gas pressure gauge has been developed to measure gas pressure in a fuel element during neutron irradiation. To perform stable measurements with high accuracy under high temperature, high pressure and high dosed environment, the coil material of LVDT was changed to MI cable. As a result of this development, the LVDT type gas pressure gauge showed high accuracy within 1.8% of a full scale, and good stability. (author)

  13. Development of in-pile instruments for fuel and material irradiation tests

    International Nuclear Information System (INIS)

    Shibata, Akira; Kitagishi, Shigeru; Kimura, Nobuaki; Saito, Takashi; Nakamura, Jinichi; Ohmi, Masao; Izumo, Hironobu; Tsuchiya, Kunihiko

    2012-01-01

    To get measurement data with high accuracy for fuel and material behavior studies in irradiation tests, two kinds of measuring equipments have been developed; these are the Electrochemical Corrosion Potential (ECP) sensor and the Linear Voltage Differential Transformer (LVDT) type gas pressure gauge. The ECP sensor has been developed to determine the corrosive potential under high temperature and high pressure water conditions. The structure of the joining parts was optimized to avoid stress concentration. The ECP sensor showed enough performance at 288degC and at 9MPa conditions. The LVDT type rod inner gas pressure gauge has been developed to measure gas pressure in a fuel element during neutron irradiation. To perform stable measurements with high accuracy under high temperature, high pressure and high dosed environment, the coil material of LVDT was changed to MI cable. As a result of this development, the LVDT type gas pressure gauge showed high accuracy within 1.8% of a full scale, and good stability. (author)

  14. Study of In-Pile test facility for fast reactor safety research: performance requirements and design features

    Energy Technology Data Exchange (ETDEWEB)

    Nonaka, N.; Kawatta, N.; Niwa, H.; Kondo, S.; Maeda, K

    1996-12-31

    This paper describes a program and the main design features of a new in-pile safety facility SERAPH planned for future fast reactor safety research. The current status of R and D on technical developments in relation to the research objectives and performance requirements to the facility design is given.

  15. Verification of tritium production evaluation procedure using Monte Carlo code MCNP for in-pile test of fusion blanket with JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Y. E-mail: nagao@jmtr.oarai.jaeri.go.jp; Nakamichi, K.; Tsuchiya, M.; Ishitsuka, E.; Kawamura, H

    2000-11-01

    To evaluate exactly the total amount of tritium production in tritium breeding materials during in-pile test with JMTR, the 'tritium monitor' has been produced and evaluation of total tritium generation was done by using 'tritium monitor' in preliminary in-pile mock-up, and verification of procedure concerning tritium production evaluation was conducted by using Monte Carlo code MCNP and nuclear cross section library of FSXLIBJ3R2. Li-Al alloy (Li 3.4 wt.%, 95.5% enrichment of {sup 6}Li) was selected as tritium monitor material for the evaluation on the total amount of tritium production in high {sup 6}Li enriched materials. From the results of preliminary experiment, calculated amounts of total tritium production at each 'tritium monitor', which was installed in the preliminary in-pile mock-up, were about 50-290% higher than the measured values. Concerning tritium measurement, increase of measurement error in tritium leak form measuring system to measure small amount of tritium (0.2-0.7 mCi in tritium monitor) was found in the results of present experiment. The tendency for overestimation of calculated thermal neutron flux in the range of 1-6x10{sup 13} n cm{sup -2} per s was found in JMTR and the reason may be due to the beryllium cross section data base in JENDL3.2.

  16. Verification of tritium production evaluation procedure using Monte Carlo code MCNP for in-pile test of fusion blanket with JMTR

    International Nuclear Information System (INIS)

    Nagao, Y.; Nakamichi, K.; Tsuchiya, M.; Ishitsuka, E.; Kawamura, H.

    2000-01-01

    To evaluate exactly the total amount of tritium production in tritium breeding materials during in-pile test with JMTR, the 'tritium monitor' has been produced and evaluation of total tritium generation was done by using 'tritium monitor' in preliminary in-pile mock-up, and verification of procedure concerning tritium production evaluation was conducted by using Monte Carlo code MCNP and nuclear cross section library of FSXLIBJ3R2. Li-Al alloy (Li 3.4 wt.%, 95.5% enrichment of 6 Li) was selected as tritium monitor material for the evaluation on the total amount of tritium production in high 6 Li enriched materials. From the results of preliminary experiment, calculated amounts of total tritium production at each 'tritium monitor', which was installed in the preliminary in-pile mock-up, were about 50-290% higher than the measured values. Concerning tritium measurement, increase of measurement error in tritium leak form measuring system to measure small amount of tritium (0.2-0.7 mCi in tritium monitor) was found in the results of present experiment. The tendency for overestimation of calculated thermal neutron flux in the range of 1-6x10 13 n cm -2 per s was found in JMTR and the reason may be due to the beryllium cross section data base in JENDL3.2

  17. In-pile and out-of-pile testing of a molybdenum-uranium dioxide cermet fueled themionic diode

    Science.gov (United States)

    Diianni, D. C.

    1972-01-01

    The behavior of Mo-UO2 cermet fuel in a diode for thermionic reactor application was studied. The diode had a Mo-0.5 Ti emitter and niobium collector. Output power ranged from 1.4 to 2.8 W/cm squared at emitter and collector temperatures of 1500 deg and 540 C. Thermionic performance was stable within the limits of the instrumentation sensitivity. Through 1000 hours of in-pile operation the emitter was dimensionally stable. However, some fission gases (15 percent) leaked through an inner clad imperfection that occurred during fuel fabrication.

  18. Pre-test prediction and post-test analysis of PWR fuel rod ballooning in the MT-3 in-pile LOCA simulation experiment in the NRU reactor

    International Nuclear Information System (INIS)

    Donaldson, A.T.; Horwood, R.A.; Healey, T.

    1983-01-01

    The USNRC and the UKAEA have jointly funded a series of in-pile LOCA simulation experiments in the Canadian NRU reactor in order to secure further information on the thermal hydraulic and clad deformation response of PWR fuel rod bundles. Test MT-3 in the series was performed using reflood rate and rod internal pressure conditions specified by the UK nuclear industry. The parameters were selected to ensure the development of a near-isothermal clad temperature history during which zircaloy was required to balloon and rupture near the alpha-alpha/beta phase transition. Specification of the reflood rate conditions was assisted by the performance of a precursor test on an unpressurised rod bundle and by complementary application of appropriate thermal hydraulic analyses. Identification of the rod internal pressure needed to cause ballooning and rupture was achieved using a creep deformation model, BALLOON, in conjunction with the clad thermal history defined by the prior thermal hydraulic test. This paper presents the basis of the BALLOON analysis and describes its application in calculating the fill gas pressure for rods MT-3, their axial ballooning profile and the clad temperature at peak radial strain elevations. (author)

  19. Test-element assembly and loading parameters for the in-pile test of HCPB ceramic pebble beds

    Energy Technology Data Exchange (ETDEWEB)

    Laan, J.G. van der E-mail: vanderlaan@nrg-nl.com; Boccaccini, L.V.; Conrad, R.; Fokkens, J.H.; Jong, M.; Magielsen, A.J.; Pijlgroms, B.J.; Reimann, J.; Stijkel, M.P.; Malang, S

    2002-11-01

    In the framework of developing the helium cooled pebble-bed (HCPB) blanket an irradiation test of pebble-bed assemblies is prepared at the HFR Petten. The test objective is to concentrate on the effect of neutron irradiation on the thermal-mechanical behaviour of the HCPB breeder pebble-bed at DEMO representative levels of temperature and defined thermal-mechanical loads. The paper reports on the project status, and presents the results of pre-tests, material characteristics, the manufacturing of the pebble-bed assemblies, and the nuclear and thermo-mechanical loading parameters.

  20. Manufacturing of In-Pile Test Section(IPS) Mock-up for the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. M.; Park, K. N.; Chi, D. Y. (and others)

    2005-10-15

    Manufacturing process of IPS Mock-up was initiated in late of 2003 with DAEWOO Precision industries Company. Manufacturing drawings due to detail drawings are composed of Outer assembly and Inner assembly. Welding of IPS Mock-up was performed by the GMAW(Gas Metal Arc Welding) process. After the welding process, non-destructive examination was conducted. Leak test was performed to the Main cooling water part and Neon gas inter-space gap part by the He gas injection with the pressure of 6.0 kg{sub f}/cm{sup 2} and 30 minutes holding time. the result was shown that there was no leak at the Neon gas inter-space gap part but leak was occurred at Main cooling water part according to imperfect screw of purge plug. so, it was re-finished and test was performed to certify the leak tightness. To satisfy the HANARO Limiting Operation Condition, IPS should be tested ahead of installation at the HANARO reactor by the use of test facilities. IPS Mock-up and its test facilities will be designed and used for the test of 'HANARO flow tube pressure drop', 'IPS inner pressure drop' and 'IPS inner vibration'.

  1. In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

    1983-01-01

    Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio

  2. In-pile test results of U-silicide or U-nitride coated U-7Mo particle dispersion fuel in Al

    Science.gov (United States)

    Kim, Yeon Soo; Park, J. M.; Lee, K. H.; Yoo, B. O.; Ryu, H. J.; Ye, B.

    2014-11-01

    U-silicide or U-nitride coated U-Mo particle dispersion fuel in Al (U-Mo/Al) was in-pile tested to examine the effectiveness of the coating as a diffusion barrier between the U-7Mo fuel kernels and Al matrix. This paper reports the PIE data and analyses focusing on the effectiveness of the coating in terms of interaction layer (IL) growth and general fuel performance. The U-silicide coating showed considerable success, but it also provided evidence for additional improvement for coating process. The U-nitride coated specimen showed largely inefficient results in reducing IL growth. From the test, important observations were also made that can be utilized to improve U-Mo/Al fuel performance. The heating process for coating turned out to be beneficial to suppress fuel swelling. The use of larger fuel particles confirmed favorable effects on fuel performance.

  3. In-pile test results of U-silicide or U-nitride coated U-7Mo particle dispersion fuel in Al

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Park, J.M.; Lee, K.H.; Yoo, B.O. [Korea Atomic Energy Research Institute, 989-111 Daedeokdaero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Ryu, H.J. [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Ye, B. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2014-11-15

    U-silicide or U-nitride coated U-Mo particle dispersion fuel in Al (U-Mo/Al) was in-pile tested to examine the effectiveness of the coating as a diffusion barrier between the U-7Mo fuel kernels and Al matrix. This paper reports the PIE data and analyses focusing on the effectiveness of the coating in terms of interaction layer (IL) growth and general fuel performance. The U-silicide coating showed considerable success, but it also provided evidence for additional improvement for coating process. The U-nitride coated specimen showed largely inefficient results in reducing IL growth. From the test, important observations were also made that can be utilized to improve U-Mo/Al fuel performance. The heating process for coating turned out to be beneficial to suppress fuel swelling. The use of larger fuel particles confirmed favorable effects on fuel performance.

  4. The MOX fuel behaviour test IFA-597.4/.5/.6/.7; Summary of in-pile fuel temperature and gas release data

    Energy Technology Data Exchange (ETDEWEB)

    Koike, Hisashi

    2003-11-15

    It is considered important to study the in-reactor behaviour of MOX fuel in order to enhance the database on such fuel. For this reason, IFA-597.4/.5/.6/.7 were included in the joint research programme of the Halden Project. The series of tests, containing two MIMAS-MOX fuel rods, both equipped with a fuel centre thermocouple and a pressure bellows transducer, has been irradiated in the Halden Reactor since July 1997 under HBWR conditions. The objectives of the test series were to study the thermal and fission gas release (FGR) behaviour of MOX fuel and to explore potential differences in behaviour between solid and hollow pellets. One of the rods had mainly solid pellets, while the other contained only hollow pellets. Both rods had an initial Pu-fissile enrichment of 6.07%. The cladding outside diameter was 9.50 mm, and the initial fuel-clad gap was 180 mum. In the course of the test, power upratings for FGR studies of the MOX fuel were planned at burnup intervals of about 10 MWd/kg MOX. The power uprating was successfully performed at approx10 MWd/kg MOX, where the estimated fuel peak temperature of the solid pellets exceeded the FGR threshold temperature for UO{sub 2} fuel, while that of the hollow pellets remained below the threshold. For the solid fuel, the temperature at onset of FGR was consistent with the empirical threshold temperature for UO{sub 2} fuel. For the hollow fuel, gas release was observed at temperatures below the threshold. FGRs at the end-of-life were approx17% for the solid pellet rod and approx14% for the hollow pellet rod, respectively. As a result of discussions in HPG meetings, IFA-597.7 was unloaded in January 2002. PIE was carried out to check in-pile pressure measurements and examine fuel structural characteristics. The discharge burn-up of the MOX fuel was 32 MWd/kg MOX as determined from in-pile power data. This report supersedes HWR-712 (June 2002) previously issued on in-pile data from IFA-597.4/5/6/7. (Author)

  5. In-pile testing of ITER first wall mock-ups at relevant thermal loading conditions in the LVR-15 nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kysela, Jan [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Entler, Slavomir, E-mail: slavomir.entler@cvrez.cz [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Vsolak, Rudolf; Klabik, Tomas [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Zlamal, Ondrej [CEZ, Duhova 2/1444, 140 53 Praha 4 (Czech Republic); Bellin, Boris; Zacchia, Francesco [Fusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2015-10-15

    Highlights: • Irradiated thermal fatigue testing of the ITER primary first wall mock-ups. • Cyclic heat flux of 0.5 MW/m{sup 2} in the neutron field of the nuclear reactor core. • 17,040 thermal cycles. • Radiation damage in the range of 0.41–1.17 dpa depending on the material. - Abstract: The TW3 in-pile rig enabled the thermal fatigue testing of ITER primary first wall mock-ups in the core of the nuclear reactor. This experiment investigated the neutron irradiation influence on the design performance under high heat flux testing. A thermal flux of 0.5 MW/m{sup 2} in the neutron field of the core of the LVR-15 nuclear reactor was applied. Within the scope of the tests with simultaneous neutron irradiation, the TW3 rig reached a record of 17,040 thermal cycles with the radiation damage in the range of 0.41–1.17 dpa depending on the material. Even after a high number of thermal cycles, while being irradiated by neutrons, no damage of the tested mock-ups was visually observed. Further testing and analysis will follow in the Forschungszentrum Juelich.

  6. Simulation of the in-pile test Phebus-FPT3 using ASTEC V2 and ATHLET-CD 2.1A

    Energy Technology Data Exchange (ETDEWEB)

    Kruse, Philipp; Koch, Marco K. [Bochum Univ. (Germany). Chair of Energy Systems and Energy Economics

    2011-07-01

    The Phebus-FPT programme, initiated in 1988 by the 'Institut de Radioprotection et de Surete Nucleaire' (IRSN) and the Joint Research Centre (JRC) of the European Commission (EC), was performed in the Phebus facility operated by 'Commissariat a'Energie Atomique' (CEA). The facility represent a 900 MWe Pressurized Water Reactor (PWR) scaled down by a factor 1:5000 which objective is to study fuel degradation and the subsequent release, transport and retention of fission products, structure, control rod and fuel materials, in case of a severe accident. The Phebus-FPT programme consists of integral in-pile tests, varying the fuel burn-up and geometry, the control rod nature, the thermal hydraulic conditions in the bundle and through the experimental circuit as well as in the containment. In primary, the integral experiments should outline a detailed description of the main phenomena of core degradation, fission product release and transport as well as radionuclide interactions. Due to that it is possible to analyse the physical and chemical processes due to a severe accident. With the ascertained data, an evaluation of the accident management measures could be made as well. A secondary aim of the Phebus tests was to enable model development and evaluation of severe accident codes such like ASTEC and ATHLET-CD. (orig.)

  7. The effect of thermocycling on tensile bond strength of two soft liners.

    Science.gov (United States)

    Geramipanah, Farideh; Ghandari, Masoumeh; Zeighami, Somayeh

    2013-09-01

    Failure of soft liners depends mostly on separation from the denture base resin; therefore measurement of the bond strength is very important. The purpose of this study was to compare the tensile bond strength of two soft liners (Acropars, Molloplast-B) to denture base resin before and after thermocycling. Twenty specimens fromeach of the two different soft liners were processed according to the manufacturer's instructions between two polymethyl methacrylate (PMMA) sheets. Ten specimens in each group were maintained in 37°C water for 24 hours and 10 were thermocycled (5000 cycles) among baths of 5° and 55°C. The tensile bond strength was measured using a universal testing machine at a crosshead speed of 5 mm/min. Mode of failure was determined with SEM (magnification ×30). Two-way ANOVA was used to analyze the data. The mean and standard deviation of tensile bond strength of Acropars and Molloplast-B before thermocycling were 6.59±1.85 and1.51±0.22 MPa, respectively and 5.89±1.52 and1.37±0.18 MPa, respectively after thermocycling. There was no significant difference before and after thermocycling. Mode of failure in Acropars and Molloplast-B were adhesive and cohesive, respectivley. The bond strength of Acropars was significantly higher than Molloplast-B (P<0.05).

  8. Microleakage under Orthodontic Metal Brackets Bonded with Three Different Bonding Techniques with/without Thermocycling

    Directory of Open Access Journals (Sweden)

    Berahman Sabzevari

    2013-12-01

    Full Text Available Introduction: The aim of this study was to compare the microleakage of beneath the orthodontic brackets bonded with 3 different bonding techniques and evaluate the effect of thermocycling. Methods: One hundred and twenty premolars were randomly divided into 6 groups, received the following treatment: group 1: 37% phosphoric acid gel+Unite primer+Unite adhesive, group 2: 37% phosphoric acid gel+ Transbond XT primer+Transbond XT adhesive, group 3: Transbond plus Self Etching Primer (TSEP+Transbond XT adhesive. Groups 4, 5, and 6 were similar to groups 1, 2, and 3, respectively. Evaluation of microleakage was done following to thermocycling test. After bonding, the specimens were sealed with nail varnish except for 1 mm around the brackets and then stained with 0.5% basic fuchsine. The specimens were sectioned at buccolingual direction in 2 parallel planes and evaluated under a stereomicroscope to determine the amount of microleakage at bracket-adhesive and adhesive-enamel interfaces from gingival and occlusal margins. Results: Microleakage was observed in all groups, and increased significantly after thermocycling at some interfaces of Unite adhesive group and conventional etching+Transbond XT adhesive group, but the increase was not significant in any interface of TSEP group. With or without thermocycling, TSEP displayed more microleakage than other groups. In most groups, microleakage at gingival margin was significantly higher than occlusal margin. Conclusion: Thermocycling and type of bonding technique significantly affect the amount of microleakage.

  9. Status of the in-pile test of HCPB pebble-bed assemblies in the HFR Petten

    Energy Technology Data Exchange (ETDEWEB)

    Laan, J.G. van der; Fokkens, J.H.; Hofmans, H.E.; Jong, M.; Magielsen, A.J.; Pijlgroms, B.J.; Stijkel, M.P. [NRG, Petten (Netherlands); Conrad, R. [JRC, Inst. for Energy, Petten (Netherlands); Malang, S.; Reimann, J. [FZK, Karlsruhe (Germany); Roux, N. [CEA Saclay (France)

    2002-06-01

    In the framework of developing the helium cooled pebble-bed (HCPB) blanket an irradiation test of pebble-bed assemblies is prepared at the HFR Petten. The test objective is to concentrate on the effect of neutron irradiation on the thermal-mechanical behaviour of the HCPB breeder pebble-bed at DEMO representative levels of temperature and defined thermal-mechanical loads. The basic test elements are EUROFER-97 cylinders with a horizontal bed of ceramic breeder pebbles sandwiched between two beryllium beds. The pebble beds are separated by EUROFER-97 steel plates. The heat flow is managed such as to have a radial temperature distribution in the ceramic breeder pebble-bed as flat as reasonably possible. The paper reports on the project status, and presents the results of pre-tests, material characteristics, the manufacturing of the pebble-bed assemblies, and the nuclear and thermo-mechanical loading parameters. (orig.)

  10. Neutronic design of pulse operation simulating device for in-pile functional test of fusion blanket by MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Yoshiharu; Nakamichi, Masaru; Kawamura, Hiroshi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan)

    2000-03-01

    The pulse operation of a fusion reactor can be simulated in a fission reactor by controlling the neutron flux entering a test section by using a rotating 'hollow cylinder with window' made of hafnium. The rotating cylinder is installed between the test section and the fixed outer neutron absorber cylinder and is also made of hafnium with an opening in the direction to the core center. For gathering engineering data for the tritium breeding blanket such as characteristics of temperature change, tritium release and recovery, etc., it is desirable that the ratio of minimum to maximum thermal neutron fluxes is greater than 1:10. Design calculations were performed for the test assembly which considered local neutronic effects and the mechanical constraints of the device. From the results of these calculations, the ratio of minimum to maximum thermal neutron flux under irradiation would be about 1:10 using a pulse operation simulating device which has a thickness of 6.5 mm and a 150deg window angle for the rotating hollow cylinder and 5.0 mm in thickness of fixed neutron absorber. (author)

  11. Lay out and materials for in pile tritium transport testing of breeder-inside-tube pin assemblies

    International Nuclear Information System (INIS)

    Alvani, C.; Casadio, S.; Mancini, M.R.; Nannetti, C.A.; Avon, J.; Ravel, S.; Pruzzo, G.; Terrosi; Roux, N.; Terlain, A.; and others.

    1994-01-01

    An irradiation experiment (90 FPD in SILOE reactor) has been designed in order to evaluate the in-situ effect of red-ox power of sweeping gas (helium with 100 vpm of H 2 /H 2 O with relative concentrations varying from pure H 2 to pure H 2 O ) on tritium removal from LiAlO 2 and Li 2 ZrO 3 ; and tritium permeation through AlSl-316L SS tubes with bare and coated surfaces. The conditions and materials explored were selected in order to test possible improvements with respect to critical issues for the 'Breeder Inside Tube' (BIT) blanket concept development. (author) 6 refs.; 4 figs.; 2 tabs

  12. Visual in-pile fuel disruption experiments

    International Nuclear Information System (INIS)

    Cano, G.L.; Ostensen, R.W.; Young, M.F.

    1978-01-01

    In a loss-of-flow (LOF) accident in an LMFBR, the mode of disruption of fuel may determine the probability of a subsequent energetic excursion. To investigate these phenomena, in-pile disruption of fission-heated irradiated fuel pellets was recorded by high speed cinematography. Instead of fuel frothing or dust-cloud breakup (as used in the SAS code) massive and very rapid fuel swelling, not predicted by analytical models, occurred. These tests support massive fuel swelling as the initial mode of fuel disruption in a LOF accident. (author)

  13. Postirradiation thermocyclic loading of ferritic-martensitic structural materials

    Science.gov (United States)

    Belyaeva, L.; Orychtchenko, A.; Petersen, C.; Rybin, V.

    Thermonuclear fusion reactors of the Tokamak-type will be unique power engineering plants to operate in thermocyclic mode only. Ferritic-martensitic stainless steels are prime candidate structural materials for test blankets of the ITER fusion reactor. Beyond the radiation damage, thermomechanical cyclic loading is considered as the most detrimental lifetime limiting phenomenon for the above structure. With a Russian and a German facility for thermal fatigue testing of neutron irradiated materials a cooperation has been undertaken. Ampule devices to irradiate specimens for postirradiation thermal fatigue tests have been developed by the Russian partner. The irradiation of these ampule devices loaded with specimens of ferritic-martensitic steels, like the European MANET-II, the Russian 05K12N2M and the Japanese Low Activation Material F82H-mod, in a WWR-M-type reactor just started. A description of the irradiation facility, the qualification of the ampule device and the modification of the German thermal fatigue facility will be presented.

  14. In-pile Instrumentation Development

    International Nuclear Information System (INIS)

    Vermeeren, L.

    2005-01-01

    Advanced irradiations in research reactors require the on-line monitoring of crucial parameters like neutron fluxes, gamma dose rates, central fuel rod temperatures, fission gas release pressures and small geometry changes. Our activities in this field aim at a detailed understanding of the sensor behaviour in the irradiation conditions in order to extract reliable real-time information. The objectives of work performed by SCK-CEN are to study of the on-line in-pile measurement of gamma and neutron fluxes in real time and to investigate parasitic radiation-induced signals in instrumentation cables

  15. New trends in pile safety instrumentation

    International Nuclear Information System (INIS)

    Furet, J.

    1961-01-01

    This report addresses the protection of nuclear piles against damages due to operation incidents. The author discusses the current trends in the philosophy of safety of atomic power piles, identifies the parameters which define safety systems, presents tests to be performed on safety chains, comments the relationship between safety and the decrease of the number of pile inadvertent shutdowns, discusses the issues of instrument failures and chain multiplicity, comments the possible improvement of the operation of elements which build up safety chains (design simplification, development of semiconductors, replacement of electromechanical relays by static relays), the role of safety logical computers and the development of automatics in pile safety, presents automatic control as a safety factor (example of automatic start-up), and finally comments the use of fuses

  16. Qualification of power determination and in-pile measurements in the UO{sub 2} Gd{sub 2} 0{sub 2} fuel irradiation test IFA 636

    Energy Technology Data Exchange (ETDEWEB)

    Tverberg, T.; Volkov, B.; Kim, J-C.

    2004-04-15

    IFA-S36 is irradiated with the main objective of extending the database on the performance of UO{sub 2}Gd{sub 2}O{sub 2} fuel (with 8% absorbing gadolinia isotopes) compared with commercial UO{sub 2}. The rig carries 6 rods in the lower cluster (including three Gd-doped fuel rods) and 3 rods in the upper cluster (one rod with Gd-doped fuel). The rods are instrumented with expansion thermometers (ETs), fuel and cladding elongation detectors (EFs and ECs) and pressure transducers (PFs). Repeated calorimetric power measurements, physics calculations by the HELIOS code and gamma scans of selected rods in both clusters enabled the power and burnup determination to be qualified and corrected. The data suggest that as of May 2004 the power ratings in both fuels are much alike and burnups are about 30 and 34 MW/kgUO{sub 2} in the Gd-doped and ordinary UO{sub 2} rods respectively. Analysis of in-pile measurements compared with calculations shows that neutron absorption affects fuel temperature, power and burnup radial distributions in Gd-doped fuel at BOL compared with UO{sub 2} fuel. Sensitivity analyses performed with the HELIOS and FTEMP3 codes show that fuel centreline temperature in Gd-doped fuel is influenced by radial power depression, depletion of fissile materials and absorbing Gd isotopes as well as thermal conductivity of the fuel matrix and its degradation during irradiation. Analysis of the fuel dimension changes revealed densification only in the UO{sub 2} fuel whereas fuel elongation measurements in the Gd-doped fuel rods indicated essentially constant swelling with burnup. At burnups above 5 MWd/kgUO{sub 2} the swelling rate was about 0.5-O.fi % DELTAV/V per 10 MWd/kgUO{sub 2} for both fuel types. Internal pressure measured in the Gd-doped rod at BOL showed slight fuel densification and possibly He gas absorption, whereas derived swelling rate was somewhat Iarger than values obtained from the fuel elongation measurements. Cladding elongation measurements

  17. Phase transformation changes in thermocycled nickel-titanium orthodontic wires.

    Science.gov (United States)

    Berzins, David W; Roberts, Howard W

    2010-07-01

    In the oral environment, orthodontic wires will be subject to thermal fluctuations. The purpose of this study was to investigate the effect of thermocycling on nickel-titanium (NiTi) wire phase transformations. Straight segments from single 27 and 35 degrees C copper NiTi (Ormco), Sentalloy (GAC), and Nitinol Heat Activated (3M Unitek) archwires were sectioned into 5mm segments (n=20). A control group consisted of five randomly selected non-thermocycled segments. The remaining segments were thermocycled between 5 and 55 degrees C with five randomly selected segments analyzed with differential scanning calorimetry (DSC; -100150 degrees C at 10 degrees C/min) after 1000, 5000, and 10,000 cycles. Thermal peaks were evaluated with results analyzed via ANOVA (alpha=0.05). Nitinol HA and Sentalloy did not demonstrate qualitative or quantitative phase transformation behavior differences. Significant differences were observed in some of the copper NiTi transformation temperatures, as well as the heating enthalpy with the 27 degrees C copper NiTi wires (p<0.05). Qualitatively, with increased thermocycling the extent of R-phase in the heating peaks decreased in the 35 degrees C copper NiTi, and an austenite to martensite peak shoulder developed during cooling in the 27 degrees C copper NiTi. Repeated temperature fluctuations may contribute to qualitative and quantitative phase transformation changes in some NiTi wires. Copyright 2010 Academy of Dental Materials. All rights reserved.

  18. Microtensile dentin bond strength of fifth with five seventh-generation dentin bonding agents after thermocycling: An in vitro study

    Directory of Open Access Journals (Sweden)

    Bruhvi Poptani

    2012-01-01

    Full Text Available Objectives: The objective of this in vitro study was to compare the microtensile dentin bond strength (μTBS of five seventh-generation dentin bonding agents (DBA with fifth-generation DBA before and after thermocycling. Materials and Methods: Ten extracted teeth were assigned to fifth generation control group (optibond solo and each of the five experimental groups namely, Group I (G-Bond ,Group II (S 3 Clearfil, Group III (One Coat 7.0, Group IV (Xeno V, and Group V (Optibond all in one. The crown portions of the teeth were horizontally sectioned below the central groove to expose the dentin. The adhesive resins from all groups were bonded to the teeth with their respective composites. Specimens of sizes 1 × 1 × 6 mm 3 were obtained. Fifty specimens that bonded to dentin from each group were selected. Twenty-five of the specimens were tested for debonding without thermocycling and the remaining were subjected to thermocycling followed by μTBS testing. The data were analyzed with one-way ANOVA and Dunnett′s-test for comparison with the reference group(Vth Generation. Results: There was no significant difference (P > 0.05 between the fifth- and seventh-generation adhesives before and after thermocycling. The results of our study showed significantly higher value (P < 0.05 of μTBS of seventh-generation Group II (Clearfil S 3 compared to the fifth-generation before and after thermocycling. Conclusion: The study demonstrated that the Clearfil S 3 bond had the highest μTBS values. In addition, of the five tested seventh-generation adhesive resins were comparable to the fifth-generation DBA.

  19. The effect of thermocycling on the sealing ability of White Mineral Trioxide Aggregate: an in vitro study

    Directory of Open Access Journals (Sweden)

    Mohammad Ali Saghiri

    2016-01-01

    Full Text Available OBJECTIVE: The purpose of this study was to evaluate the effect of thermocycling on the sealing ability of White Mineral Trioxide Aggregate (WMTA after application for management of furcation perforation. MATERIALS AND METHOD: Thirty two human permanent mandibular molar teeth were chosen and after root amputation, the coronal parts were further trimmed and conventional access cavities were prepared. Furcation perforations were made with diamond bur and Peeso drills. Samples were divided into 3 experimental groups (n=10 and two control groups (n=1. The perforations were filled with WMTA in the experimental groups; in the control groups samples remained unfilled. Samples in the first group remained without further treatment, while in the second and third groups, teeth were thermocycled 500 and 800 times between 5 to 55 °C prior to leakage testing. Microleakage testing was done by using bovine serum albumin for 90 days. The number of days for color change was used as an indicator of protein leakage. Data were analyzed by using one-way analysis of variance and a post hoc Tukey test at a significance level of p<0.05. RESULTS: Non-thermocycled teeth showed significantly the longest time necessary for protein leakage to occur in comparison with the other two thermocycled groups (p<0.0001. The samples after 800 cycles showed the lowest resistance to protein leakage, while samples after 500 cycles indicated significantly more resistance against leakage (p<0.0001. CONCLUSION: Thermocycling can remarkably influence the microleakage property of WMTA. Thermal changes occurring inside the oral cavity might jeopardize the sealing property of the applied cement, which can lead to microleakage and possible failure of treatment in a clinical scenario.

  20. Durability of feldspathic veneering ceramic on glass-infiltrated alumina ceramics after long-term thermocycling.

    Science.gov (United States)

    Mesquita, A M M; Ozcan, M; Souza, R O A; Kojima, A N; Nishioka, R S; Kimpara, E T; Bottino, M A

    2010-01-01

    This study compared the bond strength durability of a feldspathic veneering ceramic to glass-infiltrated reinforced ceramics in dry and aged conditions. Disc shaped (thickness: 4 mm, diameter: 4 mm) of glass-infiltrated alumina (In-Ceram Alumina) and glass-infiltrated alumina reinforced by zirconia (In-Ceram Zirconia) core ceramic specimens (N=48, N=12 per groups) were constructed according to the manufacturers' recommendations. Veneering ceramic (VITA VM7) was fired onto the core ceramics using a mold. The core-veneering ceramic assemblies were randomly divided into two conditions and tested either immediately after specimen preparation (Dry) or following 30000 thermocycling (5-55 ºC±1; dwell time: 30 seconds). Shear bond strength test was performed in a universal testing machine (cross-head speed: 1 mm/min). Failure modes were analyzed using optical microscope (x20). The bond strength data (MPa) were analyzed using ANOVA (α=0.05). Thermocycling did not decrease the bond strength results for both In-Ceram Alumina (30.6±8.2 MPa; P=0.2053) and In-Ceram zirconia (32.6±9 MPa; P=0.3987) core ceramic-feldspathic veneering ceramic combinations when compared to non-aged conditions (28.1±6.4 MPa, 29.7±7.3 MPa, respectively). There were also no significant differences between adhesion of the veneering ceramic to either In-Ceram Alumina or In-Ceram Zirconia ceramics (P=0.3289). Failure types were predominantly a mixture of adhesive failure between the veneering and the core ceramic together with cohesive fracture of the veneering ceramic. Long-term thermocycling aging conditions did not impair the adhesion of the veneering ceramic to the glass-infiltrated alumina core ceramics tested.

  1. Bond strength of three luting agents to zirconia ceramic - influence of surface treatment and thermocycling

    Directory of Open Access Journals (Sweden)

    Ahmed Attia

    2011-08-01

    Full Text Available OBJECTIVE: This in vitro study aimed to evaluate the influence of different surface treatments, 3 luting agents and thermocycling on microtensile bond strength (µTBS to zirconia ceramic. Material and METHODS: A total of 18 blocks (5x5x4 mm were fabricated from zirconia ceramic (ICE Zirkonia and duplicated into composite blocks (Alphadent. Ceramic blocks were divided into 3 groups (n=6 according to the following surface treatments: airborne-particle abrasion (AA, silica-coating, (SC (CoJet and silica coating followed by silane application (SCSI (ESPE Sil. Each group was divided into 3 subgroups (n=2 according to the 3 luting agents used. Resin-modified glass-ionomer cement (RMGIC, Ketac Cem Plus, self-adhesive resin cement (UN, RelyX Unicem and adhesive resin cement (ML, MultiLink Automix were used for bonding composite and zirconia blocks. Each bonding assembly was cut into microbars (10 mm long and 1±0.1 mm². Seven specimens of each subgroup were stored in water bath at 37ºC for 1 week. The other 7 specimens were stored in water bath at 37ºC for 30 days then thermocycled (TC for 7,500 cycles. µTBS values were recorded for each specimen using a universal testing machine. Statistical analyses were performed using a 3-way ANOVA model followed by serial 1-way ANOVAs. Comparison of means was performed with Tukey's HSD test at (α=0.05. RESULTS: µTBS ranged from 16.8 to 31.8 MPa after 1 week and from 7.3 to 16.4 MPa after 30 days of storage in water and thermocycling. Artificial aging significantly decreased µTBS (p<0.05. Considering surface treatment, SCSI significantly increased µTBS (p<0.05 compared to SC and AA. Resin cements (UN and ML demonstrated significantly higher µTBS (p<0.05 compared to RMGIC cement. CONCLUSIONS: Silica coating followed by silane application together with adhesive resin cements significantly increased µTBS, while thermocycling significantly decreased µTBS.

  2. NEET In-Pile Ultrasonic Sensor Enablement-Final Report

    Energy Technology Data Exchange (ETDEWEB)

    J. Daw; J. Rempe; J. Palmer; P. Ramuhalli; R. Montgomery; H.T. Chien; B. Tittmann; B. Reinhardt; P. Keller

    2014-09-01

    Ultrasonic technologies offer the potential to measure a range of parameters during irradiation of fuels and materials, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes under harsh irradiation test conditions. There are two primary issues that currently limit in-pile deployment of ultrasonic sensors. The first is transducer survivability. The ability of ultrasonic transducer materials to maintain their useful properties during an irradiation must be demonstrated. The second issue is signal processing. Ultrasonic testing is typically performed in a lab or field environment, where the sensor and sample are accessible. The harsh nature of in-pile testing and the variety of desired measurements demand that an enhanced signal processing capability be developed to make in-pile ultrasonic sensors viable. To address these issues, the NEET ASI program funded a three year Ultrasonic Transducer Irradiation and Signal Processing Enhancements project, which is a collaborative effort between the Idaho National Laboratory, the Pacific Northwest National Laboratory, the Argonne National Laboratory, and the Pennsylvania State University. The objective of this report is to document the objectives and accomplishments from this three year project. As summarized within this document, significant work has been accomplished during this three year project.

  3. As-built measurement of the in-pool structure for the installation of In-Pile Test Section in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Park, K. N.; Cho, Y. G.; Lee, Y. S.; Sim, B. S.; Lee, J. M.; Chi, D. Y.; Park, S. K.; Lee, H. H.; Whang, D. K.; Lee, C. Y. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    Fuel Test Loop (FTL) is designed at the operation condition of power reactor such as high temperature, high pressure and neutron flux etc. As the design of the FTL has been completed, purchasing and manufacturing hardware are underway at present. Installation of the facility is going to do during reactor shutdown period in 2006. This paper describes the preparation of measurement and as-built measurement about in pool structure.

  4. Parametric study of the behaviour of a pre irradiated BWR fuel rod under conditions of LOCA simulated in the halden in pile test system with the FALCON code

    Energy Technology Data Exchange (ETDEWEB)

    Khvostov, G.; Zimmermann, M. A. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, Villigen (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, Leibstadt (Switzerland); Kolstad, E. [Institute for Energy Technology - OECD Halden Reactor Project, Halden (Norway); Montgomery, R. O. [Anatech Corporation, San Diego (United States)

    2008-10-15

    A new LOCA test at Halden was planned as the first experiment within the Halden LOCA program addressing the behaviour of commercially irradiated BWR fuel of medium burn up with burst of the cladding expected to occur at a temperature of about 1050.deg.C, which is essentially higher than in the preceding experiments. The specific measures to be adopted have been suggested based upon a parametric study using the FALCON fuel behaviour code and aimed at an optimized design of the test fuel rod for the given high target cladding temperature of 1150 .deg. C (peak local). The analysis has shown a reasonable agreement with the fundamental experimental findings, such as correlations of NUREG 0630, as well as consistency with the data from Halden LOCA testing available so far. Thus, a general conclusion is drawn about the applicability of the methodology developed at PSI to the analysis of LWR fuel rod behaviour during LOCA, in consideration of the effects of fuel burn up.

  5. Tape cast isotropic, fine-grained tungsten for thermo-cyclic loading applications

    Energy Technology Data Exchange (ETDEWEB)

    Sommerer, Mathias, E-mail: Mathias.Sommerer@tum.de [Lehrstuhl für Werkstoffkunde und Werkstoffmechanik, Technische Universität München, Boltzmannstr. 15, 85748 Garching (Germany); Li, Muyuan [Max-Planck-Institut für Plasma Physik, Boltzmannstraße 2, 85748 Garching (Germany); Werner, Ewald [Lehrstuhl für Werkstoffkunde und Werkstoffmechanik, Technische Universität München, Boltzmannstr. 15, 85748 Garching (Germany); Dewitz, Hubertus von; Walter, Steffen; Lampenscherf, Stefan [Siemens AG, Corporate Technology, Otto-Hahn-Ring 6, 81730 München (Germany); Arnold, Thomas [Siemens Healthcare GmbH, Henkestr. 127, 91052 Erlangen (Germany)

    2016-04-15

    Highlights: • The tape casting process for tungsten is described. • A set-up of a HHF test facility for standing anodes is presented. • The thermo-cyclic behavior of tape cast tungsten and a reference is investigated. • The evolution of crack patterns is described in dependency of HHF-loadings. • The surface roughness of X-ray anodes is related to the microstructural evolution. - Abstract: This paper introduces tape casting as a new route for the production of isotropic and fine-grained tungsten components. Microstructural and thermal properties of tape cast tungsten samples are determined. Thermal shock behavior according to the thermo-cyclic loading of standing X-ray anodes is investigated and compared to the behavior of a rolled tungsten grade. The development of surface roughness during the thermal shock loading is discussed in relation to the development of the grain structure and crack pattern. The fine-grained and stable microstructure of the tape cast material exhibits less roughening under such test conditions.

  6. Effect of storage in water and thermocycling on hardness and roughness of resin materials for temporary restorations

    Directory of Open Access Journals (Sweden)

    Jerusa Cleci de Oliveira

    2010-09-01

    Full Text Available PURPOSE: This study evaluated the effect of storage in water and thermocycling on hardness and roughness of resin materials for temporary restorations. MATERIAL AND METHODS: Three acrylic resins (Dencor-De, Duralay-Du, and Vipi Cor-VC were selected and one composite resin (Opallis-Op was used as a parameter for comparison. The materials were prepared according to the manufacturers' instructions and were placed in stainless steel moulds (20 mm in diameter and 5 mm thick. Thirty samples of each resin were made and divided into three groups (n = 10 according to the moment of Vickers hardness (VHN and roughness (Ra analyses: C (control group: immediately after specimen preparation; Sw: after storage in distilled water at 37 °C for 24 hours; Tc: after thermocycling (3000 cycles; 5-55 °C, 30 seconds dwell time. Data were submitted to 2-way ANOVA followed by Tukey's test (α = 0.05. RESULTS: Op resin had higher surface hardness values (p 0.05 in roughness among materials (De = 0.31 ± 0.07; Du = 0.51 ± 0.20; VC = 0.41 ± 0.15; Op = 0.42 ± 0.18. Storage in water did not change hardness and roughness of the tested materials (p > 0.05. There was a significant increase in roughness after thermocycling (p < 0.05, except for material Du, which showed no significant change in roughness in any evaluated period (p = 0.99. CONCLUSION: Thermocycling increased the roughness in most tested materials without affecting hardness, while storage in water had no significant effect in the evaluated properties.

  7. Effect of Endodontic Irrigants on Microtensile Bond Strength to Dentin After Thermocycling and Long-Term Water Storage

    Directory of Open Access Journals (Sweden)

    Daniel Galafassi

    2013-01-01

    Full Text Available Objective: The bond strength of adhesives in irrigated dentin behaves differently over time. The aim of this study was to evaluate the influence of long-term water storage and thermocycling on the microtensile bond strength of adhesive systems to dentin irrigated with endodontic solutions.Materials and Methods: Sixty human molars were used after removal of the occlusal portion and exposure of the dentin by grinding. The specimens were irrigated with 2.5% NaOCl for 30 minutes and then 17% EDTA for 5 minutes and assigned to six groups according to the adhesive system (n=10: G1 and G2–Clearfil SE Bond; G3 and G4–Single Bond 2; and G5 and G6–XP Bond. The teeth were restored with composite and were subjected to water storage for different time periods. G1, G3 and G5 were stored for 24 h; G2, G4 and G6 were stored for 6 months and were subjected to thermocycling (12,000 cycles, 5°C to 55°C, 500 cycles per week for 6 months. After storage, the tooth/restoration assembly was sectioned to obtain four sticks of approximately 1 mm2, for microtensile bond strength testing. The results were analyzed by two-way ANOVA and Tukey’s test.Results: Significant differences were observed among the adhesives (p<0.01. No significant differences were observed in the microtensile bond strength between samples after 24 hours of storage without thermocycling and after 6-month storage with 12,000 cycles (p<0.05.Conclusion: The bond strengths of G5 and G6 after irrigation with 2.5% NaOCl and 17% EDTA were significantly different from those of other groups. Long-term water storage/thermocycling had no effect on bond strength to dentin.

  8. SAS validation and analysis of in-pile TUCOP experiments

    International Nuclear Information System (INIS)

    Morman, J.A.; Tentner, A.M.; Dever, D.J.

    1985-01-01

    The validation of the SAS4A accident analysis code centers on its capability to calculate the wide range of tests performed in the TREAT (Transient Reactor Test Facility) in-pile experiments program. This paper presents the SAS4A analysis of a simulated TUCOP (Transient-Under-Cooled-Over-Power) experiment using seven full-length PFR mixed oxide fuel pins in a flowing sodium loop. Calculations agree well with measured thermal-hydraulic, pin failure time and post-failure fuel motion data. The extent of the agreement confirms the validity of the models used in the SAS4A code to describe TUCOP accidents

  9. In-pile experiments on fuel rod behaviour during a LOCA

    International Nuclear Information System (INIS)

    Sepold, E.H.; Karb, E.H.; Pruessmann, M.

    1981-07-01

    This report describes the results of the Test Series G2/3 within the in-pile experimental program for the investigation of LWR fuel rod behavior. The results were obtained with single rods of a PWR design in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechanisms of fuel rod failure were being performed with irradiated and unirradiated rods. The main parameter of the test program ist the burnup, ranging from 2500 to 35000 MWd/t. The results of test series G2/3 (35000 MWd/t) with respect to the burst data, i.e. burst temperature, burst pressure, and burst strain, do not indicate major differences from the in-pile tests with unirradiated test specimens. (orig.) [de

  10. In-pile experiemts on fuel rod behavior during a LOCA

    International Nuclear Information System (INIS)

    Pruessmann, M.; Karb, E.H.; Sepold, L.

    1981-02-01

    This report describes the results of the Test Series G1 within the in-pile experimental program for the investigation of LWR fuel rod behavior. The results were obtained with single rods of a PWR design in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechansims of fuel rod failure were being performed with irradiated and unirradiated rods. The main parameter of the test program is the burnup ranging from 2500 to 35 000 MWd/t. The results of test series G1 (35 000 MWd/t) with respect to the burst data, i.e. burst temperature, burst pressure, and burst strain, do not indicate major differences from the in-pile tests with unirradiated test specimens. (orig.) [de

  11. Color change of CAD-CAM materials and composite resin cements after thermocycling.

    Science.gov (United States)

    Gürdal, Isil; Atay, Ayse; Eichberger, Marlis; Cal, Ebru; Üsümez, Aslihan; Stawarczyk, Bogna

    2018-04-24

    The color of resin cements and computer-aided-design and computer-aided-manufacturing (CAD-CAM) restorations may change with aging. The purpose of this in vitro study was to analyze the influence of thermocycling on the color of CAD-CAM materials with underlying resin cement. Seven different CAD-CAM materials, composite resins and glass-ceramics were cut into 0.7-mm and 1.2-mm thicknesses (n=10) and cemented with a dual-polymerizing resin cement, a light-polymerizing resin cement, and a preheated composite resin (N=420). Color values were measured by using spectrophotometry. Specimens were subjected to thermocycling (5°C and 55°C; 5000 cycles). The measured color difference (ΔE) data were analyzed by using descriptive statistics. Normality of data distribution was tested by using the Kolmogorov-Smirnov test. Three-way and 1-way ANOVA followed by the Scheffé post-hoc test and unpaired 2-sample Student t test were computed to determine the significant differences among the tested parameters (α=.05). ΔE values were significantly influenced by the CAD-CAM material (η p 2 =0.85, Pcement (η P 2 =0.03, P=.003) but were not influenced by thickness (P=.179). Significant interactions were present among thickness, cement, and CAD-CAM materials (Pcement showed significantly lower ΔE values than the preheated composite resin (P=.003). Restoration materials and composite resin cement types used for cementation influence the amount of color change due to aging. Copyright © 2018 Editorial Council for the Journal of Prosthetic Dentistry. Published by Elsevier Inc. All rights reserved.

  12. In pile helium loop ''COMEDIE''

    International Nuclear Information System (INIS)

    Abassin, J.J.; Blanchard, R.J.; Gentil, J.

    1981-01-01

    The SR1 test in the COMEDIE loop has permitted to demonstrate particularly the device operation reliability with a fuel loading. The post-irradiation examinations have pointed out the good filter efficiency and have enabled to determine the deposition profiles either for the activation products (e.g.: 51 Cr, 60 Co) or for the fission products (e.g.: sup(110m)Ag, 131 I, 134 Cs, 137 Cs). (author)

  13. In-pile experimental device for Sirene thermionic converters

    International Nuclear Information System (INIS)

    Bliaux, J.; Durand, J.; Lazare-Chopard, G.

    1969-01-01

    The irradiation device described here, was built for in pile life tests of 100 We SIRENE converters. The nuclear converter is located in a sealed vacuum chamber, which is plugged at the lower end of a coaxial tubing acting as electrical leads. The output power is available on a variable resistive load on the bank of the reactor pool. Thermal, electrical and neutronic parameters of the converter are recorded. Since 1967, two permanent devices allowed five experiments in the swimming pool TRITON (CEN-FAR) and the results, obtained till now, are presented. (authors) [fr

  14. Effect of storage time, thermocycling and resin coating on durability of dentin bonding systems

    Directory of Open Access Journals (Sweden)

    Seyed-Mostafa Mousavinasab

    2009-01-01

    Full Text Available Background:Along with development of different dental adhesives, concerns about hydrolytic deg-radation of the adhesive components have arisen. The purpose of this study was to evaluate the in-vitro influence of thermocycling, water storage and resin coating on the microshear bond strength of total etch and self etch adhesive systems to dentin. Methods:The superficial coronal dentin of eighty intact third molars were exposed and divided into 5 equal groups. Dental adhesives including Scotch Bond Multi Purpose (SBMP, Single Bond (SB, Clearfil SE Bond (CSE, Prompt L-Pop (PLP, and Prompt L-Pop plus Margin bond (PLPM were applied according to the manufacturers′ instructions on prepared surfaces in the study groups, respectively. Then composite cylinders were bonded and specimens were divided into two subgroups. One subgroup was stored in water for 24 hours. The second subgroup was subjected to 3000 thermocycle shocks and then was stored in 37°C water for 3 months. Finally, all teeth were subjected to the mi-croshear bond strength test. Data were analyzed using two-way ANOVA and Tukey HSD tests. One specimen similar to each subgroup was also prepared for SEM evaluation. Results:After one-day storage, the SBMP showed the highest bond strength followed by CSE, PLPM, SB and PLP. After three months storage, the highest bond strength was observed in SBMP followed by PLPM, CSE, SB, and PLP. Conclusion: SBMP showed the best bond strength while CSE represented acceptable bond durabil-ity. Resin coating on PLP improved bond strength and durability.

  15. In pile helium loop ''Comedie''

    International Nuclear Information System (INIS)

    Blanchard, R.J.

    1985-01-01

    The loop is located in the SILOE reactor at Centre d'Etudes Nucleaires de Grenoble. The purpose and objectives are divided into two groups, principal and secondary. The primary objective was to provide basic data on the deposition behavior of important condensable fission products on a variety of steel surfaces, i.e. temperature (sorption isotherms) and mass transfer (physical adsorption) dependencies; to provide information concerning the degree of penetration of important fission products into the metals comprising the heat exchanger-recuperator tubes as a function of alloy type and/or metal temperature; to provide complementary information on the reentrainment (liftoff) of important fission and activation products by performing out-of-pile blowdown experiments on tube samples representative of the alloy types used in the heat exchanger-recuperator and of the surface temperatures experienced during plateout. The secondary objective was to provide information concerning the migration of important fission products through graphite. To this end, concentration profiles in the web between the fuel rods containing the fission product source and the coolant channels and in the graphite diffusion sample will be measured to study the corrosion of metallic specimens placed in the conditions of high temperature gas cooled reactor. The first experiment SRO enables to determine the loop characteristics and possibilities related to thermal, thermodynamic, chemical and neutronic properties. The second experiment has been carried out in high temperature gas cooled reactor operating conditions. It enables to determine in particular the deposition axial profile of activation and fission products in the plateout section constituting the heat exchanger, the fission products balance trapped in the different filter components, and the cumulated released fraction of solid fission products. The SR1 test permits to demonstrate in particular the Comedie loop operation reliability, either

  16. Thermocyclic treatment of Be for higher stability of mechanical properties

    International Nuclear Information System (INIS)

    Neklyudov, I.M.; Papirov, I.I.; Stoev, P.I.

    2004-01-01

    The paper reports the results from studies of the effects of upper temperature and speed of thermocyclic treatment (TCT), a combined action of thermal treatment and TCT on the acoustic emission of two batches of hot-pressed beryllium having different mechanical properties. It is demonstrated that the upper temperature of treatment exerts a substantial effect on the mechanical and acoustic characteristics of hot-pressed beryllium. At an upper TCT temperature of 500degC, the materials under study exhibit the minimum activity of acoustic emission and a small total number of pulses detected. Acoustic spectra of beryllium samples were measured after the samples were subjected to the TCT with different velocity values of the process. It has been established that the treatment preceding the TCT (ageing at 650degC for 5 hours) had little effect on the mechanical and acoustic parameters of beryllium, while the treatment following the TCT (600degC, 1 hour) led to dislocation pinning and thus reduced the dislocation mobility. It has been demonstrated that the acoustic parameters can be used for choosing the optimum temperature of the TCT process, for estimating the degree of dislocation mobility and for controlling the quality of thermal treatment performed.(author)

  17. Summary on out-of-pile and in-pile properties of M5 alloy

    International Nuclear Information System (INIS)

    Zhao Wenjin

    2001-01-01

    The out-of-pile and in-pile corrosion, mechanical properties, microstructure,hydrogen absorption, creep and growth resistances of M5 alloy using as PWR fuel rod cladding materials developed by FRAMATOME in France has been summarized with reference to the literatures. The results obtained from in-pile irradiation tests show that the corrosion and hydrogen absorption resistances, creep and irradiation growth resistances of M5 alloy cladding are superior to that of the optimized Zircaloy-4. It could be estimated that the M5 alloy enables rod burnups close to 65GWd/tU to be reached

  18. Progress In Developing An In-Pile Acoustically Telemetered Sensor Infrastructure

    Energy Technology Data Exchange (ETDEWEB)

    Smith, James A.; Garrett, Steven L.; Heibel, Michael D.; Agarwal, Vivek; Heidrich, Brenden J.

    2016-09-01

    A salient grand challenge for a number of Department of Energy programs such as Fuels Cycle Research and Development ( includes Accident Tolerant Fuel research and the Transient Reactor Test Facility Restart experiments), Light Water Sustainability, and Advanced Reactor Technologies is to enhance our fundamental understanding of fuel and materials behavior under irradiation. Robust and accurate in-pile measurements will be instrumental to develop and validate a computationally predictive multi-scale understanding of nuclear fuel and materials. This sensing technology will enable the linking of fundamental micro-structural evolution mechanisms to the macroscopic degradation of fuels and materials. The in situ sensors and measurement systems will monitor local environmental parameters as well as characterize microstructure evolution during irradiation. One of the major road blocks in developing practical robust, and cost effective in-pile sensor systems, are instrument leads. If a wireless telemetry infrastructure can be developed for in-pile use, in-core measurements would become more attractive and effective. Thus to be successful in accomplishing effective in-pile sensing and microstructure characterization an interdisciplinary measurement infrastructure needs to be developed in parallel with key sensing technology. For the discussion in this research, infrastructure is defined as systems, technology, techniques, and algorithms that may be necessary in the delivery of beneficial and robust data from in-pile devices. The architecture of a system’s infrastructure determines how well it operates and how flexible it is to meet future requirements. The limiting path for the effective deployment of the salient sensing technology will not be the sensors themselves but the infrastructure that is necessary to communicate data from in-pile to the outside world in a non-intrusive and reliable manner. This article gives a high level overview of a promising telemetry

  19. Effect of thermocycling on nickel release from orthodontic arch wires: an in vitro study.

    Science.gov (United States)

    Sheibaninia, Ahmad

    2014-12-01

    The amount of daily intake of metals from orthodontic appliances over time is a matter of great concern. Nickel results in one of the most common metal-induced allergic contact dermatitis in humans; it produces more allergic reactions than all the other metals combined together. The purpose of this study was to evaluate the effects of thermocycling on the nickel release from orthodontic arch wires stored in artificial saliva with different pH values. Forty new wire pieces were selected. Each wire piece was placed in a special capillary Pyrex tube filled with artificial saliva, which was sealed and immersed in deionized water at 37 °C. The samples were divided into four groups of ten. Group I received no treatment; group II was subjected to thermocycling. The pH of storage in groups III and IV was reduced to 4.5, and group IV was subjected to thermocycling. Thermocycling was carried out between 5 and 55 °C for 500 cycles. The release of nickel ions was statistically analyzed by two-way ANOVA for the effects of two variables: pH and thermocycling. The interaction between pH and thermocycling was found to be statistically significant (F = 12.127, P = 0.001). Two-way ANOVA showed that different storage media or pH and thermocycling had a significant effect on the nickel release (F = 52.812, P nickel from orthodontic wires, while thermocycling is clearly the dominant factor.

  20. Remote-welding technique for assembling in-pile IASCC capsule in hot cell

    International Nuclear Information System (INIS)

    Kawamata, Kazuo; Ishii, Toshimitsu; Kanazawa, Yoshiharu; Iwamatsu, Shigemi; Ohmi, Masao; Shimizu, Michio; Matsui, Yoshinori; Saito, Jun-ichi; Ugachi, Hirokazu; Kaji, Yoshiyuki; Tsukada, Takashi

    2006-01-01

    In order to investigate behavior of the irradiation assisted stress corrosion cracking (IASCC) caused by the simultaneous effects of neutron irradiation and high temperature water environment in such a light water reactor (LWR), it is necessary to perform crack growth tests in an in-pile IASCC capsule irradiated in the Japan Materials Testing Reactor (JMTR). The development of the remote-welding technique is essential for remotely assembling the in-pile IASCC capsule installing the pre-irradiated CT specimens. This report describes a new remote-welding machine developed for assembling the in-pile IASCC capsule. The remote-welding technique that the capsule tube is rotated light under the fixed torch was applied to the machine for the welding of thick and large-diameter tubes. The assembly work of four in-pile IASCC capsules having pre-irradiated CT specimens in the hot cell was succeeded for performing the crack growth test under the neutron irradiation in JMTR. The irradiation test of two capsules has been already finished in JMTR without problems. (author)

  1. NEET In-Pile Ultrasonic Sensor Enablement-FY 2012 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    JE Daw; JL Rempe; BR Tittmann; B Reinhardt; P Ramuhalli; R Montgomery; HT Chien

    2012-09-01

    Several Department Of Energy-Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development, Advanced Reactor Concepts, Light Water Reactor Sustainability, and Next Generation Nuclear Plant programs, are investigating new fuels and materials for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials when irradiated. The Nuclear Energy Enabling Technology (NEET) Advanced Sensors and Instrumentation (ASI) in-pile instrumentation development activities are focused upon addressing cross-cutting needs for DOE-NE irradiation testing by providing higher fidelity, real-time data, with increased accuracy and resolution from smaller, compact sensors that are less intrusive. Ultrasonic technologies offer the potential to measure a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes, under harsh irradiation test conditions. There are two primary issues associated with in-pile deployment of ultrasonic sensors. The first is transducer survivability. The ability of ultrasonic transducer materials to maintain their useful properties during an irradiation must be demonstrated. The second issue is signal processing. Ultrasonic testing is typically performed in a lab or field environment, where the sensor and sample are accessible. Due to the harsh nature of in-pile testing, and the range of measurements that are desired, an enhanced signal processing capability is needed to make in-pile ultrasonic sensors viable. This project addresses these technology deployment issues.

  2. In-pile loop experiments in water chemistry and corrosion

    International Nuclear Information System (INIS)

    Kysela, J.; Jindrich, K.; Masarik, V.; Fric, Z.; Chotivka, V.; Hamerska, H.; Vsolak, R.; Erben, O.

    1986-08-01

    Methods and techniques used were as follows: (a) Method of polarizing resistance for remote monitoring of instantaneous rate of uniform corrosion. (b) Out-of-pile loop at the temperature 350 degC, pressure 19 MPa, circulation 20 kgs/h, testing time 1000 h. (c) High temperature electromagnetic filter with classical solenoid and ball matrix for high pressure filtration tests. (d) High pressure and high temperature in-pile water loop with coolant flow rate 10 000 kgs/h, neutron flux in active channel 7x10 13 n/cm 2 .s, 16 MPa, 330 degC. (e) Evaluation of experimental results by chemical and radiochemical analysis of coolant, corrosion products and corrosion layer on surface. The results of measurements carried out in loop facilities can be summarized into the following conclusions: (a) In-pile and out-of-pile loops are suitable means of investigating corrosion processes and mass transport in the nuclear power plant primary circuit. (b) In studying transport phenomena in the loop, it is necessary to consider the differences in geometry of the loop and the primary circuit, mainly the ratio of irradiated and non-irradiated surfaces and volumes. (c) In the experimental facility simulating the WWER-type nuclear power plant primary circuit, solid suspended particles of a chemical composition corresponding most frequently to magnetite or nickel ferrite, though with non-stoichiometric composition Me x 2+ Fe 3-x 3+ O 4 , were found. (d) Continuous filtration of water by means of an electromagnetic filter removing large particles of corrosion products leads to a decrease in radioactivity of the outer epitactic layer only. The effect of filtration on the inner topotactic layer is negligible

  3. Effect of Thermocycling on Microleakage of New Adhesive Systems on Primary Teeth: An In-Vitro Study

    Directory of Open Access Journals (Sweden)

    Ramin Atash

    2013-09-01

    Full Text Available Introduction: This study investigated the sealing ability of three different adhesives in primary bovine teeth. Methods: Facial and lingual class V cavities were prepared half in enamel and half in cementum, in 48 bovine primary mandibular incisors and randomly divided into three groups and each group divided to two subgroups. The tested adhesives were XPBond (XP, ClearfilS3 Bond (S3, and Xeno III (XE. All cavities were restored with composite and light cured. After 24 hours storage in 37°C distilled water and polishing, teeth were thermocycled and sealed with nail varnish. Then, they were stored in 2% methylene blue and dye penetration was evaluated under a stereomicroscope. Results: No significant differences were recorded in the microleakage value between three adhesives in enamel and dentin margins (p>0.0.5 before and after thermocycling. The lowest microleakage value was obtained in XE followed by XP and S3. Conclusion: There were not any differences between adhesives in enamel and dentin margins of class V cavities on primary bovine teeth.

  4. Effect of Thermocycling on Microleakage of New Adhesive Systems on Primary Teeth: An In-Vitro Study

    Directory of Open Access Journals (Sweden)

    Niloofar Shadman

    2013-10-01

    Full Text Available Introduction: This study investigated the sealing ability of three different adhesives in primary bovine teeth. Methods: Facial and lingual class V cavities were prepared half in enamel and half in cementum, in 48 bovine primary mandibular incisors and randomly divided into three groups and each group divided to two subgroups. The tested adhesives were XPBond (XP, ClearfilS3 Bond (S3, and Xeno III (XE. All cavities were restored with composite and light cured. After 24 hours storage in 37°C distilled water and polishing, teeth were thermocycled and sealed with nail varnish. Then, they were stored in 2% methylene blue and dye penetration was evaluated under a stereomicroscope. Results: No significant differences were recorded in the microleakage value between three adhesives in enamel and dentin margins (p>0.0.5 before and after thermocycling. The lowest microleakage value was obtained in XE followed by XP and S3. Conclusion: There were not any differences between adhesives in enamel and dentin margins of class V cavities on primary bovine teeth.

  5. Results from In-pile experiments on LWR fuel rod behavior under LOCA conditions with unirradiated rods

    International Nuclear Information System (INIS)

    Sepold, L.; Karb, E.H.; Pruessmann, M.

    1981-06-01

    This report summarizes the results of the FR2-in-pile tests at KfK (Kernforschungszentrum Karlsruhe) with unirradiated test rods. The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechanisms of fuel rod failure were being performed with irradiated and unirradiated single rods of a PWR design in the DK loop of the FR2 reactor. The main parameter of the test program was the burnup, ranging from 2.500 to 35.000 MWd/t. The program with unirradiated specimens comprised the series A and B with a total of 14 tests. (orig.) [de

  6. Selecting the thermo-cyclic treatment’s optimum parameters based analysis of fractal surfaces indicators

    Directory of Open Access Journals (Sweden)

    Вікторія Юріївна Іващенко

    2015-03-01

    Full Text Available Optimization of complex modes of heat treatments, in which control the properties of processed steel occurs by varying the large number of parameters, is quite time-consuming process. The influence of thermal processes on the formation of the metal structure manifested at the level of micro- and meso-sizes, which are realized qualitatively different mechanisms of destruction. Method of multi-factual description of the fracture’s surfaces, which was got after tests of mechanical properties, was used for the choice of the optimum thermo-cyclic mode with the variable temperatures Tmax and Tmin in cycles in this work. It vas founded the number of TCT-mode’s cycles and order changing Tmax affect the processes of dislocation motion and the formation of micro-voids in the metal. This work shows the relationship between these processes and fractal indices. Fractal indices of micro levels correlate to the dislocation density of the structure, and the meso-level indices - to the percentage reduction of area at fracture. It was proved that the analysis of the topography of the fracture’s surfaces using fractal indices to determine the optimal combination of processing parameters required to obtain the best mechanical properties. The new TCT-modes with variable temperature settings can be seen as reinforcing thermal technology that promotes self-organization phase-structural state of steels because it is able to generate an effective barrier to the movement of dislocations and cracks promotion

  7. An In Vitro Model for Studying Neutrophil Activation During Cardiopulmonary Bypass by Using a Polymerase Chain Reaction Thermocycler

    NARCIS (Netherlands)

    Tang, Min; Zhao, Xiao-Gang; Gu, Y. John; Chen, Chang-Zhi

    The accurate temperature control of a polymerase chain reaction (PCR) thermocycler was exploited in developing an in vitro model to study neutrophil activation during cardiopulmonary bypass. Neutrophils from 12 volunteers underwent temperature changes in a PCR thermocycler (37 degrees C for 30

  8. Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

    International Nuclear Information System (INIS)

    Obara, Toru; Sekimoto, Hiroshi

    1997-01-01

    The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors. (author)

  9. In pile AISI 316L. Low cycle fatigue. Final report

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.; Moons, F.

    1994-12-01

    In pile testing of the effect of neutron irradiation on the fatigue life of the reference material AISI 316L was performed in the framework of the European fusion technology program. The overall programme, carried out at SCK CEN (Mol,Belgium), exists of two instrumented rigs for low cycle fatigue testing, which were consecutively loaded in the BR-2 reactor during periods Jan (94) June (94) and Aug (94)-Dec(94). In each experiment, two identical samples were loaded by means of a pneumatically driven system. The samples were instrumented with thermocouples, strain gages, linear variable displacement transducers, and activation monitors. The experimental conditions are given. Type of fatigue test: load controlled, symmetric, uniaxial, triangular wave shape; stress range: about 580 MPa; sample shape: hourglass, diameter 3.2 mm, radius 12.5 mm; environment: NaK (peritectic); temperature: 250 C; maximum dpa value up to fracture: 1.7. Two of four samples were broken (one in each experiment) after having experienced 17 419 respectively 11 870 stress cycles. These new data points confirm earlier results from pile fatigue tests: irradiation causes no degradation of fatigue life of AISI 316L steel, at least for the parameters corresponding to these experiments

  10. Microleakage after Thermocycling of Three Self-Etch Adhesives under Resin-Modified Glass-Ionomer Cement Restorations

    Directory of Open Access Journals (Sweden)

    Sabine O. Geerts

    2010-01-01

    Full Text Available This study was designed to evaluate microleakage that appeared on Resin-Modified Glass-Ionomer Cement (RMGIC restorations. Sixty class V cavities (h×w×l=2mm×2mm×3mm were cut on thirty extracted third molars, which were randomly allocated to three experimental groups. All the buccal cavities were pretreated with polyacrylic acid, whereas the lingual cavities were treated with three one-step Self-Etch adhesives, respectively, Xeno III (Dentsply Detrey GmbH, Konstanz, Germany, iBond exp (Heraeus Kulzer gmbH & Co. KG, Hanau, Germany, and Adper Prompt-L-Pop (3M ESPE AG, Dental products Seefeld, Germany. All cavities were completely filled with RMGIC, teeth were thermocycled for 800 cycles, and leakage was evaluated. Results were expressed as means ± standard deviations (SDs. Microleakage scores were analysed by means of generalized linear mixed models (GLMMs assuming an ordinal logistic link function. All results were considered to be significant at the 5% critical level (<.05. The results showed that bonding RMGIC to dentin with a Self-Etch adhesive rather than using polyacrylic acid did not influence microleakage scores (=.091, except for one tested Self-Etch adhesive, namely, Xeno III (<.0001. Nevertheless, our results did not show any significant difference between the three tested Self-Etch adhesive systems. In conclusion, the pretreatment of dentin with Self-Etch adhesive system, before RMGIC filling, seems to be an alternative to the conventional Dentin Conditioner for the clinicians as suggested by our results (thermocycling and others (microtensile tests.

  11. Thermal and in-pile densification of MOX fuels: Some recent results

    International Nuclear Information System (INIS)

    Caillot, L.; Malgouyres, P.P.; Souchon, F.; Gotta, M.J.; Warin, D.; Chotard, A.; Couty, J.C.

    1997-01-01

    In-pile densification of PWR fuels is one of the main phenomena which determine the evolution of the pellet-clad gap during the first stage of the irradiation, and thus has consequences onto the thermo-mechanical behaviours of fuel rods. It can be predicted using the results of resintering tests and appropriate correlations. In this context, CEA, FRAMATOME and EDF have undertaken a joint research programme aiming to characterize the densification of MOX fuels. Different fuels were prepared by the MIMAS process using different UO 2 powders as matrix. After a detailed characterization, fuel pellets were submitted to isothermal resintering tests and analytical irradiations. Correlations between in-pile and thermal densification were established. This paper presents the results obtained with two types of MOX fuel: one fabricated wit the AUC UO 2 powder (ammonium uranyl carbonate conversion process) and another one fabricated with the SFEROX powder (peroxide conversion process). 8 refs, 8 figs

  12. Design and performance evaluation of a cryogenic condenser for an in-pile experiment

    Science.gov (United States)

    Graham, R. W.; Crum, R. J.; Hsu, Y.

    1972-01-01

    An apparatus was designed to enable in-pile irradiation of materials in liquid hydrogen at cryogenic temperatures. One of the principal components of this apparatus was a horizontal tube condenser. The performance of the condenser was evaluated by running a liquid-nitrogen prototype of the apparatus at heat loads comparable to or greater than those expected during the irradiation. The test showed that the condenser was capable of handling the design heat load and that the design procedure was sound.

  13. Preliminary results from initial in-pile debris bed experiments

    International Nuclear Information System (INIS)

    Rivard, J.B.

    1977-01-01

    An accident in a liquid metal fast breeder reactor (LMFBR) in which molten core material is suddenly quenched with subcooled liquid sodium could result in extensive fragmentation and dispersal of fuel as subcritical beds of frozen particulate debris within the reactor vessel. Since this debris will continue to generate power due to decay of retained fission products, containment of the debris is threatened if the generated heat is not removed. Therefore, the initial safety question is the capacity which debris beds may have for transfer of the decay heat to overlying liquid sodium by natural processes--i.e., without the aid of forced circulation of the coolant. Up to the present time, all experiments on debris bed behavior either have used substitute materials (e.g., sand and water) or have employed actual materials, but atypical heating methods. Increased confidence in the applicability of debris bed simulations is afforded if the heat is generated within the fuel component of the appropriate fast reactor materials. The initial series of in-pile tests reported on herein constitutes the first experiments in which the internal heating mode has been produced in particulate oxide fuel immersed in liquid sodium. Fission heating of the fully-enriched UO 2 in the experiment while it is contained within Sandia Laboratories Annular Core Pulse Reactor (ACPR), operating in its steady-state mode, approximates the decay heating of debris. Preliminary results are discussed

  14. Neutronic and thermal estimation of blanket in-pile mockup with Li2TiO3 pebbles

    International Nuclear Information System (INIS)

    Nagao, Y.; Nakamichi, M.; Tsuchiya, K.; Kawamura, H.

    2001-01-01

    To evaluate exactly temperature distribution in large volume of tritium breeding materials during the blanket in-pile tests with the JMTR, neutronic and thermal calculations were conducted by using Monte Carlo code 'MCNP' with nuclear cross section library of 'FSXLIBJ3R2' and the transient and steady-state distribution code 'TRUMP'. From the results of preliminary estimation of temperature distribution in the blanket in-pile mockup, the calculated values were 24-28% higher than the measured values. One of the reasons is due to overestimation of calculated thermal neutron flux

  15. Interpretation of the TRADE In-Pile source multiplication experiments

    International Nuclear Information System (INIS)

    Mercatali, Luigi; Carta, Mario; Peluso, Vincenzo

    2006-01-01

    Within the framework of the neutronic characterization of the TRIGA RC-1 reactor in support to the TRADE (TRiga Accelerator Driven Experiment) program, the interpretation of the subcriticality level measurements performed in static regime during the TRADE In-Pile experimental program is presented. Different levels of subcriticality have been measured using the MSA (Modified Source Approximated) method by the insertion of a standard fixed radioactive source into different core positions. Starting from a reference configuration, fuel elements were removed: control rods were moved outward as required for the coupling experiments envisioned with the proton accelerator and fission chambers were inserted in order to measure subcritical count rates. A neutron-physics analysis based on the modified formulation of the source multiplication method (MSM) has been carried out, which requires the systematic solution for each experimental configuration of the homogeneous, both in the forward and adjoint forms, and inhomogeneous Boltzmann equations. By means of such a methodology calculated correction factors to be applied to the MSA measured reactivities were produced in order to take into account spatial and energetic effects creating changes in the detector efficiencies and effective source with respect to the calibration configuration. The methodology presented has been tested against a large number of experimental states. The measurements have underlined the sensitivity of the MSA measured reactivities to core geometry changes and control rod perturbations; the efficiency of MSM factors to dramatically correct for this sensitivity is underlined, making of this technique a relevant methodology in view of the incoming US RACE program to be performed in TRIGA reactors

  16. Irradiation technology (1). Development of new in-pile instrumentation at JMTR

    International Nuclear Information System (INIS)

    Shibata, Akira; Kimura, Nobuaki; Tanimoto, Masataka; Nakamura, Jinichi; Saito, Takashi; Tsuchiya, Kunihiko

    2012-01-01

    Development of instrumentation which can use under severe accident condition is important issue for the purpose to cope with severe accident at nuclear reactors. And also to improve the quality of irradiation tests data and to increase the reliability of safety management system of reactors, the development of new instrumentation is key issue. JAEA is developing several in-pile instrumentations to conduct irradiation tests at JMTR. This study includes the developments of three new instrumentations and describes the characteristics of the instrumentations. These are ECP sensor, new water level indicator and in-reactor observation system using Cherenkov light. (author)

  17. Stress analysis of primary pipe rigid support of the in pile loop

    International Nuclear Information System (INIS)

    Hasibuan, Dj.

    1998-01-01

    Base on requirement of the safety analysis report and operation planning preparation on the in pile loop by using the fuel bundle in the test section, the stress analysis of primary pipe support has been done. The analysis was performed for the 3 (three) points of pipe support, which are chosen by random selection, i.e.: GU 2001, GU 2002, and GU 2331. The analysis result showed that the maximum allowable stress was greater then the actual stress. It is concluded that the existing supports fulfil the safety requirement

  18. Effect of brushing and thermocycling on the shade and surface roughness of CAD-CAM ceramic restorations.

    Science.gov (United States)

    Yuan, Judy Chia-Chun; Barão, Valentim Adelino Ricardo; Wee, Alvin G; Alfaro, Maria F; Afshari, Fatemeh S; Sukotjo, Cortino

    2017-09-29

    The effects of toothbrushing (B) and thermocycling (TC) on the surface texture of different materials with various fabrication processes have been investigated. However, studies of computer-aided design and computer-aided manufacturing (CAD-CAM) ceramic restorations are limited. The purpose of this in vitro study was to evaluate the effect of B and TC on the color stability and surface roughness of extrinsically characterized and glazed CAD-CAM ceramic restorations. Lithium disilicate CAD ceramic (n=90) and zirconia ceramic (n=90) were studied. All specimens were crystallized/sintered, characterized, and glazed following the manufacturer's recommendation. The specimens were divided into 9 different groups: B, TC, and a combination of B plus TC (B+TC). Brushing was performed at 50 000, 100 000, and 150 000 cycles, simulating an oral environment of 5, 10, and 15 years. Thermocycling was performed at 6000, 12 000, and 18 000 cycles, simulating an oral environment of 5, 10, and 15 years. Brushing plus TC was performed with the combination of the 50 000 cycles of B, then 6000 cycles of TC, and 10 000 cycles of B, then 12 000 cycles of TC, and 15 000 cycles of B, then 18 000 cycles of TC. The color and surface roughness of each specimen were measured before and after all interventions with simulated cycles. Color differences (ΔE) and surface roughness (ΔR a ) data were analyzed using 2-way ANOVA, followed by the least significant difference test (α=.05). The correlation between ΔE and ΔR a was statistically analyzed using the Pearson correlation analysis. Within the lithium disilicate CAD groups, intervention did not result in any significant differences in color change (P>.05). Within the zirconia groups, a 15-year clinical simulation revealed significantly higher ΔE values than a simulated 5-year exposure (P=.017). Increased simulated cycles showed significantly higher R a values for all groups. Within the zirconia groups, B revealed

  19. Report On Design And Preliminary Data Of Halden In-Pile Creep Rig

    Energy Technology Data Exchange (ETDEWEB)

    Terrani, Kurt A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Karlsen, T. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    A set of in-pile creep tests is ongoing in the Halden reactor on ORNL’s candidate accident tolerant fuel cladding materials. These tests are meant to provide essential material property information that is needed for an informed analysis of these fuel concepts under normal operating conditions. These tests provide detailed information regarding swelling, thermal creep, and irradiation creep rates of these materials. The results to date have been compared with the limited set of information available in literature that is form irradiation tests in other reactors or out-of-pile tests. Most of the results are in good agreement with prior literature, except for irradiation creep rate of SiC. To elucidate the difference between the HFIR and Halden test results continued testing is necessary. The tests describe in this progress report are ongoing and will continue for at least another year.

  20. ThermoCycle: A Modelica library for the simulation of thermodynamic systems

    DEFF Research Database (Denmark)

    Quoilin, Sylvain; Desideri, Adriano; Wronski, Jorrit

    2014-01-01

    This paper presents the results of an on-going project to develop ThermoCycle, an open Modelica library for the simulation of low-capacity thermodynamic cycles and thermal systems. Special attention is paid to robustness and simulation speed since dynamic simulations are often limited by numerical...... constraints and failures, either during initialization or during integration. Furthermore, the use of complex equations of state (EOS) to compute thermodynamic properties significantly decreases the simulation speed. In this paper, the approach adopted in the library to overcome these challenges is presented...

  1. Analysis of Sandia in-pile EOS experiments

    International Nuclear Information System (INIS)

    Breitung, W.; Gorham-Bergeron, E.; Murata, K.K.

    1979-01-01

    Preliminary analysis has been carried out of the dynamic in-pile equation-of-state measurements for UO 2 , conducted at Sandia Laboratories, aimed at reducing the uncertainties in the effective UO 2 enthalpy corresponding to the measured pressures. Of the remaining width of the p-H band of some 350 J/g, about 200 J/g originate in the uncertainties of the analytical modelling and about 150 J/g result from the scatter in the experimental data

  2. In-pile experimental facility needs for LMFR safety research

    International Nuclear Information System (INIS)

    Kawata, Norio; Niwa, Hajime

    1994-01-01

    Although the achievement of the safety research during the past years has been significant, there still exists a strong need for future research, especially when there is prospect for future LMFR commercialization. In this paper, our current views are described on future research needs especially with a new in-pile experimental facility. The basic ideas and progress are outlined of a preliminary feasibility study. (author)

  3. The MOZART in-pile tritium extraction experiment

    International Nuclear Information System (INIS)

    Briec, M.

    1990-01-01

    In-pile tritium extraction behavior of various ceramics was compared in the MOZART experiment. The influence of temperature and purge gas composition was studied. The experimental results are analyzed by taking into account the processes of diffusion in the grain and desorption at grain surface. This analysis confirms that a better knowledge of the desorption process is necessary for a satisfactory explanation of the experimental data

  4. Effect of two-step and one-step surface conditioning of glass ceramic on adhesion strength of orthodontic bracket and effect of thermo-cycling on adhesion strength.

    Science.gov (United States)

    Asiry, Moshabab A; AlShahrani, Ibrahim; Alaqeel, Samer M; Durgesh, Bangalore H; Ramakrishnaiah, Ravikumar

    2018-08-01

    The adhesion strength of orthodontic brackets bonded to dental glass ceramics was evaluated after ceramic surface was treated with two-step and one-step surface conditioning systems, and subjecting to thermo-cycling. A total of forty specimens were fabricated from silica based glass ceramic (lithium disilicate) by duplicating the buccal surface of maxillary first premolar. The specimens were randomly assigned to two experimental groups (n = 20), group one specimens were treated with two-step surface conditioning system (IPS ceramic etching gel™ and Monobond plus™) and group two specimens were treated with one-step surface conditioning system (Monobond etch and prime™). The surface roughness of the specimens after treatment with two-step and one-step surface conditioning system was measured using non-contact surface profilometer. Ten randomly selected specimens from each group were subjected to thermo-cycling and the remaining ten served as baseline. The shear bond strength of the specimens was measured using universal material testing machine. The adhesive remnant index score was calculated, and the results of surface roughness and bond strength were tabulated and subjected to analysis of variance and post hoc tukey's test at a significance level of p step conditioning system had higher surface roughness and bond strength than one-step conditioning system. The majority of the specimens treated with both two-step and one-step conditioned specimens showed adhesive failure after subjecting thermo-cycling. Traditional two-step conditioning provides better bond strength. The clinical importance of the study is that, the silane promoted adhesion significantly reduces on exposure to thermo-cycling. Copyright © 2018 Elsevier Ltd. All rights reserved.

  5. Thermal conductivity of sintered UO2 under in-pile conditions

    International Nuclear Information System (INIS)

    Stora, J.P.; Bernardy De Sigoyer, B.; Delmas, R.; Deschamps, P.; Lavaud, B.; Ringot, C.

    1964-01-01

    The temperature distribution in a stack of sintered UO 2 cylinders has been studied both in the laboratory where the heat energy is produced by an axial heating element, and in-pile, where the heating is due solely to nuclear effects. Under a high thermal gradient the UO 2 cracks both along radial planes and along pseudo-cylindrical surfaces: these latter act as thermal barriers to the heat flow, It is therefore an apparent thermal conductivity k a (T), lower than the intrinsic value k(T) of this parameter which is measured. The efficiency of these barriers decreases when the gap decreases and when the external pressure acting on the cracked stack increases: in the limiting case, for high values of the binding strain, k a (T) ≅ k(T). In the domain of phonon conduction (T ≤ 1350 deg C), the expression kw.cm -1 .C -1 =1/(11+0.024*T) accounts for the real thermal conductivity. Above 1350 deg C the thermal conductivity increases. Two in-pile measurements up to 1250 deg C carried out using cartridges fitted with thermocouples confirm, within the limits of experimental error, the above expression and the qualitative effects of the binding strains. Similar tests have been carried out-of-pile and in-pile on the real shape of the EL-4 fuel 'pencils'. Out-of-pile, the influence of the initial free gap, of the nature of the gas filing the 'pencil' and of the external pressure have been studied; the results are compatible with the above interpretation; It appears that an external pressure of 60 kg/cm 2 is insufficient to restore completely the thermal conductivity of the fuel. (authors) [fr

  6. The Use of Infrared Thermography as a Novel Approach for Real-Time Validation of PCR Thermocyclers

    DEFF Research Database (Denmark)

    Grønlund, Hugo Ahlm; Löfström, Charlotta; Helleskov, Jens Bue

    2010-01-01

    Validation of PCR thermocycler performance is crucial to obtain reliable results. In this study, infrared (IR) thermography was evaluated as a novel validation tool. After stabilisation, no significant difference in the temperatures recorded using thermography and a reference blockbased system wa...... was found. By employing IR thermography, information about the length of the time until temperature stabilisation in the sample could be obtained. This study shows the potential of using IR thermography for validation of thermocyclers.......Validation of PCR thermocycler performance is crucial to obtain reliable results. In this study, infrared (IR) thermography was evaluated as a novel validation tool. After stabilisation, no significant difference in the temperatures recorded using thermography and a reference blockbased system...

  7. Ultrasonic Thermometry for In-Pile Temperature Detection

    International Nuclear Information System (INIS)

    Daw, J.E.; Rempe, J.L.; Wilkins, S.C.

    2002-01-01

    example, signal processing can be very complicated, as multiple echoes may overlap. Contact between the sensor and solid materials can cause extraneous echoes. If a sheath is required, contact bonding at high temperatures may cause extraneous echoes or attenuation of primary echoes. The most successful materials used in previous studies, tungsten and rhenium, are unattractive for nuclear applications due to material transmutation. Clearly, in order for ultrasonic thermometers to be viable for an in-pile sensor, these issues must be resolved through the use of modern signal processing and materials technologies. As part of the INL feasibility study, all of the issues associated with UT use and proposed resolution options will be identified and evaluated. Once most promising options are proven, it is planned to produce one or more prototype ultrasonic temperature sensors for evaluation. Ultimately, a full test should include a long term installation in a high temperature test assembly installed in a high neutron flux environment, such as that found in the Idaho National Laboratory's Advanced Test Reactor.

  8. Comparative evaluation of tensile bond strength of silicone-based denture liners after thermocycling and surface treatment.

    Science.gov (United States)

    Kaur, Harsimran; Datta, Kusum

    2015-01-01

    To examine, evaluate, and compare the tensile bond strength of two silicone-based liners; one autopolymerizing and one heat cured, when treated with different chemical etchants to improve their adhesion with denture base resin. Hundred and sixty test specimens of heat-cured polymethyl methacrylate (PMMA) were fabricated; out of which 80 specimens were tested for tensile bond strength after bonding it to autopolymerizing resilient liner (Ufigel P) and rest 80 to heat-cured resilient liner (Molloplast B). Each main group was further divided into four subgroups of 20 specimens each, one to act as a control and three were subjected to surface treatment with different chemical etchants namely dichloromethane, MMA monomer, and chloroform. The two silicone-based denture liners were processed between 2 PMMA specimens (10 mm × 10 mm × 40 mm) in the space provided by a spacer of 3 mm, thermocycled (5-55°C) for 500 cycles, and then their tensile strength measurements were done in the universal testing machine. One-way ANOVA technique showed a highly significant difference in the mean tensile bond strength values for all the groups. The Student's t-test computed values of statistics for the compared groups were greater than the critical values both at 5% and at 1% levels. Surface treatment of denture base resin with chemical etchants prior to the application of silicone-based liner (Ufigel P and Molloplast-B) increased the tensile bond strength. The increase was the highest with specimens subjected to 180 s of MMA surface treatment and the lowest with control group specimens.

  9. Comparative evaluation of tensile bond strength of silicone-based denture liners after thermocycling and surface treatment

    Directory of Open Access Journals (Sweden)

    Harsimran Kaur

    2015-01-01

    Full Text Available Purpose: To examine, evaluate, and compare the tensile bond strength of two silicone-based liners; one autopolymerizing and one heat cured, when treated with different chemical etchants to improve their adhesion with denture base resin. Materials and Methods: Hundred and sixty test specimens of heat-cured polymethyl methacrylate (PMMA were fabricated; out of which 80 specimens were tested for tensile bond strength after bonding it to autopolymerizing resilient liner (Ufigel P and rest 80 to heat-cured resilient liner (Molloplast B. Each main group was further divided into four subgroups of 20 specimens each, one to act as a control and three were subjected to surface treatment with different chemical etchants namely dichloromethane, MMA monomer, and chloroform. The two silicone-based denture liners were processed between 2 PMMA specimens (10 mm × 10 mm × 40 mm in the space provided by a spacer of 3 mm, thermocycled (5-55°C for 500 cycles, and then their tensile strength measurements were done in the universal testing machine. Results: One-way ANOVA technique showed a highly significant difference in the mean tensile bond strength values for all the groups. The Student′s t-test computed values of statistics for the compared groups were greater than the critical values both at 5% and at 1% levels. Conclusion: Surface treatment of denture base resin with chemical etchants prior to the application of silicone-based liner (Ufigel P and Molloplast-B increased the tensile bond strength. The increase was the highest with specimens subjected to 180 s of MMA surface treatment and the lowest with control group specimens.

  10. In-pile critical heat flux and post-dryout heat transfer measurements – A historical perspective

    Energy Technology Data Exchange (ETDEWEB)

    Groeneveld, D.C., E-mail: degroeneveld@gmail.com

    2017-06-15

    In the 1960s’ and 1970s’ Canada was a world leader in performing in-reactor heat transfer experiments on fuel bundles instrumented with miniature sheath thermocouples. Several Critical Heat Flux (CHF) and Post-CHF experiments were performed in Chalk River’s NRU and NRX reactors on water-cooled 3-, 18-, 19-, 21-, and 36-element fuel bundles. Most experiments were obtained at steady-state conditions, where the power was raised gradually from single-phase conditions up to the CHF and beyond. Occasionally, post-dryout temperatures up to 600 °C were maintained for several hours. In some tests, the fuel behaviour during loss-of-flow and blowdown transients was investigated – during these transients sheath temperatures could exceed 2000 °C. Because of the increasingly more stringent licensing requirements for in-pile heat transfer tests on instrumented fuel bundles, no in-pile CHF and post-dryout tests on fuel bundles have been performed anywhere in the world for the past 40 years. This paper provides details of these unique in-pile experiments and describes some of their heat transfer results.

  11. In-pile gamma spectrometry and irradiation control at Osiris

    International Nuclear Information System (INIS)

    Farny, G.; Destot, M.; Corre, J.; Texier, D.; Faugere, J.L.; Mouchnino, M.

    1975-01-01

    A new gamma spectrometry facility is available near Osiris reactor core, at Saclay. This device enables nuclear fuels to be examined in loops or capsules all along their irradiation, a few minutes being sufficient to transfer the fuel from the irradiation place to the measurement bench. So, spacelike and timelike history of a lot of fission products, especially short-lived radionuclides, can be observed. Using such in-pile spectrometry device, of original design, allows to avoid radioactive decay corrections and the risks of any information less. Performance of the device is given together with some preliminary results and their interpretation [fr

  12. ECP measurements under neutron and gamma ray in in-pile loop and their data evaluation by water radiolysis calculations

    Energy Technology Data Exchange (ETDEWEB)

    Hanawa, S.; Nakamura, T.; Uchida, S. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan); Kus, P.; Vsolak, R.; Kysela, J. [Nuclear Research Inst. Rez plc, Husinec - Rez (Czech Republic)

    2010-07-01

    In order to establish reliable electrochemical corrosion potential (ECP) sensors for applying in reactor core peripherals of power plants, performance tests of sensors under irradiation were carried out in the in-pile loop of the experimental reactor, LVR-15, at the Nuclear Research Institute (NRI) in Czech Republic. Responses of different kinds of sensors under neutron and gamma irradiation conditions have been compared each other. Corrosive conditions along the in-pile loop were calculated by water radiolysis calculation code, WRAC-J and calculated corrosive conditions were compared with the measured results. As a result of the evaluation, it was confirmed that the ECP sensors could be applied to irradiation conditions of reactor peripherals, while the water radiolysis model could be also applied for evaluation of corrosive conditions of reactor peripherals. (author)

  13. New in-pile water loop facility for IASCC studies at JMTR

    International Nuclear Information System (INIS)

    Tsukada, T.; Tsuji, H.; Nakajima, H.; Komori, Y.; Ito, H.

    2002-01-01

    Irradiation assisted stress corrosion cracking (IASCC) is caused by the synergistic effects of neutron and gamma radiation, residual and applied stresses and high temperature water environment on the structural materials of vessel internals. IASCC has been studied since the beginning of the 1980's and the phenomenological knowledge on IASCC is accrued extensively. However, mainly due to the experimental difficulties, data for the mechanistic understanding and prediction of failures of the specific in-vessel components are still insufficient and further well-controlled experiments are needed [1]. In recent years, efforts to perform the in-pile materials test for IASCC study have been made at some research reactors [2-4]. At JAERI, a high temperature water loop facility was designed to install at the Japan Materials Testing Reactor (JMTR) to carry out the in-core IASCC testing. This report describes an overview of design and specification of the loop facility. (authors)

  14. IR thermocycler for centrifugal microfluidic platform with direct on-disk wireless temperature measurement system

    Science.gov (United States)

    Burger, J.; Gross, A.; Mark, D.; Roth, G.; von Stetten, F.; Zengerle, R.

    2011-06-01

    The direct on-disk wireless temperature measurement system [1,2] presented at μTAS 2010 was further improved in its robustness. We apply it to an IR thermocycler as part of a centrifugal microfluidic analyzer for polymerase chain reactions (PCR). This IR thermocycler allows the very efficient direct heating of aqueous liquids in microfluidic cavities by an IR radiation source. The efficiency factor of this IR heating system depends on several parameters. First there is the efficiency of the IR radiator considering the transformation of electrical energy into radiation energy. This radiation energy needs to be focused by a reflector to the center of the cavity. Both, the reflectors shape and the quality of the reflecting layer affect the efficiency. On the way to the center of the cavity the radiation energy will be diminished by absorption in the surrounding air/humidity and especially in the cavity lid of the microfluidic disk. The transmission spectrum of the lid material and its thickness is of significant impact. We chose a COC polymer film with a thickness of 150 μm. At a peak frequency of the IR radiator of ~2 μm approximately 85 % of the incoming radiation energy passes the lid and is absorbed within the first 1.5 mm depth of liquid in the cavity. As we perform the thermocycling for a PCR, after heating to the denaturation temperature of ~ 92 °C we need to cool down rapidly to the primer annealing temperature of ~ 55 °C. Cooling is realized by 3 ventilators venting air of room temperature into the disk chamber. Due to the air flow itself and an additional rotation of the centrifugal microfluidic disk the PCR reagents in the cavities are cooled by forced air convection. Simulation studies based upon analogous electrical models enable to optimize the disk geometry and the optical path. Both the IR heater and the ventilators are controlled by the digital PID controller HAPRO 0135 [3]. The sampling frequency is set to 2 Hz. It could be further increased up

  15. Effects of air-abrasion pressure on the resin bond strength to zirconia: a combined cyclic loading and thermocycling aging study

    Directory of Open Access Journals (Sweden)

    Eman Z. Al-Shehri,

    2017-06-01

    Full Text Available Objectives To determine the combined effect of fatigue cyclic loading and thermocycling (CLTC on the shear bond strength (SBS of a resin cement to zirconia surfaces that were previously air-abraded with aluminum oxide (Al2O3 particles at different pressures. Materials and Methods Seventy-two cuboid zirconia specimens were prepared and randomly assigned to 3 groups according to the air-abrasion pressures (1, 2, and 2.8 bar, and each group was further divided into 2 groups depending on aging parameters (n = 12. Panavia F 2.0 was placed on pre-conditioned zirconia surfaces, and SBS testing was performed either after 24 hours or 10,000 fatigue cycles (cyclic loading and 5,000 thermocycles. Non-contact profilometry was used to measure surface roughness. Failure modes were evaluated under optical and scanning electron microscopy. The data were analyzed using 2-way analysis of variance and χ2 tests (α = 0.05. Results The 2.8 bar group showed significantly higher surface roughness compared to the 1 bar group (p < 0.05. The interaction between pressure and time/cycling was not significant on SBS, and pressure did not have a significant effect either. SBS was significantly higher (p = 0.006 for 24 hours storage compared to CLTC. The 2 bar-CLTC group presented significantly higher percentage of pre-test failure during fatigue compared to the other groups. Mixed-failure mode was more frequent than adhesive failure. Conclusions CLTC significantly decreased the SBS values regardless of the air-abrasion pressure used.

  16. In-pile Thermal Conductivity Characterization with Time Resolved Raman

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xinwei [Iowa State Univ., Ames, IA (United States). Dept. of Mechanical Engineering; Hurley, David H. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2018-03-19

    The project is designed to achieve three objectives: (1) Develop a novel time resolved Raman technology for direct measurement of fuel and cladding thermal conductivity. (2) Validate and improve the technology development by measuring ceramic materials germane to the nuclear industry. (3) Conduct instrumentation development to integrate optical fiber into our sensing system for eventual in-pile measurement. We have developed three new techniques: time-domain differential Raman (TD-Raman), frequency-resolved Raman (FR-Raman), and energy transport state-resolved Raman (ET-Raman). The TD-Raman varies the laser heating time and does simultaneous Raman thermal probing, the FR-Raman probes the material’s thermal response under periodical laser heating of different frequencies, and the ET-Raman probes the thermal response under steady and pulsed laser heating. The measurement capacity of these techniques have been fully assessed and verified by measuring micro/nanoscale materials. All these techniques do not need the data of laser absorption and absolute material temperature rise, yet still be able to measure the thermal conductivity and thermal diffusivity with unprecedented accuracy. It is expected they will have broad applications for in-pile thermal characterization of nuclear materials based on pure optical heating and sensing.

  17. In-pile loop experiments in water chemistry and corrosion

    International Nuclear Information System (INIS)

    Kysela, J.

    1986-09-01

    Results on the study of Zr-1% Nb alloy corrosion, in out-of and in-pile loops simulating the working conditions of the VVER-440 reactor (Soviet, PWR type), covered the time period May 1982-April 1986 were reported, as well as, results on transport and filtration of corrosion products. Methods and techniques used in the study included remote measurement of corrosion rate by polarizing resistance, out-of-pile loop at the temperature 350 deg. C, pressure 19 MPa, circulation 20 kgs/h and in-pile water loop with constant flow rate 10,000 kgs/h, pressure 16 MPa, temperature 330 deg. C and neutron flux 7x10 13 n/cm 2 .s. It was shown that solid suspended particles with chemical composition corresponding most frequently to magnetite or nickelous ferrite, though with non-stoichiometric composition Me x 2+ Fe 3- x 3+ O 4 were found. Continuous filtration of water by means of electromagnetic filter leads to a decrease of radioactivity of the outer epitactic layer only. Effect of filtration on the inner topotactic layer is negligible. The corrosion rates for the above-mentioned parameters are given

  18. Characteristics of pressure control system on PWR/PHWR in pile loop facility

    International Nuclear Information System (INIS)

    Sarwani; Hendro, P.; Suwoto; Sutrisno

    1998-01-01

    PWR/PHWR in-pile loop facility is used for testing of fuel element bundle which is correspond to the condition of power reactor operation. So, this facility is designed at 150 bar of pressure and 350 o C of temperature. Pressure control system is one of the components of the facility and it is equipped with 6 electrical heaters (30 KW), water spray, pressure and temperature monitors. The characterization test of pressure control system has been carried out with operating of 2 electrical heaters (10 KW). The K eff calculation value is different 5.2% from pressure in the pressure control system can be increased to 160 bar within 27 hours. After the system pressure reached the nominal pressure, the pressure control system was in the steady state condition

  19. In-pile Hydrothermal Corrosion Evaluation of Coated SiC Ceramics and Composites

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, David [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Ang, Caen [Univ. of Tennessee, Knoxville, TN (United States); Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    Hydrothermal corrosion accelerated by water radiolysis during normal operation is among the most critical technical feasibility issues remaining for silicon carbide (SiC) composite-based cladding that could provide enhanced accident-tolerance fuel technology for light water reactors. An integrated in-pile test was developed and performed to determine the synergistic effects of neutron irradiation, radiolysis, and pressurized water flow, all of which are relevant to a typical pressurized water reactor (PWR). The test specimens were chosen to cover a range of SiC materials and a variety of potential options for environmental barrier coatings. This document provides a summary of the irradiation vehicle design, operations of the experiment, and the specimen loading into the irradiation vehicle.

  20. In-pile measurement of the thermal conductivity of irradiated metallic fuel

    International Nuclear Information System (INIS)

    Bauer, T.H.; Holland, J.W.

    1995-01-01

    Transient test data and posttest measurements from recent in-pile overpower transient experiments are used for an in situ determination of metallic fuel thermal conductivity. For test pins that undergo melting but remain intact, a technique is described that relates fuel thermal conductivity to peak pin power during the transient and a posttest measured melt radius. Conductivity estimates and their uncertainty are made for a database of four irradiated Integral Fast Reactor-type metal fuel pins of relatively low burnup (<3 at.%). In the assessment of results, averages and trends of measured fuel thermal conductivity are correlated to local burnup. Emphasis is placed on the changes of conductivity that take place with burnup-induced swelling and sodium logging. Measurements are used to validate simple empirically based analytical models that describe thermal conductivity of porous media and that are recommended for general thermal analyses of irradiated metallic fuel

  1. Measurement data of cesium 137 yields in primary coolant of an in-pile water loop in fission products release experiment

    International Nuclear Information System (INIS)

    Ishiwatari, Nasumi; Nagai, Hitoshi; Takeda, Tsuneo

    1979-03-01

    Series of fuel rods (UO 2 pellets sheathed with stainless steel) having an artificial pinhole were irradiated in the in-pile test section of water loop JMTR OWL-1. Presented are the results of measurements of cesium 137 yields in primary coolant of OWL-1 from 1975 to 1978. (author)

  2. Acoustic sensor for in-pile fuel rod fission gas release measurement

    International Nuclear Information System (INIS)

    Fourmentel, D.; Villard, J. F.; Ferrandis, J. Y.; Augereau, F.; Rosenkrantz, E.; Dierckx, M.

    2009-01-01

    We have developed a specific acoustic sensor to improve the knowledge of fission gas release in Pressurized Water Reactor (PWR) fuel rods when irradiated in materials testing reactors. In order to perform experimental programs related to the study of the fission gas release kinetics, the CEA (French Nuclear Energy Commission) acquired the ability to equip a pre-irradiated PWR fuel rod with three sensors, allowing the simultaneous on-line measurements of the following parameters: - fuel temperature with a centre-line thermocouple type C, - internal pressure with a specific counter-pressure sensor, - fraction of fission gas released in the fuel rod with an innovative acoustic sensor. The third detector is the subject of this paper. This original acoustic sensor has been designed to measure the molar mass and pressure of the gas contained in the fuel rod plenum. For in-pile instrumentation, the fraction of fission gas, such as Krypton and Xenon, in Helium, can be deduced online from this measurement. The principle of this acoustical sensor is the following: a piezoelectric transducer generates acoustic waves in a cavity connected to the fuel rod plenum. The acoustic waves are propagated and reflected in this cavity and then detected by the transducer. The data processing of the signal gives the velocity of the acoustic waves and their amplitude, which can be related respectively to the molar mass and to the pressure of the gas. The piezoelectric material of this sensor has been qualified in nuclear conditions (gamma and neutron radiations). The complete sensor has also been specifically designed to be implemented in materials testing reactors conditions. For this purpose some technical points have been studied in details: - fixing of the piezoelectric sample in a reliable way with a suitable signal transmission, - size of the gas cavity to avoid any perturbation of the acoustic waves, - miniaturization of the sensor because of narrow in-pile experimental devices

  3. In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

    International Nuclear Information System (INIS)

    Gunnerson, F.S.; MacDonald, P.E.

    1980-01-01

    The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made

  4. Improvement in properties of welded joints of titanium alloy VT22 by thermocyclic treatment

    International Nuclear Information System (INIS)

    Lyasotskaya, V.S.; Kulikov, F.R.; Kirillov, Yu.G.; Ravdonikas, N.Yu.

    1983-01-01

    The results of investigations of the thermocyclic treatment (TCT) effect on the structure and properties of butt welded joints of tubes (with external diameter 180 mm and wall thickness 20-25 mm) of the VT22 alloy are presented. Welded joints have been obtained by means of multipassing automatic argon-arc (ARAW) and electron-beam (ELB) welding. It is shown that TCT of welded joints of the VT22 alloy results in formation in all zones of substructure with disperse precipitations of α-phase which is analogous to the structure of near welded seam zone metal immediately after welding. As a result of TCT and subsequent TT of welded joints poligonization and recrystallization processes of α- and #betta#-phases, changes in parameters of structural components and thin phase structure take place. TCT with strengthening TT or annealing leads to strength increase, while TCT with annealing besides that improves placticity and impact strength of the VT22 alloy welded joints

  5. Droplet centrifugation, droplet DNA extraction, and rapid droplet thermocycling for simpler and faster PCR assay using wire-guided manipulations.

    Science.gov (United States)

    You, David J; Yoon, Jeong-Yeol

    2012-09-04

    A computer numerical control (CNC) apparatus was used to perform droplet centrifugation, droplet DNA extraction, and rapid droplet thermocycling on a single superhydrophobic surface and a multi-chambered PCB heater. Droplets were manipulated using "wire-guided" method (a pipette tip was used in this study). This methodology can be easily adapted to existing commercial robotic pipetting system, while demonstrated added capabilities such as vibrational mixing, high-speed centrifuging of droplets, simple DNA extraction utilizing the hydrophobicity difference between the tip and the superhydrophobic surface, and rapid thermocycling with a moving droplet, all with wire-guided droplet manipulations on a superhydrophobic surface and a multi-chambered PCB heater (i.e., not on a 96-well plate). Serial dilutions were demonstrated for diluting sample matrix. Centrifuging was demonstrated by rotating a 10 μL droplet at 2300 round per minute, concentrating E. coli by more than 3-fold within 3 min. DNA extraction was demonstrated from E. coli sample utilizing the disposable pipette tip to cleverly attract the extracted DNA from the droplet residing on a superhydrophobic surface, which took less than 10 min. Following extraction, the 1500 bp sequence of Peptidase D from E. coli was amplified using rapid droplet thermocycling, which took 10 min for 30 cycles. The total assay time was 23 min, including droplet centrifugation, droplet DNA extraction and rapid droplet thermocycling. Evaporation from of 10 μL droplets was not significant during these procedures, since the longest time exposure to air and the vibrations was less than 5 min (during DNA extraction). The results of these sequentially executed processes were analyzed using gel electrophoresis. Thus, this work demonstrates the adaptability of the system to replace many common laboratory tasks on a single platform (through re-programmability), in rapid succession (using droplets), and with a high level of

  6. Droplet centrifugation, droplet DNA extraction, and rapid droplet thermocycling for simpler and faster PCR assay using wire-guided manipulations

    Directory of Open Access Journals (Sweden)

    You David J

    2012-09-01

    Full Text Available Abstract A computer numerical control (CNC apparatus was used to perform droplet centrifugation, droplet DNA extraction, and rapid droplet thermocycling on a single superhydrophobic surface and a multi-chambered PCB heater. Droplets were manipulated using “wire-guided” method (a pipette tip was used in this study. This methodology can be easily adapted to existing commercial robotic pipetting system, while demonstrated added capabilities such as vibrational mixing, high-speed centrifuging of droplets, simple DNA extraction utilizing the hydrophobicity difference between the tip and the superhydrophobic surface, and rapid thermocycling with a moving droplet, all with wire-guided droplet manipulations on a superhydrophobic surface and a multi-chambered PCB heater (i.e., not on a 96-well plate. Serial dilutions were demonstrated for diluting sample matrix. Centrifuging was demonstrated by rotating a 10 μL droplet at 2300 round per minute, concentrating E. coli by more than 3-fold within 3 min. DNA extraction was demonstrated from E. coli sample utilizing the disposable pipette tip to cleverly attract the extracted DNA from the droplet residing on a superhydrophobic surface, which took less than 10 min. Following extraction, the 1500 bp sequence of Peptidase D from E. coli was amplified using rapid droplet thermocycling, which took 10 min for 30 cycles. The total assay time was 23 min, including droplet centrifugation, droplet DNA extraction and rapid droplet thermocycling. Evaporation from of 10 μL droplets was not significant during these procedures, since the longest time exposure to air and the vibrations was less than 5 min (during DNA extraction. The results of these sequentially executed processes were analyzed using gel electrophoresis. Thus, this work demonstrates the adaptability of the system to replace many common laboratory tasks on a single platform (through re-programmability, in rapid succession (using droplets

  7. Use of standard reliability levels in design and safety assessment of in-pile loops

    International Nuclear Information System (INIS)

    Bogani, G.; Verre, A.; Balestreri, S.; Colombo, A.G.; Luisi, T.

    1975-01-01

    This paper describes a logic-probabilistic analysis technique for a critical design review and safety assessment of in-pile loops. The examples in this paper refer to the analysis performed for the experimental loops already constructed or under construction in the ESSOR reactor of the Joint Research Centre of Ispra, as irradiation facilities for fuel element research and development tests. The proposed technique is based on the classification into categories of components and protective device malfunctions. Such subdivision into categories was agreed upon by the Italian Safety Authority and Euratom JRC, and adopted for the safety assessment of the ESSOR reactor in-pile loops. For each category, the method makes a link with a corresponding malfunction probability range (probability level). This probability level is defined taking into account design, construction, inspection and maintenance criteria as well as periodic controls; therefore the quality level and consequently the reliability level are thus also defined. The analysis is developed in the following stages: (1) definition of the analysis object (top event) and drawing of the relative fault-tree; (2) loop design analysis and preliminary optimization based on logic criteria; (3) classification into categories of the fault-tree primary events; (4) final loop design analysis and optimization based on defined component quality requirements. Stages 2 and 4 are quite different since stage 2 mainly consists of a redundance optimization, while stage 4 acts on the component quality level in such a way that each minimum cut-set leading to the top has an acceptable probability level. During analysis development, use is made of computer codes which, among other things enable the verification of fault-tree logic makeup, the listing of the minimum cut-sets with and without event categorization, and the evaluation of each cut-set order. (author)

  8. Fuel disruption mechanisms determined in-pile in the ACRR

    International Nuclear Information System (INIS)

    Wright, S.A.; Fischer, E.A.

    1984-09-01

    Over thirty in-pile experiments were performed to investigate fuel disruption behavior for LMFBR loss of flow (LOF) accidents. These experiments reproduced the heating transients for a variety of accidents ranging from slow LOF accidents to rapid LOF-driven-TOP accidents. In all experiments the timing and mode of the fuel disruption were observed with a high speed camera, enabling detailed comparisons with a fuel pin code, SANDPIN. This code transient intra- and inter-granular fission gas behavior to predict the macroscopic fuel behavior, such as fission gas induced swelling and frothing, cracking and breakup of solid fuel, and fuel vapor pressure driven dispersal. This report reviews the different modes of fuel disruption as seen in the experiments and then describes the mechanism responsible for the disruption. An analysis is presented that describes a set of conditions specifying the mode of fuel disruption and the heating conditions required to produce the disruption. The heating conditions are described in terms of heating rate (K/s), temperature gradient, and fuel temperature. A fuel disruption map is presented which plots heating rate as a function of fuel temperature to illustrate the different criteria for disruption. Although this approach to describing fuel disruption oversimplifies the fission gas processes modeled by SANDPIN, it does illustrate the criteria used to determine which fuel disruption mechanism is dominant and on what major fission gas parameters it depends

  9. Evaluation of the in-pile pressure data from instrumented fuel assemblies IFA-431 and IFA-432

    International Nuclear Information System (INIS)

    Bradley, E.R.; Cunningham, M.E.; Lanning, D.D.; Williford, R.E.

    1979-10-01

    This report includes results of the examination of the in-pile pressure data from instrumented test assemblies IFA-431 and 432. The pressure data have been used to estimate the fission gas release fraction as a function of fuel burnup. Included are comparisons of the estimated release functions and those predicted by three fission gas release models using the experimental temperature histories of the fuel rods. These comparisons show that fuel temperature is the primary factor in determining fission gas release and that burnup-enhanced fission gas release is not important in UO 2 fuels irradiated to 1700 GJ/kgU

  10. Study of a device for the direct measurement of the fission gas pressure inside an in-pile fuel element

    International Nuclear Information System (INIS)

    Lavaud, B.; Uschanoff, S.

    1964-01-01

    The fission gas pressure inside a fuel element made of a refractory fuel constitutes an important limiting factor for the burn-up. Although it is possible to calculate approximately the volume of gas produced outside the fuel during its life-time; it is nevertheless very difficult to evaluate the pressure since the volume allowed to the fission gases, as well as their temperature are known only very approximately. This physical value, which is essential for the technologist, can only be known by direct in-pile measurement of the pressure. The report describes the equipment which has been developed for this test. (authors) [fr

  11. In-pile creep behaviour of Zry-4 and ZrNb3Sn1 cladding under uniaxial and biaxial stress

    International Nuclear Information System (INIS)

    Boehner, G.; Wildhagen, B.; Wilhelm, H.

    1987-01-01

    An irradiation programme - started in 1977 - was performed at the research reactor FRG-2 at Geesthacht, Germany, as a joint project of GKSS and KWU in order to study the in-pile creep behaviour of zirconium alloy cladding tubes of PWR fuel rods. The test objective was to establish a data base which allows refined modelling of the in-pile creep phenomenon. A wide test matrix was realized in which each of the precisely monitored test conditions (hoop stress, temperature, fast neutron flux) was varied separately. Different cladding materials (Zircaloy-4 and Zirconium-Niob-Tin alloy ZrNb3Sn1) were subjected to those varying test conditions. Cladding tube specimens of 10.75 mm outer diameter were irradiated in test capsules under various stress conditions and levels up to approx. 6000 h, at temperatures ranging from 300 0 C to 400 0 C and fast neutron flux (E > 1 MeV) of approx. 3x10 13 cm -2 .s -1 . Diametrical and/or axial creep deformation of all tubes were measured in the Hot Cells several times in the course of the tests. In order to extract the irradiation induced creep strain some out-pile experiments were carried out under the very same test conditions as the in-pile tests concerned. (orig./GL)

  12. A review of radionuclide release and transport in recent in-pile experiments

    International Nuclear Information System (INIS)

    Harman, N.F.; Clough, P.N.

    1992-01-01

    The experimental series, reviewed in this work, are LOFT-LP-FP-2, PBF SFD ST to 1-4, Treat/STEP I to 4, and ACRR ST-1 and ST-2. These have the common features of in-core heating of a test fuel bundle to high temperatures (usually with some fuel melting) in an experimental reactor core, and of collecting and analysing the released fission products. They were designed to provide detailed information on the release from fuel of fission products and other radionuclides under LWR severe accident conditions, and on the chemical and physical forms and transport of the fission products. The main aim of this review is to bring together, in a systematic way, information on the conduct of the tests, on their successes and failures, and particularly on the information they generated on the chemical and physical behaviour of released fission products. By examining and analysing the data from all of the tests together, patterns of fission product behaviour may become apparent and insights may be gained, which would not be arrived at from individual test results. Moreover, important lessons may be learned, and useful guidance obtained, relating to the aims and conduct of future experimental programmes of fission product release from fuel and transport behaviour. The conclusions should be particularly relevant to the imminent Phebus-FP in-pile test series at Cadarache

  13. Development of In-pile Plug Assembly and Primary Shutter for Cold Neutron Guide System

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Jin Won; Cho, Yeong Garp; Ryu, Jeong Soo; Lee, Jung Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    The HANARO, a 30 MW multi-purpose research reactor in Korea, will be equipped with a neutron guide system, in order to transport cold neutrons from the neutron source to the neutron scattering instruments in the neutron guide hall near the reactor building. The neutron guide system of HANARO consists of the in-pile plug assembly with in-pile guides, the primary shutter with in-shutter guides, the neutron guides in the guide shielding room with dedicated secondary shutters, and the neutron guides connected to the instruments in the neutron guide hall. The functions of the in-pile plug assembly are to shield the reactor environment from a nuclear radiation and to support the neutron guides and maintain them precisely oriented. The primary shutter is a mechanical device to be installed just after the in-pile plug assembly, which stops neutron flux on demand. This report describes the mechanical design, fabrication, and installation procedure of the in-pile plug assembly and the primary shutter for the neutron guide system at HANARO. A special tool and procedure for a replacement of in-pile plug and guide cassette is also presented with the interface condition in the reactor hall.

  14. Innovations for In-Pile Measurements in the Framework of the CEA-SCK•CEN Joint Instrumentation Laboratory

    Science.gov (United States)

    Villard, Jean-Francois; Schyns, Marc

    2010-12-01

    Optimizing the life cycle of nuclear systems under safety constraints requires high-performance experimental programs to reduce uncertainties on margins and limits. In addition to improvement in modeling and simulation, innovation in instrumentation is crucial for analytical and integral experiments conducted in research reactors. The quality of nuclear research programs relies obviously on an excellent knowledge of their experimental environment which constantly calls for better online determination of neutron and gamma flux. But the combination of continuously increasing scientific requirements and new experimental domains -brought for example by Generation IV programsnecessitates also major innovations for in-pile measurements of temperature, dimensions, pressure or chemical analysis in innovative mediums. At the same time, the recent arising of a European platform around the building of the Jules Horowitz Reactor offers new opportunities for research institutes and organizations to pool their resources in order to face these technical challenges. In this situation, CEA (French Nuclear Energy Commission) and SCK'CEN (Belgian Nuclear Research Centre) have combined their efforts and now share common developments through a Joint Instrumentation Laboratory. Significant progresses have thus been obtained recently in the field of in-pile measurements, on one hand by improvement of existing measurement methods, and on the other hand by introduction in research reactors of original measurement techniques. This paper highlights the state-of-the-art and the main requirements regarding in-pile measurements, particularly for the needs of current and future irradiation programs performed in material testing reactors. Some of the main on-going developments performed in the framework of the Joint Instrumentation Laboratory are also described, such as: - a unique fast neutron flux measurement system using fission chambers with 242Pu deposit and a specific online data processing

  15. Nuclear heating measurements by in-pile calorimetry: prospective works for a microsensor design

    Energy Technology Data Exchange (ETDEWEB)

    Reynard-Carette, C.; Carette, M.; Aguir, K.; Bendahan, M.; Fiorido, T. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Lyoussi, A.; Fourmentel, D.; Villard, J.F. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 (France); Barthes, M.; Lanzetta, F.; Layes, G.; Vives, S. [FEMTO-ST, UMR 6174, Departement ENERGIE, Universite de Franche-Comte, 90000, Belfort (France)

    2015-07-01

    Since 2009 works have been performed in the framework of joint research programs between CEA and Aix-Marseille University. The main aim of these programs is to design and develop in-pile instrumentations, advanced calibration procedure and accurate measurement methods in particular for the new Material Testing Reactor (MTR) under construction in the South of France: Jules Horowitz Reactor (JHR). One major sensor is a specific radiometric calorimeter, which was studied out-of-pile from a thermal point of view and in-pile during irradiation campaigns. This sensor type is dedicated to measurements of nuclear heating (energy deposition rate per mass unit induced by interactions between nuclear rays and matter) inside experimental channels of MTRs. This kind of in-pile calorimeter corresponds to heat flux calorimeter exchanging with the external cooling fluid. This thermal running mode allows the establishment of steady thermal conditions inside the sensor to carry out online continuous measurements inside the reactor (core or reflector). Two main types of calorimeters exist. The first type consists of a single cell calorimeter. It is divided into a sample of material to be tested and a jacket instrumented with two thermocouples or a single thermocouple (Gamma Thermometer). The second, called a differential calorimeter, is composed of two superposed twin cells (a measurement cell containing a sample of material, and a reference cell to remove the heating of the cell body) instrumented with four thermocouples and two electrical heaters. Contrary to a single-cell calorimeter, a differential calorimeter allows the compensation of the parasite nuclear heating of the sensor body or jacket. Moreover, it possesses interesting advantages: thanks to the heaters embedded in the cells, three different measurement methods can be applied during irradiations to quantify nuclear heating. The first one is based on the use of out-of-pile calibration curves obtained by generating a heat

  16. In-Pile thermal fatigue of First Wall mock-ups under ITER relevant conditions

    International Nuclear Information System (INIS)

    Blom, F.; Schmalz, F.; Kamer, S.; Ketema, D.J.

    2006-01-01

    The objective of this study is to perform in-pile thermal fatigue testing of three actively cooled First Wall (FW) mock-ups to check the effect of neutron irradiation on the Be/CuCrZr joints under representative FW operation conditions. Three FW mock-ups with Beryllium armor tiles will be neutron irradiated at 1 dpa (in Be) with parallel thermal fatigue testing for 30,000 cycles. The temperatures, stress distributions and stress amplitudes at the Be/CuCrZr interface of the mock-ups will be as close as possible to the values calculated for ITER FW panels. For this objective the PWM mocks-up subjected to thermal fatigue will be integrated with high density (W) plates on the Be-side to provide heat flux by nuclear heating. The assembly will be placed in the pool-side facility of the HFR and thermal cycling is then arranged by mechanical movement towards and from the core box. As the thermal design of the irradiation rig is very critical a pilot-irradiation will be performed to cross check the models used in the thermal design of the rig. The project is currently in the design phase of both the pilot and actual irradiation rig. The irradiation of the actual rig is planned to start at mid 2007 and last for two years. (author)

  17. In-pile experiments on fuel rod behavior during a LOCA

    International Nuclear Information System (INIS)

    Karb, E.; Pruessmann, M.; Sepold, L.

    1980-05-01

    This report describes the results of the Test Series F, Tests F 1 through F 5, in the in-pile experimental program with single rods in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The research is part of the Nuclear Safety Project's (PNS) fuel behavior program. The main objective of the FR2-LOCA tests is to provide information about the effects of a nuclear environment on the mechanisms of fuel rod failure in the second heatup phase of a LOCA. The test rods have a heated length of 50 cm, and their radial dimensions are identical with those of a commercial German PWR. The main parameter of the FR2-LOCA test program is the burnup. The F tests were perfomed from Oct. 25, 1977 to Nov. 22, 1977. They were the first tests in this program to use pre-irradiated fuel rods. The nominal burnup of the test rods was 20 000 MWd/t. During the transient test, the test rods were subjected to rod powers between 36 and 41 W/cm and were pressurized with He to hot internal pressures between 46 and 83 bar. The test rods during the heatup phase at pressures of 56, 53, 42, 72 and 60 bar, respectively. The burst temperatures were determined to be 890, 893, 932, 835 and 880 0 C for test F 1 through F 5. The maximum total circumferential elongations amount to 59, 38, 27, 34 and 41%, respectively. The F tests revealed a fragmentation of the fuel after the irradiation (prior to the tests) and a disintegration of the fuel pellet column after the transient tests due to cladding ballooning. The post-test results indicated a significant reduction of the pellet stack length for all five test rods. The burst data of the F tests did not reveal any difference between tests with unirradiated fuel rods and the irradiated fuel rods of this test series. (orig./HP) [de

  18. Evaluation of the surface hardness, roughness, gloss and color of composites after different finishing/polishing treatments and thermocycling using a multitechnique approach.

    Science.gov (United States)

    Pala, Kanşad; Tekçe, Neslihan; Tuncer, Safa; Serim, Merve Efe; Demirci, Mustafa

    2016-01-01

    The objectives of this study were to evaluate the mechanical and physical properties of resin composites. The materials evaluated were the Clearfil Majesty Posterior, Filtek Z550 and G-aenial Posterior composites. A total of 189 specimens were fabricated for microhardness, roughness, gloss and color tests. The specimens were divided into three finishing and polishing systems: Enhance, OneGloss and Sof-Lex Spiral. Microhardness, roughness, gloss and color were measured after 24 h and after 10,000 thermocycles. Two samples from each group were evaluated using SEM and AFM. G-aenial Posterior exhibited the lowest microhardness values. The mean roughness ranged from 0.37 to 0.61 µm. The smoothest surfaces were obtained with Sof-Lex Spiral for each material. G-aenial Posterior with Enhance was determined to be the glossiest surfaces. All of the materials exhibited similar ΔE values ranging between 1.69 and 2.75. Sof-Lex Spiral discs could be used successfully to polish composites.

  19. In-pile creep strain and failure of CW 316 Ti pressurized tubes

    International Nuclear Information System (INIS)

    Boutard, J.L.; Maillard, A.; Carteret, Y.; Levy, V.; Meny, L.

    1984-06-01

    The in-pile creep and failure behavior of CW 316 Ti pressurized tubes irradiated in the same rig at 660-680 0 C and 81.4 dpaF max in Phenix is presented and compared to monitors of the same heat. The in-pile plastic strains are of the same order of what is expected from the monitors and are rather independent of the dose rate in the range 4 to 9 x 10 -3 dpaF/h. Such a behavior supports the assumption that the out-of-pile deformation mechanisms are operative in pile and a certain balance occurs between modification of the microstructure, dynamic hardening and deformation mechanisms due to irradiation. Examinations by fractography and optical micrography, show that the failures are intergranular either in-pile or out-of-pile. In both cases the damage consists in intergranular wedge cracks, and no cavitation can be observed by transmission electron microscopy. Then the in-pile embrittlement which gives lower failure strain and time is to be associated with a decrease of the surface energy of grain-boundaries rather then growth and coalescence of cavities

  20. Viability of Pushrod Dilatometry Techniques for High Temperature In-Pile Measurements

    Energy Technology Data Exchange (ETDEWEB)

    J. E. Daw; J. L. Rempe; D. L. Knudson; K. G. Condie; J. C. Crepeau

    2008-03-01

    To evaluate the performance of new fuel, cladding, and structural materials for use in advanced and existing nuclear reactors, robust instrumentation is needed. Changes in material deformation are typically evaluated out-of-pile, where properties of materials are measured after samples were irradiated for a specified length of time. To address this problem, a series of tests were performed to examine the viability of using pushrod dilatometer techniques for in-pile instrumentation to measure deformation. The tests were performed in three phases. First, familiarity was gained in the use and accuracy of this system by testing samples with well defined thermal elongation characteristics. Second, high temperature data for steels, specifically SA533 Grade B, Class 1 (SA533B1) Low Alloy Steel and Stainless Steel 304 (SS304), found in Light Water Reactor (LWR) vessels, were aquired. Finally, data were obtained from a short pushrod in a horizontal geometry to data obtained from a longer pushrod in a vertical geometry, the configuration likely to be used for in-situ measurements. Results of testing show that previously accepted data for the structural steels tested, SA533B1 and SS304, are inaccurate at high temperatures (above 500 oC) due to extrpolation of high temperature data. This is especially true for SA533B1, as previous data do not account for the phase transformation of the material between 730 oC and 830 oC. Also, comparison of results for horizontal and vertical configurations show a maximum percent difference of 2.02% for high temperature data.

  1. Acoustic Emission Signal Processing Technique to Characterize Reactor In-Pile Phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Vivek Agarwal; Magdy Samy Tawfik; James A Smith

    2014-07-01

    Existing and developing advanced sensor technologies and instrumentation will allow non-intrusive in-pile measurement of temperature, extension, and fission gases when coupled with advanced signal processing algorithms. The transmitted measured sensor signals from inside to the outside of containment structure are corrupted by noise and are attenuated, thereby reducing the signal strength and signal-to-noise ratio. Identification and extraction of actual signal (representative of an in-pile phenomenon) is a challenging and complicated process. In this paper, empirical mode decomposition technique is proposed to reconstruct actual sensor signal by partially combining intrinsic mode functions. Reconstructed signal corresponds to phenomena and/or failure modes occurring inside the reactor. In addition, it allows accurate non-intrusive monitoring and trending of in-pile phenomena.

  2. In-Pile Loop Safety in Integrated with the Multipurpose Reactor in the case of in-Pile Loop Leakage at the Core Position

    International Nuclear Information System (INIS)

    Suharno; Sugianto; Giarno; Aliq; Widodo, Surip; Aji, Bintoro; Purba, Julwan Hendry; Karyanta, Edy

    1999-01-01

    In-Pile Loop Safety analysis in integrated with the multipurpose reactor in the case of In-Pile Loop leakage at the core position has been conducted which intended to evaluate the failure of fuel element. By considering design of In-Pile Loop and the highest possibility position of of leakage, the failure of fuel element is emphasized on mechanical aspect. The thermal hydraulic aspect is not taken into account due to the condition that when the leakage occurred the reactor has been in shut down condition. It is determined that the spray attacks the top position of fuel element, and to be calculated the force, of spray that produces 1 cm deflection on the single fuel element. Using that four (4) fuel elements is calculated because in the real condition 4 fuel elements will undergo deflection of 43.8 kg is obtained that producing 1 cm deflection and the force of 1228 kg that causes failure on the bottom of fuel element as shear force is also obtained. Whatever the force, high or low, the damage of fuel element occurred at the bottom part or at the position of grid plate. Therefore there is no damage on the fuel part (uranium meat) and the releasing of radioactive material from fuel plate is not happened

  3. In-Pile Qualification of the Fast-Neutron-Detection-System

    Science.gov (United States)

    Fourmentel, D.; Villard, J.-F.; Destouches, C.; Geslot, B.; Vermeeren, L.; Schyns, M.

    2018-01-01

    In order to improve measurement techniques for neutron flux assessment, a unique system for online measurement of fast neutron flux has been developed and recently qualified in-pile by the French Alternative Energies and Atomic Energy Commission (CEA) in cooperation with the Belgian Nuclear Research Centre (SCK•ECEN). The Fast-Neutron-Detection-System (FNDS) has been designed to monitor accurately high-energy neutrons flux (E > 1 MeV) in typical Material Testing Reactor conditions, where overall neutron flux level can be as high as 1015 n.cm-2.s-1 and is generally dominated by thermal neutrons. Moreover, the neutron flux is coupled with a high gamma flux of typically a few 1015 γ.cm-2.s-1, which can be highly disturbing for the online measurement of neutron fluxes. The patented FNDS system is based on two detectors, including a miniature fission chamber with a special fissile material presenting an energy threshold near 1 MeV, which can be 242Pu for MTR conditions. Fission chambers are operated in Campbelling mode for an efficient gamma rejection. FNDS also includes a specific software that processes measurements to compensate online the fissile material depletion and to adjust the sensitivity of the detectors, in order to produce a precise evaluation of both thermal and fast neutron flux even after long term irradiation. FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCK•CEN in Belgium. Then a second test was recently completed at ISIS reactor operated by CEA in France. FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%.

  4. 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

  5. Microshear Bond Strength of OptiBond All-in-One Self-adhesive Agent to Er:YAG Laser Treated Enamel After Thermocycling and Water Storage.

    Science.gov (United States)

    Kasraei, Shahin; Yarmohammadi, Ebrahim; Ghazizadeh, Mohammad Vahid

    2016-01-01

    Introduction: This study aimed to compare the microshear bond strength of composite to enamel treated with Erbium-Doped Yttrium Aluminum Garnet (Er:YAG) laser using a self-etch one step bonding agent. Methods: Seventy-six enamel surfaces were prepared from 38 sound human third molar teeth. Specimens were randomly divided into four groups of 18. The enamel surface in half the specimens was irradiated with Er:YAG laser. One extra specimen from each group was evaluated under a scanning electron microscope (SEM). Composite micro-cylinders were bonded to the specimen surfaces using OptiBond All-In-One (OB) adhesive agent and stored in distilled water for 24 hours. Half the specimens were thermocycled (2000 cycles) and stored in distilled water at 37°C for three months (TW). The microshear bond strength of composite to enamel was measured using a universal testing machine at a crosshead speed of 1 mm/min. The fractured surfaces were evaluated under a stereomicroscope at ×40 magnification to determine the mode of failure. Data were analyzed using repeated measures analysis of variance (ANOVA) and t test. Results: The mean values (±standard deviation) were 17.96 ± 2.92 MPa in OB group, 22.29 ± 4.25 MPa in laser + OB group, 18.11 ± 3.52 MPa in laser + OB + TW group and 9.42 ± 2.47 MPa in OB + TW group. Repeated measures ANOVA showed that laser irradiation increased the microshear bond strength ( P Enamel surface preparation with Er:YAG laser is recommended to enhance the durability of the bond of self-etch bonding systems to enamel.

  6. New trends in pile safety instrumentation; Les tendances nouvelles dans l'instrumentation de securite des piles

    Energy Technology Data Exchange (ETDEWEB)

    Furet, J.

    1961-04-19

    This report addresses the protection of nuclear piles against damages due to operation incidents. The author discusses the current trends in the philosophy of safety of atomic power piles, identifies the parameters which define safety systems, presents tests to be performed on safety chains, comments the relationship between safety and the decrease of the number of pile inadvertent shutdowns, discusses the issues of instrument failures and chain multiplicity, comments the possible improvement of the operation of elements which build up safety chains (design simplification, development of semiconductors, replacement of electromechanical relays by static relays), the role of safety logical computers and the development of automatics in pile safety, presents automatic control as a safety factor (example of automatic start-up), and finally comments the use of fuses.

  7. 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

  8. In-Pile Section(IPS) Inner Assembly Manufacturing Report

    International Nuclear Information System (INIS)

    Lee, Jong Min; Shim, Bong Sik; Lee, Chung Yong

    2009-12-01

    The objective of this report is to present the manufacturing, assembling and testing process of IPS Inner Assembly used in Fuel Test Loop(FTL) pre-operation test. The majority of the manufactured components are test fuels, inner assembly structures and subsidiary tools that is needed during the assembly process. In addition, Mock-up test for the welding and brazing is included at this stage. Lower structure, such as test fuels, fuel carrier legs are assembled and following structures, such as fuel carrier stem in the middle structure, top flange in the top structure are assembled together each other. To Verify the Reactor Coolant Pressure Boundary(RCPB) function in IPS Inner Assembly helium leak test and hydraulic test is performed with its acceptance criteria. According to the ASME III code Authorized Nuclear Inspector(ANI) is required during the hydraulic test. As-built measurement and insulation resistance test are performed to the structures and instrumentations after the test process. All requirements are satisfied and the IPS Inner Assembly was loaded in HANARO IR-1 hole in September 25, 2009

  9. Studies of Nuclear Fuel Performance Using On-site Gamma-ray Spectroscopy and In-pile Measurements

    International Nuclear Information System (INIS)

    Matsson, Ingvar

    2006-01-01

    Presently there is a clear trend of increasing demands on in-pile performance of nuclear fuel. Higher target burnups, part length rods and various fuel additives are some examples of this trend. Together with an increasing demand from the public for even safer nuclear power utilisation, this implies an increased focus on various experimental, preferably non-destructive, methods to characterise the fuel. This thesis focuses on the development and experimental evaluation of such methods. In its first part, the thesis presents a method based on gamma-ray spectroscopy with germanium detectors that have been used at various power reactors in Europe. The aim with these measurements is to provide information about the thermal power distribution within fuel assemblies in order to validate core physics production codes. The early closure of the Barsebaeck 1 BWR offered a unique opportunity to perform such validations before complete depletion of burnable absorbers in Gd-rods had taken place. To facilitate the measurements, a completely submersible measuring system, LOKET, was developed allowing for convenient in-pool measurements to be performed. In its second part, the thesis describes methods that utilise in-pile measurements. These methods have been used in the Halden test-reactor for determination of fission gas release, pellet-cladding interaction studies and fuel development studies. Apart from the power measurements, the LOKET device has been used for fission gas release (FGR) measurements on single fuel rods. The significant reduction in fission gas release in the modern fuel designs, in comparison with older designs, has been demonstrated in a series of experiments. A FGR database covering a wide range of burnup, power histories and fuel designs has been compiled and used for fuel performance analysis. The fission gas release has been measured on fuel rods with average burnups well above 60 MWd/kgU. The comparison between core physics calculations (PHOENIX-4/POLCA

  10. Post-Irradiation Examination and In-Pile Measurement Techniques for Water Reactor Fuels

    International Nuclear Information System (INIS)

    2009-12-01

    in the 1960s when the construction of NPPs was being started. Evidently it can be assumed that infrastructure with basic unique equipments is old enough, both morally and physically, and needs to be up-graded or replaced. Thus, a sharp increase of the hydrocarbon fuel cost, green-house effect, necessity to construct more safe and efficient NPPs, justification of the lifetime prolongation of the existing NPPs, moral and physical ageing of the hot labs and research reactors equipment lead to the strong necessity to develop more perfect and more precise methods and equipment to examine irradiated components of nuclear reactors, first of all the most expensive one - nuclear fuel. Now the national hot laboratories and material testing reactors usually act as individual independent research establishments without any common and coordinated technical and business strategy towards the future needs and challenges. Even if there are not many joint programs for the development of nuclear power engineering in different countries, the method base and accumulated experience of the in- and post-reactor experiments should be widely shared so as to decrease the cost of this base in each country and to enforce its development. Thus, both problems and results of the application of new techniques to examine nuclear reactor components, as well as the conditions of separate labs should be discussed at the international level. The IAEA technical meetings are one of the most convenient means of arranging such discussion on the problems of the hot labs and research reactors development and application of new original techniques for examination of reactor materials properties. This publication represents a summary and proceedings of the two technical meetings (TMs) organized by IAEA on the subjects of Hot Cell Post-Irradiation Examination (PIE) Techniques and Pool Side Inspection of Water Reactor Fuel Assemblies and Fuel Rod Instrumentation and In-Pile Measurement Techniques. The first TM was

  11. Evaluation of Candidate In-Pile Thermal Conductivity Techniques

    International Nuclear Information System (INIS)

    Fox, B.; Ban, H.; Daw, J.; Condie, K.; Knudson, D.; Rempe, J.

    2009-01-01

    Thermophysical properties of materials must be known for proper design, test, and application of new fuels and structural properties in nuclear reactors. In the case of nuclear fuels during irradiation, the physical structure and chemical composition change as a function of time and position within the rod. Typically, thermal conductivity changes, as well as other thermophysical properties being evaluated during irradiation in a materials and test reactor, are measured out-of-pile in 'hot-cells'. Repeatedly removing samples from a test reactor to make out-of-pile measurements is expensive, has the potential to disturb phenomena of interest, and only provide understanding of the sample's end state at the time each measurement is made. There are also limited thermophysical property data for advanced fuels. Such data are needed for the development of next generation reactors and advanced fuels for existing nuclear plants. Having the capacity to effectively and quickly characterize fuels and material properties during irradiation has the potential to improve the fidelity of nuclear fuel data and reduce irradiation testing costs

  12. In-pile vapor pressure measurements on UO2 and (U,Pu)O2

    International Nuclear Information System (INIS)

    Breitung, W.; Reil, K.O.

    1985-08-01

    The Effective-Equation-of-State (EEOS) experiments investigated the saturation vapor pressures of ultra pure UO 2 , reactor grade UO 2 , and reactor grade (Usub(.77)Pusub(.23))O2 using newly developed in-pile heating techniques. For enthalpies between 2150 and 3700 kJ/kg (about 4700 to 8500 K) vapor pressures from 1.3 to 54 MPa were measured. The p-h curves of all three fuel types were identical within the experimental uncertainties. An assessment of all published p-h measurements showed that the p-h saturation curve of UO 2 appears now well established by the EEOS and the CEA in-pile data. Using an estimate for the heat capacity of liquid UO 2 , the in-pile results were also compared to earlier p-T measurements. The assessments lead to proposal of two equations. Equation I, which includes a factor-of-2 uncertainty band, covers all p-T equilibrium evaporation measurements. Equation I yields 3817 K for the normal boiling point, 415.4 kJ/mol for the corresponding heat of vaporization, and 1.90 MPa for the vapor pressure at 5000 K. Equations I and II, which represent a parametric form of the p-h curve (T=parameter), also give a good description of the EEOS and CEA in-pile data. Thus the proposed equations allow a consistent representation of both p-T and p-h measurements, they are sufficiently precise for CDA analyses and cover the whole range of interest (3120-8500 K, 1400-3700 kJ/kg). (orig./HP) [de

  13. Aluminum cladding oxidation of prefilmed in-pile fueled experiments

    Energy Technology Data Exchange (ETDEWEB)

    Marcum, W.R., E-mail: marcumw@engr.orst.edu [Oregon State University, School of Nuclear Science and Engineering, 116 Radiation Center, Corvallis, OR 97331 (United States); Wachs, D.M.; Robinson, A.B.; Lillo, M.A. [Idaho National Laboratory, Nuclear Fuels & Materials Department, 2525 Fremont Ave., Idaho Falls, ID 83415 (United States)

    2016-04-01

    A series of fueled irradiation experiments were recently completed within the Advanced Test Reactor Full size plate In center flux trap Position (AFIP) and Gas Test Loop (GTL) campaigns. The conduct of the AFIP experiments supports ongoing efforts within the global threat reduction initiative (GTRI) to qualify a new ultra-high loading density low enriched uranium-molybdenum fuel. This study details the characterization of oxide growth on the fueled AFIP experiments and cross-correlates the empirically measured oxide thickness values to existing oxide growth correlations and convective heat transfer correlations that have traditionally been utilized for such an application. This study adds new and valuable empirical data to the scientific community with respect to oxide growth measurements of highly irradiated experiments, of which there is presently very limited data. Additionally, the predicted oxide thickness values are reconstructed to produce an oxide thickness distribution across the length of each fueled experiment (a new application and presentation of information that has not previously been obtainable in open literature); the predicted distributions are compared against experimental data and in general agree well with the exception of select outliers. - Highlights: • New experimental data is presented on oxide layer thickness of irradiated aluminum fuel. • Five oxide growth correlations and four convective heat transfer correlations are used to compute the oxide layer thickness. • The oxide layer thickness distribution is predicted via correlation for each respective experiment. • The measured experiment and predicted distributions correlate well, with few outliers.

  14. An In Vitro Evaluation of Leakage of Two Etch and Rinse and Two Self-Etch Adhesives after Thermocycling

    Science.gov (United States)

    Geerts, Sabine; Bolette, Amandine; Seidel, Laurence; Guéders, Audrey

    2012-01-01

    Our experiment evaluated the microleakage in resin composite restorations bonded to dental tissues with different adhesive systems. 40 class V cavities were prepared on the facial and lingual surfaces of each tooth with coronal margins in enamel and apical margins in cementum (root dentin). The teeth were restored with Z100 resin composite bonded with different adhesive systems: Scotchbond Multipurpose (SBMP), a 3-step Etch and Rinse adhesive, Adper Scotchbond 1 XT (SB1), a 2-step Etch and Rinse adhesive, AdheSE One (ADSE-1), a 1-step Self-Etch adhesive, and AdheSE (ADSE), a 2-step Self-Etch adhesive. Teeth were thermocycled and immersed in 50% silver nitrate solution. When both interfaces were considered, SBMP has exhibited significantly less microleakage than other adhesive systems (resp., for SB1, ADSE-1 and ADSE, P = 0.0007, P adhesives, microleakage was found greater at enamel than at dentin interfaces (for ADSE, P = 0.024 and for ADSE-1, P adhesive systems, there was no significant difference between enamel and dentin interfaces; (3) SBMP was found significantly better than other adhesives both at enamel and dentin interfaces. In our experiment Etch and Rinse adhesives remain better than Self-Etch adhesives at enamel interface. In addition, there was no statistical difference between 1-step (ADSE-1) and 2-step (ADSE) Self-Etch adhesives. PMID:22675358

  15. Synergistic Smart Fuel For In-pile Nuclear Reactor Measurements

    Energy Technology Data Exchange (ETDEWEB)

    James A. Smith; Dale K. Kotter; Randall A. Ali; Steven L . Garrett

    2013-10-01

    In March 2011, an earthquake of magnitude 9.0 on the Richter scale struck Japan with its epicenter on the northeast coast, near the Tohoku region. In addition to the immense physical destruction and casualties across the country, several nuclear power plants (NPP) were affected. It was the Fukushima Daiichi NPP that experienced the most severe and irreversible damage. The earthquake brought the reactors at Fukushima to an automatic shutdown and because the power transmission lines were damaged, emergency diesel generators (EDGs) were activated to ensure that there was continued cooling of the reactors and spent fuel pools. The situation was being successfully managed until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to the reactors.2 At this point, the situation became critical. There was a loss of the sensors and instrumentation within the reactor that could have provided valuable information to guide the operators to make informed decisions and avoid the unfortunate events that followed. In the light of these events, we have developed and tested a potential self-powered thermoacoustic system, which will have the ability to serve as a temperature sensor and can transmit data independently of electronic networks. Such a device is synergistic with the harsh environment of the nuclear reactor as it utilizes the heat from the nuclear fuel to provide the input power.

  16. 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.

  17. 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.

  18. Thermocyclic behaviour of microstructurally modified EB-PVD thermal barrier coatings

    International Nuclear Information System (INIS)

    Schulz, U.; Fritscher, K.; Raetzer-Scheibe, H.-J.; Kaysser, W.A.; Peters, M.

    1997-01-01

    This paper focuses on the combined effects of substrate temperature and rotation during electron-beam physical vapor deposition (EB-PVD) on the columnar microstructure of yttria partially stabilized zirconia (YPSZ) thermal barrier coatings. Diameter and degree of ordering of the columns and the density of the coatings are sensitive to the processing parameters. Results are discussed in the frame of common structural zone models for PVD processes. The models are extended to consider the rotational effect. EB-PVD YPSZ TBCs of different column diameters were deposited on top of an EB-PVD NiCoCrAlY bondcoat on IN 100 superalloy test bars. The performance of the TBCs was investigated in a cyclic oxidation furnace test rig between 1100 C and 130 C and in a burner rig under hot gas corrosion conditions at a maximum temperature of 900 C. Results showed a correlation between cyclic lifetime and the various microstructures of the TBCs. Samples having a non-regular arrangement of columns performed best in both tests. (orig.)

  19. Theoretical interpretation of SCARABEE single pin in-pile boiling experiments

    International Nuclear Information System (INIS)

    Struwe, D.; Bottoni, M.; Fries, W.; Elbel, H.; Angerer, G.

    1977-01-01

    In the framework of LMFBR safety analysis a theoretical interpretation of some of the most representative of the single pin experiments of the in-pile SCARABEE project has been performed from both viewpoints of thermohydraulic and fuel behaviour using the computer codes CAPRI-2 and SATURN-1. The analysis is aimed at investigating the pin behavior from the preirradiation history, through the observed sequence of events following a coolant mass flow reduction from boiling inception up to pin breakdown. A comparison of theoretical results with experimentally recorded data has allowed a deeper insight into the peculiar features of the experiments and enabled a valuable code verification. (Auth.)

  20. In-pile intragranular densification of oxide fuels (AWBA Development Program)

    International Nuclear Information System (INIS)

    Dollins, C.C.; Nichols, F.A.

    1977-10-01

    This report proposes a model to describe in-pile densification of oxide fuels, by both vacancy boil-off due to thermal excitation and vacancy knockout by the passage of fission fragments through the pores. The model includes the migration rates of both vacancies and interstitials to pores and the production of vacancy-rich damage cascades by fission fragments. It has been coupled with a previously reported swelling and gas release model so that it can predict the total dimensional changes of the fuel as well as predicting intragranular densification for both ThO 2 and UO 2 fuels for advanced water breeder reactor applications development effort

  1. Analytical out-of-pile and in-pile experiments on gadolinia bearing fuels

    International Nuclear Information System (INIS)

    Bruet, M.; Francois, B.; Do, Q.; Bergeron, J.; Trotabas, M.

    1986-06-01

    New fuel management schemes in PWRs can be achieved through the use of burnable poisons like gadolinia bearing fuel rods. However, the introduction of such a design has required a qualification program, which has been performed in collaboration between CEA, FRAGEMA and/or FRAMATOME by specialized teams in CEA facilities. The main scoops of this program concern: the fabrication process; the out of pile physical properties determination: the in pile thermomechanical behaviour and fission product release; the neutronic studies in view to validate the Computed Gd efficiency and the LBP depletion calculation schemes and to analyse and assess various schemes of core calculations

  2. In pile measurement of creep rate of stainless steel cladding tubes for fast reactor pins

    International Nuclear Information System (INIS)

    Calza Bini, A.; Cosoli, G.; Filacchioni, G.; Lanchi, M.; Nobili, A.; Pesce, E.; Rocca, U.V.; Rotoloni, P.L.

    1975-01-01

    Results are reported of a direct in pile measurement of creep on a cladding sample of 10cm length, under tensile stress of 22.82kg/mm 2 at a temperature of 550 0 during about 500 hours, up to an integrated flux of 2.6.10 20 n/cm 2 . Two identical samples were irradiated in the same temperature and flux conditions to be submitted to out of pile creep measurements together with other unirradiated samples. The aim of this first experiment was mainly to set up the device and to evaluate the kind and the quality of the available data

  3. Safety analysis report for packaging: the ORNL in-pile capsule shipping cask

    International Nuclear Information System (INIS)

    Evans, J.H.; Chipley, K.K.; Haynie, C.B.; Crowley, W.K.; Just, R.A.

    1977-11-01

    The ORNL in-pile capsule shipping cask is used to transport irradiated experimental capsules and spent fuel elements. The cask was analytically evaluated to determine its compliance with the applicable regulations governing containers in which radioactive materials are transported. Computational procedures were used to determine the structural integrity and thermal behavior of the cask relative to the general standards for normal conditions of transport and the standards for the hypothetical accident conditions. The results of the evaluation show that the cask is in compliance with the applicable regulations

  4. In-pile experimental device for Sirene thermionic converters; Dispositif d'experimentation en pile des convertisseurs thermoioniques sirene

    Energy Technology Data Exchange (ETDEWEB)

    Bliaux, J; Durand, J; Lazare-Chopard, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1969-07-01

    The irradiation device described here, was built for in pile life tests of 100 We SIRENE converters. The nuclear converter is located in a sealed vacuum chamber, which is plugged at the lower end of a coaxial tubing acting as electrical leads. The output power is available on a variable resistive load on the bank of the reactor pool. Thermal, electrical and neutronic parameters of the converter are recorded. Since 1967, two permanent devices allowed five experiments in the swimming pool TRITON (CEN-FAR) and the results, obtained till now, are presented. (authors) [French] Le dispositif d'irradiation SIRENE decrit ici a ete concu en vue d'une etude statistique de performances de convertisseurs thermoioniques nucleaires de puissance unitaire 100 We. Le dispositif doit assurer la bonne marche du convertisseur en pile, permettre le changement de la position verticale du convertisseur dans le coeur, sortir du coeur la puissance electrique convertie sans degradation notable et enregistrer les differents parametres thermiques, electriques et neutroniques du convertisseur. Depuis 1967, deux dispositifs fonctionnent en permanence et ont permis de faire cinq experiences dans le reacteur piscine TRITON du CEN-FAR. Les resultats obtenus jusqu'a present, sont presentes. (auteurs)

  5. Results of postirradiation examination of the in-pile blockage experiments MOL-7C/4 and MOL-7C/5

    International Nuclear Information System (INIS)

    Weimar, P.; Schleisiek, K.

    1991-01-01

    The Mol-7C in-pile local blockage experiments are performed in the BR-2 reactor at Mol, Belgium as a joint project of Kernforchungszentrum Karlsruhe (KfK) and Studiecentrum voor Kernenergie/Centre d'Etude de l'Energie Nuclearire-Mol. The main objective is to investigate the consequences of local cooling disturbances in liquid-metal-cooled reactor (LMR) fuel subassemblies. In the tests Mol-7C/4 and MOL-7C/5, fuel pins from KNK II are used with a burnup of 5 and 1.7%, respectively. An active central porous blockage is used to simulate the cooling disturbance. During irradiation, the blockage causes significant local damage, including melting of cladding and fuel. Extensive postirradiation examinations (PIE) are performed to investigate the extent of damage. In this paper a description and interpretation of results of the destructive PIE performed at the Hot Cells Laboratory at KfK is given, along with some conclusions related to LMR safety

  6. Research on in-pile release of fission products from coated particle fuels

    International Nuclear Information System (INIS)

    Fukuda, K.; Iwamoto, K.

    1985-01-01

    Coated particle fuels fabricated in accordance with VHTR (Very High Temperature gas-cooled Reactor) fuel design have been irradiated by both capsules and an in-pile gas loop (OGL-1), and data on the fission products release under irradiation were obtained for loose coated particles, fuel compacts and fuel rods in the temperature range between 800 deg. C and 1600 deg. C. For the fission gases, temperature- and time dependences of the fractional release(R/B) were measured. Relation between release and failure fraction of the coated particles was elucidated on the VHTR reference fuels. Also measured was tritium concentration in the helium coolant of OGL-1. In-pile release behavior of the metallic fission products was studied by measuring the activities of the fission products adsorbed in the graphite sleeves of the OGL-1 fuel rods and the graphite fuel container of the sweep gas capsules in the PIE. Investigation on palladium interaction with SiC coating layer was included. (author)

  7. Beryllium data base for in-pile mockup test on blanket of fusion reactor, (1)

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Hiroshi; Ishitsuka, Etsuo (Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment); Sakamoto, Naoki; Kato, Masakazu; Takatsu, Hideyuki.

    1992-11-01

    Beryllium has been used in the fusion blanket designs with ceramic breeder as a neutron multiplier to increase the net tritium breeding ratio (TBR). The properties of beryllium, that is physical properties, chemical properties, thermal properties, mechanical properties, nuclear properties, radiation effects, etc. are necessary for the fusion blanket design. However, the properties of beryllium have not been arranged for the fusion blanket design. Therefore, it is indispensable to check and examine the material data of beryllium reported previously. This paper is the first one of the series of papers on beryllium data base, which summarizes the reported material data of beryllium. (author).

  8. Design computations and safety report of a cell for in pile irradiation tests

    International Nuclear Information System (INIS)

    Verri, A.

    1987-01-01

    The criteria adopted in positioning the irradiation cell within the 1Mw TRIGA reactor of the ENEA Casaccia are reported. Maximum heat which can be released by the cell is then evaluated. The final configuration of the cell as a whole, the heating system for the sample under irradiation, the procedure used in the calculations, are also reported. The selection and the design of the safety system, including auxiliary equipments are discussed

  9. In-pile test of tritium release from tritium breeding materials (VOM-21H experiment)

    International Nuclear Information System (INIS)

    Kurasawa, Toshimasa; Takeshita, Hidefumi; Watanabe, Hitoshi; Yoshida, Hiroshi.

    1986-10-01

    Material development and blanket design of lithium-based ceramics such as lithium oxide, lithium aluminate, lithium silicate and lithium zirconate have been performed in Japan, United State of America and Europian Communities. Lithium oxide is a most attractive candidate for tritium breeding materials because of its high lithium density, high thermal conductivity and good tritium release performance. This work has been done to clarify the characteristics of tritium release and recovery from Li 2 O by means of in-situ tritium release measurement. The effects of temperature and sweep gas composition on the tritium release were investigated in this VOM-21H Experiment. Good measurement of tritium release was achieved but there were uncertainties in reproduciblity of data. The experimental results show that the role of surface adsorption/desorption makes a significant contribution to the tritium release and tritium inventory. Also, it is necessary to define the rate limiting process either diffusion or surface adsorption/desorption. (author)

  10. Current in-pile absorbed dose measurements at the Boris Kidric Institute of nuclear sciences - Vinca, Status report

    Energy Technology Data Exchange (ETDEWEB)

    Draganic, G I [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1966-11-15

    So far in-pile absorbed dose measurements have been limited only to experiments in the RA reactor at the Boris Kidric Institute of Nuclear Sciences at Vinca (6.5 D{sub 2}O moderated and 2% enriched uranium). The methods used for absorbed dose and neutron flux measurements were 1,2 discussed in some earlier reports at the IAEA meetings. The purpose of the present report is to illustrate the further development of methods of determining in-pile absorbed doses (author)

  11. In-pile study of the reaction between breeder fuel and sodium

    International Nuclear Information System (INIS)

    Hugot, J.P.

    1982-10-01

    Studies carried out until now show that the determinant parameter of fuel can failure evolution is the development of the reaction between mixed uranium and plutonium dioxide and sodium. The parameters of the reaction are presented from results of an out of pile study, as also results obtained from examination on pins failed in reactors. The best way to study in pile the development of the reaction was to irradiate at a constant power a fuel pin containing sodium. In the experiment, the pin was equipped with a central thermocouple. It shows, that the reaction is developing intergranularly, from cracks and interpellet spaces, in an internal fringe of the fuel before spreading to the periphery. An overheating of the pin is associated to the development of the reaction as also a modification of the fuel pin geometry and a reduction of the oxide [fr

  12. ACTIV, Sandwich Detector Activity from In-Pile Slowing-Down Spectra Experiment

    International Nuclear Information System (INIS)

    Bozzi, L. and others

    1978-01-01

    1 - Nature of physical problem solved: Calculates the activities of a sandwich detector, to be used for in-pile measurements in slowing-down spectra below a few keV. The effect of scattering with energy degradation in the filter and in the detectors has been included to a first approximation. 2 - Method of solution: An iterative procedure is used: the calculation starts with a flux guess in which one assumes that each measured reactivity difference depends on the principal resonance only. The secondary resonance contribution is computed through the iterative process. For self-shielded cross-section calculations the model of Pearlstein and Weinstock (ref. 3) is used. The neutron spectrum can optionally be constant or 1/E inside each finite energy group relative to the resonance considered. 3 - Restrictions on the complexity of the problem: Maximum number of energy groups : 60

  13. State-of-the-art review of some artificial intelligence applications in pile foundations

    Directory of Open Access Journals (Sweden)

    Mohamed A. Shahin

    2016-01-01

    Full Text Available Geotechnical engineering deals with materials (e.g. soil and rock that, by their very nature, exhibit varied and uncertain behavior due to the imprecise physical processes associated with the formation of these materials. Modeling the behavior of such materials in geotechnical engineering applications is complex and sometimes beyond the ability of most traditional forms of physically-based engineering methods. Artificial intelligence (AI is becoming more popular and particularly amenable to modeling the complex behavior of most geotechnical engineering applications because it has demonstrated superior predictive ability compared to traditional methods. This paper provides state-of-the-art review of some selected AI techniques and their applications in pile foundations, and presents the salient features associated with the modeling development of these AI techniques. The paper also discusses the strength and limitations of the selected AI techniques compared to other available modeling approaches.

  14. Investigation of in-pile grown corrosion films on zirconium-based alloys

    International Nuclear Information System (INIS)

    Gebhardt, O.; Hermann, A.; Bart, G.; Blank, H.; Ray, I.L.F.

    1996-01-01

    In-pile grown corrosion films on different fuel rod claddings (standard Zircaloy-4, extra low tin Zircaloy (ELS), and Zr2.5Nb) have been studied using a variety of experimental techniques. The aim of the investigations was to find out common features and differences between the corrosion layers grown on zirconium alloys having different composition. Methods applied were scanning and transmission electron microscopy (SEM, TEM), electrochemical impedance spectroscopy (EIS), and electrochemical anodization. The morphological differences have been observed between the specimens that could explain the irradiation enhancement of corrosion of Zircaloy-4. The features of the compact oxide close to the oxide/metal interface have been characterized by electrochemical methods. The relationship between the thickness of this protective oxide and the overall oxide thickness has been investigated by EIS. It was found that this relation is dependent on the location of the oxide along the fuel rod and on the corrosion rate

  15. State-of-the-art review of some artificial intelligence applications in pile foundations

    Institute of Scientific and Technical Information of China (English)

    Mohamed A. Shahin

    2016-01-01

    Geotechnical engineering deals with materials (e.g. soil and rock) that, by their very nature, exhibit varied and uncertain behavior due to the imprecise physical processes associated with the formation of these materials. Modeling the behavior of such materials in geotechnical engineering applications is complex and sometimes beyond the ability of most traditional forms of physically-based engineering methods. Artificial intelligence (AI) is becoming more popular and particularly amenable to modeling the complex behavior of most geotechnical engineering applications because it has demonstrated superior predictive ability compared to traditional methods. This paper provides state-of-the-art review of some selected AI techniques and their applications in pile foundations, and presents the salient features associated with the modeling development of these AI techniques. The paper also discusses the strength and limitations of the selected AI techniques compared to other available modeling approaches.

  16. MCNP simulation to optimise in-pile and shielding parts of the Portuguese SANS instrument.

    Science.gov (United States)

    Gonçalves, I F; Salgado, J; Falcão, A; Margaça, F M A; Carvalho, F G

    2005-01-01

    A Small Angle Neutron Scattering instrument is being installed at one end of the tangential beam tube of the Portuguese Research Reactor. The instrument is fed using a neutron scatterer positioned in the middle of the beam tube. The scatterer consists of circulating H2O contained in a hollow disc of Al. The in-pile shielding components and the shielding installed around the neutron selector have been the object of an MCNP simulation study. The quantities calculated were the neutron and gamma-ray fluxes in different positions, the energy deposited in the material by the neutron and gamma-ray fields, the material activation resulting from the neutron field and radiation doses at the exit wall of the shutter and around the shielding. The MCNP results are presented and compared with results of an analytical approach and with experimental data collected after installation.

  17. Visual observations of fuel disruption in in-pile LMFBR accident experiments

    International Nuclear Information System (INIS)

    Wright, S.A.; Mast, P.K.

    1982-01-01

    Sandia National Laboratories has been investigating initiation phase phenomena in a series of Fuel Disruption (FD) experiments since 1977. In this program high speed cinematography is used to observe fuel disruption in in-pile experiments that simulate loss of flow accidents. Thus, these experiments provide high resolution measurements of initial fuel and clad motion with prototypic materials and prototypic heating conditions. The main objective of the FD experiment is to determine the timing (relative to fuel temperature) and the mode of fuel disruption under LOF heating conditions. Observed modes of disruption include fuel swelling, solid state breakup, cracking, ejection of a molten fuel jet, slumping, and rapid expansion of small particles. Because the temperature and character of the fuel at disruption are known, disruption can be correlated with the mechanisms driving the disruption such as fuel vapor pressure, molten fuel expansion, fission gases, and impurity gases

  18. Analyses with the FSTATE code: fuel performance in destructive in-pile experiments

    International Nuclear Information System (INIS)

    Bauer, T.H.; Meek, C.C.

    1982-01-01

    Thermal-mechanical analysis of a fuel pin is an essential part of the evaluation of fuel behavior during hypothetical accident transients. The FSTATE code has been developed to provide this required computational ability in situations lacking azimuthal symmetry about the fuel-pin axis by performing 2-dimensional thermal, mechanical, and fission gas release and redistribution computations for a wide range of possible transient conditions. In this paper recent code developments are described and application is made to in-pile experiments undertaken to study fast-reactor fuel under accident conditions. Three accident simulations, including a fast and slow ramp-rate overpower as well as a loss-of-cooling accident sequence, are used as representative examples, and the interpretation of STATE computations relative to experimental observations is made

  19. An In-Pile Kinetic Method for Determining the Delayed Neutron Fraction βeff

    International Nuclear Information System (INIS)

    Gilad, E.; Rivin, O.; Ettedgui, H.; Yaar, I.; Geslot, B.; Pepino, A.; Di Salvo, J.; Gruel, A.; Blaise, P.

    2014-01-01

    Delayed neutrons are of fundamental importance in the field of nuclear reactor dynamics and control. Although only a small fraction of the neutrons emitted by fission are not prompt, the knowledge of the delayed neutrons parameters is essential for transient analysis, such as startup or shutdown of the reactor, as well as for accidents analysis and control system design [1]. One of the main delayed neutron parameters used in the point reactor model equations is the effective delayed neutron fraction, which incorporates both delayed neutron spectral properties and core geometrical configuration [1,2]. Additional delayed neutron parameters include the fraction of fission neutrons emitted in each delayed group, and the delayed neutron precursors decay constants . Experimental efforts aimed at determining the value ofβ, which provide experimental support for the evaluation of delayed neutron parameters, are extremely valuable. This is due to the fact that unlike other fields in reactor physics, e.g. criticality safety or shielding, the availability of experimental data and benchmark problems for validating delayed neutron parameters and its implementation in different models is highly limited. Furthermore, the existing experimental data exhibit significant discrepancies between the different sets of parameter, which lead to substantial disparity in the analysis of kinetic experiments and reactor dynamic behavior]. In this work, a method for determining the effective delayed neutron fraction using in-pile reactivity oscillation and Fourier analysis is presented. The method is based on measurements of the reactor's power response to small periodic in-pile reactivity perturbations and utilizes Fourier analysis for reconstruction of the reactor zero power transfer function. Knowledge of the reactor transfer function enables the estimation of theβ value using multi-parameter nonlinear fit. The method accounts for higher harmonics, which are excited by the trapezoidal

  20. Evaluation of the in pile performance of boron containing fuel pellets

    International Nuclear Information System (INIS)

    Jeong, Gwanyoon; Sohn, Dongseong

    2012-01-01

    The world rare earth resource are heavily concentrated in certain area and if these natural resources are weaponized by a country, we may confront serious difficulty because rare earth element gadolinium(Gd) is used as burnable poison material in some nuclear power plants (NPP) in Korea. Gd is used as a neutron absorbing material in Gd 2 O 3 form and mixed with UO 2 When boron is used as burnable poison in nuclear fuel, in fuel pellets. The burnable poison mixed in the fuel pellets is called integral burnable absorber (BA) design which differentiates it from the old separate BA design. In the old separate BA design, boron(B) was used in borosilicate glass (PYREX) form and placed in guide tubes. With the development of the concern over the availability of rare earth material Gd, B is considered as a candidate material replacing Gd for the case when the rare earth material is weaponized. However the idea for new boron BA design is integral type because the integral type BA design has several benefits over the separate BA design, such as reduction of radioactive waste, more positions for BA location, etc. 10 B absorbs a neutron and produces helium by the following reaction: 10 B + n → 7 Li + 4 He The helium produced by the nuclear reaction may cause the increase of rod internal pressure and change the gap conductivity if the significant amount of helium gas is released to the gap between the pellet and the cladding. Thus, it is necessary to investigate the in-pile behaviors of B containing pellet. However, few experiment have been carried out so far on the behavior of in-pile produced helium in UO 2 fuel pellets, especially for the cases boron compound is mixed with UO 2 In this paper, we will evaluate the production and the release of helium depending on fuel. 10 B concentration in the fuel

  1. Thermal-hydraulic calculation and analysis on helium cooled ceramic breeder pebble bed assembly for in-pile irradiation and in-situ tritium extraction

    International Nuclear Information System (INIS)

    Guo Chunqiu; Xie Jiachun; Liu Xingmin

    2013-01-01

    In-pile irradiation and in-situ tritium extraction experiment is one of associated domestic research projects in ITER special program. According to the technical requirements of in-pile irradiation experiment of helium cooled ceramic breeder (ceramic) pebble bed assembly in a research reactor, the feasibility of the design for the in-pile irradiation and in-situ tritium extraction experiment of ceramic pebble bed assembly was evaluated. By conducting thermal-hydraulic design calculation with different in-pile irradiation channels, locations and structure parameters for ceramic pebble bed assembly, a reasonable design scheme of ceramic pebble bed assembly satisfying the design requirements for in-pile irradiation was obtained. (authors)

  2. MTR and PWR/PHWR in-pile loop safety in integration with the operation of multipurpose reactor - GAS

    International Nuclear Information System (INIS)

    Suharno; Aji, Bintoro; Sugiyanto; Rohman, Budi; Zarkasi, Amin S.; Giarno

    1998-01-01

    MTR and PWR/PHWR In-Pile Loop safety analysis in integration with the operation of Multipurpose Reactor - Gas has been carried out and completed. The assessment is emphasized on the function of the interface systems from the dependence of the operation and the evaluation to the possibility of leakage or failure of the in-pile part inside the reactor pool and reactor core. The analysis is refers to the logic function of the interface system and the possibility of leakage or failure of the in-pile part inside reactor pool and reactor core to consider the integrity of the core qualitatively. The results show that in normal and in transient conditions , the interface system meet the function requirement in safe integrated operation of in-pile loop and reactor. And the results of the possibility analysis of the leakage shows that the possibility based on mechanically assessment is very low and the impact to core integrity is nothing or can be eliminated. The possible position for leakage is on the flen on which one meter above the top level of the core, therefore no influence of leakage to the core

  3. Modification of Electrical Properties of Thin La0.67Ca0.33MnO3 Films by Pulsed Thermocycling

    Directory of Open Access Journals (Sweden)

    Fiodoras ANISIMOVAS

    2012-09-01

    Full Text Available Highly resistive states were formed in nonhomogeneous thin La0.67Ca0.33MnO3 films at 80 K temperature after resistance switching induced by the pulsed thermocycling. Heating up to room temperature does not destroy the resistive states. They demonstrate high values of electroresistance at applied pulsed electric field. It was registered formation of novel highly resistive state by resistance switching at 130 K in Tm region. We suppose that local temperature increase in the film is responsible for the formation of the resistive states in both cases and present a plausible explanation of the obtained results. The method of cyclic nanosecond temperature increase and decrease can be useful for modification of material properties having strongly correlated electron system and of ferroelectrics. Questions of practical realization of proposed method are discussed.DOI: http://dx.doi.org/10.5755/j01.ms.18.3.2427

  4. Numerical Signal Analysis of Thermo-Cyclically Operated MOG Gas Sensor Arrays for Early Identification of Emissions from Overloaded Electric Cables

    Directory of Open Access Journals (Sweden)

    Rolf Seifert

    2015-10-01

    Full Text Available A thermo-cyclically operated multi metal oxide gas sensor (MOG array is introduced together with a novel signal analysis approach (SimSens for identifying the emissions from overheated isolation cable materials thereby detecting the fires originated in electrical cabinets at early stages. The MOG array can yield specific conductance signatures appropriate to specifically identify gases. The obtained results bear good capability for detection and identification of pyrolysis gas emissions at relatively low sample heating temperatures even before a visible color-change of the polyvinyl chloride (PVC-isolation material. The dynamic conductance signals were evaluated using SimSens, a numerical analysis tool designed for simultaneous evaluation of conductance profiles. The results show promising pyrolysis gas identification and concentration determination capabilities in relation to the conductance profile shapes of model gases like carbon monoxide (CO and propene.

  5. In-Pile Instrumentation Multi- Parameter System Utilizing Photonic Fibers and Nanovision

    Energy Technology Data Exchange (ETDEWEB)

    Burgett, Eric [Idaho State Univ., Pocatello, ID (United States)

    2015-10-13

    An advanced in-pile multi-parameter reactor monitoring system is being proposed in this funding opportunity. The proposed effort brings cutting edge, high fidelity optical measurement systems into the reactor environment in an unprecedented fashion, including in-core, in-cladding and in-fuel pellet itself. Unlike instrumented leads, the proposed system provides a unique solution to a multi-parameter monitoring need in core while being minimally intrusive in the reactor core. Detector designs proposed herein can monitor fuel compression and expansion in both the radial and axial dimensions as well as monitor linear power profiles and fission rates during the operation of the reactor. In addition to pressure, stress, strain, compression, neutron flux, neutron spectra, and temperature can be observed inside the fuel bundle and fuel rod using the proposed system. The proposed research aims at developing radiation-hard, harsh-environment multi-parameter systems for insertion into the reactor environment. The proposed research holds the potential to drastically increase the fidelity and precision of in-core instrumentation with little or no impact in the neutron economy in the reactor environment while providing a measurement system capable of operation for entire operating cycles.

  6. In-Pile Instrumentation Multi- Parameter System Utilizing Photonic Fibers and Nanovision

    International Nuclear Information System (INIS)

    Burgett, Eric

    2015-01-01

    An advanced in-pile multi-parameter reactor monitoring system is being proposed in this funding opportunity. The proposed effort brings cutting edge, high fidelity optical measurement systems into the reactor environment in an unprecedented fashion, including in-core, in-cladding and in-fuel pellet itself. Unlike instrumented leads, the proposed system provides a unique solution to a multi-parameter monitoring need in core while being minimally intrusive in the reactor core. Detector designs proposed herein can monitor fuel compression and expansion in both the radial and axial dimensions as well as monitor linear power profiles and fission rates during the operation of the reactor. In addition to pressure, stress, strain, compression, neutron flux, neutron spectra, and temperature can be observed inside the fuel bundle and fuel rod using the proposed system. The proposed research aims at developing radiation-hard, harsh-environment multi-parameter systems for insertion into the reactor environment. The proposed research holds the potential to drastically increase the fidelity and precision of in-core instrumentation with little or no impact in the neutron economy in the reactor environment while providing a measurement system capable of operation for entire operating cycles.

  7. Pneumatic capsule with a measuring system for in-pile irradiation

    International Nuclear Information System (INIS)

    Oshima, Keiichi; Yamazaki, Yasaburo; Hirata, Mitsuho; Ishii, Toshio; Shimozawa, Ryohei.

    1967-01-01

    A prior-art in-pile irradiation apparatus wherein a rabbit containing an irradiation specimen therein is inserted into and removed from a pile by a pneumatic system does not include means for measuring the temperature and pressure of the specimen under irradiation. When the rabbit is deformed during irradiation, it cannot be reliably recovered. A pneumatic capsule assembly with a measuring system according to this invention has a double structure which consists of an inner capsule containing the specimen therein and an outer capsule evacuated or filled with a gas. A thermocouple lace wire and a strain gauge are welded on the outside surface of the inner capsule as detection terminals for measuring the temperature and pressure. A rupture plate which bursts when the pressure in the inner capsule reaches a predetermined value is provided at a part of the inner capsule, and a fin for heat transmission is provided between the inner and outer capsules to thus prevent the deformation of the pneumatic capsule assembly as a whole. (Takasuka, S.)

  8. Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

    Energy Technology Data Exchange (ETDEWEB)

    Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

    2014-09-15

    Highlights: • Gas as a coolant in Gen-IV reactors, history and development. • Main physical parameters comparison of gas coolants: carbon dioxide, helium, hydrogen with water. • Forced convection in turbulent pipe flow. • Gas cooled fast reactor concept comparisons to very high temperature reactor concept. • High temperature helium loop: concept, development, mechanism, design and constraints. - Abstract: Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuel supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a very high temperature reactor (VHTR). The VHTR concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behavior within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Center Rez Ltd. One of the topics analyzed in this article are also physical characteristic and benefits of gas

  9. Novel approach for assessing performance of PCR cyclers used for diagnostic testing

    DEFF Research Database (Denmark)

    Schoder, D.; Schmalwieser, A.; Schauberger, G.

    2005-01-01

    of thermal homogeneities became most evident at the denaturation level during the first 15 s. At the time point zero, the accurate cyclers showed temperature deviations of 0.5 to 1.5 degrees C, whereas less-accurate cyclers failed to reach the set temperature by 13 to 20 degrees C. Consequently, the two less......As part of a large international project for validation and standardization of PCR, the influence of thermocyclers on PCR was tested. Six brand-new, Peltier technology-driven 96-well thermocyclers were subjected to a novel and stringent in-tube (not block) physical testing. The temperature...

  10. Fabrication, characteristics, and in-pile performance of UO2 pellets prepared from dry route powder

    International Nuclear Information System (INIS)

    Chotard, A.; Ledac, A.; Bernardin, M.

    1991-01-01

    The dry route conversion process of UF 6 to sinterable UO 2 powder has been used in France on a large scale for more than 10 years for the fabrication of PWR fuels. Thus, our fabrication and irradiation experience relates to more than 10,000 tons of fuel. As everyone knows, the dry route conversion process only involves gas-gas and gas-solid reactions which present the advantage of producing very little contaminated wastes and no liquid effluents. Powders obtained by this process are characterized by: - a very high purity, - a low specific surface area (around 2 m 2 /g), therefore a high resistance to spontaneous oxidation, - a good compressibility, - a very high sinterability (.98% T.D.), - a very high reproducibility. This powder also shows a high fineness which leads to very homogeneous blends with additives like pore former, U 3 O 8 or Gd 2 O 3 . On the other hand this fineness requires a granulation step which is actually not a disadvantage since it allows to adjust the granulate size to optimize the filling of press dies and so as to guarantee a good stability of the pellet dimensions and density. This pelletizing process leads to pellets characterized by: - a good thermal stability (0.5% T.D. after 34 hours at 1700degC), - no open porosity, - low H 2 content (0,3 ppm), - an homogeneous microstructure (grain size and porosity). Such characteristics mean that the UO 2 pellets from dry route conversion present an excellent in pile behaviour for high burnup up to 58,000 MWd/MtU in commercial plant, with: - low fission gas release, - good dimensional stability (densification, swelling), of which examples and results of PIE are described in the paper. The qualities of the dry route conversion powder and its flexibility of use make it possible to consider adjustment of the pellet characteristics, mainly: density, grain size and pore size distribution for specific uses or performance upgrade. (orig.)

  11. Results of water chemistry control in the in-pile ''Callisto'' loop (an experimental PWR rig installed in the BR2 reactor)

    International Nuclear Information System (INIS)

    Weber, M.; Benoit, P.; Dekeyser, J.; Verwimp, A.

    1994-01-01

    Since June 1992, a new experimental facility, called CALLISTO, is being irradiated in the BR2 materials testing reactor at Mol, Belgium. The main objective of the present test campaign is to study the behaviour of advanced fuel to high burn-up rates in a realistic PWR environment. Three in-pile sections, containing each 9 fuel rods, are loaded inside the reactor vessel and are connected to a common out-of-pile pressurized water circulation loop (ref.1). The later is branched-off into a purification circuit (feed-bleed concept) and further equipped with safety and auxiliary systems. To cope with the test programme, the equipments are designed so that the guidelines of a PWR primary water chemistry can be followed (ref.2). Real steady-state conditions cannot be observed because the typical BR2 cycle (3 weeks running/3 weeks shut-down) is much shorter and because the rig is cooled down during each reactor shut-down. The purpose of this poster is to provide results of chemical parameters recorded during the cycling behaviour of the CALLISTO primary water. (authors). 4 figs., 1 tab., 2 refs

  12. In-pile behavior of controlled beta-quenched fuel channels

    Energy Technology Data Exchange (ETDEWEB)

    Moeckel, Andreas; Pflaum, Wolfgang; Cremer, Ingo [AREVA NP GmbH, Erlangen (Germany); Zbib, Ali A. [AREVA NP Inc., Richland, WA (United States)

    2011-07-01

    of a German boiling water reactor (BWR) nuclear power plant in 2004. Since then, additional beta-quenched lead fuel channels have been placed in the core of some other European BWRs to broaden the in-pile experience with such channels in different nuclear power plants. (orig.)

  13. Out-of-pile and in-pile temperature noise investigations: a survey of methods results and models

    International Nuclear Information System (INIS)

    Dentico, G.; Giovannini, R.; Marseguerra, M.; Pacilio, N.; Taglienti, S.; Tosi, V.; Vigo, A.; Oguma, R.

    1982-01-01

    A review is given of the main results obtained from temperature noise measurements performed in out-of-pile sodium loops on fast fuel element mock-ups. Sources of data were thermocouples placed in the central axis of the channel downstream from the bundle end. Autoregressive moving average (ARMA) models have been applied to several temperature time series; the analysis shows that a simple ARMA (3, 2) model adequately accounts for the observed fluctuations. Finally, highlights of a heat transfer stochastic model are also reported together with a preliminary validation against in-pile experimental data. (author)

  14. In-pile observations of fuel and clad relocation during LMFBR initiation phase accident experiments - the STAR experiments

    International Nuclear Information System (INIS)

    Wright, S.A.; Schumacher, G.; Henkel, P.R.; Royl, P.

    1987-01-01

    A series of seven in-pile experiments (the STAR experiments) were performed in which clad motion and fuel dispersal were observed in small pin bundles with high-speed cinematography. The experimental heating conditions reproduced a range of Loss of Flow (LOF) accident scenarios for the lead subassemblies in LMFBRs. The experiments show strong tendencies for limited clad motion in multiple pin bundles, early fuel disruption and dispersal (prior to fuel melting) in moderate power transients having simultaneous clad melting and fuel disruption. The more recent experiments indicate a possibility of steel vapor driven fuel dispersal after fuel breakup and intimate fuel/steel mixing. (author)

  15. Multi-channel mechanical test machine for HANARO (I)

    International Nuclear Information System (INIS)

    Song, M. S.; Choi, Y.; Cho, M. S.; Kim, B. G.; Kang, Y. H.

    2004-01-01

    Design and fabrication of multi-channel mechanical test machine is useful and important for the study of in-pile test of nuclear materials in HANARO. The dimension and shape of the multi-channel mechanical test machine should be fixed to a test reactor and their objectives. KAERI successfully developed a non-instrumented multi-channel mechanical test machine for material irradiation tests in a domestic research reactor, HANARO. This results in strongly stimulating and accelerating irradiation tests of materials in domestic industry and research fields with HANARO. Although various types of in-pile creep capsule were made for well installation in each test reactor, there is no in-pile creep multi-channel mechanical test machine for HANARO. Hence, the objectives of this study are to fabricate and test a multi-channel mechanical test machine of HANARO

  16. Status of the EXOTIC-8 programme and first in-pile results for Li{sub 2}TiO{sub 3} pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Van der Laan, J G; Stijkel, M P [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Conrad, R

    1998-03-01

    After renewal of the Tritium Measuring Station the HFR is again fully operational for in-pile breeder irradiations. The EXOTIC-8 series has started with first three experiments on June 12, 1997. First in-pile results have been obtained for Li{sub 2}TiO{sub 3}-pebbles supplied by CEA: preliminary analyses indicate satisfactory in-pile behaviour with fast recovery from transient conditions. Five further experiments have been defined which implies that in the present planning EXOTIC-8 is filled completely up to Fall`98 and 2 of 4 positions are occupied up to Spring`99. P.I.E. results will be obtained from Spring`98 onwards. (J.P.N.)

  17. Alpha radiation and in-pile annealing effects on the fracture properties of a sintered alumino borosilicate glass

    International Nuclear Information System (INIS)

    Bevilacqua, Arturo M.; Prado, Miguel O.; Messi de Bernasconi, Norma B.; Heredia, Arturo D.; Sanfilippo, Miguel

    1999-01-01

    The alpha radiation and the in-pile during irradiation effects on the hardness, the crack nucleation and the fracture toughness of the German alumino borosilicate glass SG7 were investigated by using the Vickers indentation. Cold pressed and sintered samples were irradiated with thermal neutrons, in the Argentine nuclear reactors RA-3 and RA-6, to produce alpha particles in the whole volume of the glass by means of the (n, alpha)-reaction with B-10. The Vickers hardness, the crack nucleation, as 50 percent fracture probability load, plotted as the Weibull's fracture probability distribution function and the fracture toughness, as critical stress intensity factor K Ic , were correlated to the four cumulative disintegration values. It was ascertained that: a) the Vickers hardness decreases from 5.6 GPa for the non-irradiated sample up to 4.7 GPa for the sample irradiated 70 h at the lower neutron flux (4.0 x 10 - sup 18 - alpha disintegration per cm - sup 3 -), b) the 50 % fracture probability load increases from 1.4 N for the non-irradiated sample up to 4.7 g for the sample irradiated 22 h at the higher flux (6.8 x 10 - sup 18 - alpha disintegration per cm - sup 3 -), and c) the stress intensity factor increases from 0.80 MPa.m - sup 1/2 - for the non irradiated sample up to 0.86 MPa.m - sup 1/2 - for the sample mentioned in b). The in-pile annealing was analyzed by comparing the crack nucleation after irradiation with data obtained by heavy ion irradiation followed by thermal annealing. Results for the SG7 glass were compared to those for soda-lime and borosilicate glasses. (author)

  18. In-Pile Qualification of the Fast-Neutron-Detection-System

    Directory of Open Access Journals (Sweden)

    Fourmentel D.

    2018-01-01

    FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCK•CEN in Belgium. Then a second test was recently completed at ISIS reactor operated by CEA in France. FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%.

  19. IN-PILE INSTRUMENTATION TO SUPPORT FUEL CYCLE RESEARCH AND DEVELOPMENT - FY12 STATUS REPORT

    Energy Technology Data Exchange (ETDEWEB)

    J. . Rempe; J. Daw; D. Knudson; R. Schley

    2012-09-01

    As part of the FCRD program objective to emphasize science-based, goal-oriented research, a strategic research program is underway to develop new sensors that can be used to obtain the high fidelity, real-time, data required for characterizing the performance of new fuels during irradiation testing. The overarching goal of this initiative is to develop new test vehicles with new sensors of unprecedented accuracy and resolution that can obtain the required data. Prior laboratory testing and, as needed, irradiation testing of sensors in these capsules will be completed as part of this initiative to give sufficient confidence that the irradiation tests will yield the required data. This report documents FY12 progress in this initiative.

  20. Design data for the 1/4-in. thick alloy 718 in-pile tube

    International Nuclear Information System (INIS)

    Reuter, W.G.

    1977-05-01

    Design data applicable to the proposed 0.64-cm (1/4-in.) thick inpile tube (IPT) used in the experimental test loop of the Power Burst Facility (PBF) are reported, together with fracture toughness results which were also desired to complement the design data. These data were required for temperatures ranging from room temperature (RT) to 811 K (1000 0 F). Tests were conducted on transverse-oriented specimens for base metal and longitudinal- and transverse-oriented weld metal specimens. Tensile test results from base metal (Heats 3 and 6) and weld metal (Heat C), for test temperatures ranging from RT to 811 K (1000 0 F), were used to develop curves for ultimate and yield tensile strength. The design fatigue curves for transverse-oriented base metal and longitudinal- and transverse-oriented weld metal specimens were developed using the tensile test results plus an analytical approach. Experimental fatigue data were generated to augment the analytically determined values. The design curves for cycles to failure ranging from 10 1 to 10 6 and for temperatures for RT to 811 K (1000 0 F) are contained in this report. Instrumented precracked-Charpy specimens were tested over a range of temperatures to provide some estimate of the dynamic fracture toughness (K/sub Id/). On the basis of these tests, K/sub Id/ values for the base metal and weldments are quite satisfactory

  1. Parametric analysis of LIBRETTO-4 and 5 in-pile tritium transport model on EcosimPro

    Energy Technology Data Exchange (ETDEWEB)

    Alcalde, Pablo Martínez, E-mail: pablomiguel.martinez@externos.ciemat.es [Universidad Nacional de Educación a Distancia (UNED), c/Juan del Rosal 12, 28040 Madrid (Spain); Moreno, Carlos; Ibarra, Ángel [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain)

    2014-10-15

    Highlights: • Introduction of a new tritium transport model of LIBRETTO-4 and 5 on EcosimPro{sup ®}. • Analysis of model input parameter and variable sensitivities and effects on tritium simulated fluxes. • Demonstrations of high tritium out-flux dependencies on lead-lithium parameters. • Rough fitting achievements proposed by Li17Pb solubility or recombination increase. - Abstract: A new model for LIBRETTO-4/1, 4/2 and 5 experiments have been developed on ECOSIMPro{sup ©} tool to simulate tritium in-pile breeding and transport into two separate purge gas channels with He + 0.1%H{sub 2}. Release from lead lithium eutectic plenum with coupled permeation through an austenitic steel wall on the first and single permeation through EUROFER-97 in the temperature ranges of 300–550 °C can be simulated tuning the transport parameters involved. A parametric study has been performed to reduce the degrees of freedom and to determine the error caused in the simulation due to the uncertainty in experimental input data. The information obtained is essential for the experimental benchmarking. The Tritium Permeation Percentage (TPP) is an output calculated parameter with low variations between 2 and 6% along the whole experimental time easy to compare (730 Full Power Days for LIBRETTO-4 and 520 for 5). Tritium transport parameter ranges verifying this output are defined herein.

  2. Some aspects of in-pile swelling of fissile materials, 1. part: non-alloyed α uranium

    International Nuclear Information System (INIS)

    Mikailoff, H.

    1964-01-01

    An examination has been carried out of non-alloyed uranium samples, having various structural states, cold-worked and recrystallized, as-cast and β-treated, and irradiated at temperatures of between 450 and 600 C and with burn-ups from 1300 to 5500 MW days/metric ton. These samples swelled because of precipitation of the fission gases the porosity thus produced has a morphology depending mainly on the type of deformation to which the metal has been subjected and which is due to in-pile growth. The most homogeneous distribution of pores, and thus that leading to the minimum swelling, is only observed in the material having a marked [010] texture in which the growth and perhaps the thermal cycling introduce little or no strain. For other materials the deformation /swelling association causes a more rapid destruction of the samples either by cracking when the deformation is due to twinning, or by pronounced swelling localized in the bands when deformation is due to slipping. Finally the fission-gas precipitation considerably facilitates, above 500 C, the germination and growth of the intergranular cracks which can then develop at low stresses. (author) [fr

  3. Analysis of ULOF accident in Monju reflecting the knowledge from CABRI in-pile experiments and others

    International Nuclear Information System (INIS)

    Sato, Ikken; Tobita, Yoshiharu; Suzuki, Tohru; Kawada, Ken-ichi; Fukano, Yoshitaka; Fujita, Satoshi; Kamiyama, Kenji; Ishikawa, Makoto; Nonaka, Nobuyuki; Usami, Shin

    2007-05-01

    In the safety evaluation in the original licensing procedure of the prototype FBR Monju, mechanical energy release during an unprotected loss-of-flow (ULOF) event, one of technically inconceivable events postulated beyond design basis, was evaluated and shown that the radiological consequence is suitably limited. Since the sodium leakage accident in the secondary heat-transport system, the Monju plant has been kept under the stand-by condition for more than ten years and the composition of fuel in the current reactor core has changed due to decay of a plutonium isotope into americium. This change in fuel composition affects the neutronic characteristics of the core, requiring assessment of its effect on safety. In this study, event sequences of ULOF were analyzed to assess the effect of the change in neutronic characteristics on the mechanical energy release during ULOF. A significant advance has been made in the safety analysis codes reflecting the knowledge obtained through extensive safety research programs in the last decades, notably the CABRI in-pile experiments. The present results with the advanced analysis codes showed that the mechanical energy release in the current Monju core with the changed neutronic characteristics would not exceed the value evaluated formerly. (author)

  4. Risk analysis for a radiolysis gas detonation in an in-pile loop with supercritical water

    International Nuclear Information System (INIS)

    Zeiger, T.; Raque, M.; Kuznetsov, M.; Redlinger, R.; Schulenberg, T.

    2012-01-01

    The SCWR (supercritical water reactor) -FQT project is a cooperation between European and Chinese partners aimed to test the fuel SCWR elements under reactor conditions. In the frame of this work the risk of radiolysis gas production in the active range of the test track was assessed. The radiolysis gas could accumulate in an emergency cooling system with stagnating coolant. The ignition of this radiolysis gas could cause pressure peaks that are able to damage the primary coolant circuit. Pressure increase and deformations in case of ignition of accumulated gas were investigated. As piping material the Ti stabilized austenitic steel 08Ch18N10T was assumed, the simulation was performed using the ANSYS code. The results show that pipes without significant wall thickness enhancement cannot withstand the radiolysis gas detonation.

  5. Tensile bond strength of adhesive systems: effects of primer and thermocycling Resistência à tração de sistemas adesivos: efeitos do “ primer” e dos ciclos térmicos

    Directory of Open Access Journals (Sweden)

    Luciana Tibiriçá AGUILAR

    2002-03-01

    Full Text Available The objective of this study was to evaluate the effects of primer and thermocycling on the bond strength of multi-purpose adhesive systems applied to enamel, under tensile stress. The following bonding systems were applied, according to the manufacturers' instructions, on unground enamel buccal surfaces of 96 premolars, with or without the application of primer: Scotchbond MP, OptiBond FL, Amalgambond Plus and OptiBond (dual-cure. A composite resin (Z100, 3M was applied and light-cured in a cast metal hollow cone, which was previously fixed to the enamel surfaces. Half of the sample was subjected to 3,000 thermocycles (5-37ºC; 37-55ºC, dwell time of 60 s, and the other half was stored in water at 37ºC for the same period. The data were treated by means of ANOVA and no significant effects were detected, which indicates that tensile bond strength was not affected by the adhesive systems, application of primer or thermocycling.O objetivo desta pesquisa foi o de verificar o efeito do "primer" e dos ciclos térmicos na resistência da união entre adesivos multiuso e esmalte dental, sob ensaios de tração. Os seguintes sistemas adesivos foram aplicados, de acordo com as instruções dos fabricantes, na superfície vestibular (sem desgaste de 96 pré-molares com ou sem a aplicação prévia do "primer": Scotchbond MP, OptiBond FL, Amalgambond Plus e OptiBond - "dual cure". Após a aplicação do sistema adesivo, foi confeccionado um cone de resina composta (Z100, 3M, e fotoativado dentro de um molde metálico. Metade do total de espécimes foi submetida a 3.000 ciclos térmicos (5-37ºC; 37-55ºC, 60 s de imersão; a outra metade permaneceu imersa em água a 37ºC pelo mesmo tempo dispensado no procedimento anterior. Os dados foram submetidos a uma análise de variância (p = 0,05 e nenhum efeito significante foi detectado, indicando que a resistência de união não foi afetada pelo sistema adesivo, pela aplicação do "primer" ou pelos ciclos térmicos.

  6. Use of Kalman filter methods in analysis of in-pile LMFBR accident simulations

    International Nuclear Information System (INIS)

    Meek, C.C.; Doerner, R.C.

    1983-01-01

    Kalman filter methodology has been applied to inpile liquid-metal fast breeder reactor simulation experiments to obtain estimates of the fuel-clad thermal gap conductance. A transient lumped parameter model of the experiment is developed. An optimal estimate of the state vector chosen to characterize the experiment is obtained through the use of the Kalman filter. From this estimate, the fuel-clad thermal gap conductance is calculated as a function of time into the test and axial position along the length of the fuel pin

  7. Summarizing evaluation of the results of in-pile experiments for the transient fission gas release under accidental conditions of fast breeders

    International Nuclear Information System (INIS)

    Fischer, E.A.; Vaeth, L.

    1989-04-01

    The transient fission gas behaviour and the fission gas induced fuel motion were studied in in-pile experiments in different countries, under conditions typical for hypothetical accidents. This report summarizes first the different experiment series and the main results. Then, a comparative evaluation is given, which provides a basis for the choice of the fission gas parameters in the accident code SAS3D

  8. Thermal annealing behaviour of sulphur-35 produced in pile-irradiated mixed crystals AlCl/sub 3/-FeCl/sub 3/

    Energy Technology Data Exchange (ETDEWEB)

    Dyakovich, V; Todorovski, D S; Kostadinova, Z D [Sofia Univ. (Bulgaria). Khimicheski Fakultet

    1983-12-19

    The regression analysis of the experimental results on the thermal annealing behaviour of /sup 35/S produced in pile-irradiated mixed crystals AlCl/sub 3/-FeCl/sub 3/ confirms some suppositions made in a previous paper. The chemical state of /sup 35/S is defined by the target prehistory and the iron concentration. The influence of Fe/sup 3 +/ can be observed indirectly through its influence on the defect structure formed.

  9. Test Pile Reactivity Loss Due to Trichloroethylene

    International Nuclear Information System (INIS)

    Plumlee, K.E.

    2001-01-01

    The presence of trichloroethylene in the test pile caused a continual decrease in pile reactivity. A system which removed, purified, and returned 12,000 cfh helium to the pile has held contamination to a negligible level and has permitted normal pile operation

  10. In-pile loop studies of the effect of PWR coolant pH on corrosion product radionuclide deposition

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Harling, O.K.; Kohse, G.E.

    1992-02-01

    An in-pile loop which simulates the primary coolant system of a PWR has been constructed and operated in the MIT research reactor. A total of seven one-month-long irradiations have been carried out to evaluate the effect of coolant pH controlled by variation in LiOH/H 3 BO 3 concentrations. With the exception of one run at zero boron, all employed 800 ppm B; pH 300degreesC values of 6.5, 7.0, 7.2, 7.5 were studied, and two runs each at 7.0 and 7.2 were carried out. Finally, one of the runs at a pH 300degreesC of 7.2 was conducted with special care to exclude zinc because of its potential effects on cobalt deposition. The results show the expected benefits of high pH in reducing the rate of activity deposition on plant surfaces, but pH 300degreesC = 7.2 is approximately as effective as 7.5, while pH 300degreesC = 6.5 exhibits much larger activity transport and qualitatively different deposition behavior. Significant heat flux effects not predicted by current models have been consistently observed. While not as extensively studied, the zero-boron run suggests that the presence of boron species, at fixed pH, may reduce the net amount of activity deposited on ex-core surfaces. Neutron activation analysis of a variety of samples ruled out Zircaloy as an important source of Co-60, since its cobalt content is less than one ppm, considerably less than the applicable ASTM specification of ≤ 20 ppm. Amendment of the latter has been recommended

  11. DeBeNe Test Facilities for Fast Breeder Development

    International Nuclear Information System (INIS)

    Storz, R.

    1980-10-01

    This report gives an overview and a short description of the test facilities constructed and operated within the collaboration for fast breeder development in Germany, Belgium and the Netherlands. The facilities are grouped into Sodium Loops (Large Facilities and Laboratory Loops), Special Equipment including Hot Cells and Reprocessing, Test Facilities without Sodium, Zero Power Facilities and In-pile Loops including Irradiation Facilities

  12. New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

    International Nuclear Information System (INIS)

    Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

    1996-01-01

    The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

  13. Effect of thermocycling on the bond strength of a glass-infiltrated ceramic and a resin luting cement Efeito da ciclagem térmica sobre a resistência de união entre uma cerâmica infiltrada com vidro e um cimento resinoso

    Directory of Open Access Journals (Sweden)

    Osvaldo Daniel Andreatta Filho

    2003-03-01

    Full Text Available The aim of the present study was to evaluate the effect of thermocycling on the bond strength between the surface of the glass-infiltrated alumina ceramic In-Ceram (VITA and the Panavia F resin cement (Kuraray CO.. Four 5x6x6mm In-Ceram blocks were obtained. One of the 6x6mm faces of each block was conditioned with Cojet - System (tribochemical silica coating, ESPE-3M and then luted under a constant 750g pressure with Panavia F cement to another identical face of a resin composit block (Clearfil AP-X, Kuraray obtained by reproduction of the ceramic one from Express (3M addition curing silicone impressions. The four sets so formed by ceramic, luting cement and resin have been each one serially sectioned in 20 sticks so that the adhesive surface in each presented 1mm² of area. The samples were divided in 2 groups (n=10: G1- stored for 7 days in deionized water at 36 ± 2ºC; G2 - thermocycled 1500 times between 5 and 55ºC dwell times. The microtensile tests were accomplished in an universal testing machine (EMIC at a crosshead speed of 0,5 mm/min. The results showed that the mean tensile bond strength values (MPa for the group G2: (22,815 ± 5,254 had not statistically differ of the values of group G1: (25,628 ± 3,353 (t = 1,427; gl = 18; p-value = 0,171, at the level of a= 5%. It can be concluded that the thermocycling technique used in the present experiment had not produced statistically significant differences between the bond strength results of the specimens obtained by the two used techniques.O objetivo deste estudo foi avaliar o efeito da ciclagem térmica sobre a resistência de adesão entre a superfície da cerâmica In-Ceram Alumina (VITA e o cimento resinoso Panavia F (Kuraray. Foram confeccionados quatro blocos de cerâmica In-Ceram com dimensões de 5x6x6mm. Uma das faces com 6x6mm de cada bloco cerâmico, após condicionamento com o sistema Cojet (ESPE-3M (jateamento com óxido de alumínio/jateamento com óxido de s

  14. Part 1: Logging residues in piles - Needle loss and fuel quality. Part 2: Nitrogen leaching under piles of logging residues

    International Nuclear Information System (INIS)

    Lehtikangas, P.; Lundkvist, H.

    1991-01-01

    Part 1: Experimental piles were built in three geographical locations during May-Sept. 1989. Logging residues consisted of 95% spruce and 5% pine. Height of the piles varied between 80 and 230 cm. Needles were collected by placing drawers under 40 randomely chosen piles. The drawers were emptied every two weeks during the storage period. Natural needle loss was between 18 and 32% of the total amount of needles after the first two months of storage. At the end of the storage period, 24-42% of the needles had fallen down to the drawers. At the end of the experiment the total needle fall was 95-100% in the shaken piles. According to the results of this study piles smaller than 150 cm had the most effective needle fall. Piles should be placed on open places where the air and sun heat penetrate and dry them. Needles were the most sensitive fraction to variations in precipitation compared to the other components, such as branches. Piles usually dried quickly, but they also rewet easily. This was especially true in the smaller piles. The lowest moisture content was measured at the end of June. The ash content in needles varied between 4 and 8%. 16 refs., 15 figs. Part 2: Three field experiments were equipped with no-tension humus lysimeters. Pairs of lysimeters with the same humus/field layer vegetation material were placed in pairs, one under a pile of felling residues and another in the open clear felling. Leaching of nitrogen as well as pH and electric conductivity in the leachate was followed through sampling of the leachate at regular intervals. The results from the investigation show that: * the amount of leachate was higher in lysimeters in the open clear felling, * pH in the leachate was initially lower under piles of felling residues, * the amount of nitrogen leached was higher in the open clear felling. Thus, storing of felling residues in piles during the summer season did not cause any increase in nitrogen leaching, which had been considered to be a risk

  15. Thermal conductivity of sintered UO{sub 2} under in-pile conditions; Conductibilite thermique de l'UO{sub 2} fritte dans les conditions d'utilisation en pile

    Energy Technology Data Exchange (ETDEWEB)

    Stora, J P; Bernardy De Sigoyer, B; Delmas, R; Deschamps, P; Lavaud, B; Ringot, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The temperature distribution in a stack of sintered UO{sub 2} cylinders has been studied both in the laboratory where the heat energy is produced by an axial heating element, and in-pile, where the heating is due solely to nuclear effects. Under a high thermal gradient the UO{sub 2} cracks both along radial planes and along pseudo-cylindrical surfaces: these latter act as thermal barriers to the heat flow, It is therefore an apparent thermal conductivity k{sub a}(T), lower than the intrinsic value k(T) of this parameter which is measured. The efficiency of these barriers decreases when the gap decreases and when the external pressure acting on the cracked stack increases: in the limiting case, for high values of the binding strain, k{sub a}(T) {approx_equal} k(T). In the domain of phonon conduction (T {<=} 1350 deg C), the expression kw.cm{sup -1}.C{sup -1}=1/(11+0.024*T) accounts for the real thermal conductivity. Above 1350 deg C the thermal conductivity increases. Two in-pile measurements up to 1250 deg C carried out using cartridges fitted with thermocouples confirm, within the limits of experimental error, the above expression and the qualitative effects of the binding strains. Similar tests have been carried out-of-pile and in-pile on the real shape of the EL-4 fuel 'pencils'. Out-of-pile, the influence of the initial free gap, of the nature of the gas filing the 'pencil' and of the external pressure have been studied; the results are compatible with the above interpretation; It appears that an external pressure of 60 kg/cm{sup 2} is insufficient to restore completely the thermal conductivity of the fuel. (authors) [French] La distribution de temperature dans un empilement de cylindres d'UO{sub 2} fritte est etudiee a la fois au laboratoire, ou l'energie calorifique est produite par un element chauffant axial, et en pile, ou l'echauffement est uniquement nucleaire. Sous gradient thermique eleve, l'UO{sub 2} se fracture a la fois suivant des plans radiaux et

  16. Advanced instrumentation and analysis methods for in-pile thermal and nuclear measurements: from out-of-pile studies to irradiation campaigns

    International Nuclear Information System (INIS)

    Reynard-Carette, C.; Lyoussi, A.

    2015-01-01

    Research and development on nuclear fuel behavior under irradiations and accelerated ageing of structure materials is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. A new Material Testing Reactor (MTR), the Jules Horowitz Reactor (JHR) currently under construction in the South of France in the CEA Cadarache research centre will offer a real opportunity to perform R and D programs and hence will crucially contribute to the selection, optimization and qualification of innovative materials and fuels. To perform such programs advanced accurate and innovative experiments, irradiation devices that contain material and fuel samples are required to be set up inside or beside the reactor core. These experiments needs beforehand in situ and on line sophisticated measurements to accurately reach specific and determining parameters such as thermal and fast neutron fluxes, nuclear heating and temperature conditions to precisely monitor and control the conducted assays. Consequently, since 2009 CEA and Aix-Marseille University collaborate in order to design and develop a new multi-sensor device which will be dedicated to measuring profiles of such conditions inside the experimental channels of the JHR. These works are performed in the framework of two complementary joint research programs called MAHRI-BETHY and INCORE. These programs couple experimental studies carried out both out-of nuclear fluxes (in laboratory) and under irradiation conditions (in OSIRIS MTR reactor in France and MARIA MTR reactor in Poland) with numerical works realized by thermal simulations (CAST3M code) and Monte Carlo simulations (MCNP code). These programs deal with three main aims. The first one corresponds to the design and/or the test of new in-pile instrumentation. The second one concerns the development of advanced calibration procedures in particular in the case of one specific sensor: a differential calorimeter used to quantify

  17. Advanced instrumentation and analysis methods for in-pile thermal and nuclear measurements: from out-of-pile studies to irradiation campaigns

    Energy Technology Data Exchange (ETDEWEB)

    Reynard-Carette, C. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 (France)

    2015-07-01

    Research and development on nuclear fuel behavior under irradiations and accelerated ageing of structure materials is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. A new Material Testing Reactor (MTR), the Jules Horowitz Reactor (JHR) currently under construction in the South of France in the CEA Cadarache research centre will offer a real opportunity to perform R and D programs and hence will crucially contribute to the selection, optimization and qualification of innovative materials and fuels. To perform such programs advanced accurate and innovative experiments, irradiation devices that contain material and fuel samples are required to be set up inside or beside the reactor core. These experiments needs beforehand in situ and on line sophisticated measurements to accurately reach specific and determining parameters such as thermal and fast neutron fluxes, nuclear heating and temperature conditions to precisely monitor and control the conducted assays. Consequently, since 2009 CEA and Aix-Marseille University collaborate in order to design and develop a new multi-sensor device which will be dedicated to measuring profiles of such conditions inside the experimental channels of the JHR. These works are performed in the framework of two complementary joint research programs called MAHRI-BETHY and INCORE. These programs couple experimental studies carried out both out-of nuclear fluxes (in laboratory) and under irradiation conditions (in OSIRIS MTR reactor in France and MARIA MTR reactor in Poland) with numerical works realized by thermal simulations (CAST3M code) and Monte Carlo simulations (MCNP code). These programs deal with three main aims. The first one corresponds to the design and/or the test of new in-pile instrumentation. The second one concerns the development of advanced calibration procedures in particular in the case of one specific sensor: a differential calorimeter used to quantify

  18. Chemical effects of 13N produced by recoil protons and deuterons in pile-irradiated methanol and methanol-d4

    International Nuclear Information System (INIS)

    Sensui, Y.; Tomura, K.; Matsuura, T.

    1982-01-01

    The stabilized chemical forms of 13 N resulting from the reactions 13 C(p,n) 13 N by a recoil proton and 12 C(d,n) 13 N by a recoil deuteron, were studied in pile-irradiated methanol and methanol-d 4 in the temperature range from 77 to 295 K. Contrary to the target of benzene, cyclohexane, acetone and diethyl ether previously studied, the relative yield of 13 N-compounds did not depend on the irradiation temperature in the present media. In the yield of 13 N-compounds no marked change was observed between methanol and methanol-d 4 , differing from the results between benzene and benzene-d 6 . A mechanism is proposed to explain the results. (author)

  19. Absolute on-line in-pile measurement of neutron fluxes using self-powered neutron detectors: Monte Carlo sensitivity calculations

    Energy Technology Data Exchange (ETDEWEB)

    Vermeeren, L. [SCK/CEN, B-2400 Mol (Belgium)

    2001-07-01

    Self-powered neutron detectors (SPND) are well suited to monitor continuously the neutronic operating conditions of driver fuel of research reactors and to follow its burnup evolution. This is of particular importance when advanced or new MTR fuel designs need to be qualified. We have developed a detailed MCNP-4B based Monte Carlo approach for the calculation of neutron sensitivities of SPNDs. Results for the neutron sensitivity of a Rh SPND are in excellent agreement with experimental data recently obtained at the BR2 research reactor. A critical comparison of the Monte Carlo results with results from standard analytical methods reveals an important deficiency of the analytical methods in the description of the electron transport efficiency. Our calculation method allows a reliable on-line determination of the absolute in-pile neutron flux. (author)

  20. Absolute on-line in-pile measurement of neutron fluxes using self-powered neutron detectors: Monte Carlo sensitivity calculations

    International Nuclear Information System (INIS)

    Vermeeren, L.

    2001-01-01

    Self-powered neutron detectors (SPND) are well suited to monitor continuously the neutronic operating conditions of driver fuel of research reactors and to follow its burnup evolution. This is of particular importance when advanced or new MTR fuel designs need to be qualified. We have developed a detailed MCNP-4B based Monte Carlo approach for the calculation of neutron sensitivities of SPNDs. Results for the neutron sensitivity of a Rh SPND are in excellent agreement with experimental data recently obtained at the BR2 research reactor. A critical comparison of the Monte Carlo results with results from standard analytical methods reveals an important deficiency of the analytical methods in the description of the electron transport efficiency. Our calculation method allows a reliable on-line determination of the absolute in-pile neutron flux. (author)

  1. Irradiation experiments of 3rd, 4th and 5th fuel assemblies by an in-pile gas loop, OGL-1

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kobayashi, Fumiaki; Hayashi, Kimio; Minato, Kazuo; Kikuchi, Teruo; Adachi, Mamoru; Iwamoto, Kazumi; Ikawa, Katsuichi; Itami, Hiroharu.

    1986-07-01

    Three irradiation experiments for 3rd, 4th and 5th fuel assemblies which had been composed of VHTR reference coated particle fuels and graphite components were carried out by an in-pile gas loop, OGL-1 during 1979 and 1982. The main purposes of these experiments were to study on bowing of the fuel rod by irradiation for the 3rd fuel assembly, to study on fuel behavior under relatively low burnup irradiation for the 4th fuel assembly, and to study on fuel behavior up to full burnup of VHTR design for the 5th fuel assembly. For understanding in-pile fuel behavior, fractional releases of fission gases from each fuel assembly were estimated by measuring the fission gas concentrations in the primary loop of OGL-1. The post-irradiation examination (PIE) was carried out extensively on the fuel block, the fuel rods and the fuel compacts in Tokai Hot Laboratory. Also, made were the measurements of metallic fission product distributions in the fuel assemblies and the fuel rods. The results in these experiments were given as follows ; bowing of the fuel rod in the 3rd fuel assembly was 0.7 mm, but integrity of the rod was kept under irradiation. Fractional release of the fission gas from the 4th fuel assembly remained in the order of 10 -7 during irradiation, suggesting that the fuel performance was excellent. The fractional release from the 5th fuel assembly, on the other hand, was in the order of 10 -5 which was the same level in the VHTR design. (author)

  2. Preparation of a thermal-hydraulic design method for driver core fuel pins of a new in-pile experimental reactor for FBR safety research

    International Nuclear Information System (INIS)

    Mizuno, Masahiro; Yamaguchi, Katsuhisa; Uto, Nariaki

    1999-07-01

    A design study of a new in-pile experimental reactor, SERAPH (Safety Engineering Reactor for Accident PHenomenology), for FBR safety research has progressed at JNC (Japan Nuclear Cycle Development Institute). SERAPH is intended for various in-pile experiments to be performed under quasi-steady state and various transient operation modes. In order to evaluate the driver core performance in conducting such experiments, clarify the relating design issues to be resolved and refine the experimental needs, it is indispensable to comprehend the allowable margin for the thermal-hydraulic fuel pin design since it largely affects the strategy for the driver core design. This report presents a thermal-hydraulic design method for the driver core fuel pins, which is a combination of a two-dimensional time-dependent heat transfer analysis code TAC-2D and a general non-linear finite-element structural analysis code FINAS. In TAC-2D, the allowable spatial mesh and the time step sizes are evaluated. The code is modified so as to treat time-dependent thermal properties, include an improved gap heat-transfer model and treat the change of intra-pin gap width under transient modes, for the purpose of improving the accuracy of evaluating heat transfer characteristics which gives a significant impact on the thermal-hydraulic design. As for FINAS, the number of element nodes and spatial meshes required to obtain adequate accuracy for the thermal stress characteristics of a fuel pellet during transient modes are investigated. In addition, post-processing tools are newly developed to process the calculation results obtained from these codes. The results of this work contribute to advancing the fuel pin design study for SERAPH as well with the investigation on the technique of manufacturing fuel pins. (author)

  3. Safety test facilities - status, needs, future directions

    International Nuclear Information System (INIS)

    Heusener, G.; Cogne, F.

    1979-08-01

    A survey is given of the in-pile programs which are presently or in the near future being performed in the DeBeNe-area and in France. Only those in-pile programs are considered which are dealing with severe accidents that might lead to disruption of major parts of the core. By comparing the needs with the goals of the present programs points are identified which are not sufficiently well covered up till now. The future procedure is described: the existing facilities will be used to the largest possible extent. Whenever it is necessary, upgrading and improvement will be foreseen. Studies of a Test Facility allowing the transient testing of large pin bundles should be continued. The construction of such a facility in Europe in the near future however seems premature

  4. Fabrication of irradiation capsule for IASCC irradiation tests (2). Irradiation capsule for crack propagation test (Joint research)

    International Nuclear Information System (INIS)

    Ide, Hiroshi; Matsui, Yoshinori; Kawamata, Kazuo; Taguchi, Taketoshi; Kanazawa, Yoshiharu; Onuma, Yuichi; Watanabe, Hiroyuki; Inoue, Shuichi; Izumo, Hironobu; Ishida, Takuya; Saito, Takashi; Ishitsuka, Etsuo; Kawamura, Hiroshi; Kaji, Yoshiyuki; Ugachi, Hirokazu; Tsukada, Takashi

    2008-03-01

    It is known that irradiation Assisted Stress Corrosion Cracking (IASCC) occurs when austenitic stainless steel components used for light water reactor (LWR) are irradiated for a long period. In order to evaluate the high aging of the nuclear power plant, the study of IASCC becomes the important problem. The specimens irradiated in the reactor were evaluated by post irradiation examination in the past study. For the appropriate evaluation of IASCC, it is necessary to test it under the simulated LWR conditions; temperature, water chemistry and irradiation conditions. In order to perform in-pile SCC test, saturated temperature capsule (SATCAP) was developed. There are crack growth test, crack propagation test and so on for in-pile SCC test. In this report, SATCAP for crack propagation test is reported. (author)

  5. Fabrication of irradiation capsule for IASCC irradiation tests (1). Irradiation capsule for crack growth test (Joint research)

    International Nuclear Information System (INIS)

    Ide, Hiroshi; Matsui, Yoshinori; Kawamata, Kazuo; Taguchi, Taketoshi; Kanazawa, Yoshiharu; Onuma, Yuichi; Watanabe, Hiroyuki; Inoue, Shuichi; Izumo, Hironobu; Ishida, Takuya; Saito, Takashi; Ishitsuka, Etsuo; Kawamura, Hiroshi; Kaji, Yoshiyuki; Ugachi, Hirokazu; Tsukada, Takashi

    2008-03-01

    It is known that Irradiation Assisted Stress Corrosion Cracking (IASCC) occurs when austenitic stainless steel components used for light water reactor (LWR) are irradiated for a long period. In order to evaluate the high aging of the nuclear power plant, the study of IASCC becomes the important problem. The specimens irradiated in the reactor were evaluated by post irradiation examination in the past study. For the appropriate evaluation of IASCC, it is necessary to test it under the simulated LWR conditions; temperature, water chemistry and irradiation conditions. In order to perform in-pile SCC test, saturated temperature capsule (SATCAP) was developed. There are crack growth test, crack propagation test and so on for in-pile SCC test. In this report, SATCAP for crack growth test is reported. (author)

  6. Finite Element-Assisted Assessment of the Thermo-cyclic Characteristics of Leads Soldered with SnAgCu(+Bi,In) Alloys

    Science.gov (United States)

    Lis, Adrian; Nakanishi, Kohei; Matsuda, Tomoki; Sano, Tomokazu; Minagawa, Madoka; Okamoto, Masahide; Hirose, Akio

    2017-07-01

    Solder joints between leads and printed circuit boards in thin small outline packages were produced with conventional Sn1.0Ag0.7Cu (SAC107) and Sn3.0Ag0.7Cu (SAC305) solders as well as various solder alloys with gradually increasing amounts of Bi (up to 3.0 wt.%) and In (up to 1.0 wt.%) within the SAC107 base solder. The reliability of soldered leads in temperature cycle (TC) tests improved most with solder alloys containing both Bi (1.6 wt.%) and In (0.5 wt.%). Microindentation and electron probe microanalysis mappings revealed that the effect originates from a combination of solution and precipitation strengthening of the initial SAC alloy. The distribution of inelastic strain accumulation (ISA), as a measure for degradation, was determined in the solder joints by finite element calculations. It was shown that defects in the solder proximal to the lead (60-75 μm), which was underpinned by similar cracking characteristics along the lead-solder interface. The ISA was confirmed to be lower in SAC+Bi/In alloys owing to their enhanced elasto-plastic properties. Moreover, the addition of a thin Cu coating on the leads could improve the joint reliability, as suggested by the calculation of the ISA and the acceleration factor.

  7. In-pile Tritium Permeation through F82H Steel with and without a Ceramic Coating of Cr2O3-SiO2 Including CrPO4

    International Nuclear Information System (INIS)

    Nakamichi, M.; Hayashi, K.; Kulsartov, T.V.; Afanasyev, S.E.; Shestakov, V.P.; Chikhray, Y.V.; Kenzhin, E.A.; Kolbaenkov, A.N.

    2006-01-01

    Development of coating on blanket structural materials with significant reduction capability of tritium permeation is highly required in order to realize a reasonable design of a tritium recovery and processing system of demonstration (DEMO) fusion reactors. An effective coating has been developed in Japan Atomic Energy Agency (JAEA) using a ceramic material of Cr 2 O 3 -SiO 2 including CrPO 4 . In previous out-of-pile deuterium permeation experiments at 600 o C [T.V. Kulsartov et al., Fusion Eng. Des. 81 (2006) 701], a significant permeation reduction factor (PFR) of about 300 was obtained for the coating on the inner-side surface of tubular diffusion cells made by ferritic steel (F82H). In the present study, in-pile experiments on tritium permeation were conducted for F82H steel with and without the same coating, using a testing reactor IGV-1M in Kazakhstan. The tritium source used was liquid lithium-lead eutectics, Pb17Li, which was poured into a space around a tubular diffusion cell (specimen) of F82H steel with or without the coating on the inner side the cell. The irradiation time was about 4 hours, which corresponds to a fast-neuron fluence of about 2x10 21 m -2 (E > 1.1 MeV). The permeation reduction factor (PRF) was obtained by comparison of kinetics curves of tritium permeation through the diffusion cell of F82H steel with and without the coating. The PRFs at 600 and 500 o C were 292 and 30, respectively. These values are close to corresponding PRF values of 307 and 45, which had been obtained at 600 and 500 o C, respectively, in the previous out-of-pile experiments [T.V. Kulsartov et al., Fusion Eng. Des. 81 (2006) 701]. (author)

  8. Irradiation Testing of Ultrasonic Transducers

    International Nuclear Information System (INIS)

    Daw, J.; Rempe, J.; Palmer, J.; Tittmann, B.; Reinhardt, B.; Kohse, G.; Ramuhalli, P.; Montgomery, R.; Chien, H.T.; Villard, J.F.

    2013-06-01

    Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of numerous parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of single, small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. Other efforts include an ultrasonic technique to detect morphology changes (such as crack initiation and growth) and acoustic techniques to evaluate fission gas composition and pressure. These efforts are limited by the lack of existing knowledge of ultrasonic transducer material survivability under irradiation conditions. To address this need, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate promising magnetostrictive and piezoelectric transducer performance in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 10 21 n/cm 2 (E> 0.1 MeV). This test will be an instrumented lead test; and real-time transducer performance data will be collected along with temperature and neutron and gamma flux data. By characterizing magnetostrictive and piezoelectric transducer survivability during irradiation, test results will enable the development of novel radiation tolerant ultrasonic sensors for use in Material and Test Reactors (MTRs). The current work bridges the gap between proven out-of-pile ultrasonic techniques and in-pile deployment of ultrasonic sensors by acquiring the data necessary to demonstrate the performance of ultrasonic transducers. (authors)

  9. Capsule development and utilization for material irradiation tests

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

    The development program of advanced nuclear structural and fuel materials includes the in-pile tests using the instrumented capsule at HANARO. The tests were performed in the in-core test holes of CT, IR 1 and 2 and OR 4 and 5 of HANARO. Extensive efforts have also 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. Since the first instrumented capsule(97M-01K) had been designed and successfully fabricated, five tests were done to support the users and provided the economic benefits to user by generating the essential in-pile information on the performance and structural integrity of materials. This paper describes the present status and future plans of these R and D activities for the development of the instrumented capsule including in-situ material property measurement capsules and nuclear fuel test capsules.

  10. Capsule development and utilization for material irradiation tests

    International Nuclear Information System (INIS)

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

    2000-05-01

    The development program of advanced nuclear structural and fuel materials includes the in-pile tests using the instrumented capsule at HANARO. The tests were performed in the in-core test holes of CT, IR 1 and 2 and OR 4 and 5 of HANARO. Extensive efforts have also 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. Since the first instrumented capsule(97M-01K) had been designed and successfully fabricated, five tests were done to support the users and provided the economic benefits to user by generating the essential in-pile information on the performance and structural integrity of materials. This paper describes the present status and future plans of these R and D activities for the development of the instrumented capsule including in-situ material property measurement capsules and nuclear fuel test capsules

  11. Capsule development and utilization for material irradiation tests

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Hwan; Kim, B G; Joo, K N [and others

    2000-05-01

    The development program of advanced nuclear structural and fuel materials includes the in-pile tests using the instrumented capsule at HANARO. The tests were performed in the in-core test holes of CT, IR 1 and 2 and OR 4 and 5 of HANARO. Extensive efforts have also 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. Since the first instrumented capsule(97M-01K) had been designed and successfully fabricated, five tests were done to support the users and provided the economic benefits to user by generating the essential in-pile information on the performance and structural integrity of materials. This paper describes the present status and future plans of these R and D activities for the development of the instrumented capsule including in-situ material property measurement capsules and nuclear fuel test capsules.

  12. Accident analysis of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.; Chi, D. Y

    1998-03-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. The HANARO fuel test loop was designed to match the CANDU and PWR fuel operating conditions. The accident analysis was performed by RELAP5/MOD3 code based on FTL system designs and determined the detail engineering specification of in-pile test section and out-pile systems. The accident analysis results of FTL system could be used for the fuel and materials designer to plan the irradiation testing programs. (author). 23 refs., 20 tabs., 178 figs.

  13. Test

    DEFF Research Database (Denmark)

    Bendixen, Carsten

    2014-01-01

    Bidrag med en kortfattet, introducerende, perspektiverende og begrebsafklarende fremstilling af begrebet test i det pædagogiske univers.......Bidrag med en kortfattet, introducerende, perspektiverende og begrebsafklarende fremstilling af begrebet test i det pædagogiske univers....

  14. Study of thermocycling effect on the bond strength between an aluminous ceramic and a resin cement Estudo do efeito da ciclagem térmica na resistência da união adesiva entre uma cerâmica aluminizada e um cimento resinoso

    Directory of Open Access Journals (Sweden)

    Osvaldo Daniel Andreatta Filho

    2005-03-01

    Full Text Available This study evaluated the effect of thermocycling on the bond strength between Procera AllCeram (Nobel-Biocare and a resin cement (Panavia F, Kuraray CO. Nine ceramic blocks with dimensions of 5x6x6mm were conditioned at one face with Rocatec System (Espe. After, they were luted with Panavia F to composite resin blocks (Clearfil AP-X, Kuraray CO. The nine groups formed by ceramic, cement and composite resin were split up obtaining 75 samples with dimensions of 12x1x1mm and adhesive surface presenting 1mm²±0.1mm² of area. The samples were divided into 3 groups (n=25: G1 - 14 days in distilled water at 37ºC; G2 - 6,000 cycles in water (5ºC - 55ºC - 30s; G3 - 12,000 cycles in water (5ºC - 55ºC - 30s. The samples were tested in a universal testing machine (EMIC at a crosshead speed of 1mm/min. Data were analyzed by ANOVA and Tukey tests. The results indicated that mean values of rupture tension (MPa of G1 (10.71 ± 3.54 did not differ statistically (p Este trabalho avaliou o efeito da ciclagem térmica sobre a resistência adesiva entre a cerâmica aluminizada (Procera AllCeram, Nobel-Biocare e um cimento resinoso (Panavia F, Kuraray CO. Nove blocos de cerâmica, com dimensões de 5x6x6mm, foram condicionados em uma de suas faces com o Sistema Rocatec (ESPE. A seguir foram cimentados a blocos de resina composta (Clearfil AP-X, Kuraray CO. Os conjuntos cerâmica-cimento-resina foram cortados em 75 corpos-de-prova com formato retangular com dimensões de 12x1x1mm e superfície adesiva apresentando 1mm² ± 0,1mm². Os corpos-de-prova foram divididos em três grupos (n=25: G1 - 14 dias em água destilada a 37ºC, G2 - 6000 ciclos em água (5ºC e 55ºC e G3 - 12000 ciclos em água (5ºC e 55ºC. Os corpos-de-prova foram ensaiados sob velocidade de 1mm/min em máquina de ensaio universal (EMIC. Os dados foram tratados estatisticamente com os testes de Anova e Tukey e indicaram que os valores médios de tensão de rupturas (MPa de G1 (10,71 ± 3

  15. Effect of in-pile degradation of the meat thermal conductivity on the maximum temperature of the plate-type U-Mo dispersion fuels

    International Nuclear Information System (INIS)

    Medvedev, Pavel G.

    2009-01-01

    Effect of in-pile degradation of thermal conductivity on the maximum temperature of the plate-type research reactor fuels has been assessed using the steady-state heat conduction equation and assuming convection cooling. It was found that due to very low meat thickness, characteristic for this type of fuel, the effect of thermal conductivity degradation on the maximum fuel temperature is minor. For example, the fuel plate featuring 0.635 mm thick meat operating at heat flux of 600 W/cm2 would experience only a 20 C temperature rise if the meat thermal conductivity degrades from 0.8 W/cm-s to 0.3 W/cm-s. While degradation of meat thermal conductivity in dispersion-type U-Mo fuel can be very substantial due to formation of interaction layer between the particles and the matrix, and development of fission gas filled porosity, this simple analysis demonstrates that this phenomenon is unlikely to significantly affect the temperature-based safety margin of the fuel during normal operation.

  16. Thermal conductivity of Na3(U/sub 1-y/Pu/sub y/)O4: A preliminary in-pile determination

    International Nuclear Information System (INIS)

    Lee, M.J.; Lambert, J.D.B.; Ukai, S.; Odo, T.

    1987-01-01

    During Run-Beyond-Cladding-Breach (RBCB) operation in an oxide LMR, the performance of a breached fuel element is intimately associated with the formation of fuel-sodium reaction product (FSRP), Na 3 (U/sub 1-y/Pu/sub y/)O 4 . In-pile experiments coupled with destructive examinations of breached fuel have consistently revealed noticeable changes in fuel structure accompanying FSRP formation at the fuel surface. Previous analyses have also indicated a significant impact of FSRP on fuel centerline temperature. Successful modeling of breached fuel thermal behavior therefore requires a reasonably accurate knowledge of the thermal properties of the FSRP, especially its thermal conductivity. But laboratory investigations have been scarce and limited to the Na/UO 2 system because of the toxicity of plutonium and hygroscopicity of the FSRP. Hence, post-irradiation observations of fuel samples remain the most amenable way of deriving the thermal conductivity of the FSRP. Such work is a spin-off of the RBCB program in the Experimental Breeder Reactor-II (EBR-II), a program jointly sponsored by the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan

  17. In-Pile Assemblies for Investigation of Tritium Release from Li2TiO3 Lithium Ceramic

    International Nuclear Information System (INIS)

    Shestakov, V.; Tazhibayeva, I.; Kawamura, H.; Kenzhin, Y.; Kulsartov, T.; Chikhray, Y.; Kolbaenkov, A.; Arinkin, F.; Gizatulin, Sh.; Chakrov, P.

    2005-01-01

    The description of algorithm to design in-pipe experimental ampoule devices (IPAD) is presented here, including description of IPAD design for irradiation tests of highly enriched lithium ceramics at WWR-K reactor. The description of the system for registration of tritium release from ceramics during irradiation is presented as well. Typical curve of tritium release from the IPAD during irradiation under various temperatures of the samples is shown here

  18. Program for in-pile qualification of high density silicide dispersion fuel at IPEN/CNEN-SP

    International Nuclear Information System (INIS)

    Silva, Jose E.R. da; Silva, Antonio T. e; Terremoto, Luis A.A.; Durazzo, Michelangelo

    2009-01-01

    The development of high density nuclear fuel (U 3 Si 2 -Al) with 4,8 gU/cm 3 is on going at IPEN, at this time. This fuel has been considered to be utilized at the new Brazilian Multipurpose Reactor (RMB), planned to be constructed up to 2014. As Brazil does not have hot-cell facilities available for post-irradiation analysis, an alternative qualifying program for this fuel is proposed based on the same procedures used at IPEN since 1988 for qualifying its own U 3 O 8 -Al (1,9 and 2,3 gU/cm 3 ) and U 3 Si 2 -Al (3,0 gU/cm 3 ) dispersion fuels. The fuel miniplates and full-size fuel elements irradiations should be tested at IEA-R1 core. The fuel characterization along the irradiation time should be made by means of non-destructive methods, including periodical visual inspections with an underwater video camera system, sipping tests for fuel elements suspected of leakage, and underwater dimensional measurements for swelling evaluation, performed inside the reactor pool. This work presents the program description for the qualification of the high density nuclear fuel (U 3 Si 2 -Al) with 4,8 gU/cm 3 , and describes the IPEN fuel fabrication infrastructure and some basic features of the available systems for non-destructive tests at IEA-R1 research reactor. (author)

  19. Post irradiation examination of type 316 stainless steels for in-pile Oarai water loop No.2 (OWL-2)

    International Nuclear Information System (INIS)

    Shibata, Akira; Kimura, Tadashi; Nagata, Hiroshi; Aoyama, Masashi; Kanno, Masaru; Ohmi, Masao

    2010-11-01

    The Oarai water loop No.2 (OWL-2) was installed in JMTR in 1972 for the purpose of irradiation experiments of fuel element and component material for light water reactors. Type 316 stainless steels (SSs) were used for tube material of OWL-2 in the reactor. But data of mechanical properties of highly irradiated Type 316 SSs has been insufficient since OWL-2 was installed. Therefore surveillance tests of type 316 SSs which were irradiated up to 3.4x10 25 n/m 2 in fast neutron fluence (>1 MeV) were performed. Meanwhile type 316 stainless steel (SS) is widely used in JMTR such as other irradiation apparatus and irradiation capsule, and additional data of type 316 SSs irradiated higher is required. Therefore post irradiation examinations of surveillance specimens made of type 316 SSs which were irradiated up to 1.0x10 26 n/m 2 in fast neutron fluence were performed and reported in this paper. In this result of surveillance tests of type 316 SSs irradiated up to 1.0x10 26 n/m 2 , tensile strength increase with increase of Neutron fluence and total elongation decreased with increase of Neutron fluence compared to unirradiated specimens and specimens irradiated up to 3.4x10 25 n/m 2 . This tendency has good agreement with results of 10 24 - 10 25 n/m 2 in fast neutron fluence. More than 37% in total elongation was confirmed in all test conditions. It was confirmed that type 316 SS irradiated up to 1.0x10 26 n/m 2 in fast neutron fluence has enough ductility as structure material. (author)

  20. Risk analysis of the delayed ettringite formation in pile caps foundation in the metropolitan region of Recife - PE - Brasil

    Directory of Open Access Journals (Sweden)

    I. F. Torres

    Full Text Available ABSTRACT Currently, there is an awareness that is critical to assess the durability characteristics of concrete with as much attention as the mechanical properties. The durability of concrete structures can often be affected by chemical attacks, jeopardizing its performance and security. When concrete is subjected to high temperature at early ages, many physical and chemical changes in hardened concrete may occur. It iswidely accepted that concrete subjected to these conditions of temperature and exposed to moisture is prone to cracking due to Delayed Ettringite Formation (DEF. This work aims at providing a DEF risk analysis on foundation pile caps at the Metropolitan Region of Recife - PE. Temperature rise measurement was performed in situ at 5 different caps through datalogger and thermocouples equipments. Furthermore, the Duggan test was performed in order to assess the level of expansion of 3 cements studied: X (CP II E 40, Y (CP II F 32 and Z (CP V ARI RS. Simultaneously, the chemical compositions of these cements and their respective clinkers were quantified by analysis of X-ray fluorescence (XRF. The cement X (CP II E 40 showed the chemical characteristics favoring with more intensity DEF and, as a result, higher level of expansion in the test Duggan. It is noteworthy that incorporation of metakaolin (8% and 16% and silica fume (5% and 10% showed mitigating potential of expansions. It is important to point out that all factors related to thermal properties and chemical composition of the concrete used in the region converge to a condition of ideal susceptibility for triggering DEF. Therefore, it is essential at least minimum and basic requirements in the design specification in order to avoid high temperatures in the massive concrete elements, preventing them from delayed ettringite formation.

  1. Development of 3-Pin Fuel Test Loop and Utilization Technology

    International Nuclear Information System (INIS)

    Lee, Chung Young; Sim, B. S.; Lee, C. Y.

    2007-06-01

    The principal contents of this project are to design, fabricate and install the steady-state fuel test loop in HANARO for nuclear technology development. Procurement and, fabrication of main equipment, licensing and installation for fuel test loop have been performed. Following contents are described in the report. 1. Design - Design of the In-pile system and Out pile system 2. Fabrication and procurement of the equipment - Fabrication of the In-pile system and In-pool piping - Fabrication and procurement of the equipment of the out-pile system 3. Acquisition of the license - Preparation of the safety analysis report and acquisition of the license - Pre-service inspection of the facility 4. Installation and commissioning - Installation of the FTL - Development of the commissioning procedure

  2. Safety test facilities. Needs and concepts. A French evaluation

    International Nuclear Information System (INIS)

    Tretiakoff, O.; Bailly, J.

    1976-01-01

    The fuel behaviour of LMFBRs in the event of an accident has been tested in-pile in the SCARABEE program (local blockage, sudden flow reduction and pump coast-down at constant power). These tests will be carried on in the framework of an international cooperation on irradiated fuels: this is the purpose of the CABRI and SCARABEE N programs. All those studies should enable to assess safety margins between accident conditions and the technical specifications of the reactor. The paper explains how a logical set of simple observations has led to the present state of the Cadarache in-pile experimental safety program and how it may help to find the way in a dense forest of both technical and psychological difficulties

  3. Safety test facilities. Needs and concepts. A French evaluation

    International Nuclear Information System (INIS)

    Tretiakoff, O.; Bailly, J.

    1976-01-01

    The fuel behavior of LMFBRs in the event of an accident has been tested in-pile in the SCARABEE program. These tests will be carried on in the framework of an international cooperation on irradiated fuels: this is the purpose of the CABRI and SCARABEE N programs. All those studies should enable to assess safety margins between accident conditions and the technical specifications of the reactor. The purpose of this paper is to explain how a logical set of simple observations has led us to the present state of the Cadarache in-pile experimental safety program and how it may help us to find our way in a dense forest of both technical and psychological difficulties

  4. Optical Fier Based System for Multiple Thermophysical Properties for Glove Box, Hot Cell and In-Pile Application

    Energy Technology Data Exchange (ETDEWEB)

    Ban, Heng

    2017-11-30

    Thermal diffusivity of materials is of interest in nuclear applications at temperatures in excess of 2000°C. Commercial laser flash apparatus (LFA) that heats samples with a furnace typically do not reach these elevated temperatures nor are they easily adapted to a glove-box or hot cell environment. In this research, we performed work on an experimental technique using single laser surface heating, i.e. heating the disk sample only at its front surface with the continuous wave (CW) laser, to allow measurement of thermal diffusivity at very high temperatures within a small chamber. Thermal diffusivity is measured using a separate pulsed laser on the front side and IR detector on the rear side. The new way of heating provides easy operation in comparison to other heating methods. The measurement of sample reference temperature is needed for the measured thermal diffusivity. A theoretical model was developed to describe transient heat transfer across the sample due to the laser pulse, starting from the steady state temperature of the sample heated by the CW laser. The experimental setup was established with a 500W CW laser and maximum 50 Joule pulse laser irradiated at the front surface of the sample. The induced temperature rise at the rear surface, along with the steady-state temperature at the front surface, was recorded for the determination of thermal diffusivity and the sample temperature. Three samples were tested in vacuum over a wide temperature range of 500°C to 2100°C, including graphite, Inconel 600 and tungsten. The latter two samples were coated with sprayed graphite on their front surfaces in order to achieve surface absorption/emission needs, i.e. high absorptivity of the front surface against relatively low emissivity of the rear surface. Thermal diffusivity of graphite determined by our system are within a 5% difference of the commercial LFA data at temperatures below 1300°C and agree well with its trend at higher temperatures. Good agreement

  5. High heat flux tests of mock-ups for ITER divertor application

    International Nuclear Information System (INIS)

    Giniatulin, R.; Gervash, A.; Komarov, V.L.; Makhankov, A.; Mazul, I.; Litunovsky, N.; Yablokov, N.

    1998-01-01

    One of the most difficult tasks in fusion reactor development is the designing, fabrication and high heat flux testing of actively cooled plasma facing components (PFCs). At present, for the ITER divertor project it is necessary to design and test components by using mock-ups which reflect the real design and fabrication technology. The cause of failure of the PFCs is likely to be through thermo-cycling of the surface with heat loads in the range 1-15 MW m -2 . Beryllium, tungsten and graphite are considered as the most suitable armour materials for the ITER divertor application. This work presents the results of the tests carried out with divertor mock-ups clad with beryllium and tungsten armour materials. The tests were carried out in an electron beam facility. The results of high heat flux screening tests and thermo-cycling tests in the heat load range 1-9 MW m -2 are presented along with the results of metallographic analysis carried out after the tests. (orig.)

  6. Fuel-rod response during the large-break LOCA Test LOC-6

    International Nuclear Information System (INIS)

    Vinjamuri, K.; Cook, B.A.; Hobbins, R.R.

    1981-01-01

    The large break Loss of Coolant Accident (LOCA) Test LOC-6 was conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory by EG and G Idaho, Inc. The objectives of the PBF LOCA tests are to obtain in-pile cladding ballooning data under blowdown and reflood conditions and assess how well out-of-pile ballooning data represent in-pile fuel rod behavior. The primary objective of the LOC-6 test was to determine the effects of internal rod pressures and prior irradiation on the deformation behavior of fuel rods that reached cladding temperatures high in the alpha phase of zircaloy. Test LOC-6 was conducted with four rods of PWR 15 x 15 design with the exception of fuel stack length (89 cm) and enrichment (12.5 W% 235 U). Each rod was surrounded by an individual flow shroud

  7. Study of a device for the direct measurement of the fission gas pressure inside an in-pile fuel element; Etude d'un dispositif pour la mesure directe de la pression des gaz de fission a l'interieur d'un element combustible en pile

    Energy Technology Data Exchange (ETDEWEB)

    Lavaud, B; Uschanoff, S [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The fission gas pressure inside a fuel element made of a refractory fuel constitutes an important limiting factor for the burn-up. Although it is possible to calculate approximately the volume of gas produced outside the fuel during its life-time; it is nevertheless very difficult to evaluate the pressure since the volume allowed to the fission gases, as well as their temperature are known only very approximately. This physical value, which is essential for the technologist, can only be known by direct in-pile measurement of the pressure. The report describes the equipment which has been developed for this test. (authors) [French] La pression des gaz de fission a l'interieur d'un element combustible a combustible refractaire constitue une des limitations importantes du taux de combustion. Si on peut approcher par calcul la determination du volume, des gaz degages hors du combustible au cours de sa vie, il est par contre tres difficile d'evaluer la pression car le volume alloue aux gaz de fission et leur temperature sont tres mal connus. Cette donnee essentielle pour le technologue ne peut etre atteinte que par une mesure directe en pile de la pression. Le rapport decrit l'appareillage qui a ete mis au point pour cet essai. (auteurs)

  8. Axial location of cladding failure during a slow transient overpower TREAT test

    International Nuclear Information System (INIS)

    Page, R.J.; Murphy, W.F.; Holland, J.W.

    1983-01-01

    The axial location of cladding failure following a transient overpower accident is of importance in fast reactor safety studies in that it is a determining factor in the relocation of fuel, and therefore in the possibility of inherent neutronic shutdown of the reactor. In-pile experimental data on the axial location of cladding failure of fuel in bundles of pins is sparse since, in general, the experimental fuel pin bundles are largely destroyed during the in-pile test. The post-test examination work has been completed for TREAT test J1. It was found that damage to the fuel elements during the irradiation was low enough for an accurate observation of the location of cladding failure to be made for each of the seven pins

  9. Investigation of Be/Cu joints via HHF tests of small-scale mockups

    Energy Technology Data Exchange (ETDEWEB)

    Giniatulin, R.; Gervash, A.; Komarov, V.L.; Litunovsky, N.; Mazul, I.; Yablokov, N. [Efremov Inst., St. Petersburg (Russian Federation)

    1998-01-01

    Beryllium-copper (Be/Cu) joints in divertor components work under cyclic heat loads. To develop reliable joints small-scale mockups are fabricated by divertor technologies and tested under the divertor conditions. One of the critical damaging factors that exist in the divertor and have to be simulated is thermocyclic heat loads in the range of 1-15 MW/m{sup 2}. This work presents the divertor mockups that have beryllium tiles with different dimensions (5 x 5 - 44 x 44) mm{sup 2} brazed with copper alloy heat sink. The electron beam was used to braze these mockups so as to decrease the formation of brittle intermetallic layers. The description of mockups design, geometry of armour tiles and fabrication techniques are presented in the paper. The results of screening and thermocyclic tests of these mockups in the heat flux range of 2-12 MW/m{sup 2} with a number of cycles {approx}10{sup 3} are presented. The results of metallographic analysis are also presented. The results of fabrication and testing with small-scale mockups for first wall application are also described. (author)

  10. DN detection during SLSF tests

    International Nuclear Information System (INIS)

    Braid, T.H.; Harper, H.A.; Wilson, R.E.; Baldwin, R.D.; Gilbert, D.M.; Baxter, D.E.; Gillins, R.L.; Jeffries, G.L.

    1982-01-01

    During two tests in the Sodium Loop Safety Facility (W2 and P4), two systems were operated to detect delayed neutrons from exposed fuel. One monitored directly the sodium in the in-pile loop with a transit delay from the flux region of approx.2 seconds; the other conducted a sample stream of sodium to external detectors with a transit delay which could be varied from approx.10 to approx. 40 seconds. Detectors of a wide range of sensitivity were operated in pulse and current mode; DN signals varying from 2 recoil to many grams of molten fuel could be recorded reliably without saturation. During the W2 and P4 tests a continuous DN record was made. Massive fuel failure signals were observed during reactor transients in both tests, including events interpreted as fuel melting and exposure of large areas. The steady signal from the blockage in P4 was studied as a function of reactor power and sodium temperature

  11. Measurement of the in-pile core temperature of an EL-4 pencil element, first charge (can of type-347 stainless steel, 0.4 mm thick, UO{sub 2} fuel, 11 mm diameter). Determination of the apparent thermal conductivity integral of in-pile UO{sub 2}; Mesure de la temperature a coeur en pile d'un crayon EL-4 1er jeu (gaine acier inoxydable, nuance 347 - epaisseur 0,4 mm - combustible UO{sub 2} - diametre 11 mm). Determination de l'integrale de conductibilite thermique apparente de l'UO{sub 2} en pile

    Energy Technology Data Exchange (ETDEWEB)

    Lavaud, B; Ringot, C; Vignesoult, N [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1966-11-01

    The core temperature of a pencil fuel element depends on the thermal conductivity of the UO{sub 2}, and on the UO{sub 2}-can contact. This temperature may be known accurately only if in-pile tests using the actual geometry are carried out. The test described concerns the measurement of the core- temperature of an EL-4 fuel element, first charge, having a stainless steel can. This temperature is measured at the center of the in-pile pencil element using a high-temperature thermocouple (W-Re with Ta sheath). The element is subjected to operating conditions similar to those of EL-4, both for the specific power and the can temperature and for the pressure acting on the can. The specific power is obtained in the EL-3 reactor using a slightly higher enrichment for the UO{sub 2} than that planned for EL-4. The required can temperature and pressure are obtained using a Zircaloy-2 irradiation container filled with NaK, adapted for use in the EL-3 reactor. The core temperatures of the UO{sub 2}, and that of the can surface are measured. The power is calculated from the heat exchanges in the container calibrated in the laboratory. The temperature drop at the UO{sub 2}-can interface is deduced from laboratory measurements carried out under comparable heat flux conditions, and in a gas atmosphere corresponding to the beginning of the life-time of the fuel element. It is possible to draw an integral conductivity curve. It is also possible to check the temperature distribution in the oxide, as deduced from the thermal conductivity integral, by micro-graphic examination of the oxide structure. (authors) [French] La temperature a coeur d'un crayon combustible est fonction de la conductibilite thermique de l'UO{sub 2}, mais aussi du contact UO{sub 2}-gaine. Les essais de mesure en geometrie reelle en pile sont les seuls qui permettent d'avoir une connaissance exacte de cette valeur. L'essai dont il est question dans ce rapport a trait a la mesure de la temperature a coeur d

  12. Current status and development of mechanical test techniques in the course of irradiation in MIR, SM and RBT-6 reactors

    International Nuclear Information System (INIS)

    Burukin, A.; Izhutov, A.; Nuzhdov, A.; Palachev, P.

    2015-01-01

    The reason for development and advancement of in-pile mechanical test methods for the materials and products of nuclear power engineering is that the irradiation-induced changes of some properties cannot be explained based on the results of preirradiation and post-irradiation testing. The need for such methods grows due to strengthening the requirements to performance, reliability, safety, licensing and competitive performance of nuclear fuel. The nuclear reactor operation under the power maneuvering has become more challenging. Nowadays the nuclear fuel burnup increase is the pressing issue. The nuclear power engineering gave rise to innovative projects. The phenomena being examined called for development of models in the fame of physical and mechanical concepts which provide for methodology including a full set of in-pile test methods. The strategy of irradiation testing is targeted at demonstrating irradiation feasibility of different projects in the field of nuclear power engineering with the use of conceptual approach to the property of the material or product under examination. Such an approach provides for the following: Development of the general phenomenon concept based on the available data on its mechanism under irradiation; Development of research methodology based on the general concept providing for the similarity condition by employing RRs and test methods; In-pile testing in accordance with the research methodology; Development of the phenomenon model; Model verification with the use of data obtained in the course of irradiation testing in other RRs and examination of nuclear core components; Integration of the model into the computer code intended for demonstrating feasibility of nuclear reactor design concept. The present paper describes the in-pile test methods of mechanical properties available at JSC “SSC RIAR” and applied in research reactors MIR, SM and RBT-6. Provided here is also information on their further development and some

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  14. Feasibility study on the transient fuel test loop installation

    International Nuclear Information System (INIS)

    Kim, J. Y.; Lee, C. Y.

    1997-02-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. The objective of this study is to investigate and analyze the test capsules and loops in research reactors of the other countries and to design preliminarily the eligible transient fuel test facility to be installed in HANARO. The principle subjects of this study are to analyze the contents, kinds and scopes of the irradiation test facilities for nuclear technology development. The guidances for the basic and detail design of the transient fuel test facility in the future are presented. The investigation and analysis of various kinds of test facilities that are now in operation at the research reactors of nuclear advanced countries are carried out. Based on the design data of HANARO the design materials for an eligible transient fuel test facility comprises two pacts : namely, in pile test fuel in reactor core site, and out of pile system regulates the experimental conditions in the in pile test section. Especially for power ramping and cycling selection of the eligible power variation equipment in HANARO is carried out. (author). 13 refs., 4 tabs., 46 figs

  15. LMFBR post accident heat removal testing needs and conceptual design of a test facility

    International Nuclear Information System (INIS)

    Kleefeldt, K.; Kuechle, M.; Royl, P.; Werle, H.; Boenisch, G.; Heinzel, V.; Mueller, R.A.; Schramm, K.; Smidt, D.

    1977-03-01

    A study has been carried out in which the needs and requirements for a test facility were derived, enabling detailed investigation of key phenomena anticipated during the post accident heat removal (PAHR) phase as a consequence of a postulated LMFBR whole core accident. Part I of the study concentrates on demonstrating the PAHR phenomena and related testing needs. Three types of experiments were identified which require in-pile testing, ranging from 10 to 70 cm test bed diameter and correspondingly, 30 to 5 W/g minimum power density in the test fuel. In part II a conceptual design for a test facility is presented, emphasizing the capability for accomodating large test beds. This is achieved by a below-reactor-vessel testing device, neutronically coupled to a 100 MWt sodium cooled fast reactor. (orig.) [de

  16. Superfluid He testing of titanium-stainless steel transitions fabricated by explosive welding

    International Nuclear Information System (INIS)

    Budagov, Yu.; Sabirov, B.; Shirkov, G.

    2009-01-01

    An experimental setup was constructed to test in liquid He bimetallic (titanium-stainless steel) tube joints which were manufactured by an explosive welding method. The leak levels of the samples tested at room temperature 7.5·10 -10 and 7.5·10 -9 Torr·1/s at 77 K, correspondingly, measured at FNAL (Batavia, USA) after the thermocycling have coincided with the earlier results obtained at JINR (Dubna, Russia) and INFN (Pisa, Italy) data for the same samples. For the liquid helium test the tubes were welded in pairs by their titanium ends. At the room temperature the leak level of the three tested samples was 4.9·10 -10 Torr·l/s. At the first cryogenic tests (4-6 K) one of the samples manifested a leak. The investigation will be continued since the explosive welding seems to be a very perspective new generation technology

  17. Study and modelling of the in-pile densification of the UO{sub 2} and MO{sub x} nuclear oxides; Etude et modelisation de la densification en pile des oxydes nucleaires UO{sub 2} et MO{sub x}

    Energy Technology Data Exchange (ETDEWEB)

    Boulore, A

    2001-03-01

    Amongst the many phenomena which take place in the course of the irradiation of UO{sub 2} or (U, Pu)O{sub 2} nuclear fuels, one of them involves the elimination of a fraction of the as-fabricated porosity. In-pile densification or sintering can reach 2.5%, i.e. approximately half the initial volume of pores is likely to disappear. Our literature survey indicates that the amplitude and kinetics of the phenomenon are both heavily dependent on the initial fuel microstructure. Micro-structural characterisation techniques of oxide fuels have therefore been developed in conjunction with quantitative image analysis methods. The ensuing methodology enables a quantitative comparison of micro-structural features in different fuels and has been applied to ascertaining the influence of the local fission rate and temperature on in-pile densification. It is thus revealed that in-pile operation eliminates a significant fraction of pores smaller than 3 microns in diameter. The experimental data generated has been used to set up a semi-empirical and a mechanistic model. The former is based on experimental results and is not essentially predictive. The inability of this model to predict the in-pile densification of oxide fuels is illustrated by the fact that the maximum fraction of pores that disappears is proportional to an empirical function of fission rate, and temperature. The proportionality factor appears to be difficult to correlate quantitatively to any given micro-structural feature. The model has however been applied to the interpretation of an in-pile densification experiment carried out in the Halden reactor (Norway). The latter model is mechanistic, i.e. it is based on the solution to a set of equations that describe the coupled temperature and radiation induced phenomena which occur in-pile. These can broadly be broken down into three categories: the fission fragment-pore interaction, the creation of point defects as the fission fragments slow down, and the diffusion

  18. Gas Test Loop Booster Fuel Hydraulic Testing

    International Nuclear Information System (INIS)

    Gas Test Loop Hydraulic Testing Staff

    2006-01-01

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

  19. Gas Test Loop Booster Fuel Hydraulic Testing

    Energy Technology Data Exchange (ETDEWEB)

    Gas Test Loop Hydraulic Testing Staff

    2006-09-01

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

  20. Testing Testing Testing.

    Science.gov (United States)

    Deville, Craig; O'Neill, Thomas; Wright, Benjamin D.; Woodcock, Richard W.; Munoz-Sandoval, Ana; Gershon, Richard C.; Bergstrom, Betty

    1998-01-01

    Articles in this special section consider (1) flow in test taking (Craig Deville); (2) testwiseness (Thomas O'Neill); (3) test length (Benjamin Wright); (4) cross-language test equating (Richard W. Woodcock and Ana Munoz-Sandoval); (5) computer-assisted testing and testwiseness (Richard Gershon and Betty Bergstrom); and (6) Web-enhanced testing…

  1. Analysis of the LBLOCAs in the HANARO pool for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

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

    2004-12-01

    The Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Large Break Loss of Coolant Accidents (LBLOCAs) in HANARO pool for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the LBLOCAs. The location of the pipe break is assumed at the hill taps connecting the cold and hot legs in HANARO pool to the inlet and outlet nozzles of the In-Pile test Section (IPS). Double ended guillotine break is assumed for the large break loss of coolant accidents. The discharge coefficients of 0.1, 0.33, 0.67, 1.0 are investigated for the LBLOCAs. The test fuels for PWR and CANDU test modes are not heated up for the LBLOCAs caused by the double ended guillotine break in the HANARO pool. The reason is that the sufficient emergency cooling water to cool down the test fuels is supplied continuously to the in-pile test section. Therefore the PCTs for the LBLOCAs in the HANARO pool meet the design criterion of commercial PWR fuel that maximum PCT is lower than 1204 .deg. C

  2. Shear bond strength of different adhesives tested in accordance with DIN 13990-1/-2 and using various methods of enamel conditioning.

    Science.gov (United States)

    Richter, C; Jost-Brinkmann, P-G

    2015-03-01

    The purpose of this work was to analyze the shear bond strength (SBS) of different adhesives for orthodontic brackets in accordance with DIN 13990-1/-2, also taking into consideration potential effects arising from different scenarios of enamel conditioning and specimen storage. A total of 390 experiments were performed, with groups of 10 specimens subjected to identical treatments. Three adhesives were tested: Transbond™ XT (3M Unitek, Monrovia, USA), Beauty Ortho Bond (Shofu, Kyoto, Japan), and Fuji Ortho LC (GC Europe, Leuven, Belgium). SBS was evaluated separately at the bracket-adhesive and adhesive-enamel interfaces, as well as the total (enamel-adhesive-bracket) interface. The brackets were metal brackets for upper right central incisors (Discovery® from Dentaurum, Ispringen, Germany). A universal testing machine (Zwick Z010, Ulm, Germany) was used for testing the SBS after 15 min, or after storage in distilled water at 37 °C for 24 h, or after 24 h followed by 500 thermocycles alternating between 5 and 55 °C. Transbond™ XT produced the highest levels of SBS. The least favorable performance was observed with Fuji Ortho LC after enamel conditioning with 10 % polyacrylic acid. Thermocycling did not have a significant influence. Transbond™ XT and Beauty Ortho Bond (but not Fuji Ortho LC) yielded levels of SBS adequate for clinical application (≥ 7 MPa).

  3. Considerations on a PAHR test facility

    International Nuclear Information System (INIS)

    Boenisch, G.; Groetzbach, G.; Heinzel, V.; Kleefeld, K.; Kuechle, M.; Mueller, R.A.; Royl, P.; Schramm, K.; Smidt, D.; Werle, H.

    1976-01-01

    On the basis of a hypothetical core disruptive accident (HCDA) analysis the phenomena of the post accident phase are first identified which require experimental investigations and can only be studied in pile. Then the experimental requests for both debris bed and molten fuel pool studies are specified and grouped into three categories. For two of the categories the requests can be satisfied with loop experiments in thermal reactors. For the third category a 70 cm diameter test bed is needed and here the proposal is to use a flat core fast reactor with the test bed located below the core heated by axial leakage neutrons. Finally a conceptual design for such a reactor is presented where the test bed is loaded into an ex-vessel device and is removable on a carriage to a hot cell building. Maintenance and safety problems are briefly discussed and alternative solutions are mentioned

  4. Verification test for radiation reduction effect and material integrity on PWR primary system by zinc injection

    Energy Technology Data Exchange (ETDEWEB)

    Kawakami, H.; Nagata, T.; Yamada, M. [Nuclear Power Engineering Corp. (Japan); Kasahara, K.; Tsuruta, T.; Nishimura, T. [Mitsubishi Heavy Industries, Ltd. (Japan); Ishigure, K. [Saitama Inst. of Tech. (Japan)

    2002-07-01

    Zinc injection is known to be an effective method for the reduction of radiation source in the primary water system of a PWR. There is a need to verify the effect of Zn injection operation on radiation source reduction and materials integrity of PWR primary circuit. In order to confirm the effectiveness of Zn injection, verification test as a national program sponsored by Ministry of Economy, Trade and Industry (METI) was started in 1995 for 7-year program, and will be finished by the end of March in 2002. This program consists of irradiation test and material integrity test. Irradiation test as an In-Pile-Test managed by AEAT Plc(UK) was performed using the LVR-15 reactor of NRI Rez in Check Republic. Furthermore, Out-of-Pile-Test using film adding unit was also performed to obtain supplemental data for In-Pile-Test at Takasago Engineering Laboratory of NUPEC. Material Integrity test was planned to perform constant load test, constant strain test and corrosion test at the same time using large scale Loop and slow strain extension rate testing (SSRT) at Takasago Engineering Laboratory of NUPEC. In this paper, the results of the verification test for Zinc program at present are discussed. (authors)

  5. A pilot application of the RELAP file to the steady state and transient analysis of a test section inside the BR2 reactor

    International Nuclear Information System (INIS)

    Ferri, M. G.; D'Auria, F.; Forasassi, G.; Giot, M.

    2000-01-01

    BR2 is a material test reactor sited in the Belgian Nuclear Research Centre in Mol. The main research programs carried out in BR2 are related to the safety of nuclear reactor structural materials and fuels, in normal and accidental conditions, plant lifetime evaluation and ageing of components. In this framework, a computer program that allows the performance of detailed, steady state analysis of several kinds of in-pile sections with an axisymmetrical geometry has been developed. Furthermore, comparing its results with those of the well known, extensively used, Relap5/Mod 3.2 code on a test problem has validated this program. This was performed in three steps: 1. modalisation development of a subsystem of a typical in-pile section. 2. steady state analysis and comparison with the above-mentioned program. 3. transient simulation of the same system; the considered transient consists of a loss of coolant flow. (author)

  6. The development of an electrically compensated differential calorimeter for the measurement of in-pile heat evolution (1962); Etude et realisation d'un calorimetre differentiel a compensation electrique pour la mesure des echaufpements en pile (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Ayela, F; Derrien, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1962-07-01

    It has been possible with the calorimeter of which we give the principle and details in the present report, to determine the calorific power produced by nuclear radiation in a given sample using simple electrical measurement. It also renders unnecessary all absolute measurements of temperature, all possible calibration, and the use of the thermal or thermoelectric constants of the constituent elements. The possible uses of the apparatus are fairly varied, as shown by the first in-pile experiments which we have carried out and for which we give results. (authors) [French] Le calorimetre, dont nous decrivons le principe et les details dans le present rapport, nous permet de determiner, par de simples mesures electriques, la puissance calorifique liberee par les rayonnements nucleaires dans un echantillon donne. Il nous affranchit aussi, de toute mesure absolue de temperature, de tout etalonnage prealable, et de l'utilisation des constantes thermiques ou thermoelectriques de ses elements constitutifs. Les possibilites d'emploi de l'appareil sont assez variees comme nous le montrent les premiers essais que nous avons effectues en pile et dont nous donnons les resultats. (auteurs)

  7. Fission gas behaviour and interdiffusion layer growth in in-pile and out-of-pile irradiated U-Mo/Al nuclear fuels

    International Nuclear Information System (INIS)

    Zweifel, Tobias

    2014-01-01

    Worldwide, research and test reactors are to convert their fuel from highly towards lower enriched uranium, among them the FRM II. One prospective fuel is an alloy of uranium and molybdenum (abbr. U-Mo). Test irradiations showed an insufficient irradiation behavior of this new fuel due to the growth of an interdiffusion layer (abbr. IDL) between the U-Mo fuel and the surrounding Al matrix. Furthermore, this layer accumulates fission gases. In this work, heavy ion irradiated U-Mo/Al layer systems were studied and compared to in-reactor irradiated fuel to study the fission gas dynamics. It is demonstrated that the gas behavior is identical for both in-reactor and out-of-reactor approaches.

  8. Instrumentation to Enhance Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  9. Instrumentation to Enhance Advanced Test Reactor Irradiations

    International Nuclear Information System (INIS)

    Rempe, J.L.; Knudson, D.L.; Condie, K.G.; Daw, J.E.; Taylor, S.C.

    2009-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  10. Influência da ciclagem térmica na resistência à flexão de resinas laboratoriais = The effect of thermo-cycling on flexural strength of laboratorial resins

    Directory of Open Access Journals (Sweden)

    Oliveira, Karla Mychellyne Costa

    2007-01-01

    Full Text Available Introdução: Uma nova geração de resinas laboratoriais contendo micro-partículas cerâmicas em sua composição tem sido introduzida no mercado com a finalidade de melhorar suas propriedades mecânicas. Objetivo: O objetivo deste estudo foi avaliar o efeito da termociclagem na resistência à flexão de três resinas laboratoriais. Materiais e método: Foram confeccionadas 48 barras com medidas de 2 mm × 2 mm × 25 mm, utilizando as resinas laboratoriais: G1 - RESILAB MASTER (Wilcos, Brasil, G2 - VITA VM LC (Vita Zanhfabrik, Germany, G3 - VITA ZETA LC (Vita Zanhfabrik, Germany. A metade das barras de cada grupo foi submetida a 3000 ciclos térmicos (5ºC/55ºC ± 1, com banhos de 30 s. Os ensaios mecânicos foram efetuados em máquina de ensaio universal EMIC com capacidade de carga de 100 kg e velocidade de 0,5 mm/min. Resultados: Os dados obtidos foram analisados estatisticamente pelo teste de ANOVA (two-way e Tukey (p < 0,05. As médias (± DP observadas para os grupos sem ciclagem foram: G1(112,6 ± 25,39, G2(116,5 ± 18,51 e G3(136,1 ± 24,34; e para os grupos com ciclagem: G1(91,8 ± 18,86, G2(101,9 ± 20,61 e G3(89,9 ± 23,05. Conclusão: Concluiu-se que a diminuição da resistência à flexão em MPa está diretamente dependente da ciclagem térmica nos grupos testados, sendo que os maiores valores foram obtidos pela resina VITA VM LC. Entretanto, na condição sem ciclagem térmica a VITA ZETA apresentou os maiores valores

  11. RIA testing capability of the transient reactor test facility

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  12. Autoclave Testing on Zirconium Alloy Materials

    International Nuclear Information System (INIS)

    Hoffmann, Petra-Britt; Sell, Hans-Juergen; Garzarolli, Friedrich

    2012-09-01

    The corrosion of Zirconium components like fuel rod claddings and spacer grids is limiting lifetime and duty of these components. In Pressurized and Boiling Water Reactors (PWR and BWR), different corrosion phenomena are of interest. Although in-pile experience is the final proof for a material development, significant experience was gained by autoclave tests, trying to simulate in-pile conditions but reducing time for return of experience by increased temperatures. For PWR application, the uniform corrosion is studied in water at up to 370 deg. C and in high pressure steam at 400 deg. C, and for BWR, the nodular corrosion is studied in high pressure steam at 500-520 deg. C. Particular attention has to be given to the corrosion media, because oxidative traces in the water can significantly affect the corrosion response. An extensive air removal is thus important for all corrosion tests. This links to the different water chemistry conditions that have been investigated as separate effects otherwise difficult to separate under in-pile conditions. Uniform corrosion in 350 deg. C water is usually a cyclic process with repeated rate transitions. In addition, at high exposure times an acceleration of corrosion can occur, e.g. for Zr-Sn alloys with a high Sn content. In 400 deg. C steam, corrosion rate decreases somewhat with increasing time. Uniform corrosion rate of Zr alloys depends on their Sn- and Fe+Cr contents as well as on their annealing parameters with a similar trend as in PWR and on their yield strength, however with an opposite trend compared to BWR conditions. Nodular corrosion of BWR alloys depends on the annealing parameter with a similar trend as in PWR and out-of-reactor also significantly on the Fe+Cr content. The hydrogen pickup fraction (HPUF) depends largely on details of the water chemistry and can particularly depend on autoclave degassing and probably also on autoclave contaminations. Thus any HPUF value from out-of- pile corrosion tests is only

  13. Updated Results of Ultrasonic Transducer Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Daw, Joshua; Palmer, Joe [Idaho National Laboratory, P.O. Box 1625, MS 4112, Idaho Falls, ID, 38415-3840 (United States); Ramuhalli, Pradeep; Keller, Paul; Montgomery, Robert [Pacific Northwest National Laboratory, 902 Battelle Blvd. Richland, WA, 99354 (United States); Chien, Hual-Te [Argonne National Laboratory, 9700 S. Cass Avenue Argonne, IL, 60439 (United States); Tittmann, Bernhard; Reinhardt, Brian [Pennsylvania State University, 212 Earth and Engr. Sciences Building, University Park, PA, 16802 (United States); Kohse, Gordon [Massachusetts Institute of Technology, 77 Massachusetts Ave. Cambridge, MA 02139 (United States); Rempe, Joy [Rempe and Associates, LLC, 360 Stillwater, Idaho Falls, ID 83404 (United States); Villard, J.F. [Commissariat a l' energie atomique et aux energies alternatives, Centre d' etudes de Cadarache, 13108 Saint-Paul-lez-Durance (France)

    2015-07-01

    Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. These efforts are limited by the lack of identified ultrasonic transducer materials capable of long term performance under irradiation test conditions. To address this need, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate the performance of promising magnetostrictive and piezoelectric transducers in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 10{sup 21} n/cm{sup 2}. A multi-National Laboratory collaboration funded by the Nuclear Energy Enabling Technologies Advanced Sensors and Instrumentation (NEET-ASI) program also provided initial support for this effort. This irradiation, which started in February 2014, is an instrumented lead test and real-time transducer performance data are collected along with temperature and neutron and gamma flux data. The irradiation is ongoing and will continue to approximately mid-2015. To date, very encouraging results have been attained as several transducers continue to operate under irradiation. (authors)

  14. 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.

  15. Release of fission products from a fuel rod with an artificial hole through cladding irradiated in an in-pile water loop, (2)

    International Nuclear Information System (INIS)

    Ishiwatari, Nasumi

    1978-11-01

    To make clear the iodine spiking phenomenon from a defective fuel rod into the primary coolant, the fuel rod (UO 2 pellets, with stainless steel sheath) with an artificial pin hole was irradiated in the inpile test section of water loop JMTR.OWL-1. Experimental conditions were depressurization and temperature drop of the primary loop coolant and diameter and position of the pin hole. Iodine 131 and cesium 137 in loop coolant were measured under various coolant conditions. The inventory and translation rate of iodine 131 in fuel rod related to irradiation histories were calculated. The levels of I-131 and Cs-137 released to loop coolant from fuel rod were compared. Comparison of the results with LWRs was made by way of the spiked amount and release rate of iodine 131. (author)

  16. 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%.

  17. Review of WWER fuel and material tests in the Halden reactor

    International Nuclear Information System (INIS)

    Volkov, B.; Kolstad, E.

    2006-01-01

    A review of the tests with WWER fuels and materials conducted in HBWR over the years of cooperation with Russia is presented. The first test with old generation WWER-440 fuel and PWR specification fuel was carried out from 1995 to 1998. Some differences between these fuels regarding irradiation induced densification and pellet design as well as similar fuel thermal behaviour, swelling and FGR were revealed during the test. The data from this test are reviewed and compared with PIE recently performed to confirm the in-pile measurements. The second test was started in March 1999 with the main objective to study different modified WWER fuels also in comparison with PWR fuel. The results indicated that all these modified WWER fuels exhibit improved densification properties relative to earlier tested fuel. In-pile data on fuel densification have been analysed with respect to as fabricated fuel microstructure and can be used for verification of fuel behaviour models. Corrosion and creep tests in the Halden reactor encompass WWER cladding alloys and some results are given. Prospective WWER fuel and material tests foreseen within the frame of the joint program of OECD HRP are also presented. (authors)

  18. Research reactors for power reactor fuel and materials testing - Studsvik's experience

    International Nuclear Information System (INIS)

    Grounes, M.

    1998-01-01

    Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

  19. In-pile tritium permeation experiment

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Miller, L.G.; Watts, K.D.; Kershner, C.J.; Rogers, M.L.

    1982-01-01

    The experiments in progress are examining various aspects of the permeation of hydrogen isotopes through fusion materials. Of particular importance will be the measurement of permeation due to ion implantation in the presence of a neutron radiation field. Theoretical and early experimental results for these experiments have suggested that sufficient tritium will permeate fusion reactor interior structures that development of a permeation barrier will be needed. (orig.)

  20. In-pile tritium permeation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Longhurst, G.R.; Miller, L.G.; Watts, K.D. (Idaho National Engineering Lab., Idaho Falls (USA)); Kershner, C.J.; Rogers, M.L. (Monsanto Research Corp., Miamisburg, OH (USA). Mound Facility)

    The experiments in progress are examining various aspects of the permeation of hydrogen isotopes through fusion materials. Of particular importance will be the measurement of permeation due to ion implantation in the presence of a neutron radiation field. Theoretical and early experimental results for these experiments have suggested that sufficient tritium will permeate fusion reactor interior structures that development of a permeation barrier will be needed.

  1. Boron carbide in pile behaviour Rapsodie experience

    International Nuclear Information System (INIS)

    Kryger, B.; Colin, M.

    1983-04-01

    Results concerning boron carbide irradiation experiments performed in RAPSODIE up to 10 22 .cm - 3 capture density in the temperature range 600-1100 0 lead to the following main conclusions: initial density and grain size lowering contribute to swelling decrease but density is the major parameter for swelling limitation; swelling rate can vary in a wide range (ratio 1 to 3) according to combinations of density (1.8 to 2.3) and grain size (10 to 50 μm) values; a swelling balance reveals that the most important contribution to swelling should be a high density of helium small bubbles (<400 A); helium retention increases with density and grain size and decreases with temperature elevation. A diffusion law is proposed to describe the rate of helium release

  2. Specific methods on in-pile gammametry

    International Nuclear Information System (INIS)

    Pointud, M.-L.; Michel, Francois.

    1979-01-01

    The gammametry technique in nuclear research reactors has evolved by the adequation of its means and the quality of its results since its beginnings in 1972. We do not propose here to make a detailed presentation, nor describe the kinds of well known results that can henceforth be attained with it in a conventional manner; our intention, on the other hand, is to describe a few specific methods developed for using it in the SILOE reactor of the CEN/G [fr

  3. In-pile tritium permeation experiment

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Miller, L.G.; Watts, K.D.; Kershner, C.J.; Rogers, M.L.

    1982-01-01

    To examine radiation and implantation effects simultaneously, an experiment has been designed which makes use of the Coupled Fast Reactivity Measurement Facility (CFRMF), a small pool reactor at the INEL. The neutron flux is low in this reactor, but the high cross section (5300 b) for the 3 He(n,p) 3 H reaction with thermal neutrons gives a sufficiently intense flux of protons and tritons to a simulated fusion first wall for meaningful results

  4. Out-of-pile test of the crud separator system. (I)

    International Nuclear Information System (INIS)

    Takasaki, Akito; Iimura, Katsumichi; Tanaka, Isao

    1991-01-01

    The JMTR Project has been developing the crud separator system since 1981, and the advanced system has been fabricated for the in-pile test to be performed in the HBWR (Norway). The crud separator system removes magnetized crud circulating in a primary circuit by the permanent magnet assembly surrounded inside and outside of the separator vessel. Before the in-pile test, out-of-pile test was carried out in the JMTR Project under the condition of atmospheric pressure and room temperature, and the simplified theoretical analysis for crud separation mechanism was also carried out. The out-of-pile test results suggested that separation factor increased with increasing magnetic susceptibility of crud and crud particle diameter, and decreased with increasing flow rate. These results were in good agreement with the theoretical analysis. The test results also showed that the crud size enlarger was effective in lower separation factor region, which related to lower magnetic susceptibility of crud, smaller crud diameter and higher flow rate. (author)

  5. Tests for validation of fast neutron reactors safety

    International Nuclear Information System (INIS)

    Nagata, T.; Yamashita, H.

    2001-01-01

    Japanese scientific research and design enterprises in cooperation with industrial and power generating corporations implement a project on creating a fast neutron reactor of the ultimate safety. One of the basic expected results from such a development is creation of a reactor core structure that is able to eliminate recriticality occurrence in the course of reactor accident involving fuel melting. One of the possible ways to solve this problem is to include pipes (meant for specifying directed (controlled) molten fuel relocation) into fuel assembly structure. In the course of conduction and subsequent implementation of such a design the basic issue is to experimentally confirm the operating capacity of FA having such a structure and that is called FAIDUS. Within EAGLE Project on experimental basis of IAE NNC RK an activity has been started on preparation and conduction of out-of-pile and in-pile tests. During tests a sodium coolant will be used. Studies are conducted by NNC RK in cooperation with the Japanese corporations JAPC and JNC. Basic objective of out-of-pile tests was to obtain preliminary information on fuel relocation behavior under conditions simulating accident involving melting of core consisting of FAIDUS FA, which will help to clarify simulation criteria and to develop the most optimum structure of the experimental channel for reactor experiments conduction. The basic objective of in-pile tests was the experimental confirmation of operating capacity of FAIDUS FA model under reactor conditions. According to the program two tests are planned to be performed at IGR reactor: tests for validation of fast neutron reactor safety, and out-of-pile tests at EAGLE experimental facility without sodium coolant

  6. Handbook of materials testing reactors and ancillary hot laboratories in the European Community

    International Nuclear Information System (INIS)

    1977-01-01

    The purpose of this Handbook is to make available to those interested in 'in-pile' irradiation experiments important data on Materials Testing Reactors in operation in the European Community. Only thermal reactors having a power output of more than 5 MW(th) are taken into consideration. In particular, detailed technical information is given on the experimental irradiation facilities of the reactors, their specialized irradiation devices (loops and instrumented capsules), and the associated hot cell facilities for post-irradiation examination of samples

  7. Some aspects of in-pile swelling of fissile materials, 1. part: non-alloyed {alpha} uranium; Quelques aspects du gonflement en pile des materiaux fissiles. 1. partie: uranium {alpha} non allie

    Energy Technology Data Exchange (ETDEWEB)

    Mikailoff, H [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-07-01

    An examination has been carried out of non-alloyed uranium samples, having various structural states, cold-worked and recrystallized, as-cast and {beta}-treated, and irradiated at temperatures of between 450 and 600 C and with burn-ups from 1300 to 5500 MW days/metric ton. These samples swelled because of precipitation of the fission gases the porosity thus produced has a morphology depending mainly on the type of deformation to which the metal has been subjected and which is due to in-pile growth. The most homogeneous distribution of pores, and thus that leading to the minimum swelling, is only observed in the material having a marked [010] texture in which the growth and perhaps the thermal cycling introduce little or no strain. For other materials the deformation /swelling association causes a more rapid destruction of the samples either by cracking when the deformation is due to twinning, or by pronounced swelling localized in the bands when deformation is due to slipping. Finally the fission-gas precipitation considerably facilitates, above 500 C, the germination and growth of the intergranular cracks which can then develop at low stresses. (author) [French] On a examine des echantillons d'uranium non allie, de divers etats structuraux, marteles et recristallises, bruts de coulee et traites {beta}, irradies a des temperatures comprises entre 450 et 600 C, et a des taux de combustion allant de 1300 a 5500 MWj/t. Ces echantillons ont gonfle par suite de la precipitation de gaz de fission: la porosite ainsi fournie a une morphologie qui depend principalement des modes de deformation subie par le metal et due a la croissance en pile. La repartition la plus homogene des pores, donc celle qui donnera le gonflement minimum, est observee seulement dans le materiau a forte texture [010] dans lequel la croissance et eventuellement le cyclage thermique introduisent peu ou pas de contraintes. Dans les autres materiaux l'association deformation/gonflement rend plus rapide

  8. Behavior of metallic fuel in treat transient overpower tests

    International Nuclear Information System (INIS)

    Bauer, T.H.; Wright, A.E.; Robinson, W.R.; Klickman, A.E.

    1988-01-01

    Results and analyses are reported for TREAT in-pile transient overpower tests of margin to cladding failure and pre-failure axial expansion of metallic fuel. In all cases the power rise was exponential on an 8 s period until either incipient or actual cladding failure was achieved. Test fuel included EBR-II driver fuel and ternary alloy, the reference fuel of the Intergral Fast Reactor concept. Test pin burnup spanned the widest range available. The nature of the observed cladding failure and resultant fuel dispersals is described. Simple models are presented which describe observed cladding failures and pre-failure axial expansions yet are general enough to apply to all metal fuel types

  9. Testing Systems and Results for Advanced Nuclear Fuel Materials

    International Nuclear Information System (INIS)

    Rooyen, I.J. van; Griffith, G.W.; Garnier, J.E.

    2012-01-01

    Light Water Reactor Sustainability (LWRS) Program Advanced LWR Nuclear Fuel Development (ALFD) Pathway. Development and testing of high performance fuel cladding identified as high priority to support: enhancement of fuel performance, reliability, and reactor safety. One of the technologies being examined is an advanced fuel cladding made from ceramic matrix composites (CMC) utilizing silicon carbide (SiC) as a structural material supplementing a commercial Zircaloy-4 (Zr-4) tube. A series of out-of-pile tests to fully characterize the SiC CMC hybrid design to produce baseline data. The planned tests are intended to either produce quantitative data or to demonstrate the properties required to achieve two initial performance conditions relative to standard zircaloybased cladding: decreased hydrogen uptake (corrosion) and decreased fretting of the cladding tube under normal operating and postulated accident conditions. These two failure mechanisms account for approximately 70% of all in-pile failures of LWR commercial fuel assemblies

  10. Non-destructive test for VHTR fuel using 160kV X-ray system in Hotcell

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jun; Yoo, Boung Ok; Choo, Yong sun; Baik Sang youl; Kim, Hee Moon; Ahn, Sang Bok [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The research for VHTR which is one of the next generation reactor has been actively carried out. As a part of the research for VHTR, an irradiation examination for the VHTR fuel was performed to confirm an in-pile behavior in HANARO. The non-destructive test for the irradiated fuel is very important to understand the in-pile behavior of the fuel. Especially, the X-ray system is useful to observe the fuel shape without destruction. A dimensional change and defect of the fuel can be confirmed thorough the Xray system. Also, using the 3-D software and CT technology, the fuel shape can be intuitionally observed. The 450kV and 160kV X-ray system were installed and operated in IMEF hotcell. The 160kV X-ray system relatively using a low voltage is suitable to a small scale sample. And high resolution images can be obtained. In this study, the non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. Through these test, the possibility for the X-ray inspection of irradiated fuel was confirmed. The non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. The clear images of the irradiated coated particle were produced without the radiation damage during the Xray inspection. The X-ray images of the VHTR fuel will be utilized as the in-pile performance validation data.

  11. Non-destructive test for VHTR fuel using 160kV X-ray system in Hotcell

    International Nuclear Information System (INIS)

    Kim, Young Jun; Yoo, Boung Ok; Choo, Yong sun; Baik Sang youl; Kim, Hee Moon; Ahn, Sang Bok

    2016-01-01

    The research for VHTR which is one of the next generation reactor has been actively carried out. As a part of the research for VHTR, an irradiation examination for the VHTR fuel was performed to confirm an in-pile behavior in HANARO. The non-destructive test for the irradiated fuel is very important to understand the in-pile behavior of the fuel. Especially, the X-ray system is useful to observe the fuel shape without destruction. A dimensional change and defect of the fuel can be confirmed thorough the Xray system. Also, using the 3-D software and CT technology, the fuel shape can be intuitionally observed. The 450kV and 160kV X-ray system were installed and operated in IMEF hotcell. The 160kV X-ray system relatively using a low voltage is suitable to a small scale sample. And high resolution images can be obtained. In this study, the non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. Through these test, the possibility for the X-ray inspection of irradiated fuel was confirmed. The non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. The clear images of the irradiated coated particle were produced without the radiation damage during the Xray inspection. The X-ray images of the VHTR fuel will be utilized as the in-pile performance validation data.

  12. Postirradiation examination results for the Irradiation Effects Scoping Test 2

    International Nuclear Information System (INIS)

    Mehner, A.S.

    1977-01-01

    The postirradiation examination results are reported for two rods from the second scoping test (IE-ST-2) of the Nuclear Regulatory Commission Irradiation Effects Program. The rods were irradiated in the in-pile test loop of the Power Burst Facility at the Idaho National Engineering Laboratory. Rod IE-005 was fabricated from fresh fuel and cladding previously irradiated in the Saxton Reactor. Rod IE-006, fabricated from fresh fuel and unirradiated cladding, was equipped with six developmental cladding surface thermocouples. The rods were preconditioned, power ramped, and then subjected to film boiling operation. The performance of the rods and the developmental thermocouples are evaluated from the post irradiation examination results. The effects of prior irradiation damage in cladding are discussed in relation to fuel rod behavior during a power ramp and subsequent film boiling operation

  13. PIE of test assembly of Qinshan nuclear power plant

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  14. A study for the development of the capsule assembly machine for the re-irradiation test

    International Nuclear Information System (INIS)

    Kang, Y. H.; Kim, J. K.; Yeom, K. Y.; Yoon, K. B.; Choi, M. H.; Kim, B. K.

    2004-01-01

    A series of in-pile tests are being carried out to support the advanced fuel development programs at the HANARO reactor. There are still some limitations for satisfying the test requirements. To meet the demands for the high burnup test at HANARO, new capsule assembling technology is required. This paper describes the design requirements, design and fabrication of the mockup, and pre-operational tests performed for the development of the new capsule assembly machine. The mockup manufactured consists of a base plate, a capsule stand, a capsule guide pipe and clamping device and is 1m in outer diameter, 1.8m in height and 136kg in weight. From the pre-operation tests, the optimum clamping torque was 450kgf·cm for preventing rotation and shaking of the capsule main body during assembling capsule main body and protection tube, and this remote assembling procedure can be applicable to the high burnup test

  15. GRIST-2 preliminary test plan and requirements for fuel fabrication and preirradiation

    International Nuclear Information System (INIS)

    Tang, I.M.; Harmon, D.P.; Torri, A.

    1978-12-01

    The preliminary version of the GRIST-2 test plan has been developed for the planned initial 5 years (1984 to 1989) of TREAT-Upgrade in-pile tests. These tests will be employed to study the phenomenology and integral behavior of GCFR core disruptive accidents (CDAs) and to support the Final Safety Analysis Report (FSAR) CDA analyses for the demonstration plant licensing. The preliminary test plan is outlined. Test Phases I and II are for the fresh fuel (preconditioned or not) CDA behavior at the beginning-of-life (BOL) reactor state. Phase III is for the reactor state that contains irradiated fuel with a saturated content of helium and fission gas. Phase IV is for larger bundle tests and scaling effects

  16. Power-cooling mismatch test series. Test PCM-2A; test results report

    International Nuclear Information System (INIS)

    Cawood, G.W.; Holman, G.W.; Martinson, Z.R.; Legrand, B.L.

    1976-09-01

    The report describes the results of an in-pile experimental investigation of pre- and postcritical heat flux (CHF) behavior of a single 36-inch-long, pressurized water reactor (PWR) type, UO 2 -fueled, zircaloy-clad fuel rod. The nominal coolant conditions for pressure and temperature were representative of those found in a commercial PWR. Nine separate departures from nucleate boiling (DNB) cycles were performed by either of two different methods: (a) decreasing the coolant flow rate while the fuel rod power was held constant, or (b) increasing the fuel rod power while the coolant flow rate was held constant. DNB was obtained during eight of the nine cycles performed. For the final flow reduction, the mass flux was decreased to 6.1 x 10 5 lb/hr-ft 2 at a constant peak linear heat generation rate of 17.8 kW/ft. The fuel rod was allowed to remain in film boiling for about 210 seconds during this final flow reduction. The fuel rod remained intact during the test. Results of on-line measurements of the fuel rod behavior are presented together with discussion of instrument performance. A comparison of the data with Fuel Rod Analysis Program-Transient 2 (FRAP-T2) computer code calculations is included

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

    International Nuclear Information System (INIS)

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

    2004-09-01

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

  18. Aero thermal test results obtained on the n. C 5 EL 4 Cluster in the atmospheric pressure cell

    International Nuclear Information System (INIS)

    Gasc, B.

    1964-01-01

    In the framework of thermal studies on the EL-4 cluster, the full-scale tests at atmospheric pressure are designed to permit measurement of local values of the wall temperature, of the velocity and of the temperature in the fluid. The experimental results, obtained with the help of an original measuring apparatus, make it possible to follow the changes in these values along the cluster and to predict in much detail the in-pile thermal behaviour. In particular it is shown that changes in the wall temperature along the cluster are greatly influenced by disruption of the flow caused by grids and supports. (author) [fr

  19. 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

  20. Characteristics of severely damaged fuel from PBF tests and the TMI-2 accident

    International Nuclear Information System (INIS)

    Osetek, D.J.; Cook, B.A.; Dallman, R.J.; Broughton, J.M.

    1986-01-01

    As a result of the TMI-2 reactor accident, the US Nuclear Regulatory Commission initiated a research program to investigate phenomena associated with severe fuel damage accidents. This program is sponsored by several countries and includes in-pile and out-of-pile experiments, separate effects studies, and computer code development. The principal in-pile testing portion of the program includes four integral severe fuel damage (SFD) tests in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). The INEL is also responsible for examining the damaged core in the Three Mile Island-Unit 2 (TMI-2) reactor, which offers the unique opportunity to directly compare the findings of an experimental program to those of an actual reactor accident. The principal core damage phenomena which can occur during a severe accident are discussed, and examples from the INEL research programs are used to illustrate the characteristics of these phenomena. The preliminary results of the programs are presented, and their impact on plant operability during severe accidents is discussed

  1. Postirradiation examination results for the irradiation effects scoping test 1

    International Nuclear Information System (INIS)

    Mehner, A.S.; Quapp, W.J.; Goetzmann, O.; Hobbins, R.R.

    1976-09-01

    A zircaloy-clad UO 2 fuel rod was operated above its critical heat flux within the in-pile test loop of the Power Burst Facility and later examined in the hot cells. The results of the postirradiation examinations are presented in this report. A Zr-UO 2 reaction at the fuel-cladding interface embrittled nearly as much of the cladding wall thickness as the Zr-water reaction on the exterior. Data on both the internal and external reactions, and cladding and fuel microstructures, are presented. Cladding embrittlement and rod failure are compared with several rod fragmentation criteria, and conclusions concerning fuel rod failure propagation in a power reactor system are made

  2. SHAKING TABLE TESTS ON SEISMIC DEFORMATION OF PILE SUPPORTED PIER

    Science.gov (United States)

    Fujita, Daiki; Kohama, Eiji; Takenobu, Masahiro; Yoshida, Makoto; Kiku, Hiroyoshi

    The seismic deformation characeteristics of a pile supported pier was examined with the shake table test, especially focusing on the pier after its deformation during earthquakes. The model based on the similitude of the fully-plastic moment in piles was prepared to confirm the deformation and stress characteristic after reaching the fully-plastic moment. Moreover, assuming transportation of emergency supplies and occurrence of after shock in the post-disaster period, the pile supported pier was loaded with weight after reaching fully-plastic moment and excited with the shaking table. As the result, it is identified that the displacement of the pile supported pier is comparatively small if bending strength of piles does not decrease after reaching fully-plastic moment due to nonoccourrence of local backling or strain hardening.

  3. Postirradiation examination results for the irradiation effects scoping test 1

    International Nuclear Information System (INIS)

    Mehner, A.S.; Quapp, W.J.; Goetzmann, O.; Hobbins, R.R.

    1976-09-01

    A zircaloy-clad UO 2 fuel rod was operated above its critical heat flux within the in-pile test loop of the Power Burst Facility and later examined in the hot cells. The results of the postirradiation examinations are presented. A Zr-UO 2 reaction at the fuel-cladding interface embrittled nearly as much of the cladding wall thickness as the Zr-water reaction on the exterior. Data on both the internal and external reactions and the cladding and fuel microstructures are presented. Cladding embrittlement and rod failure are compared with several rod fragmentation criteria, and conclusions concerning fuel rod failure propagation in a power reactor system are made

  4. Postirradiation examination results for the irradiation effects scoping test 1

    Energy Technology Data Exchange (ETDEWEB)

    Mehner, A.S.; Quapp, W.J.; Goetzmann, O.; Hobbins, R.R.

    1976-09-01

    A zircaloy-clad UO/sub 2/ fuel rod was operated above its critical heat flux within the in-pile test loop of the Power Burst Facility and later examined in the hot cells. The results of the postirradiation examinations are presented in this report. A Zr-UO/sub 2/ reaction at the fuel-cladding interface embrittled nearly as much of the cladding wall thickness as the Zr-water reaction on the exterior. Data on both the internal and external reactions, and cladding and fuel microstructures, are presented. Cladding embrittlement and rod failure are compared with several rod fragmentation criteria, and conclusions concerning fuel rod failure propagation in a power reactor system are made.

  5. Characteristic test technology for PWR fuel and its components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho; Jeong, Yong Hwan; Park, Sang Yoon; Kim, Kyeng Ho; Nam, Cheol; Baek, Jong Hyuk; Lee, Myung Ho; Choi, Byoung Kwon; Song, Kun Woo; Kang, Ki Won; Kim, Keon Sik; Kim, Jong Hun; Kim, Young Min; Yang, Jae Ho; Song, Kee Nam; Kim, Hyung Kyu; Kang, Heung Seok; Yoon, Kyung Ho; Chun, Tae Hyun; In, Wang Kee; Oh, Dong Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-01-01

    Characteristic tests of fuel assembly and its components being developed in the Advanced LWR Fuel Development Project supported by the mid-long term nuclear R and D program are described in this report. Performance verification of fuel and its components by the characteristic tests are essential to their development. Fuel components being developed in the Advanced LWR Fuel Development Project are zirconium alloy cladding, UO{sub 2} and burnable absorber pellets, spacer grid and top and bottom end pieces. Detailed test plans for those fuel components are described in this report, and test procedures of cladding and pellet are also described in the Appendix. Examples of the described tests are in- and out-of- pile corrosion and mechanical tests such as creep and burst tests for the cladding, in-pile capsule and ramp tests for the pellet, mechanical tests such as strength and vibration, and thermal-hydraulic tests such as pressure drop and critical heat flux for the spacer grid and top and bottom end pieces. It is expected that this report could be used as the standard reference for the performance verification tests in the development of LWR fuel and its components. 11 refs., 9 figs., 2 tabs. (Author)

  6. NRI experimental facility for the testing of irradiation assisted stress corrosion cracking

    International Nuclear Information System (INIS)

    Ruscak, M.; Chvatal, P.; Zamboch, M.

    1998-01-01

    IASCC influencing reactor internals of both BWR and PWR reactors is a complex phenomenon covering influences of material structure, neutron fluence, neutron flux, chemistry of environment, gamma radiation and mechanical stress. To evaluate such degradation, tests should be performed under conditions similar to those in real structure. Nuclear Research Institute has built several experimental facilities in order to be able to test IASCC degradation of materials. Basically, reactor water loops, both PWR and BWR, could be used to model environmental conditions including gamma and neutron irradiation. Pre-irradiation can be done in irradiation channels under well controlled temperature conditions. During the experiment, in-pile conditions can be compared with those out of pile. It enables to clarify pure influence of irradiation. For testing of irradiated specimens, hot cell facility has been developed for slow strain rate tests. The paper will show all above mentioned facilities as well as some of the results observed with them. (author)

  7. LMFBR. Off normal, transient test facilities and programs in the USA

    International Nuclear Information System (INIS)

    Herbst, R.J.

    1985-01-01

    The United States fast breeder reactor development program has included operational transient analyses and experiments to verify the predicted performance of core components. Operational transient testing has focused on off-normal operation during Plant Protection System terminated transient-overpower events. In-pile and out-of-pile tests have been used to simulate predicted thermal and mechanical strain cycles and measure component response. The spectrum of reactivity ramp rates investigated in TOP tests has recently been expanded to include rates of less than $0.1/s. These slow ramp rate studies are being done in cooperation with the Japanese. The US has also cooperated with the UK in the transient testing of Prototype Fast Reactor fuel pins

  8. Prediction of the Long Term Cooling Performance for the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-12-15

    In the long term cooling phase that the emergency cooling water injection ends, the performance of the residual heat removal for the 3-pin fuel test loop has been predicted by a simplified heat transfer model. In the long term cooling phase the residual heat is 1323W for PWR fuel test mode and 1449W for CANDU fuel test mode. The each residual heat is assumed as 2% of the fission power of the test fuel used in the anticipated operational occurrence and design basis accident analyses. The each fission power used for the analyses is 105% of the rated fission power in the normal operation. In the long term cooling phase the residual heat is removed to the HANARO pool through the double pressure vessels of the in-pile test section. Saturate pooling boiling is assumed on the test fuel and condensation heat transfer is expected on the inner wall of the fuel carrier and the flow divider. Natural convection heat transfer on a heated vertical wall is also assumed on the outer wall of the outer pressure vessel. The conduction heat transfer is only considered in the gap between the double pressure vessels charged with neon gas and in the downcomer filled with coolant. The heat transfer rate between the coolant temperature of 152 .deg. C in the in-pile test section and the water temperature of 45 .deg. C in the HANARO pool is predicted as about 1666W. The 152 .deg. C is the saturate temperature of the coolant pressure predicted from the MARS code. The cooling capacity of 1666W is greater than the residual heats of 1323W and 1449W. Consequently the long term cooling performance of the 3-pin fuel test loop is sufficient for the anticipated operational occurrences and design basis accidents.

  9. Comparison of the cladding deformation measured during the Power Burst Facility loss-of-coolant accident in-pile experiments with recent Oak Ridge National Laboratory out-of-pile results

    International Nuclear Information System (INIS)

    Broughton, J.M.; McCardell, R.K.; MacDonald, P.E.

    1981-01-01

    A series of four large break loss-of-coolant accident fuel behavior experiments have been performed in the Power Burst Facility. The results of these experiments are briefly reviewed and compared with results from the ORNL multirod burst test program. The effect of cladding burst temperature and prior irradiation were investigated. The cladding strain of the previously irradiated test rods was more uniformly distributed around the cladding circumference and larger than for similar unirradiated test rods. The ORNL out-of-pile single rod test results are in good agreement with the Power Burst Facility (PBF) test results with unirradiated test rods, and the ORNL out-of-pile, single-rod test results with heated shrouds and the PBF test results with previously irradiated test rods are comparable

  10. Power-Cooling-Mismatch Test Series. Test PCM-2: postirradiation examination

    International Nuclear Information System (INIS)

    Seiffert, S.L.

    1977-03-01

    The report describes the results of the postirradiation examination of four 0.91-m long pressurized water reactor (PWR) type, UO 2 -fueled zircaloy-clad fuel rods tested in an in-pile experimental investigation of pre- and post-departure from nucleate boiling (DNB) behavior of previously unirradiated fuel rods. None of the four fuel rods tested failed during testing or during reactor shut down. Visual and metallographic examination of the individual fuel rods indicated that the length of the high temperature zone of film boiling varied from rod to rod. Two of the four fuel rods showed evidence of damage by film boiling, characterized by oxide spalling and cladding collapse. Metallographic examination of these fuel rods showed internal cladding oxidation resulting from fuel-cladding reaction. Cladding embrittlement by oxidation is evaluated. A comparison of the cladding surface temperatures estimated metallographically for the separate fuel rods with cladding surface temperatures measured during testing and calculated from the Fuel Rod Analysis Program-Transient (FRAP-T) computer code is included

  11. The Assembly and Test of Pressure Vessel for Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kook Nam; Lee, Jong Min; Youn, Young Jung; June, Hyung Kil; Ahn, Sung Ho; Lee, Kee Hong; Kim, Young Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kennedy, Timothy C. [Oregon State University, Corvallis (United States)

    2009-02-15

    The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts: the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.

  12. The Assembly and Test of Pressure Vessel for Irradiation

    International Nuclear Information System (INIS)

    Park, Kook Nam; Lee, Jong Min; Youn, Young Jung; June, Hyung Kil; Ahn, Sung Ho; Lee, Kee Hong; Kim, Young Ki; Kennedy, Timothy C.

    2009-01-01

    The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts: the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature

  13. LEU WWR-M2 fuel assemblies burnable test

    International Nuclear Information System (INIS)

    Kirsanov, G.A.; Konoplev, K.A.; Pikulik, R.G.; Sajkov, Yu. P.; Tchmshkyan, D.V.; Tedoradze, L.V.; Zakharov, A.S.

    2000-01-01

    The results of in-pile irradiation tests of LEU WWR-M2 fuel assemblies with reduced enrichment of fuel are submitted in the report. The tests are made according to the Russian Program on Reduced Enrichment for Research and Test Reactors (RERTR). United States Department of Energy and the Ministry of Atomic Energy of Russian Federation jointly fund this Program. The irradiation tests of 5 WWR-M2 experimental assemblies are carried out at WWR-M reactor of the Petersburg Nuclear Physics Institute (PNPI). The information on assembly design and technique of irradiation tests is presented. In the irradiation tests the integrity of fuel assemblies is periodically measured. The report presents the data for the integrity maintained during the burnup of 5 fuel assemblies up to 45%. These results demonstrate the high reliability of the experimental fuel assemblies within the guaranteed burnup limits specified by the manufacturer. The tests are still in progress; it is planned to test and analyze the change in integrity for burnup of up to 70% - 75% or more. LEU WWR-M2 fuel assemblies are to be offered for export by their Novosibirsk manufacturer. Currently, HEU WWR-M2 fuel assemblies are used in Hungary, Ukraine and Vietnam. LEU WWR-M2 fuel assemblies were designed as a possible replacement for the HEU WWR-M2 fuel assemblies in those countries, but their use can be extended to other research reactors. (author)

  14. The Design and Manufacturing Report of Plug Type Non-Instrumented Rig for Irradiation Test in HANARO OR Hole

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Bang, Je Geon; Lim, Ik Sung; Kim, Sun Ki; Yang, Yong Sik; Song, Kun Woo

    2008-09-15

    This project is developed the plug type non-instrumented irradiation test rig of the advanced nuclear fuel in HANARO for pursuit advanced performance in High Performance Fuel Technology Development as a part Nuclear Mid and Long-term R and D Program. This irradiation rig was confirmed the integrity and HANARO core compatibility by the optimum design and the thermal hydraulic out-pile test in FIVPET. The characteristic of plug type non-instrument rig is to possible irradiation test of variable in-pile condition and reduced the wastes for reusable as function. This plug type non-instrumented rig was satisfied the quality assurance requirements and written out the end of manufacturing report. This plug type non-instrumented rig is adopt to the irradiation test for nuclear fuel irradiation test in HANARO OR hole.

  15. Qualification of an out-of-pile Thermohydraulic test Bench (BETHY) developed to calibrate calorimetric cells under specific JHR experimental conditions

    International Nuclear Information System (INIS)

    De Vita, C.; Brun, J.; Carette, M.; Reynard-Carette, C.; Lyoussi, A.; Fourmentel, D.; Villard, J.F.; Guimbal, P.; Malo, J.Y.

    2013-06-01

    Online in-pile measurement methods are crucial during irradiations in material testing reactors to better understand the behavior of materials under accelerated ageing conditions and of nuclear fuels under high irradiation levels. Thus, the construction of a new Material Testing Reactor such as the Jules Horowitz Reactor (JHR) leads to new research and development programs devoted to innovative instrumentation and measurement methods. The presented works are performed in the framework of the IN-CORE program, 'Instrumentation for Nuclear radiations and Calorimetry Online in Reactor', between CEA and Aix-Marseille University. The program aim is to develop experimental devices and test bench to quantify more precisely the nuclear heating parameter in the JHR experimental channels. This in-pile parameter is usually measured by means of calorimeter or gamma thermometer. This paper focuses on a new out-of-pile test bench called BETHY. This bench was developed to study the response of a differential calorimeter during its preliminary calibration step according to specific thermal and hydraulic conditions occurring inside one type of JHR core channel. The first section of this paper is dedicated to a detailed description of the bench. The second part presents the study of the thermal characteristics established in the bench for two main thermal running modes. The last one concerns the calibration curve of the reference cell of the differential calorimeter in the case of homogenous temperature. (authors)

  16. Thermal Conductivity Analysis and Lifetime Testing of Suspension Plasma-Sprayed Thermal Barrier Coatings

    Directory of Open Access Journals (Sweden)

    Nicholas Curry

    2014-08-01

    Full Text Available Suspension plasma spraying (SPS has become an interesting method for the production of thermal barrier coatings for gas turbine components. The development of the SPS process has led to structures with segmented vertical cracks or column-like structures that can imitate strain-tolerant air plasma spraying (APS or electron beam physical vapor deposition (EB-PVD coatings. Additionally, SPS coatings can have lower thermal conductivity than EB-PVD coatings, while also being easier to produce. The combination of similar or improved properties with a potential for lower production costs makes SPS of great interest to the gas turbine industry. This study compares a number of SPS thermal barrier coatings (TBCs with vertical cracks or column-like structures with the reference of segmented APS coatings. The primary focus has been on lifetime testing of these new coating systems. Samples were tested in thermo-cyclic fatigue at temperatures of 1100 °C for 1 h cycles. Additional testing was performed to assess thermal shock performance and erosion resistance. Thermal conductivity was also assessed for samples in their as-sprayed state, and the microstructures were investigated using SEM.

  17. Test of Flow Characteristics in Tubular Fuel Assembly I - Establishment of test loop and measurement validation test

    International Nuclear Information System (INIS)

    Park, Jong Hark; Chae, H. T.; Park, C.; Kim, H.

    2005-12-01

    Tubular type fuel has been developed as one of candidates for Advanced HANARO Reactor(AHR). It is necessary to test the flow characteristics such as velocity in each flow channels and pressure drop of tubular type fuel. A hydraulic test-loop to examine the hydraulic characteristics for a tubular type fuel has been designed and constructed. It consists of three parts; a) piping-loop including pump and motor, magnetic flow meter and valves etc, b) test-section part where a simulated tubular type fuel is located, and 3) data acquisition system to get reading signals from sensors or instruments. In this report, considerations during the design and installation of the facility and the selection of data acquisition sensors and instruments are described in detail. Before doing the experiment to measure the flow velocities in flow channels, a preliminary tests have been done for measuring the coolant velocities using pitot-tube and for validating the measurement accuracy as well. Local velocities of the radial direction in circular tubes are measured at regular intervals of 60 degrees by three pitot-tubes. Flow rate inside the circular flow channel can be obtained by integrating the velocity distribution in radial direction. The measured flow rate was compared to that of magnetic flow meter. According to the results, two values had a good agreement, which means that the measurement of coolant velocity by using pitot-tube and the flow rate measured by the magnetic flow meter are reliable. Uncertainty analysis showed that the error of velocity measurement by pitot-tube is less than ±2.21%. The hydraulic test-loop also can be adapted to others such as HANARO 18 and 36 fuel, in-pile system of FTL(Fuel Test Loop), etc

  18. The influence of chemistry on severe accident phenomena in integral tests

    International Nuclear Information System (INIS)

    Hobbins, R.R.; Osetek, D.J.; Hagrman, D.L.

    1988-01-01

    The influence of chemical processes on severe accident phenomena in integral tests is reviewed and recommendations for areas of additional work are made. The results reviewed include those from tests conducted in the in-pile facilities at ACRR, PBF, and TREAT and the TMI-2 accident. Progress has been made in understanding the influence of chemistry on important severe accident phenomena such as core melt progression, hydrogen generation, aerosol generation and transport, and fission product release and transport (including revaporization). An example is the chemistry of volatile fission products, especially iodine and tellurium. Areas where understanding is inadequate are also apparent, such as chemical interactions between fission product vapors and aerosols. Influential chemical processes reviewed include oxidation by steam and interactions among control, structural, fuel, fission product, and aerosol materials

  19. Status on the construction of the fuel irradiation test facility

    International Nuclear Information System (INIS)

    Park, Kook Nam; Sim, Bong Shick; Lee, Chung Young; Yoo, Seong Yeon

    2005-01-01

    As a facility to examine general performance of nuclear fuel under irradiation condition in HANARO, Fuel Test Loop(FTL) has been developed which can accommodate 3 fuel pins at the core irradiation hole(IR1 hole) taking consideration user's test requirement. 3-Pin FTL consists of In-Pile Test Section (IPS) and Out-of- Pile System (OPS). Test condition in IPS such as pressure, temperature and the water quality, can be controlled by OPS. 3-Pin FTL Conceptual design was set up in 2001 and had completed detail design including a design requirement and basic Piping and Instrument Diagram (P and ID) in 2004. The safety analysis report was prepared and submitted in early 2005 to the regulatory body(KINS) for review and approval of FTL. In 2005, the development team is going to purchase and manufacture hardware and make a contract for construction work. In 2006, the development team is going to install an FTL system performance test shall be done as a part of commissioning. After a 3-Pin FTL development which is expected to be finished by the 2007, FTL will be used for the irradiation test of the new PWR-type fuel and the usage of HANARO will be enhanced

  20. Out-pile test of non-instrumented capsule for the advanced PWR fuel pellets in HANARO irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Lee, C. B.; Oh, D. S.; Bang, J. K.; Kim, Y. M.; Yang, Y. S.; Jeong, Y. H.; Jeon, H. K.; Ryu, J. S. [KAERI, Taejon (Korea, Republic of)

    2002-05-01

    Non-instrumental capsule were designed and fabricated to irradiate the advanced pellet developed for the high burn-up LWR fuel in the HANARO in-pile capsule. This capsule was out-pie tested at Cold Test Loop-I in KAERI. From the pressure drop test results, it is noted that the flow velocity across the non-instrumented capsule of advanced PWR fuel pellet corresponding to the pressure drop of 200 kPa is measured to be about 7.45 kg/sec. Vibration frequency for the capsule ranges from 13.0 to 32.3 Hz. RMS displacement for non-instrumented capsule of advanced PWR fuel pellet is less than 11.6 {mu}m, and the maximum displacement is less that 30.5 {mu}m. The flow rate for endurance test were 8.19 kg/s, which was 110% of 7.45 kg/s. And the endurance test was carried out for 100 days and 17 hours. The test results found not to the wear satisfied to the limits of pressure drop, flow rate, vibration and wear in the non-instrumented capsule.

  1. Some Remarks on Foundation Pile Testing Procedures

    Science.gov (United States)

    Rybak, Jarosław

    2017-10-01

    This work presents the review of pile capacity testing techniques. In an overview, the key points in pile designing are: determination of the appropriate computational schemes, reliable data on loads and the properties of structural materials (in particular, of the soil mass, which is marked by the greatest variability). The procedure of constructing a pile foundation should include: carrying out soil tests in the scope that ensures safe designing, selecting a piling technology that is relevant both to geotechnical conditions and expected loads, drafting a piling design together with the design of load tests, setting up a testing station for further load tests, static and/or dynamic tests of pile load capacity, preceded by supplementary soil tests when the conditions of test pile installation fail to comply with the design assumptions or when the pile length exceeds the depth of the previously investigated soil, making documentation of load capacity tests (with an additional correction of the piling design), the actual piling (ongoing analysis of pile driving logs and, if necessary, testing the piles’ integrity), drawing up the as-built documentation. Unfortunately, the design is corrected after the load test have been conducted only if the piles fail to show the designed bearing capacity. The designer is then obliged to revise the design assumptions on the basis of tests results. If the test results account for the a greater bearing capacity than necessary and it would be recommendable to limit the extent of the planned (i.e. set out in the contract) piling works, usually neither the contractor nor the designer, nor even the Construction Site Supervisor, acting for the benefit of the Investor, are willing to take on the responsibility for reducing the scope of the piling works. The necessity of conducting additional control tests before and during the implementation of the construction project is often treated by the investors as an attempt at extorting extra

  2. Irradiation facilitates at the advanced test reactor

    International Nuclear Information System (INIS)

    Grover, Blaine S.

    2006-01-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC - formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950's with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens. The paper has the following contents: ATR description and capabilities; ATR operations, quality and safety requirements; Static capsule experiments; Lead experiments; Irradiation test vehicle; In-pile loop experiments; Gas test loop; Future testing; Support facilities at RTC; Conclusions. To summarize, the ATR has a long history in fuel and material irradiations, and will be fulfilling a critical role in the future fuel and material testing necessary to develop the next generation reactor systems and advanced fuel cycles. The

  3. Effect of curing and silanizing on composite repair bond strength using an improved micro-tensile test method.

    Science.gov (United States)

    Eliasson, Sigfus Thor; Dahl, Jon E

    2017-01-01

    Objectives: To evaluate the micro-tensile repair bond strength between aged and new composite, using silane and adhesives that were cured or left uncured when new composite was placed. Methods: Eighty Filtek Supreme XLT composite blocks and four control blocks were stored in water for two weeks and thermo-cycled. Sandpaper ground, etched and rinsed specimens were divided into two experimental groups: A, no further treatment and B, the surface was coated with bis-silane. Each group was divided into subgroups: (1) Adper Scotchbond Multi-Purpose, (2) Adper Scotchbond Multi-Purpose adhesive, (3) Adper Scotchbond Universal, (4) Clearfil SE Bond and (5) One Step Plus. For each adhesive group, the adhesive was (a) cured according to manufacturer's instructions or (b) not cured before repair. The substrate blocks were repaired with Filtek Supreme XLT. After aging, they were serially sectioned, producing 1.1 × 1.1 mm square test rods. The rods were prepared for tensile testing and tensile strength calculated at fracture. Type of fracture was examined under microscope. Results: Leaving the adhesive uncured prior to composite repair placement increased the mean tensile values statistically significant for all adhesives tested, with or without silane pretreatment. Silane surface treatment improved significantly ( p strength values for all adhesives, both for the cured and uncured groups. The mean strength of the control composite was higher than the strongest repair strength ( p strength. Not curing the adhesive before composite placement increased the tensile bond strength.

  4. Temperatures, strains and crack behavior during local thermal shock tests on the RPV-cylinder of the HDR

    International Nuclear Information System (INIS)

    Neubrech, G.E.; Goerner, F.; Siebler, T.

    1987-01-01

    This report summarises and critically discusses the results obtained from thermal shocks locally applied to the inner surface of the RPV-cylinder. This evaluation is based on on-line measurements (temperatures and strains at the RPV-wall during the thermal shock loading, non-destructive-testing), on materials investigations, and on theoretical investigations (finite element calculations, fracture mechanics analyses). The comparison between the corresponding measured and calculated results serves as a basis for subsequent assessments. It was the object of these tests to achieve the following primary aims: - Investigation of the loading conditions produced by local thermal shocks during realistic cooling processes. - A better understanding of the physical processes involved in crack initiation and propagation resulting from thermocyclic loading. - Assessment of non-destructive-testing methods with respect to detection and analysis of cracks as a basis for fracture mechanical evaluations. - Assessment of the reliability of the applied structural analysis methods. - Production of naturally formed deep cracks on the inner surface of the RPV-cylinder by means of excessive cooling processes. (orig./HP)

  5. LIMBO computer code for analyzing coolant-voiding dynamics in LMFBR safety tests

    International Nuclear Information System (INIS)

    Bordner, G.L.

    1979-10-01

    The LIMBO (liquid metal boiling) code for the analysis of two-phase flow phenomena in an LMFBR reactor coolant channel is presented. The code uses a nonequilibrium, annular, two-phase flow model, which allows for slip between the phases. Furthermore, the model is intended to be valid for both quasi-steady boiling and rapid coolant voiding of the channel. The code was developed primarily for the prediction of, and the posttest analysis of, coolant-voiding behavior in the SLSF P-series in-pile safety test experiments. The program was conceived to be simple, efficient, and easy to use. It is particularly suited for parametric studies requiring many computer runs and for the evaluation of the effects of model or correlation changes that require modification of the computer program. The LIMBO code, of course, lacks the sophistication and model detail of the reactor safety codes, such as SAS, and is therefore intended to compliment these safety codes

  6. Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

    2016-08-01

    A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

  7. The Hydrogen Pickup Behavior for Zirconium-based Alloys in Various Out-of-pile Corrosion Test Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Aomi, M.; Etoh, Y.; Ishimoto, S.; Une, K. [Nippon Nuclear Fuel Development, Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki-ken, 311-1313 (Japan); Ito, K. [Global Nuclear Fuel Japan Co., Ltd., 3-1 Uchikawa 2-chome, Yokosuka-shi, Kanagawa-ken, 239-0836 (Japan)

    2009-06-15

    An acceleration of hydrogen absorption in zirconium alloy claddings at high burnups is one of the most important issues limiting the fuel performance from the viewpoint of cladding integrity. In this context, advanced cladding materials with higher corrosion resistant and lower hydrogen absorption properties have been widely searched in various organizations. In this study, four kinds of zirconium-based alloys, whose in-pile data had been acquired [1,2] were subjected to comprehensive out-of-pile corrosion tests with various temperature and atmosphere conditions in order to investigate the correlation between in-pile and out-of-pile corrosion and hydrogen pick-up behavior, i.e. Zry-2, GNF-Ziron (Zry-2-based alloy with {approx}0.25 wt % of Fe), Hi-FeNi Zircaloy (Zry-2-based alloy with {approx}0.25 wt % of Fe and {approx}0.1 wt% Ni), and VB (Zr-based alloy containing Sn, Cr, and {approx}0.5 wt % of Fe). All the alloys were annealed in RXA condition. The out-of-pile corrosion tests were carried out in three different conditions of 400 deg. C steam, 475 deg. C supercritical water, and 290 deg. C LiOH aqueous solution. In addition to these alloys, several Zry-2-based alloys with various iron contents were tested in 290 deg. C LiOH aqueous solution. Among the four corrosion conditions, the 290 deg. C LiOH aqueous solution test well screened the hydrogen pick-up behavior of the alloys. The hydrogen absorption decreased with higher iron contents in the alloys in both the out-of-pile and in-pile conditions. Especially, the distinct suppression of hydrogen absorption was observed for VB with the highest iron content. The similar dependence of iron content on the hydrogen pick-up fraction was also obtained for the Zry-2-based alloys with different iron contents, which were corroded in the 290 deg. C LiOH aqueous solution condition. As for the corrosion behavior in the 290 deg. C LiOH aqueous solution condition, the weight gains of Zry-2, GNF-Ziron and VB followed the 1

  8. Test Architecture, Test Retrofit

    Science.gov (United States)

    Fulcher, Glenn; Davidson, Fred

    2009-01-01

    Just like buildings, tests are designed and built for specific purposes, people, and uses. However, both buildings and tests grow and change over time as the needs of their users change. Sometimes, they are also both used for purposes other than those intended in the original designs. This paper explores architecture as a metaphor for language…

  9. Irradiation capability of Japanese materials test reactor for water chemistry experiments

    International Nuclear Information System (INIS)

    Hanawa, Satoshi; Hata, Kuniki; Chimi, Yasuhiro; Nishiyama, Yutaka; Nakamura, Takehiko

    2012-09-01

    Appropriate understanding of water chemistry in the core of LWRs is essential as chemical species generated due to water radiolysis by neutron and gamma-ray irradiation govern corrosive environment of structural materials in the core and its periphery, causing material degradation such as stress corrosion cracking. Theoretical model calculation such as water radiolysis calculation gives comprehensive understanding of water chemistry at irradiation field where we cannot directly monitor. For enhancement of the technology, accuracy verification of theoretical models under wide range of irradiation conditions, i.e. dose rate, temperature etc., with well quantified in-pile measurement data is essential. Japan Atomic Energy Agency (JAEA) has decided to launch water chemistry experiments for obtaining data that applicable to model verification as well as model benchmarking, by using an in-pile loop which will be installed in the Japan Materials Testing Reactor (JMTR). In order to clarify the irradiation capability of the JMTR for water chemistry experiments, preliminary investigations by water radiolysis / ECP model calculations were performed. One of the important irradiation conditions for the experiments, i.e. dose rate by neutron and gamma-ray, can be controlled by selecting irradiation position in the core. In this preliminary study, several representative irradiation positions that cover from highest to low absorption dose rate were chosen and absorption dose rate at the irradiation positions were evaluated by MCNP calculations. As a result of the calculations, it became clear that the JMTR could provide the irradiation conditions close to the BWR. The calculated absorption dose rate at each irradiation position was provided to water radiolysis calculations. The radiolysis calculations were performed under various conditions by changing absorption dose rate, water chemistry of feeding water etc. parametrically. Qualitatively, the concentration of H 2 O 2 , O 2 and

  10. Postirradiation examination results for the Irradiation Effects Test IE-5

    International Nuclear Information System (INIS)

    Cook, T.F.; Ploger, S.A.; Hobbins, R.R.

    1978-03-01

    The results are presented of the postirradiation examination of four pressurized water reactor type fuel rods which were tested in-pile under a fast power ramp and film boiling operation during Irradiation Effects (IE) Test 5. The major objectives of this test were to evaluate the effects of simulated fission products on fuel rod behavior during a fast power ramp, to determine the effects of high initial internal pressure on a fuel rod during film boiling, and to assess fuel rod property changes that occur during film boiling in a fuel rod with previously irradiated cladding. The overall condition of the rods and changes that occurred in fuel and cladding as a result of the power ramp and film boiling operation, as determined from the postirradiation examination, are reported and analyzed. Effects of the simulated fission products on fuel rod behavior during a power ramp are discussed. The effect of high internal pressure on rod behavior during film boiling is evaluated. Cladding temperatures are estimated at various axial and circumferential locations. Cladding embrittlement by oxidation is also assessed

  11. 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

  12. Development of high temperature reference electrodes for in-pile application: Part I. Feasibility study of the external pressure balanced Ag/AgCl reference electrode (EPBRE) and the cathodically charged Palladium hydrogen electrode

    International Nuclear Information System (INIS)

    Bosch, R.W.; Van Nieuwenhove, R.

    1998-10-01

    The main problems connected with corrosion potential measurements at elevated temperatures and pressures are related to the stability and lifetime of the reference electrode and the correct estimation of the potential related to the Standard Hydrogen Scale (SHE). Under Pressurised Water Reactor (PWR) conditions of 300 degrees Celsius and 150 bar, the choice of materials is also a limiting factor due to the influence of radiation. Investigations on two reference electrodes that can be used under PWR conditions are reported: the cathodically charged palladium hydrogen electrode, and the external pressure balanced silver/silver chloride electrode. Preliminary investigations with the Pd-electrode were focused on the calculation of the required charging time and the influence of dissolved oxygen. High temperature applications are discussed on the basis of results reported in the literature. Investigations with the silver/silver chloride reference electrode mainly dealt with the salt bridge which is necessary to connect the reference electrode with the testing solution. It is shown that the thermal junction potential is independent of the length of the salt bridge. In addition, the high temperature contributes to an increase of the conductivity of the solution, which is beneficial for the salt bridge connection

  13. Experimental apparatus for in-pile studies of: creep of nuclear fuels, and Young's-modulus of structural materials; Dispositifs experimentaux pour etudes en pile du fluage des materiaux combustibles, et du module d'elasticite des materiaux de structure

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, A; Le Bret, P; Alfille, L; Pesenti, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    Creep test under compression: the apparatus described allows to study, in an horizontal beam hole of a research reactor such as EL2, the creep behaviour of nuclear fuel samples under neutron flux. The maximum stress applied on the specimens is a constant compression chooses between 0.200 and 0.400 kg/mm{sup 2} (285 psi and 570 psi). - Young's Modulus measurement: in another horizontal beam hole of such a reactor, an apparatus allows to study the irradiation effect on Young's Modulus of a structural material specimen. (author)Fren. [French] Essai de fluage en compression: l'appareillage decrit permet d'etudier, dans un canal horizontal d'une pile experimentale type EL2, le fluage en compression d'eprouvettes de materiaux fissiles sous le flux de neutrons, sous une contrainte maximum de 500 g/mm{sup 2}. - Mesure du module d'Young: dans un canal identique au precedent, un appareillage permet de suivre l'influence du rayonnement sur le module d'elasticite d'une eprouvette d'un materiau de structure. (auteur)

  14. Development of Start-up and Shutdown Procedure for the HANARO Fuel Test Loop

    International Nuclear Information System (INIS)

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

    2009-06-01

    A start-up and shutdown procedure for the HANARO fuel test loop has been developed. This is a facility for fuel and material irradiation tests. The facility provides experimental conditions similar to the normal operational pressures and temperatures of commercial PWR and CANDU plants. The normal operation modes of the HANARO fuel test loop are classified into loop shutdown, cold stand-by 1, cold stand-by 2, hot stand-by, and hot operation. The operation modes depend on the fission power of test fuels and the coolant temperature at the inlet of the in-pile test section. The HANARO must maintain a shutdown mode if the HANARO fuel test loop is loop shutdown, cold stand-by 1, cold stand-by 2, or hot stand-by. As the HANARO becomes power operation mode, the operation mode of the HANARO fuel test loop comes to hot operation from hot stand-by. The procedure for the HANARO fuel test loop consists of four main parts such as check of initial conditions, stat-up operation procedure, shutdown operation procedure, and check lists for operations. Several hot test operations ensure that the procedure is appropriate

  15. Technology Implementation Plan: Irradiation Testing and Qualification for Nuclear Thermal Propulsion Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rader, Jordan D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    This document is a notional technology implementation plan (TIP) for the development, testing, and qualification of a prototypic fuel element to support design and construction of a nuclear thermal propulsion (NTP) engine, specifically its pre-flight ground test. This TIP outlines a generic methodology for the progression from non-nuclear out-of-pile (OOP) testing through nuclear in-pile (IP) testing, at operational temperatures, flows, and specific powers, of an NTP fuel element in an existing test reactor. Subsequent post-irradiation examination (PIE) will occur in existing radiological facilities. Further, the methodology is intended to be nonspecific with respect to fuel types and irradiation or examination facilities. The goals of OOP and IP testing are to provide confidence in the operational performance of fuel system concepts and provide data to program leadership for system optimization and fuel down-selection. The test methodology, parameters, collected data, and analytical results from OOP, IP, and PIE will be documented for reference by the NTP operator and the appropriate regulatory and oversight authorities. Final full-scale integrated testing would be performed separately by the reactor operator as part of the preflight ground test.

  16. Current and prospective fuel test programmes in the MIR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, A.L.; Burukin, A.V.; Iljenko, S.A.; Ovchinnikov, V.A.; Shulimov, V.N.; Smirnov, V.P. [State Scientific Centre of Russia Research Institute of Atomic Reactors, Ulyanovsk region (Russian Federation)

    2007-07-01

    MIR reactor is a heterogeneous thermal reactor with a moderator and a reflector made of metal beryllium, it has a channel-type design and is placed in a water pool. MIR reactor is mainly designed for testing fragments of fuel elements and fuel assemblies (FA) of different nuclear power reactor types under normal (stationary and transient) operating conditions as well as emergency situations. At present six test loop facilities are being operated (2 PWR loops, 2 BWR loops and 2 steam coolant loops). The majority of current fuel tests is conducted for improving and upgrading the Russian PWR fuel, these tests involve issues such as: -) long term tests of short-size rods with different modifications of cladding materials and fuel pellets; -) further irradiation of power plant re-fabricated and full-size fuel rods up to achieving 80 MW*d/kg U; -) experiments with leaking fuel rods at different burnups and under transient conditions; -) continuation of the RAMP type experiments at high burnup of fuel; and -) in-pile tests with simulation of LOCA and RIA type accidents. Testing of the LEU (low enrichment uranium) research reactor fuel is conducted within the framework of the RERTR programme. Upgrading of the gas cooled and steam cooled loop facilities is scheduled for testing the HTGR fuel and sub-critical water-cooled reactor, correspondingly. The present paper describes the major programs of the WWER high burn-up fuel behavior study in the MIR reactor, capabilities of the applied techniques and some results of the performed irradiation tests. (authors)

  17. Testing Significance Testing

    Directory of Open Access Journals (Sweden)

    Joachim I. Krueger

    2018-04-01

    Full Text Available The practice of Significance Testing (ST remains widespread in psychological science despite continual criticism of its flaws and abuses. Using simulation experiments, we address four concerns about ST and for two of these we compare ST’s performance with prominent alternatives. We find the following: First, the 'p' values delivered by ST predict the posterior probability of the tested hypothesis well under many research conditions. Second, low 'p' values support inductive inferences because they are most likely to occur when the tested hypothesis is false. Third, 'p' values track likelihood ratios without raising the uncertainties of relative inference. Fourth, 'p' values predict the replicability of research findings better than confidence intervals do. Given these results, we conclude that 'p' values may be used judiciously as a heuristic tool for inductive inference. Yet, 'p' values cannot bear the full burden of inference. We encourage researchers to be flexible in their selection and use of statistical methods.

  18. First analysis of AGS0, LT2 and E9 CABRI tests with the new SFR safety code ASTEC-Na

    International Nuclear Information System (INIS)

    Perez-Martin, Sara; Bandini, Giacomino; Matuzas, Vaidas; Buck, Michael; Girault, Nathalie

    2015-01-01

    Within the framework of the European JASMIN project, the ASTEC-Na code is being developed for safety analysis of severe accidents in SFR. In the first phase of validation of the ASTEC-Na fuel thermo-mechanical models three in-pile tests conducted in the CABRI experimental reactor have been selected to be analysed. We present here the preliminary results of the simulation of two Transient Over Power tests and one power ramp test (AGS0, LT2 and E9, respectively) where no pin failure occurred during the transient. We present the comparison of ASTEC-Na results against experimental data and other safety code results for the initial steady state conditions prior to the transient onset as well as for the fuel pin behaviour during the transients. (author)

  19. Oxidation of aluminum alloy cladding for research and test reactor fuel

    Science.gov (United States)

    Kim, Yeon Soo; Hofman, G. L.; Robinson, A. B.; Snelgrove, J. L.; Hanan, N.

    2008-08-01

    The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

  20. Oxidation of aluminum alloy cladding for research and test reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: yskim@anl.gov; Hofman, G.L. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Robinson, A.B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Snelgrove, J.L.; Hanan, N. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2008-08-31

    The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

  1. Seismic testing

    International Nuclear Information System (INIS)

    Sollogoub, Pierre

    2001-01-01

    This lecture deals with: qualification methods for seismic testing; objectives of seismic testing; seismic testing standards including examples; main content of standard; testing means; and some important elements of seismic testing

  2. Schirmer test

    Science.gov (United States)

    Tear test; Tearing test; Dry eye test; Basal secretion test; Sjögren - Schirmer; Schirmer's test ... used when the eye doctor suspects you have dry eye. Symptoms include dryness of the eyes or excessive ...

  3. Pinworm test

    Science.gov (United States)

    Oxyuriasis test; Enterobiasis test; Tape test ... diagnose this infection is to do a tape test. The best time to do this is in ... lay their eggs at night. Steps for the test are: Firmly press the sticky side of a ...

  4. Analysis of a beryllium-copper diffusion joint after HHF test

    International Nuclear Information System (INIS)

    Guiniatouline, R.N.; Mazul, I.V.; Gorodetsky, A.E.; Zalavutdinov, R.Kh.; Rybakov, S.Yu.; Savenko, V.I.

    1996-01-01

    The development of beryllium-copper joints which can withstand relevant ITER divertor conditions is one of the important tasks at the present time. One of the main problems associated with these joints is the intermetallic layers. The strength and life of these joints significantly depend on the width and contents of the intermetallic layers. The objective of this work is to study the diffusion joint of TGP-56 beryllium to OFHC copper after thermal response and thermocyclic tests with a beryllium-copper mock-up. The HHF test was performed on the e-beam facility (EBTS, SNLA). The following methods were used for analysis: roentgenographic analysis, X-ray spectrum analysis and fracture analysis. During the investigation the following studies were undertaken: the analysis of the diffusion boundary layer, which was obtained at the cross-section of one of the tiles, the analysis of the debonded surfaces of several beryllium tiles and corresponding copper parts and the analysis of the upper surface of one of the tiles after HHF tests. The joint roentgenographic and element analyses revealed the following phases in the diffusion zone: Cu 2 Be (∝170 μm), CuBe (∝30 μm), CuBe 2 (∝1 μm) and a solid solution of copper in beryllium. The phases Cu 2 Be, CuBe and the solid solution of copper in beryllium were detected by the quantitative microanalysis and the phases CuBe, CuBe 2 and CuBe, by the roentgenographic analysis. The fracture (origin) is located in the central part of the tiles. This crack was caused by residual stresses and thermal fatigue testing. This analysis provides important data on the joint quality and may be used for all types of joints used for ITER applications. (orig.)

  5. Mobile Sensor System AGaMon for Breath Control: Numerical Signal Analysis of Ternary Gas Mixtures and First Field Tests

    Directory of Open Access Journals (Sweden)

    Rolf Seifert

    2018-01-01

    Full Text Available An innovative mobile sensor system for breath control in the exhaled air is introduced. In this paper, the application of alcohol control in the exhaled air is considered. This sensor system operates semiconducting gas sensor elements with respect to the application in a thermo-cyclic operation mode. This operation mode leads to so-called conductance-over-time-profiles (CTPs, which are fingerprints of the gas mixture under consideration and can be used for substance identification and concentration determination. Especially for the alcohol control in the exhaled air, ethanol is the leading gas component to be investigated. But, there are also other interfering gas components in the exhaled air, like H2 and acetone, which may influence the measurement results. Therefore, a ternary ethanol-H2-acetone gas mixture was investigated. The establishing of the mathematical calibration model and the data analysis was performed with a newly developed innovative calibration and evaluation procedure called ProSens 3.0. The analysis of ternary ethanol-H2-acetone gas samples with ProSens 3.0 shows a very good substance identification performance and a very good concentration determination of the leading ethanol component. The relative analysis errors for the leading component ethanol were in all considered samples less than 9 %. First field test performed with the sensor system AGaMon shows very promising results.

  6. Fabrication of High Temperature and High Pressure Vessel for the Fuel Test

    International Nuclear Information System (INIS)

    Park, Kook Nam; Lee, Jong Min; Sim, Bong Shick; Shon, Jae Min; Ahn, Seung Ho; Yoo, Seong Yeon

    2007-01-01

    The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR and CANDU nuclear power plants has been developed and installed in HANARO, KAERI. It is consisted of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS which is located inside the pool is divided into 3-parts; they are in-pool pipes, IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The localization of the IVA is achieved by manufacturing through local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique of the instrument lines has been checked for its functionality and yield. A IVA has been manufactured by local technique and will be finally tested under out of the high temperature and high pressure test

  7. Preliminary reactor physics calculations for Exxon LWR fuel testing in the power burst facility

    International Nuclear Information System (INIS)

    Olson, W.O.; Nigg, D.W.

    1981-05-01

    The PFB reactor is being considered as an irradiation facility to test LWR fuel rods for Exxon Nuclear Company. Requested test conditions are 18 kW/ft axial peak steady state power in 2.5% initial enrichment, 20,000 MWd/Tu exposed rods. Multigroup transport theory calculations (S/sub n/ and Monte Carlo) showed that this was unattainable in the standard PBF test loop. Thus, a flux multiplier was developed in the form of a Zr-2-clad 0.15-inch thick cylindrical shell of 35% enriched, 88% T.D. UO 2 replacing the flow divider, surrounding the rod within the in-pile tube in PFB. With this flux multiplier installed and assuming an average water density of 0.86 g/cm 3 within the test loop, a Figure of Merit (FOM) for a single-rod test assembly of 0.86 kW/ft-MW +- 5% (at 95% confidence level) was calculated. This FOM is the axial peak linear test rod power per megawatt of reactor power. A reactor power of about 21 megawatts will therefore be required to supply the requested linear test rod axial peak heating rate of 18 kW/ft

  8. Ultrasonic Testing

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyeong Jun; Kuk, Jeong Han

    2002-02-15

    This book introduces ultrasonic testing, which tells of outline of ultrasonic testing, principle of ultrasonic testing, prosperities of ultrasonic waves, radiographic test and ultrasonic test, basic theory on ultrasonic testing, mode conversion, transmission and diffraction, ultrasonic flaw detection and probe, standard test piece and reference test piece, like KS(JIS) ASME and ASTM, classification and properties of ultrasonic testing, straight beam method, angle beam method, ASME SEC.V.Art.5 ASTMA 388 and KS B 0817 Korean industrial standard.

  9. In-pile irradiation of rock-like oxide fuels

    International Nuclear Information System (INIS)

    Nitani, N.; Kuramoto, K.; Yamashita, T.; Nakano, Y.; Akie, H.

    2001-01-01

    Five kinds of ROX fuels were prepared and irradiated using 20% enriched U instead of Pu. Non-destructive and destructive post-irradiation examinations were carried out. FP gas release rates of the particle-dispersed type fuels and homogeneously-blended type fuels were larger than that of the Yttria-stabilized zirconia containing UO 2 single phase fuel. From results of SEM and EPMA, decomposition of the spinel was observed. The decomposition of the spinel is probably avoided by lowering the irradiation temperature, less than 1700 K. The regions suffering the irradiation damage of the particle dispersed type fuels were less than those of the homogeneously-blended type fuels. (author)

  10. In-pile irradiation of rock-like oxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Nitani, N.; Kuramoto, K.; Yamashita, T.; Nakano, Y.; Akie, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2001-07-01

    Five kinds of ROX fuels were prepared and irradiated using 20% enriched U instead of Pu. Non-destructive and destructive post-irradiation examinations were carried out. FP gas release rates of the particle-dispersed type fuels and homogeneously-blended type fuels were larger than that of the Yttria-stabilized zirconia containing UO{sub 2} single phase fuel. From results of SEM and EPMA, decomposition of the spinel was observed. The decomposition of the spinel is probably avoided by lowering the irradiation temperature, less than 1700 K. The regions suffering the irradiation damage of the particle dispersed type fuels were less than those of the homogeneously-blended type fuels. (author)

  11. Synergistic smart fuel for in-pile nuclear reactor measurements

    Energy Technology Data Exchange (ETDEWEB)

    Smith, J.A.; Kotter, D.K. [Idaho National Laboratories, Idaho Falls (United States); Ali, R.A.; Garrett, S.L. [Penn State University, University Park, State College, PA 16801 (United States)

    2013-07-01

    The thermo-acoustic fuel rod sensor developed in this research has demonstrated a novel technique for monitoring the temperature within the core of a nuclear reactor or the temperature of the surrounding heat-transfer fluid. It uses the heat from the nuclear fuel to generate sustained acoustic oscillations whose frequency will be indicative of the temperature. Converting a nuclear fuel rod into this type of thermo-acoustic sensor simply requires the insertion of a porous material (stack). This sensor has demonstrated a synergy with the elevated temperatures that exist within the nuclear reactor using materials that have only minimal susceptibility to high-energy particle fluxes. When the sensor is in operation, the sound waves radiated from the fuel rod resonator will propagate through the surrounding cooling fluid. The frequency of these oscillations is directly correlated with an effective temperature within the fuel rod resonator. This device is self-powered and is operational even in case of total loss of power of the reactor.

  12. European in-pile investigations on fast breeder reactor safety

    International Nuclear Information System (INIS)

    Bailly, J.; Penet, F.; Teague, H.J.

    1977-01-01

    Because of the difficulties faced by the various organisations responsible for the design and safety analysis of fast reactors, in the conception and execution of perfectly representative experiments corresponding to the various hypothetical accidents considered, the European experts in this field have chosen to adopt a more fundamental approach. By the coordination of the efforts in the different countries and pooling the results, it appears that sufficient data can be assembled to permit them to: isolate the influence of the principle parameters on the course of an accident and to have an appreciation of the possibilities for detection and propagation; and validate calculational models of the essential phenomena, and determine the values of the adjustable parameters used in accident codes. The present paper does not propose to give an exhaustive description of the European experimental programmes, nor of the possibilities for exploiting the results in safety analysis. The aim is simply to show with the aid of results already obtained or expected from the immediate programme, that it should be possible to answer certain fundamental questions concerning those accidents at present considered most important. Particular consideration is given to local blockages (SCARABEE, DFR and MOL 7C programmes), unprotected loss of pumps leading to cooling failure, and inlet blockage (SCARABEE and CABRI), and reactivity excursions (CABRI and VIPER). In addition it is planned to study, as part of the SCARABEE programme, the long-term post-accident behaviour of fuel in some simple configurations. It is shown that these various results should lead to a great improvement in our understanding of the progress of the hypothetical accidents taken into account in the design of fast breeder reactors [fr

  13. Analysis of in-pile tritium release experiments

    International Nuclear Information System (INIS)

    Kopasz, J.P.; Tam, S.W.; Johnson, C.E.

    1992-01-01

    The objective of this work is to characterize tritium release behavior from lithium ceramics and develop insight into the underlying tritium release mechanisms. Analysis of tritium release data from recent laboratory experiments with lithium aluminate has identified physical processes which were previously unaccounted for in tritium release models. A new model that incorporates the recent data and provides for release from multiple sites rather than only one site was developed. Calculations of tritium release using this model are in excellent agreement with the tritium release behavior reported for the MOZART experiment

  14. Trichomonas Testing

    Science.gov (United States)

    ... Genetic Tests for Targeted Cancer Therapy Glucose Tests Gonorrhea Testing Gram Stain Growth Hormone Haptoglobin hCG Pregnancy ... With some NAATs, samples collected for testing of gonorrhea and chlamydial infections can also be used to ...

  15. Coombs test

    Science.gov (United States)

    Direct antiglobulin test; Indirect antiglobulin test; Anemia - hemolytic ... No special preparation is necessary for this test. ... There are 2 types of the Coombs test: Direct Indirect The direct ... that are stuck to the surface of red blood cells. Many diseases ...

  16. Ham test

    Science.gov (United States)

    Acid hemolysin test; Paroxysmal nocturnal hemoglobinuria - Ham test; PNH - Ham test ... BJ. In: Chernecky CC, Berger BJ, eds. Laboratory Tests and Diagnostic Procedures . 6th ed. Philadelphia, PA: Elsevier ...

  17. New nucleic acid testing devices to diagnose infectious diseases in resource-limited settings.

    Science.gov (United States)

    Maffert, P; Reverchon, S; Nasser, W; Rozand, C; Abaibou, H

    2017-10-01

    Point-of-care diagnosis based on nucleic acid testing aims to incorporate all the analytical steps, from sample preparation to nucleic acid amplification and detection, in a single device. This device needs to provide a low-cost, robust, sensitive, specific, and easily readable analysis. Microfluidics has great potential for handling small volumes of fluids on a single platform. Microfluidic technology has recently been applied to paper, which is already used in low-cost lateral flow tests. Nucleic acid extraction from a biological specimen usually requires cell filtration and lysis on specific membranes, while affinity matrices, such as chitosan or polydiacetylene, are well suited to concentrating nucleic acids for subsequent amplification. Access to electricity is often difficult in resource-limited areas, so the amplification step needs to be equipment-free. Consequently, the reaction has to be isothermal to alleviate the need for a thermocycler. LAMP, NASBA, HDA, and RPA are examples of the technologies available. Nucleic acid detection techniques are currently based on fluorescence, colorimetry, or chemiluminescence. For point-of-care diagnostics, the results should be readable with the naked eye. Nowadays, interpretation and communication of results to health professionals could rely on a smartphone, used as a telemedicine device. The major challenge of creating an "all-in-one" diagnostic test involves the design of an optimal solution and a sequence for each analytical step, as well as combining the execution of all these steps on a single device. This review provides an overview of available materials and technologies which seem to be adapted to point-of-care nucleic acid-based diagnosis, in low-resource areas.

  18. Irradiation tests of a small-sized motor with radiation resistance

    International Nuclear Information System (INIS)

    Nakamichi, M.; Ishitsuka, E.; Shimakawa, S.; Kan, S.

    2007-01-01

    In the Test Blanket Module (TBM) of the International Thermonuclear Experimental Reactor (ITER), tritium production and release behavior will be studied using neutrons from fusion reactions, as the blanket development for a demonstration (DEMO) reactor. For development of the TBM, in-pile functional tests are planned, including an integrated irradiation experiment of a fusion blanket mock-up for pulsed operation simulating the ITER operation mode, using the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Agency (JAEA).Due to be installed in an irradiation rig, a small-sized motor has to be developed for rotating a neutron absorber with a window to realize the simulated pulse operation. Since degradation of materials of the motor may be caused by radiation damage due to neutron and gamma-ray irradiation, it is important to examine the soundness of the motor materials under the neutron and gamma irradiation.In the present study, a small-sized motor with increased radiation resistance was developed as follows. A design of a commercial alternate current (AC) servomotor was adopted in the base structure, and some components of the motor were replaced by those made of radiation-proof materials, through elimination of organic materials. Polyester-coated wire for field coil and epoxy for fixed resin were replaced by polyimide-coated wire and polysiloxane filled with MgO and Al 2 O 3 , respectively. Furthermore, inorganic lubricant (Mo-based coating of 4 micro meter in thickness) was treated on the surface of a gear, instead of organic (polyphenylether) oil.Radiation-induced degradation of the components of the developed small-sized motor was examined using JMTR and the Japan Research Reactor No.4 (JRR-4) of JAEA. The motor was operating normally up to a gamma-ray dose of 7 x 10 8 Gy, a fast neutron (E>1 MeV) fluence of 2 x 10 21 m -2 and a thermal neutron (E 22 m -2 . The irradiated gamma-ray dose for this motor is about 700 times as high as the operation

  19. Magnesium Test

    Science.gov (United States)

    ... PF4 Antibody Hepatitis A Testing Hepatitis B Testing Hepatitis C Testing HER2/neu Herpes Testing High-sensitivity C-reactive Protein (hs-CRP) Histamine Histone Antibody HIV Antibody and HIV Antigen (p24) HIV Antiretroviral Drug Resistance Testing, Genotypic HIV Viral Load HLA Testing HLA- ...

  20. Lipase Test

    Science.gov (United States)

    ... PF4 Antibody Hepatitis A Testing Hepatitis B Testing Hepatitis C Testing HER2/neu Herpes Testing High-sensitivity C-reactive Protein (hs-CRP) Histamine Histone Antibody HIV Antibody and HIV Antigen (p24) HIV Antiretroviral Drug Resistance Testing, Genotypic HIV Viral Load HLA Testing HLA- ...

  1. Rubella Test

    Science.gov (United States)

    ... PF4 Antibody Hepatitis A Testing Hepatitis B Testing Hepatitis C Testing HER2/neu Herpes Testing High-sensitivity C-reactive Protein (hs-CRP) Histamine Histone Antibody HIV Antibody and HIV Antigen (p24) HIV Antiretroviral Drug Resistance Testing, Genotypic HIV Viral Load HLA Testing HLA- ...

  2. Mono Test

    Science.gov (United States)

    ... PF4 Antibody Hepatitis A Testing Hepatitis B Testing Hepatitis C Testing HER2/neu Herpes Testing High-sensitivity C-reactive Protein (hs-CRP) Histamine Histone Antibody HIV Antibody and HIV Antigen (p24) HIV Antiretroviral Drug Resistance Testing, Genotypic HIV Viral Load HLA Testing HLA- ...

  3. AMA Test

    Science.gov (United States)

    ... PF4 Antibody Hepatitis A Testing Hepatitis B Testing Hepatitis C Testing HER2/neu Herpes Testing High-sensitivity C-reactive Protein (hs-CRP) Histamine Histone Antibody HIV Antibody and HIV Antigen (p24) HIV Antiretroviral Drug Resistance Testing, Genotypic HIV Viral Load HLA Testing HLA- ...

  4. In-core materials testing under LWR conditions in the Halden reactor

    International Nuclear Information System (INIS)

    Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A.

    2002-01-01

    The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

  5. Towards Intelligent Interpretation of Low Strain Pile Integrity Testing Results Using Machine Learning Techniques.

    Science.gov (United States)

    Cui, De-Mi; Yan, Weizhong; Wang, Xiao-Quan; Lu, Lie-Min

    2017-10-25

    Low strain pile integrity testing (LSPIT), due to its simplicity and low cost, is one of the most popular NDE methods used in pile foundation construction. While performing LSPIT in the field is generally quite simple and quick, determining the integrity of the test piles by analyzing and interpreting the test signals (reflectograms) is still a manual process performed by experienced experts only. For foundation construction sites where the number of piles to be tested is large, it may take days before the expert can complete interpreting all of the piles and delivering the integrity assessment report. Techniques that can automate test signal interpretation, thus shortening the LSPIT's turnaround time, are of great business value and are in great need. Motivated by this need, in this paper, we develop a computer-aided reflectogram interpretation (CARI) methodology that can interpret a large number of LSPIT signals quickly and consistently. The methodology, built on advanced signal processing and machine learning technologies, can be used to assist the experts in performing both qualitative and quantitative interpretation of LSPIT signals. Specifically, the methodology can ease experts' interpretation burden by screening all test piles quickly and identifying a small number of suspected piles for experts to perform manual, in-depth interpretation. We demonstrate the methodology's effectiveness using the LSPIT signals collected from a number of real-world pile construction sites. The proposed methodology can potentially enhance LSPIT and make it even more efficient and effective in quality control of deep foundation construction.

  6. In-core materials testing under LWR conditions in the Halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A. [OECD Halden Reactor Project (Norway)

    2002-07-01

    The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

  7. Accident situations tests HTR fuel with the device Kufa

    International Nuclear Information System (INIS)

    Kellerbauer, A. I.; Freis, D.

    2010-01-01

    The ceramic and ceramic-like coating materials in modern high-temperature reactor fuel are designed to ensure mechanical stability and retention of fission products under normal and transient conditions, regardless of the radiation damage sustained in-pile. In hypothetical depressurization and loss-of-forced-circulation (D LOFC) accidents, fuel elements of modular high-temperate reactors are exposed to temperatures several hundred degrees higher than during normal operation, causing increased thermo-mechanical stress on the coating layers. At the Institute for Transuranium Elements of the European Commission, a vigorous experimental program is being pursued with the aim of characterizing the performance of irradiated HTR fuel under such accident conditions. A cold finger device (Kufa), operational in ITUs hot cells since 2006, has been used to perform heating experiments on eight irradiated HTR fuel pebbles from the AVR experimental reactor and from dedicated irradiation campaigns at the High-Flux Reactor in Petten, the Netherlands. Gaseous fission products are collected in a cryogenic charcoal trap, while volatiles,are plated out on a water-cooled condensate plate. A quantitative measurement of the release is obtained by gamma spectroscopy. We highlight experimental results from the Kufa testing as well as the on-going development of new experimental facilities. (Author) 9 refs.

  8. Final report of fuel dynamics Test E7

    International Nuclear Information System (INIS)

    Doerner, R.C.; Murphy, W.F.; Stanford, G.S.; Froehle, P.H.

    1977-04-01

    Test data from an in-pile failure experiment of high-power LMFBR-type fuel pins in a simulated $3/s transient-overpower (TOP) accident are reported and analyzed. Major conclusions are that (1) a series of cladding ruptures during the 100-ms period preceding fuel release injected small bursts of fission gas into the flow stream; (2) gas release influenced subsequent cladding melting and fuel release [there were no measurable FCI's (fuel-coolant interactions), and all fuel motion observed by the hodoscope was very slow]; (3) the predominant postfailure fuel motion appears to be radial swelling that left a spongy fuel crust on the holder wall; (4) less than 4 to 6 percent of the fuel moved axially out of the original fuel zone, and most of this froze within a 10-cm region above the original top of the fuel zone to form the outlet blockage. An inlet blockage approximately 1 cm long was formed and consisted of large interconnected void regions. Both blockages began just beyond the ends of the fuel pellets

  9. Neuropathy Tests

    Science.gov (United States)

    ... LD) Lactoferrin Lactose Tolerance Tests LDL Cholesterol LDL Particle Testing (LDL-P) Lead Legionella Testing Leptin Levetiracetam Lipase ... tests, such as computed tomography (CT) scans or magnetic resonance imaging ... suspected, additional testing may be performed to evaluate heart rate, blood ...

  10. Analyses of the Anticipated Operational Occurrences for the HANARO Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. K.; Sim, B. S.; Chi, D. Y.; Lee, C. Y.; Ahn, S. H

    2007-12-15

    The analyses of anticipated operational occurrences of the HANARO fuel test loop have been carried out by using the MARS/FTL{sub A} code, which is a modified version of the MARS code. A critical heat flux correlation on the three rods with triangular array was implemented in the MARS/FTL{sub A} code. The correlation was obtained from the critical heat fluxes measured at a test section, which is the same geometry of the in-pile test section of the HANARO fuel test loop. The anticipated operational occurrences of the HANARO fuel test loop are the inadvertent closure of the isolation valves, the over-power transient of the HANARO, the stuck open of the safety valves, and the loss of HANARO class IV power. A minimum DNBR (Departure from Nucleate Boiling Ratio) was predicted in the inadvertent closure of the isolation valves. It is indicated that the minimum DNBR of 1.85 is greater than the design limit DNBR of 1.39. The maximum coolant pressure calculated in the anticipated operational occurrences is also less than the 110 percents of the design pressure.

  11. Analyses of the Anticipated Operational Occurrences for the HANARO Fuel Test Loop

    International Nuclear Information System (INIS)

    Park, S. K.; Sim, B. S.; Chi, D. Y.; Lee, C. Y.; Ahn, S. H.

    2007-12-01

    The analyses of anticipated operational occurrences of the HANARO fuel test loop have been carried out by using the MARS/FTL A code, which is a modified version of the MARS code. A critical heat flux correlation on the three rods with triangular array was implemented in the MARS/FTL A code. The correlation was obtained from the critical heat fluxes measured at a test section, which is the same geometry of the in-pile test section of the HANARO fuel test loop. The anticipated operational occurrences of the HANARO fuel test loop are the inadvertent closure of the isolation valves, the over-power transient of the HANARO, the stuck open of the safety valves, and the loss of HANARO class IV power. A minimum DNBR (Departure from Nucleate Boiling Ratio) was predicted in the inadvertent closure of the isolation valves. It is indicated that the minimum DNBR of 1.85 is greater than the design limit DNBR of 1.39. The maximum coolant pressure calculated in the anticipated operational occurrences is also less than the 110 percents of the design pressure

  12. Pharmacogenomic Testing

    Science.gov (United States)

    ... your family Plan for the future Insurance and financial planning Transition for children Emergency preparedness Testing & Services Testing ... Support Genetic Disease Information Find a Support Group Financial Planning Who Should I Tell? Genetic Testing & Counseling Compensation ...

  13. Predictive Testing

    Science.gov (United States)

    ... your family Plan for the future Insurance and financial planning Transition for children Emergency preparedness Testing & Services Testing ... Support Genetic Disease Information Find a Support Group Financial Planning Who Should I Tell? Genetic Testing & Counseling Compensation ...

  14. Syphilis Test

    Science.gov (United States)

    ... Genetic Tests for Targeted Cancer Therapy Glucose Tests Gonorrhea Testing Gram Stain Growth Hormone Haptoglobin hCG Pregnancy ... treated for another sexually transmitted disease , such as gonorrhea Is pregnant, during the first prenatal visit and ...

  15. Randomization tests

    CERN Document Server

    Edgington, Eugene

    2007-01-01

    Statistical Tests That Do Not Require Random Sampling Randomization Tests Numerical Examples Randomization Tests and Nonrandom Samples The Prevalence of Nonrandom Samples in Experiments The Irrelevance of Random Samples for the Typical Experiment Generalizing from Nonrandom Samples Intelligibility Respect for the Validity of Randomization Tests Versatility Practicality Precursors of Randomization Tests Other Applications of Permutation Tests Questions and Exercises Notes References Randomized Experiments Unique Benefits of Experiments Experimentation without Mani

  16. Evaluation of five CAD/CAM materials by microstructural characterization and mechanical tests: a comparative in vitro study.

    Science.gov (United States)

    Sonmez, Nesrin; Gultekin, Pinar; Turp, Volkan; Akgungor, Gokhan; Sen, Deniz; Mijiritsky, Eitan

    2018-01-08

    Polymer infiltrated ceramics and nano-ceramic resins are the new restorative materials which have been developed in order to enhance the adverse properties of glass-matrix ceramics and resin composites. The aim of the present in vitro study was to evaluate the characteristics of various CAD/CAM materials through mechanical, microstructural, and SEM analysis. Five test groups (n = 22) were formed by using the indicated CAD/CAM blocks: VITA Enamic (VITA Zahnfabrik), Lava Ultimate (3 M ESPE), IPS e.max CAD (Ivoclar Vivadent), IPS Empress CAD (Ivoclar Vivadent), and VITA Mark II (VITA Zahnfabrik). Two specimens from each test group were used for XRD and EDS analysis. Remaining samples were divided into two subgroups (n = 10). One subgroup specimens were thermocycled (5 °C to 55 °C, 30s, 10,000 cycles) whereas the other were not. All of the specimens were evaluated in terms of flexural strength, Vickers hardness, and fracture toughness. Results were statistically analyzed using two-way ANOVA, one-way ANOVA, Tukey's HSD, and Student's t tests (α = .05). Fractured specimens were evaluated using SEM. The highest Vickers microhardness value was found for VITA Mark II (p CAD was found to have the highest flexural strength (p CAD was also higher than other tested block materials (p CAD groups. It should be realised that simulated aging process seem to affect ceramic-polymer composite materials more significantly than glass ceramics.

  17. Test chamber

    NARCIS (Netherlands)

    Leferink, Frank Bernardus Johannes

    2009-01-01

    A test chamber for measuring electromagnetic radiation emitted by an apparatus to be tested or for exposing an apparatus to be tested to an electromagnetic radiation field. The test chamber includes a reverberation chamber made of a conductive tent fabric. To create a statistically uniform field in

  18. 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

  19. Tensile testing

    CERN Document Server

    2004-01-01

    A complete guide to the uniaxial tensile test, the cornerstone test for determining the mechanical properties of materials: Learn ways to predict material behavior through tensile testing. Learn how to test metals, alloys, composites, ceramics, and plastics to determine strength, ductility and elastic/plastic deformation. A must for laboratory managers, technicians, materials and design engineers, and students involved with uniaxial tensile testing. Tensile Testing , Second Edition begins with an introduction and overview of the test, with clear explanations of how materials properties are determined from test results. Subsequent sections illustrate how knowledge gained through tensile tests, such as tension properties to predict the behavior (including strength, ductility, elastic or plastic deformation, tensile and yield strengths) have resulted in improvements in materals applications. The Second Edition is completely revised and updated. It includes expanded coverage throughout the volume on a variety of ...

  20. Fuel bundle examination techniques for the Phebus fission product test

    International Nuclear Information System (INIS)

    Blanc, J.Y.; Clement, B.; Hardt, P. von der

    1996-01-01

    The paper develops the non-destructive examinations, with a special emphasis on transmission tomography, performed in the Phebus facility, using a linear accelerator associated with a line scan camera based on PCD components. This particular technique enabled the high level of penetration to be obtained, necessary for this high density application. Spatial resolution is not far from the theoretical limit and the density resolution is often adequate. This technique permitted: 1) to define beforehand the cuts on a precise basis, avoiding a long step-by-step choice as in previous in-pile tests; 2) to determine, at an early stage, mass balance, material relocations (in association with axial gamma spectrometry), and FP distribution, as an input into re-calculations of the bundle events. However, classical cuttings, periscopic visual examinations, macrographies, micrographies and EPMA analyses remain essential to give oxidation levels (in the less degraded zones), phase aspect and composition, to distinguish between materials of identical density, and, if possible, to estimate temperatures. Oxidation resistance of sensors (thermocouples or ultrasonic thermometers) is also traced. The EPMA gives access to the molten material chemical analyses, especially in the molten fuel blockage area. The first results show that an important part of the fuel bundle melted (which was one of the objectives of this test) and that the degradation level is close to TIMI-2 with a molten plug under a cavity surrounded by an uranium-rich crust. In lower and upper areas fuel rods are less damaged. Complementaries between these examination techniques and between international teams involved will be major advantages in the Phebus FPT0 test comprehension. 3 refs, 9 figs

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

    International Nuclear Information System (INIS)

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

    2004-10-01

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

  2. Test quality

    International Nuclear Information System (INIS)

    Hartley, R.S.; Keller, A.E.

    1992-01-01

    Inservice testing of safety-related components at nuclear power plants is performed under the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (the Code). Subsections IWP and IWV of Section 11 of the Code state test method and frequency requirements for pumps and valves, respectively. Tests vary greatly in quality and frequency. This paper explores the concept of test quality and its relationship with operational readiness and preventive maintenance. This paper also considers the frequencies of component testing. Test quality is related to a test's ability to detect degradation that can cause component failure. The quality of the test depends on several factors, including specific parameters measured, system or component conditions, and instrument accuracy. The quality of some currently required tests for check valves, motor-operated valves, and pumps is also discussed. Suggestions are made to improve test quality by measuring different parameters, testing valves under load, and testing positive displacement pumps at high pressure and centrifugal pumps at high flow rate conditions. These suggestions can help to improve the level of assurance of component operational readiness gained from testing

  3. Test quality

    International Nuclear Information System (INIS)

    Hartley, R.S.; Keller, A.E.

    1992-01-01

    This document discusses inservice testing of safety-related components at nuclear power plants which is performed under the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (the Code). Subsections IWP and IWV of Section XI of the Code state test method and frequency requirements for pumps and valves respectively. Tests vary greatly in quality and frequency. This paper explores the concept of test quality and its relationship with operational readiness and preventive maintenance. This paper also considers the frequencies of component testing. Test quality is related to a test's ability to detect degradation that can cause component failure. The quality of the test depends on several factors, including specific parameters measured, system or component conditions, and instrument accuracy. The quality of some currently required tests for check valves, motor-operated valves, and pumps is also discussed. Suggestions are made to improve test quality by measuring different parameters, testing valves under load, and testing positive displacement pumps at high pressure and centrifugal pumps at high flow rate conditions. These suggestions can help to improve the level of assurance of component operational readiness gained from testing

  4. Randomization tests

    National Research Council Canada - National Science Library

    Edgington, Eugene S

    1980-01-01

    .... This book provides all the necessary theory and practical guidelines, such as instructions for writing computer programs, to permit experimenters to transform any statistical test into a distribution-free test...

  5. Laboratory Tests

    Science.gov (United States)

    ... Medical Devices Radiation-Emitting Products Vaccines, Blood & Biologics Animal & ... What are lab tests? Laboratory tests are medical devices that are intended for use on samples of blood, urine, or other tissues ...

  6. Triglycerides Test

    Science.gov (United States)

    ... K. Brunner & Suddarth's Handbook of Laboratory and Diagnostic Tests. 2 nd Ed, Kindle. Philadelphia: Wolters Kluwer Health, Lippincott Williams & Wilkins; c2014. Triglycerides; 491–2 p. Lab Tests ...

  7. Lactate Test

    Science.gov (United States)

    ... Plasma Free Metanephrines Platelet Count Platelet Function Tests Pleural Fluid Analysis PML-RARA Porphyrin Tests Potassium Prealbumin ... ency/article/000391.htm . (2002 January, Updated). Lactate (Liquid) Reagent Set. Pointe Scientific, Inc. [On-line Reagent ...

  8. Electrolytes Test

    Science.gov (United States)

    ... Plasma Free Metanephrines Platelet Count Platelet Function Tests Pleural Fluid Analysis PML-RARA Porphyrin Tests Potassium Prealbumin ... of potassium is found in the plasma , the liquid portion of the blood. Monitoring potassium is important ...

  9. Ferritin Test

    Science.gov (United States)

    ... normal" values. By comparing your test results with reference values, you and your healthcare provider can see if ... along with other iron tests , when a routine complete blood count (CBC) shows that a person's hemoglobin and hematocrit ...

  10. Cholesterol Test

    Science.gov (United States)

    ... artery disease. Other names for a cholesterol test: Lipid profile, Lipid panel What is it used for? If you ... Clinic [Internet]. Mayo Foundation for Medical Education and Research; c1998-2017.Cholesterol Test: Overview; 2016 Jan 12 [ ...

  11. Gonorrhea Test

    Science.gov (United States)

    ... Links Patient Resources For Health Professionals Subscribe Search Gonorrhea Testing Send Us Your Feedback Choose Topic At ... Sources Ask Us Also Known As GC Test Gonorrhea NAAT or NAT Neisseria gonorrhoeae Nucleic Acid Amplification ...

  12. Susceptibility Testing

    Science.gov (United States)

    ... Marker Bicarbonate (Total CO2) Bilirubin Blood Culture Blood Gases Blood Ketones Blood Smear Blood Typing Blood Urea ... hours depending on the method used. There are commercial tests available that offer rapid susceptibility testing and ...

  13. Nationale test

    DEFF Research Database (Denmark)

    2009-01-01

    Professor Sven Erik Nordenbo og centerleder Niels Egelund, begge DPU, i samtale om nationale test.......Professor Sven Erik Nordenbo og centerleder Niels Egelund, begge DPU, i samtale om nationale test....

  14. Nationale Test

    DEFF Research Database (Denmark)

    2009-01-01

    Hvad er egentlig formålet med de nationale test? Bliver eleverne klogere af at blive testet? Og er der en sammenhæng mellem bandekrig og nationale test? Fysisk medie: dpu.dk/tv......Hvad er egentlig formålet med de nationale test? Bliver eleverne klogere af at blive testet? Og er der en sammenhæng mellem bandekrig og nationale test? Fysisk medie: dpu.dk/tv...

  15. VDRL test

    Science.gov (United States)

    ... confirmed with another blood test to make the diagnosis of syphilis. Normal value ranges may vary slightly among different ... broken) Alternative Names Venereal disease research ... laboratory test (VDRL) – serum. In: Chernecky CC, Berger BJ, eds. Laboratory Tests and Diagnostic Procedures . 6th ed. St Louis, MO: Elsevier Saunders; ...

  16. Drug Testing

    Science.gov (United States)

    ... testing, substance abuse testing, toxicology screen, tox screen, sports doping tests What is it used for? Drug screening is used to find out whether or not a person has taken a certain drug or drugs. It ... Sports organizations. Professional and collegiate athletes usually need to ...

  17. HIV Testing

    Science.gov (United States)

    ... Abroad Treatment Basic Statistics Get Tested Find an HIV testing site near you. Enter ZIP code or city Follow HIV/AIDS CDC HIV CDC HIV/AIDS See RSS | ... All Collapse All Should I get tested for HIV? CDC recommends that everyone between the ages of ...

  18. Nondestructive Testing

    Energy Technology Data Exchange (ETDEWEB)

    Berger, Harold [Argonne National Laboratory

    1969-01-01

    A nondestructive test is an examination of an object in any manner which will not impair the future usefulness of the object. This booklet discusses a few basic methods of nondestructive testing, and some of their characteristics. In addition, it discusses possible future methods for nondestructive testing by taking a quick look at some of the methods now under study.

  19. Measuring deformation of Fuel pin in a Nuclear Fuel Test Rig

    Energy Technology Data Exchange (ETDEWEB)

    Heo, S. H.; Yang, T. H.; Hong, J. T.; Joung, C. Y.; Ahn, S. H.; Jang, S. Y.; Kim, J. H. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, an LVDT core for measuring the longitudinal displacement of fuel pellets and clad was designed and produced. A signal processing method for the prepared core was investigated. The Nuclear Fuel Test Rig is used to observe changes in the characteristics of the fuel according to the neutron irradiation at HANARO (High-flux Advanced Neutron Application Reactor), which is a research reactor. Which are the strain and internal temperature of the irradiated nuclear fuel and the internal pressure of fuel due to fission gas, the characteristics of the fuel are measured using various sensors such as a thermocouple, SPND and LVDT. In this study, two shaped LVDT (Linear Variable Differential Transformer) cores for displacement measurements were designed and manufactured in order to measure the displacement of a fuel pellet and cladding tube using LVDT sensors for measuring electrical signals by converting the physical variation such as the force and displacement into a linear motion. In addition, signals from the manufactured LVDT sensor were collected and calibrated. Moreover, a method for obtaining the displacement in the core according to the sensing signal was planned. A derived equation can used to predict the change in the position of core. A following study should be conducted to test the output signal and real variation of out-pile system. For further work, a performance verification is required for an in-pile irradiation test.

  20. Analysis of PHEBUS FPT1 test with IMPACT/SAMPSON code

    International Nuclear Information System (INIS)

    Terada, Masafumi; Ikeda, Takashi; Naitoh, Masanori

    2003-01-01

    IMPACT is a simulation software developed at the Nuclear Power Engineering Corporation, which includes the severe accident analysis code, SAMPSON. SAMPSON consists of twelve modules and is capable of simulating hypothesized severe accidents in LWR. Phebus-FPT1 test, which was selected as the International Standard Problem-46, was analyzed with SAMPSON for the verification of the code. The Phebus-FPT1 test was an integral in-pile experiment for studying mainly degradation of fuel bundle and subsequent FP behavior under a LWR severe accident condition, using irradiated fuel as a source of real FP. The following analyses of the Phebus-FPT1 test, which are also the subjects of the ISP-46, were performed: (1) In-core thermal hydraulics, core degradation and FP release from the fuel, (2) FP gas and aerosol transport in the primary circuit, (3) Thermal hydraulics and FP aerosol physics in the containment and (4) Iodine chemistry in the containment. The analysis results of the thermal hydraulics and core degradation showed good agreement with experimental data, except shroud temperatures which were higher than the experiment. The difference may be due to insufficient modeling of the gap closure in the shroud. FP release from fuel, FP transport rate in the primary circuit, FP aerosol physics and iodine chemistry in the containment were also well predicted. Through the analyses, the modules of SAMPSON used were proved to be capable for evaluating thermal hydraulics and FP behaviors under LWR severe accident conditions

  1. Fuel rod failure during film boiling (PCM-1 test in the PBF)

    International Nuclear Information System (INIS)

    Domenico, W.F.; Stanley, C.J.; Mehner, A.S.

    1978-01-01

    The Power-Cooling-Mismatch (PCM) Test, PCM-1 was conducted in the Power Burst Facility (PFB) in March of 1978. The PCM Test Series is being conducted at the Idaho National Engineering Laboratory by EG and G Idaho, Inc., under contract to the USNRC and is designed to characterize the behavior of nuclear fuel rods operating under conditions of high power or low coolant flow or both leading to departure from nucleate boiling. The PCM-1 test was performed to provide in-pile data for a ''worst case'' PCM incident. The objective of this experiment was to study the behavior of a single pressurized water reactor (PWR) fuel rod subjected to a high-power and low flow environment which would result in cladding failure at full power. The ''worst case'' conditions established for the experiment consisted of a rod peak power of 78.7 kW/m and a coolant mass flux of 1356 kg/s.m 2 . Fuel temperatures at the stipulated operating conditions were such that a significant volume of molten fuel was present when failure occurred which produced a high probability of molten fuel-coolant interaction (MFCI) with the possibility of a vapor explosion

  2. Ultrasonic testing

    Energy Technology Data Exchange (ETDEWEB)

    Song, Sung Jin [Sungkwunkwan Univ., Seoul (Korea, Republic of); Jeong, Hyun Jo [Wonkwang Univ., Iksan (Korea, Republic of)

    2004-02-15

    For the proper performance of ultrasonic testing of steel welded joints, and anisotropic material it is necessary to have sound understanding on the underlying physics. To provide such an understanding, it is beneficial to have simulation tools for ultrasonic testing. In order to address such a need, we develop effective approaches to simulate angle beam ultrasonic testing with a personal computer. The simulation is performed using ultrasonic measurement models based on the computationally efficient multi-Gaussian beams. This reach will describe the developed ultrasonic testing models together with the experimental verification of their accuracy.

  3. Troponins Test

    Science.gov (United States)

    ... lab's website in order to provide you with background information about the test(s) you had performed. You will need to return ... C, Banfi G, Guidi GC. Influence of a half-marathon run on NT-proBNP and troponin ... MedlinePlus Medical Encyclopedia [On-line information]. Available online ...

  4. Insulin Test

    Science.gov (United States)

    ... hours before blood is collected, but occasionally a health practitioner may do testing when fasting is not possible, such as when a glucose tolerance test (see Glucose ) is done. In some cases, the health practitioner may request that a person fast longer ...

  5. Genetic Testing

    Science.gov (United States)

    ... is often the only way to determine if symptoms could possibly be related to celiac disease. For a person who faces this situation, a negative gene test would indicate that symptoms are not the result of celiac disease. A positive gene test, however, does not diagnose ...

  6. Potassium test

    Science.gov (United States)

    ... hyperkalemia ) may be due to: Addison disease (rare) Blood transfusion Certain medicines Crushed tissue injury Hyperkalemic periodic paralysis ... released. This may cause a falsely high result. Alternative Names Hypokalemia test; K+ Images Blood test References Mount DB. Disorders of potassium balance. ...

  7. Nationale test

    DEFF Research Database (Denmark)

    Bundsgaard, Jeppe; Puck, Morten Rasmus

    Nationale test skubber undervisning i en forkert retning. Det er lærerne og skolelederne enige om. Men særligt skolelederne ser også muligheder for at bruge testen til at få viden om elevernes faglige kompetencer og om undervisningen. Det kommer til udtryk i rapporten Nationale test: Danske lærere...

  8. Oedometer Tests

    DEFF Research Database (Denmark)

    Thorsen, Grete

    1996-01-01

    The paper describes the results of oedometer tests carried out with samples from Eemian fresh-water deposits and the methods used to determine the preconsolidation pressure from the test results. The influence of creep in the material on the apparent preconsolidation pressure is estimated from...

  9. TORCH Test

    Science.gov (United States)

    ... Parietal Cell Antibody Partial Thromboplastin Time (PTT, aPTT) Parvovirus B19 Pericardial Fluid Analysis Peritoneal Fluid Analysis Pertussis Tests ... enterovirus, Epstein-Barr virus , varicella-zoster virus , and parvovirus B19 . How is the sample collected for testing? A ...

  10. Study of fast reactor safety test facilities. Preliminary report

    International Nuclear Information System (INIS)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods

  11. Automating Test Activities: Test Cases Creation, Test Execution, and Test Reporting with Multiple Test Automation Tools

    OpenAIRE

    Loke Mun Sei

    2015-01-01

    Software testing has become a mandatory process in assuring the software product quality. Hence, test management is needed in order to manage the test activities conducted in the software test life cycle. This paper discusses on the challenges faced in the software test life cycle, and how the test processes and test activities, mainly on test cases creation, test execution, and test reporting is being managed and automated using several test automation tools, i.e. Jira, ...

  12. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; J. E. Daw

    2011-03-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this review, recommendations were made with respect to what instrumentation is needed at the ATR; and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

  13. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; R. Schley; J. Palmer; K. Condie

    2014-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.

  14. Development of a method of testing irradiation devices by gamma scanning inside a research nuclear reactor

    International Nuclear Information System (INIS)

    Michel, Francois.

    1975-01-01

    A tridimensional experiment of spectrometry of an irradiation device located inside the reactor Siloe at a place shielded against spurious radiations, is exposed. The automatic scanning system that was developed, makes it possible to fully analyze in about 24 hours, the irradiation device (fuel pin, coolant and structures). The process combined with a 'pre-processing' program allows first partial results to be simultaneously obtained, more refined results being achieved during the next week, using the 'heavy processing'. The irradiation of the device is only interrupted during the compelling duty shutdown of the reactor, the evolution of the device during the successive irradiation cycles being thus followed up without pertubing said evolution. The reproducibility was studied at a local stage for testing the collimation of the detection unit (1% accuracy) and for testing the whole set 'processing measurement and computation' (5% accuracy). The sensitivity has been illustrated by detecting and measuring local singularities inside fuel (pellets), determining the detection efficiency dependence on the radial distribution of fission products and measuring nucleides inside the coolant flow (limiting value 10 -7 ). The accuracy of the method is evaluated at 5% for relative measurement of an experimental device during its in-pile lifetime and as 10% for quantitative absolute measurements [fr

  15. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    International Nuclear Information System (INIS)

    Rempe, J.L.; Knudson, D.L.; Daw, J.E.

    2011-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this review, recommendations were made with respect to what instrumentation is needed at the ATR; and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program's strategy and initial progress toward accomplishing program objectives. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

  16. Analysis of the LBLOCAs in the room 1 for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

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

    2004-12-01

    Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Large Break Loss of Coolant Accidents (LBLOCAs) in the Room 1 for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the LBLOCAs. The location of the pipe break is assumed at the downstream of the main cooling water pump and the upstream of the main cooler in the room 1. Double ended guillotine break is assumed for the large break loss of coolant accidents. The discharge coefficients of 0.1, 0.33, 0.67, 1.0 are investigated for the LBLOCAs. The maximum Peak Cladding Temperature (PCT) is predicted to be about 734.7 .deg. C for the PWR fuel test mode and 850.4 .deg. C for the CANDU fuel test mode respectively. The maximum peak cladding temperatures meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C

  17. The Jules Horowitz reactor, a new high performance European material testing reactor open to international users: present status and objectives

    International Nuclear Information System (INIS)

    Iracane, D.; Bignan, G.

    2010-01-01

    The development of nuclear power as a sustainable and competitive energy source will continue to require research and development of fuel and material behaviour under irradiation. This necessitates a high performance material testing reactor (MTR). Facing the obsolescence of most of the existing MTR in Europe, France decided a few years ago the construction of the RJH (Jules Horowitz reactor). RJH is designed, built and will be operated as an international user facility. A first set of experimental hosting devices is being designed. For instance, there are the in-core CALIPSO Nak integrated loop for material studies and other loops for fuel studies under nominal or off-normal or accidental conditions. The RJH international program will focus on the following subjects: -) fuel reliability, assessed through power ramps tests and post-irradiation examination; -) Loss of coolant tests done out-of-pile in a first phase and in-pile in a possible second phase; and -) source term tests addressing fission products release. The paper reports also the point of view of VATTENFALL (a Swedish power utility), as a potential European RJH user. (A.C.)

  18. Trypsinogen test

    Science.gov (United States)

    ... care provider about the meaning of your specific test results. What Abnormal Results ... vary in size, so it may be harder to get a blood sample from one person than another. Other risks associated with having ...

  19. Just testing

    International Nuclear Information System (INIS)

    Robinson, D.

    1985-01-01

    In the 1950s, most of the men who witnessed Britain's nuclear tests at Christmas Island in the Pacific were national servicemen, aged 19 or 20. Some revelled in sun and swimming, some were bored, some were too busy to be bored. How many of the twenty thousand servicemen involved in the tests suspected that they might be exposed to radiation that would reveal itself a generation later in blood and bone cancers, sterility, cataracts, or deformities in their children. The Ministry of Defence insists nobody was in danger. This book tells a different story, in the words of the servicemen, and of their medical reports, about secrets which are no longer Official. It is important not only to the victims of a government's extraordinary reluctance to face up to the tragic consequences of a programme of nuclear tests on Christmas Island and in Australia, but to anyone concerned with the damage that nuclear testing is still doing in the world today. (author)

  20. Test Ship

    Data.gov (United States)

    Federal Laboratory Consortium — The U. S. Navy dedicated the decommissioned Spruance Class destroyer ex-PAUL F. FOSTER (EDD 964), Test Ship, primarily for at sea demonstration of short range weapon...

  1. Injection Tests

    CERN Document Server

    Kain, V

    2009-01-01

    The success of the start-up of the LHC on 10th of September was in part due to the preparation without beam and injection tests in 2008. The injection tests allowed debugging and improvement in appropriate portions to allow safe, efficient and state-of-the-art commissioning later on. The usefulness of such an approach for a successful start-up becomes obvious when looking at the problems we encountered before and during the injection tests and could solve during this period. The outline of the preparation and highlights of the different injection tests will be presented and the excellent performance of many tools discussed. A list of shortcomings will follow, leading to some planning for the preparation of the run in 2009.

  2. Fungal Tests

    Science.gov (United States)

    ... Wu, A. (2006). Tietz Clinical Guide to Laboratory Tests, Fourth Edition. Saunders Elsevier, St. Louis, Missouri. Pp 1569, 1570, 1532, 1616. Forbes, B. et. al. (© 2007). Bailey & Scott's Diagnostic Microbiology, Twelfth Edition: Mosby Elsevier Press, St. Louis, Missouri. ...

  3. PTH Test

    Science.gov (United States)

    ... Gregory C. Sephel PhD FACB MT(ASCP). Lab Tests Online adjunct board member. Director Clinical Pathology, VA TN Valley Healthcare System; Associate Professor Pathology, Microbiology, Immunology, Vanderbilt University School of Medicine. Sources Used ...

  4. Phosphorus Test

    Science.gov (United States)

    ... be seen with increased intake of the mineral, hypocalcemia , and with kidney dysfunction. Someone with a mild ... usually detected because of the relationship with and effect on calcium levels . Calcium is routinely tested as ...

  5. Procalcitonin Test

    Science.gov (United States)

    ... normal" values. By comparing your test results with reference values, you and your healthcare provider can see if ... g., blood culture , urine culture ), lactate , blood gases , complete blood count (CBC) , and cerebrospinal fluid (CSF) analysis . When is ...

  6. RSV Test

    Science.gov (United States)

    ... Sex Hormone Binding Globulin (SHBG) Shiga toxin-producing Escherichia coli Sickle Cell Tests Sirolimus Smooth Muscle Antibody (SMA) ... University Medical Center, LAB Letter [On-line newsletter]. PDF available for download at http://www.stanfordhospital.com/ ...

  7. Prenatal Tests

    Science.gov (United States)

    ... tests are considered routine — that is, almost all pregnant women receiving prenatal care get them. They include things like checking urine (pee) levels for protein, sugar, or signs of infection. Other non-routine ...

  8. Progesterone Test

    Science.gov (United States)

    ... replacement therapy, or help diagnose the cause of abnormal uterine bleeding When To Get Tested? At specific times during ... receiving progesterone replacement therapy; when a woman has abnormal uterine bleeding Sample Required? A blood sample drawn from a ...

  9. Testosterone Test

    Science.gov (United States)

    ... Testing Triglycerides Troponin Tryptase Tumor Markers Uric Acid Urinalysis Urine Albumin and Albumin/Creatinine Ratio Urine Culture ... 2010) 2536–259. Centers for Disease Control Hormone Standardization website(HoST). Available online at http://www.cdc. ...

  10. Pertussis Tests

    Science.gov (United States)

    ... of Conditions Not Listed? Not Listed? Acidosis and Alkalosis Adrenal Insufficiency and Addison Disease Alcoholism Allergies Alzheimer ... tested? Pertussis, commonly called whooping cough, is a respiratory infection caused by the bacteria Bordetella pertussis . These ...

  11. Tests computarizados

    Directory of Open Access Journals (Sweden)

    R. Fernando Prialé Z.

    1983-12-01

    Full Text Available En primer lugar, se considera el impacto de las microcomputadoras en la actualidad, viéndolo como un hecho social destinado a traer profundos cambios: nos orientamos hacia una cultura informática cuyo signo es la posibilidad de tratar grandes cantidades de información. En segundo lugar; se analiza brevemente la importancia de los tests en el desarrollo de la psicología. Finalmente, se discute la posibilidad de aplicar la informática a la psicometría con el ejemplo del test de BARSIT.   The impact of microcomputers is discussed as a cultural fact that will bring profound changes in the near future: a society with an ubiquous capacity for treating big amounts of information. The importance of tests for the development of psychology is then analysed. Finaly, the possibility of applying microcomputers to psychometry is discussed trough a concrete example: The BARSIT test.

  12. VMA Test

    Science.gov (United States)

    ... spasms and rapid eye movements referred to as "dancing eyes, dancing feet." The VMA test may also be ordered ... a pheochromocytoma or neuroblastoma? People with a family history of multiple endocrine neoplasia (MEN2), a genetic condition ...

  13. Neuropsychological testing.

    Science.gov (United States)

    Zucchella, Chiara; Federico, Angela; Martini, Alice; Tinazzi, Michele; Bartolo, Michelangelo; Tamburin, Stefano

    2018-06-01

    Neuropsychological testing is a key diagnostic tool for assessing people with dementia and mild cognitive impairment, but can also help in other neurological conditions such as Parkinson's disease, stroke, multiple sclerosis, traumatic brain injury and epilepsy. While cognitive screening tests offer gross information, detailed neuropsychological evaluation can provide data on different cognitive domains (visuospatial function, memory, attention, executive function, language and praxis) as well as neuropsychiatric and behavioural features. We should regard neuropsychological testing as an extension of the neurological examination applied to higher order cortical function, since each cognitive domain has an anatomical substrate. Ideally, neurologists should discuss the indications and results of neuropsychological assessment with a clinical neuropsychologist. This paper summarises the rationale, indications, main features, most common tests and pitfalls in neuropsychological evaluation. © Article author(s) (or their employer(s) unless otherwise stated in the text of the article) 2018. All rights reserved. No commercial use is permitted unless otherwise expressly granted.

  14. Testing theories

    International Nuclear Information System (INIS)

    Casten, R F

    2015-01-01

    This paper discusses some simple issues that arise in testing models, with a focus on models for low energy nuclear structure. By way of simplified examples, we illustrate some dangers in blind statistical assessments, pointing out especially the need to include theoretical uncertainties, the danger of over-weighting precise or physically redundant experimental results, the need to assess competing theories with independent and physically sensitive observables, and the value of statistical tests properly evaluated. (paper)

  15. The analysis of beryllium-copper diffusion joint after HHF test

    International Nuclear Information System (INIS)

    Guiniatouline, R.N.; Mazul, I.V.; Rubkin, S.Y.

    1995-01-01

    The development of beryllium-copper joints which can withstand to relevant ITER divertor conditions is one of the important tasks at present time. One of the main problem for beryllium-copperjoints, is the inter-metallic layers, the strength and life time of joints significantly depends from the width and contents of the intermetallic layers. The objective of this work is to study the diffusion joint of TGP-56 beryllium to OFHC copper after thermal response and thermocyclic tests with beryllium-copper mockup. The BEY test were performed at e-beam facility (EBTS, SNLA). The following methods were used for analyses: the roentgenographic analysis; X-ray spectrum analysis; the fracture graphic analysis. During the investigation the followed studies were done: the analysis of diffusion boundary Be-Cu, which was obtained at the crossection of one of the tiles, the analysis of the debonded surfaces of a few beryllium tiles and corresponding copper parts; the analysis of upper surface of one of the tiles after HHF tests. The results of this work have showed that: the joint roentgenographic and elements analyses indicated the following phases in the diffusion zone: Cu 2 Be (∼170 μm), CuBe (∼30μm), CuBe 2 (∼1 μm) and solid solution of copper in beryllium. The phases Cu 2 Be, CuBe and solid solution of copper in beryllium were indicated using quantitative microanalysis and phases CuBe, CuBe 2 , Cu, Be - by roentgenographic analysis; the source of fracture (initial crack) is located in the central part of the tiles, the crack caused by the influence of residual stresses during cooling of a mock-up after fabrication and developed under the conditions of slow elastic-plastic growing during the process of thermal fatigue testing. The analysis gives the important data about joint's quality and also may be used for any type of joints and its comparison for ITER applications

  16. Allergy testing - skin

    Science.gov (United States)

    Patch tests - allergy; Scratch tests - allergy; Skin tests - allergy; RAST test; Allergic rhinitis - allergy testing; Asthma - allergy testing; Eczema - allergy testing; Hayfever - allergy testing; Dermatitis - allergy testing; Allergy testing; ...

  17. Adaptive test

    DEFF Research Database (Denmark)

    Kjeldsen, Lars Peter; Eriksen, Mette Rose

    2010-01-01

    Artikelen er en evaluering af de adaptive tests, som blev indført i folkeskolen. Artiklen sætter særligt fokus på evaluering i folkeskolen, herunder bidrager den med vejledning til evaluering, evalueringsværktøjer og fagspecifkt evalueringsmateriale.......Artikelen er en evaluering af de adaptive tests, som blev indført i folkeskolen. Artiklen sætter særligt fokus på evaluering i folkeskolen, herunder bidrager den med vejledning til evaluering, evalueringsværktøjer og fagspecifkt evalueringsmateriale....

  18. Tests computarizados

    OpenAIRE

    Prialé Z., R. Fernando

    1983-01-01

    En primer lugar, se considera el impacto de las microcomputadoras en la actualidad, viéndolo como un hecho social destinado a traer profundos cambios: nos orientamos hacia una cultura informática cuyo signo es la posibilidad de tratar grandes cantidades de información. En segundo lugar; se analiza brevemente la importancia de los tests en el desarrollo de la psicología. Finalmente, se discute la posibilidad de aplicar la informática a la psicometría con el ejemplo del test de BARSIT.   The im...

  19. Testing Hubbert

    International Nuclear Information System (INIS)

    Brandt, Adam R.

    2007-01-01

    The Hubbert theory of oil depletion, which states that oil production in large regions follows a bell-shaped curve over time, has been cited as a method to predict the future of global oil production. However, the assumptions of the Hubbert method have never been rigorously tested with a large, publicly available data set. In this paper, three assumptions of the modern Hubbert theory are tested using data from 139 oil producing regions. These regions are sub-national (United States state-level, United States regional-level), national, and multi-national (subcontinental and continental) in scale. We test the assumption that oil production follows a bell-shaped curve by generating best-fitting curves for each region using six models and comparing the quality of fit across models. We also test the assumptions that production over time in a region tends to be symmetric, and that production is more bell-shaped in larger regions than in smaller regions

  20. Prealbumin Test

    Science.gov (United States)

    ... T. J. (© 2007). Mosby's Diagnostic and Laboratory Test Reference 8th Edition: Mosby, Inc., Saint Louis, MO. Pp 755-756. Clarke, W. and Dufour, D. R., Editors (© 2006). Contemporary Practice in Clinical Chemistry: AACC Press, Washington, DC. Pp 197. Banh, L. ( ...

  1. Genomic Testing

    Science.gov (United States)

    ... this database. Top of Page Evaluation of Genomic Applications in Practice and Prevention (EGAPP™) In 2004, the Centers for Disease Control and Prevention launched the EGAPP initiative to establish and test a ... and other applications of genomic technology that are in transition from ...

  2. Fibrinogen Test

    Science.gov (United States)

    ... thrombotic episode ) As a follow-up to an abnormal bleeding disorder test ( prothrombin time, PT or partial thromboplastin time, ... to help diagnose disseminated intravascular coagulation (DIC) or abnormal ... status of a progressive disease (such as liver disease ) over time or, rarely, ...

  3. (stress) testing

    African Journals Online (AJOL)

    However, maximal HR was significantly higher in all groups during their sporting activities than during stress testing in the laboratory (P < 0.01). Conclusions. Maximal HR in veteran athletes during specific sporting activities was significantly higher than that attained during a routine sECG. This finding was not sport-specific, ...

  4. Trypsinogen Test

    Science.gov (United States)

    ... exocrine.html. UPCMD (1998 – 2002). Cystic Fibrosis. University Pathology Consortium, LLC [On-line information]. Available online at http://www.upcmd.com/dot/examples/00218/description.html. Sainato, D., (2002, March). Genetic Testing for CF Going Mainstream? Clinical Laboratory ...

  5. Chymotrypsin Test

    Science.gov (United States)

    ... exocrine.html. UPCMD (1998 – 2002). Cystic Fibrosis. University Pathology Consortium, LLC [On-line information]. Available online at http://www.upcmd.com/dot/examples/00218/description.html. Sainato, D., (2002, March). Genetic Testing for CF Going Mainstream? Clinical Laboratory ...

  6. MOX fuel irradiation behavior in steady state (irradiation test in HBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Kohno, S; Kamimura, K [Power Reactor and Nuclear Fuel Development Corp., Naka, Ibaraki (Japan)

    1997-08-01

    Two rigs of plutonium-uranium oxide (MOX) fuel rods have been irradiated in Halden boiling water reactor (HBWR) to investigate high burnup MOX fuel behavior for thermal reactor. The objective of irradiation tests is to investigate fuel behavior as influenced by pellet shape, pellet surface treatment, pellet-cladding gap size and MOX fuel powder preparations process. The two rigs have instrumentations for in-pile measurements of the fuel center-line temperature, plenum pressure, cladding elongation and fuel stack length change. The data, taken through in-operation instrumentation, have been analysed and compared with those from post-irradiation examination. The following observations are made: 1) PNC MOX fuels have achieved high burn-up as 59GWd/tMOX (67GWd/tM) at pellet peak without failure; 2) there was no significant difference in fission gas release fraction between PNC MOX fuels and UO{sub 2} fuels; 3) fission gas release from the co-converted fuel was lower than that from the mechanically blended fuel; 4) gap conductance was evaluated to decrease gradually with burn-up and to get stable in high burn-up region. 5) no evident difference of onset LHR for PCMI in experimental parameters (pellet shape and pellet-cladding gap size) was observed, but it decreased with burn-up. (author). 13 refs, 15 figs, 3 tabs.

  7. Development and testing of the EDF-2 reactor fuel element

    International Nuclear Information System (INIS)

    Delpeyroux, P.

    1964-01-01

    This technical report reviews the work which has been necessary for defining the EDF-2 fuel element. After giving briefly the EDF-2 reactor characteristics and the preliminary choice of parameters which made it possible to draw up a draft plan for the fuel element, the authors consider the research proper: - Uranium studies: tests on the passage into the β phase of an internal crown of a tube, bending of the tube under the effect of a localized force, welding of the end-pellets and testing for leaks. The resistance of the tube to crushing and of the pellets to yielding under the external pressure have been studied in detail in another CEA report. - Can studies: conditions of production and leak proof testing of the can, resistance of the fins to creep due to the effect of the gas flow. - Studies of the extremities of the element: creep under compression and welding of the plugs to the can. - Cartridge studies: determination of the characteristics of the can fuel fixing grooves and of the canning conditions, verification of the resistance of the fuel element to thermal cycling, determination of the temperature drop at the can-fuel interface dealt with in more detail in another CEA report. - Studies of the whole assembly: this work which concerns the graphite jacket, the support and the cartridge vibrations has been carried out by the Mechanical and Thermal Study Service (Mechanics Section). In this field the Fuel Element Study Section has investigated the behaviour of the centering devices in a gas current. The outcome of this research is the defining of the plan of the element the production process and the production specifications. The validity of ail these out-of-pile tests will be confirmed by the in-pile tests already under way and by irradiation of the elements in the EDF-2 reactor itself. In conclusion the programme is given for improving the fuel element and for defining the fuel element for the second charge. (authors) [fr

  8. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Laug, David V.; Scates, Dawn M.; Reber, Edward L.; Roybal, Lyle G.; Walter, John B.; Harp, Jason M. [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Morris, Robert N. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer A system has been developed for safety testing of irradiated coated particle fuel. Black-Right-Pointing-Pointer FACS system is designed to facilitate remote operation in a shielded hot cell. Black-Right-Pointing-Pointer System will measure release of fission gases and condensable fission products. Black-Right-Pointing-Pointer Fuel performance can be evaluated at temperatures as high as 2000 Degree-Sign C in flowing helium. - Abstract: The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 Degree-Sign C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated

  9. Radiographic Test

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H.J; Yang, S.H. [Korea Electric Power Research Institute, Taejon (Korea)

    2002-07-01

    This report contains theory, procedure technique and interpretation of radiographic examination and written for whom preparing radiographic test Level II. To determine this baseline of technical competence in the examination, the individual must demonstrate a knowledge of radiography physics, radiation safety, technique development, radiation detection and measurement, facility design, and the characteristics of radiation-producing devices and their principles of operation. (author) 98 figs., 23 tabs.

  10. Testing Understanding and Understanding Testing.

    Science.gov (United States)

    Pedersen, Jean; Ross, Peter

    1985-01-01

    Provides examples in which graphs are used in the statements of problems or in their solutions as a means of testing understanding of mathematical concepts. Examples (appropriate for a beginning course in calculus and analytic geometry) include slopes of lines and curves, quadratic formula, properties of the definite integral, and others. (JN)

  11. Credibility test; Vertrauens-Test

    Energy Technology Data Exchange (ETDEWEB)

    Fuhs, Michael

    2009-07-01

    Solar wafer producers must prove their quality standards. Q-Cells has started a marketing office and opened their test center to journalists. They are aware that adherence to standards is going only half the way of quality assurance. The other half consists in gaining the customers' trust. (orig.)

  12. Knowledge Test

    DEFF Research Database (Denmark)

    Sørensen, Ole Henning

    1998-01-01

    The knowledge test is about competing temporal and spatial expressions of the politics of technological development and national prosperity in contemporary society. The discussion is based on literature of national systems of innovation and industrial networks of various sorts. Similarities...... and differences in the disparate theories are discussed through a critical perspective on metaphor, time, space, agency and technology. It is asserted that the process of globalization is leading to a new production of space-time perceptions and practices where localization and globalization is becoming...... increasingly important. National space is being contested and nation states need to perform differently....

  13. Testing SUSY

    CERN Document Server

    Cassel, S; Ross, G G

    2010-01-01

    If SUSY provides a solution to the hierarchy problem then supersymmetric states should not be too heavy. This requirement is quantified by the Barbieri-Giudice fine tuning measure that provides a quantitative test of SUSY as a solution to the hierarchy problem. The measure is useful in correlating the impact of the various experimental measurements relevant to the search for supersymmetry and also in identifying the most sensitive measurements for testing SUSY. In this paper we apply the measure to the CMSSM, computing it to two-loop order and taking account of current experimental limits and the constraint on dark matter abundance. Using this we determine the present limits on the CMSSM parameter space and identify the measurements at the LHC that are most significant in covering the remaining parameter space. Without imposing the LEP Higgs mass bound we show that the smallest fine tuning (1:14.5) consistent with a saturation of the relic density within the 1$\\sigma$ WMAP bounds corresponds to a Higgs mass o...

  14. Results of VVER fuel rods tests in the MIR.M1 reactor under power cycling conditions

    International Nuclear Information System (INIS)

    Burukin, A.; Izhutov, A.; Ovchinnikov, V.; Kalygin, V.; Markov, D.; Pimenov, Y.; Novikov, V.; Medvedev, A.; Nesterov, B.

    2011-01-01

    The paper presents the main results of the 50 ... 60 MWd/kgU burnup VVER fuel rods tests performed in the MIR.M1 reactor loop facilities under power cycling. The non-destructive PIE results are presented as well. A series of experiments was performed, including overall measurement of fuel rod parameters test, in one of which 300 cycles were done. Irradiation under power cycling conditions and PIE of high-burnup VVER fuel rods showed the following: 1) all fuel rods claddings preserved their integrity under irradiation at linear heat rate (LHR) higher than the NPP operating one; 2) experimental data were obtained on the axial and radial cladding strain and fission gas release (FGR) from 50 ... 60 MWd/kgU burnup VVER-440 and VVER-1000 fuel rods as well as on the kinetics of the change in these parameters and fuel temperature under the power cycling; 3) non-destructive PIE results are in a satisfactory correlation with the data obtained by means of in-pile measurement gages during irradiation. (authors)

  15. Lateral loading test of full-scaled pile foundation focused on ultimate behavior; Shukyoku kyodo ni chakumokushita jitsubutsudai kui kiso no suihei saika shiken

    Energy Technology Data Exchange (ETDEWEB)

    Kosa, K. [Hanshin Expressway Public Corp., Osaka (Japan); Suzuki, N. [CTI Engineering Co. Ltd., Tokyo (Japan); Kimura, M. [Kyoto University, Kyoto (Japan); Kimura, Y. [Public Works Research Institute, Tsukuba (Japan); Morita, Y. [Kiso-Jiban Consultants Co. Ltd., Tokyo (Japan)

    1998-06-21

    Lateral loading test of full-scaled pile foundation was made up to the ultimate limit state to establish a limit state design method for pile foundations. A full-scaled 3-by-3 pile group driven as bridge foundation was used for the test. A surface reclaimed layer is nearly 13cm in thickness. The lateral unidirectional loading test was repeatedly carried out by using 6 hydraulic jacks at 30MN in maximum load capacity and 50cm in maximum allowable deformation. The test and analysis results are as follows. In a range of 0-40% in pile head displacement/pile diameter ratio, a load-carrying capacity increased gradually. The yield point of a load-displacement curve agreed with the point at which a reinforcing bar of a pile body reaches analytical yield stress. Bending moment of a pile body and shear stress of a pile head increased in the order of back, middle and front rows. It was confirmed that a framework analysis method considering the non-linearity of a pile body and bedrock is also applicable to full-scaled pile group foundations. 17 refs., 32 figs., 3 tabs.

  16. Application of self-balanced loading test to socketed pile in weak rock

    Science.gov (United States)

    Cheng, Ye; Gong, Weiming; Dai, Guoliang; Wu, JingKun

    2008-11-01

    Method of self-balanced loading test differs from the traditional methods of pile test. The key equipment of the test is a cell. The cell specially designed is used to exert load which is placed in pile body. During the test, displacement values of the top plate and the bottom plate of the cell are recorded according to every level of load. So Q-S curves can be obtained. In terms of test results, the bearing capacity of pile can be judged. Equipments of the test are simply and cost of it is low. Under some special conditions, the method will take a great advantage. In Guangxi Province, tertiary mudstone distributes widely which is typical weak rock. It is usually chosen as the bearing stratum of pile foundation. In order to make full use of its high bearing capacity, pile is generally designed as belled pile. Foundations of two high-rise buildings which are close to each other are made up of belled socketed piles in weak rock. To obtain the bearing capacity of the belled socketed pile in weak rock, loading test in situ should be taken since it is not reasonable that experimental compression strength of the mudstone is used for design. The self-balanced loading test was applied to eight piles of two buildings. To get the best test effect, the assembly of cell should be taken different modes in terms of the depth that pile socketed in rock and the dimension of the enlarged toe. The assembly of cells had been taken three modes, and tests were carried on successfully. By the self-balanced loading test, the large bearing capacities of belled socketed piles were obtained. Several key parameters required in design were achieved from the tests. For the data of tests had been analyzed, the bearing performance of pile tip, pile side and whole pile was revealed. It is further realized that the bearing capacity of belled socketed pile in the mudstone will decrease after the mudstone it socketed in has been immerged. Among kinds of mineral ingredient in the mudstone

  17. Heart failure - tests

    Science.gov (United States)

    CHF - tests; Congestive heart failure - tests; Cardiomyopathy - tests; HF - tests ... the best test to: Identify which type of heart failure (systolic, diastolic, valvular) Monitor your heart failure and ...

  18. Nuclear stress test

    Science.gov (United States)

    ... Persantine stress test; Thallium stress test; Stress test - nuclear; Adenosine stress test; Regadenoson stress test; CAD - nuclear stress; Coronary artery disease - nuclear stress; Angina - nuclear ...

  19. An Installation of IPS Bypass Line at the Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ho Young; Ahn, G. H.; Lee, M.; Kim, M. S.; Cho, S. H.; Han, J. S.; Hur, S. O. [KAERI, Daejeon (Korea, Republic of)

    2010-12-15

    The Fuel Test Loop(FTL) was installed for the national goal of self-supporting technology in the field of design and construction of nuclear power plant. The FTL with the fuel irradiation equipment is essential in developing, improving and inspecting the fuel of CANDU type or PWR type nuclear power plant. The FTL should be operated at the same conditions of commercial nuclear power plant such as temperature, pressure, flow rate, neutron flux and so on. Starting designing in December 2001, the FTL was installed from March 2007 to August 2008. Especially the In Pile Section(IPS) was installed at IR1 hole in August 2008. Until September 2009 after loading the test fuel, a series of power escalation tests (LSD, CSB1, CSB2, HSB, HOP) were conducted. And it was operated at the condition of CSB2 for the 8 cycles from October 2009 to July 2010. But it could not be normally operated in early 2010, because the high radiation released from irradiated materials due to the worn down bearing of main cooling pump. So, we removed the IPS and installed a newly designed IPS bypass line to prevent increasing high radiation. In this report we will present preliminary works, main works processes, devices of making work environments, a designing and manufacturing of IPS bypass line and a rack of IPS, installing know-hows, problems and solutions broke out during the work etc. We believe that our efforts to complete successful installing and operating of the FTL system will contribute for the efficient utilization of HANARO

  20. Fast reactor test facilities in the US safety program

    International Nuclear Information System (INIS)

    Avery, R.; Dickerman, C.E.; Lennox, D.H.; Rose, D.

    1979-01-01

    The needs for safety information derivable from in-pile programs are reviewed, and the correlation made with existing and planned capability. In view of the current status of the U.S. breeder program, emphasis is given in the review to the impact of different fast breeder options on the required program and facilities. It is concluded that facility needs are somewhat independent of specific fast breeder concept, even though the relative emphasis on the various safety issues will differ. 8 refs