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Sample records for febex mock-up test

  1. FEBEX Full-Scalle Engineered Barriers Experiment in Crystalline Host Rock Preoperational Thermo-Hydro-Mechanical (THM) Modelling of the Mock Up Test

    International Nuclear Information System (INIS)

    1998-01-01

    The object of this report is to present and discuss the results of a series of 1-D and 2-D coupled thermo-hydro-mechanical (THM) and 2-D coupled thermo-hydro-mechanical (THM) analyses modelling the FEBEX mock-up test. The analyses have been carried out during the preoperational storage of the test and attempt to incorporate all available information obtained from laboratory characterisation work. The aim is not only to offer the best estimate of test performance using current models and information but also to provide a basis for future model improvements. Both the theoretical framework adopted in the analysis and the computer code employed are briefly described. The set of parameters used in the computation is then presented with particular reference to the source from which they have been derived. Initial and boundary condition are also defined. The results of a 1-D radially symmetric analysis are used to examine the basic patterns of thermal, hydraulic and mechanical behaviour of the test. A set of sensitivity analyses has been carried out in order to check the effects that the variation of a number of important parameters has on test results. Only in this way it is possible to acquire a proper understanding of the internal structure of the problem and of the interactions between the various phenomena occurring in the buffer. A better reproduction of the geometry of the test is achieved by means of a 2-D mesh representing and axisymmetric longitudinal section. Due to two-dimensional effects, the analyses carried out using this geometry exhibit some differences when compared with the results of the 1-D case, but the basic test behaviour is very similar. The test was started with an initial flooding stage with the purpose of closing the gaps between bentonite blocks. A limited number of compilations using recently developed joint elements have been performed to assess approximately the effect of this initial step on subsequent test behaviour. The analyses reported

  2. Analysis of high heat flux testing of mock-ups

    International Nuclear Information System (INIS)

    Salavy, J.-F.; Giancarli, L.; Merola, M.; Picard, F.; Roedig, M.

    2003-01-01

    ITER EU Home Team is performing a large R and D effort in support of the development of high heat flux components for ITER. In this framework, this paper describes the thermal analyses, the fatigue lifetime evaluation and the transient VDE with material melting related to the high heat flux thermo-mechanical tests performed in the JUDITH facility. It reports on several mock-ups representative of different proposed component designs based on Be, W and CFC as armour materials

  3. Project W-314 performance mock-up test procedure

    International Nuclear Information System (INIS)

    CARRATT, R.T.

    1999-01-01

    The purpose of this Procedure is to assist construction in the pre-operational fabrication and testing of the pit leak detection system and the low point drain assembly by: (1) Control system testing of the pit leak detection system will be accomplished by actuating control switches and verifying that the control signal is initiated, liquid testing and overall operational requirements stated in HNF-SD-W314-PDS-003, ''Project Development Specification for Pit Leak Detection''. (2) Testing of the low point floor drain assembly by opening and closing the drain to and from the ''retracted'' and ''sealed'' positions. Successful operation of this drain will be to verify that the seal does not leak on the ''sealed'' position, the assembly holds liquid until the leak detector actuates and the assembly will operate from on top of the mock-up cover block

  4. Blanket Manufacturing Technologies : Thermomechanical Tests on HCLL Blanket Mocks Up

    International Nuclear Information System (INIS)

    Laffont, G.; Cachon, L.; Taraud, P.; Challet, F.; Rampal, G.; Salavy, J.F.

    2006-01-01

    In the Helium Cooled Lithium Lead (HCLL) Blanket concept, the lithium lead plays the double role of breeder and multiplier material, and the helium is used as coolant. The HCCL Blanket Module are made of steel boxes reinforced by stiffening plates. These stiffening plates form cells in which the breeder is slowly flowing. The power deposited in the breeder material is recovered by the breeder cooling units constituted by 5 parallel cooling plates. All the structures such as first wall, stiffening and cooling plates are cooled by helium. Due to the complex geometry of these parts and the high level of pressure and temperature loading, thermo-mechanical phenomena expected in the '' HCLL blanket concept '' have motivated the present study. The aim of this study, carried out in the frame of EFDA Work program, is to validate the manufacturing technologies of HCLL blanket module by testing small scale mock-up under breeder blanket representative operating conditions.The first step of this experimental program is the design and manufacturing of a relevant test section in the DIADEMO facility, which was recently upgraded with an He cooling loop (pressure of 80 bar, maximum temperature of 500 o C,flow rate of 30 g/s) taking the opportunity of synergies with the gas-cooled fission reactor R-and-D program. The second step will deal with the thermo-mechanical tests. This paper focuses on the program made to support the cooling plate mock up tests which will be carried out on the DIADEMO facility (CEA) by thermo-mechanical calculations in order to define the relevant test conditions and the experimental parameters to be monitored. (author)

  5. Experimental test campaign on an ITER divertor mock-up

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Malavasi, A.; Merola, M.; Polazzi, G.; Simoncini, M.; Zito, D.

    2002-01-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body (CB), designed with some mechanical and hydraulic simplifications with respect to the reference body and its actively cooled Dummy Armour Prototype (DAP). This DAP consists of a Vertical Target (VT), a Wing (WI) and a Dump Target (DT), manufactured by European industries, which are integrated to the Gas Box Liner (GBL) supplied by the Russian Federation ITER Home Team. In 1999, in parallel with the manufacturing activity, the ITER European Home Team decided to assign to ENEA a Task for checking the component integration and performing the thermal-hydraulic and thermal mechanical testing of the DAP and CB. In 1999-2000, ENEA performed the experimental campaign at Brasimone Labs. The present work presents the experimental results of the component integration and the thermal-hydraulic and thermo-mechanical fatigue tests

  6. Experimental test campaign on an ITER divertor mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Dell' Orco, G. E-mail: giovanni.dellorco@brasimone.enea.it; Malavasi, A.; Merola, M.; Polazzi, G.; Simoncini, M.; Zito, D

    2002-11-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body (CB), designed with some mechanical and hydraulic simplifications with respect to the reference body and its actively cooled Dummy Armour Prototype (DAP). This DAP consists of a Vertical Target (VT), a Wing (WI) and a Dump Target (DT), manufactured by European industries, which are integrated to the Gas Box Liner (GBL) supplied by the Russian Federation ITER Home Team. In 1999, in parallel with the manufacturing activity, the ITER European Home Team decided to assign to ENEA a Task for checking the component integration and performing the thermal-hydraulic and thermal mechanical testing of the DAP and CB. In 1999-2000, ENEA performed the experimental campaign at Brasimone Labs. The present work presents the experimental results of the component integration and the thermal-hydraulic and thermo-mechanical fatigue tests.

  7. Feasibility study of the thermo-siphon mock-up test

    International Nuclear Information System (INIS)

    Choi, Jung Woon; Kim, Young Jin; Lee, Kye Hong; Kim, Young Ki; Jeong, Sang Kwon

    2004-09-01

    Described is the feasibility of the thermo-siphon mock-up test for the HANARO-CNS facility. The purposes of the mock-up tests are discussed in detail as the three concepts: for the detailed design, for the operation of the CNS facility, for the safety assurance of itself. This report considers the two stages of mock-up tests in terms of the experimental schedule and plan. As the first stage, the small-size mock-up test using Argon will be implemented to obtain the experience in the cryogenic fluid and to understand the basic concept of the CNS thermo-siphon. In the second stage, two kinds of mock-up tests are discussed: the full-scale mock-up test using liquid hydrogen or the integrated final test using hydrogen outside the reactor after the full-scale mock-up test using Freon gas. The contents discussed in this report will be the basis or the guide lines for the mock-up test. In addition, the results of the mock-up test will be the foundation for the safe operation of the HANARO-CNS facility

  8. Fabrication and testing of small scale mock-ups of ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Hatano, Toshihisa; Sato, Satoshi; Suzuki, Satoshi; Yokoyama, Kenji; Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1998-12-01

    Small scale mock-ups of the primary first wall, the baffle first wall, the shield block and a partial model for the edge of the primary first wall module were designed and fabricated incorporating most of the key design features of the ITER shielding blanket. All mock-ups featured the DSCu heat sink, the built-in SS coolant tubes within the heat sink and the SS shield block. CFC tiles was used as the protection armor for the baffle first wall mock-up. The small scale shield block mock-up, integrated with the first wall, was designed to have a poloidal curvature specified in the ITER design. Fabrication routes of mock-ups were decided based on the single step solid HIP of DSCu/DSCu, DSCu/SS and SS/SS reflecting the results of previous joining techniques development and testing. For attaching the CFC tiles onto DSCu heat sink in the fabrication of the baffle first wall mock-up, a two-step brazing was tried. All mock-ups and the partial model were successfully fabricated with a satisfactory dimensional accuracy. The small scale primary first wall mock-up was thermo-mechanically tested under high heat fluxes of 5-7 MW/m{sup 2} for 2500 cycles in total. Satisfactory heat removal performance and integrity of the mock-up against cyclic high heat flux loads were confirmed by measurement during the tests and destructive examination after the tests. Similar high heat flux tests were also performed with the small scale baffle first wall mock-up under 5-10 MW/m{sup 2} for 4500 cycles in total resulting in sufficient heat removal capability and integrity confirmed by measurements during the tests. (author)

  9. Fabrication and testing of small scale mock-ups of ITER shielding blanket

    International Nuclear Information System (INIS)

    Hatano, Toshihisa; Sato, Satoshi; Suzuki, Satoshi; Yokoyama, Kenji; Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Takatsu, Hideyuki; Ohara, Yoshihiro

    1998-12-01

    Small scale mock-ups of the primary first wall, the baffle first wall, the shield block and a partial model for the edge of the primary first wall module were designed and fabricated incorporating most of the key design features of the ITER shielding blanket. All mock-ups featured the DSCu heat sink, the built-in SS coolant tubes within the heat sink and the SS shield block. CFC tiles was used as the protection armor for the baffle first wall mock-up. The small scale shield block mock-up, integrated with the first wall, was designed to have a poloidal curvature specified in the ITER design. Fabrication routes of mock-ups were decided based on the single step solid HIP of DSCu/DSCu, DSCu/SS and SS/SS reflecting the results of previous joining techniques development and testing. For attaching the CFC tiles onto DSCu heat sink in the fabrication of the baffle first wall mock-up, a two-step brazing was tried. All mock-ups and the partial model were successfully fabricated with a satisfactory dimensional accuracy. The small scale primary first wall mock-up was thermo-mechanically tested under high heat fluxes of 5-7 MW/m 2 for 2500 cycles in total. Satisfactory heat removal performance and integrity of the mock-up against cyclic high heat flux loads were confirmed by measurement during the tests and destructive examination after the tests. Similar high heat flux tests were also performed with the small scale baffle first wall mock-up under 5-10 MW/m 2 for 4500 cycles in total resulting in sufficient heat removal capability and integrity confirmed by measurements during the tests. (author)

  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. High heat flux test of tungsten brazed mock-ups developed for KSTAR divertor

    Energy Technology Data Exchange (ETDEWEB)

    Song, J.H. [National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, K.M., E-mail: kyungmin@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Hong, S.H.; Kim, H.T.; Park, S.H.; Park, H.K.; Ahn, H.J. [National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, S.K.; Lee, D.W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    The tungsten (W) brazed flat type mock-up which consists of W, OFHC-Cu (oxygen-free high conductive copper) and CuCrZr alloy has been designed for KSTAR divertor in preparation for KSTAR upgrade with 17 MW heating power. For verification of the W brazed mock-up, the high heat flux test is performed at KoHLT-EB (Korea High Heat Load Test Facility-Electron Beam) in KAERI (Korea Atomic Energy Research Institute). Three mock-ups are tested for several thousand thermal cycles with absorbed heat flux up to 5 MW/m{sup 2} for 20 s duration. There is no evidence of the failure at the bonding joints of all mock-ups after HHF test. Finite element analysis (FEA) is performed to interpret the result of the test. As a result, it is considered that the local area in the water is in the subcooled boiling regime.

  12. Thermo-mechanical tests of a CFC divertor mock-up

    International Nuclear Information System (INIS)

    Cardella, A.; Akiba, M.; Duwe, R.; Di Pietro, E.; Suzuki, S.; Satoh, K.; Reheis, N.

    1994-01-01

    Thermo-mechanical tests have been performed on a divertor mock-up consisting of a metallic tube armoured with five carbon fibre composite tiles. The tube is inserted the tiles and brazed with TiCuSil braze (monoblock concept). The tube material is TZM, a molybdenum alloy, and the armour material is SEP CARB N112, a high conductivity carbon-carbon composite. Using special surface preparation consisting of laser drilling, small (≅ 500 μm) holes in the composite have been made to increase the surface wetted by the braze and the resistance. The mock-up has been tested at the JAERI 400 kW electron beam test facility JEBIS. The aim of the test was to assess the performance of the mock-up in screening and thermal fatigue tests with particular attention to the behaviour of the armour to heat sink joint. (orig.)

  13. Mock-up test of remote controlled dismantling apparatus for large-sized vessels (contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Myodo, Masato; Miyajima, Kazutoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Okane, Shogo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-03-01

    The Remote dismantling apparatus, which is equipped with multi-units for functioning of washing, cutting, collection of cut pieces and so on, has been constructed to dismantle the large-sized vessels in the JAERI's Reprocessing Test Facility (JRTF). The apparatus has five-axis movement capability and its operation is performed remotely. The mock-up tests were performed to evaluate the applicability of the apparatus to actual dismantling activities by using the mock-ups of LV-3 and LV-5 in the facility. It was confirmed that each unit was satisfactory functioned by remote operation. Efficient procedures for dismantling the large-sized vessel was studied and various date was obtained in the mock-up tests. This apparatus was found to be applicable for the actual dismantling activity in JRTF. (author)

  14. Vacuum tests of a beamline front-end mock-up at the Advanced Photon Source

    International Nuclear Information System (INIS)

    Liu, C.; Nielsen, R.W.; Kruy, T.L.; Shu, D.; Kuzay, T.M.

    1994-01-01

    A-mock-up has been constructed to test the functioning and performance of the Advanced Photon Source (APS) front ends. The mock-up consists of all components of the APS insertion-device beamline front end with a differential pumping system. Primary vacuum tests have been performed and compared with finite element vacuum calculations. Pressure distribution measurements using controlled leaks demonstrate a better than four decades of pressure difference between the two ends of the mock-up. The measured pressure profiles are consistent with results of finite element analyses of the system. The safety-control systems are also being tested. A closing time of ∼20 ms for the photon shutter and ∼7 ms for the fast closing valve have been obtained. Experiments on vacuum protection systems indicate that the front end is well protected in case of a vacuum breach

  15. Mock-up test of remote controlled dismantling apparatus for large-sized vessels (contract research)

    International Nuclear Information System (INIS)

    Myodo, Masato; Miyajima, Kazutoshi; Okane, Shogo

    2001-03-01

    The Remote dismantling apparatus, which is equipped with multi-units for functioning of washing, cutting, collection of cut pieces and so on, has been constructed to dismantle the large-sized vessels in the JAERI's Reprocessing Test Facility (JRTF). The apparatus has five-axis movement capability and its operation is performed remotely. The mock-up tests were performed to evaluate the applicability of the apparatus to actual dismantling activities by using the mock-ups of LV-3 and LV-5 in the facility. It was confirmed that each unit was satisfactory functioned by remote operation. Efficient procedures for dismantling the large-sized vessel was studied and various date was obtained in the mock-up tests. This apparatus was found to be applicable for the actual dismantling activity in JRTF. (author)

  16. Set-up of a pre-test mock-up experiment in preparation for the HCPB Breeder Unit mock-up experimental campaign

    Energy Technology Data Exchange (ETDEWEB)

    Hernández, F., E-mail: francisco.hernandez@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany); Kolb, M. [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-WPT) (Germany); Ilić, M.; Kunze, A. [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany); Németh, J. [KFKI Research Institute for Particle and Nuclear Physics (Hungary); Weth, A. von der [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany)

    2013-10-15

    Highlights: ► As preparation for the HCPB-TBM Breeder Unit out-of-pile testing campaign, a pre-test experiment (PREMUX) has been prepared and described. ► A new heater system based on a wire heater matrix has been developed for imitating the neutronic volumetric heating and it is compared with the conventional plate heaters. ► The test section is described and preliminary thermal results with the available models are presented and are to be benchmarked with PREMUX. ► The PREMUX integration in the air cooling loop L-STAR/LL in the Karlsruhe Institute for Technology is shown and future steps are discussed. -- Abstract: The complexity of the experimental set-up for testing a full-scaled Breeder Unit (BU) mock-up for the European Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) has motivated to build a pre-test mock-up experiment (PREMUX) consisting of a slice of the BU in the Li{sub 4}SiO{sub 4} region. This pre-test aims at verifying the feasibility of the methods to be used for the subsequent testing of the full-scaled BU mock-up. Key parameters needed for the modeling of the breeder material is also to be determined by the Hot Wire Method (HWM). The modeling tools for the thermo-mechanics of the pebble beds and for the mock-up structure are to be calibrated and validated as well. This paper presents the setting-up of PREMUX in the L-STAR/LL facility at the Karlsruhe Institute of Technology. A key requirement of the experiments is to mimic the neutronic volumetric heating. A new heater concept is discussed and compared to several conventional heater configurations with respect to the estimated temperature distribution in the pebble beds. The design and integration of the thermocouple system in the heater matrix and pebble beds is also described, as well as other key aspects of the mock-up (dimensions, layout, cooling system, purge gas line, boundary conditions and integration in the test facility). The adequacy of these methods for the full-scaled BU

  17. ITER baffle module small-scale mock-ups: first wall thermo-mechanical testing results

    International Nuclear Information System (INIS)

    Severi, Y.; Giancarli, L.; Poitevin, Y.; Salavy, J.F.; Le Marois, G.; Roedig, M.; Vieider, G.

    1998-01-01

    The EU-home team is in charge of the R and D related to the ITER baffle first wall. Five small-scale mock-ups, using Be, CFC and W tiles and different armour/heat-sink material joints under development, have been fabricated and thermomechanically tested in FE200 (Le Creusot) and JUDITH (Juelich) electron beam facilities. The small-scale mock-ups have been submitted to thermo-mechanical fatigue tests (up to failure using accelerating techniques). The objective was to determine the performances of the armour material joints under high heat flux cycles. (orig.)

  18. Performance test results of mock-up test facility of HTTR hydrogen production system

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Inaba, Yoshitomo; Nishihara, Tetsuo

    2004-01-01

    For the purpose to demonstrate effectiveness of high-temperature nuclear heat utilization, Japan Atomic Energy Research Institute has been developing a hydrogen production system and has planned to connect the hydrogen production system to High Temperature Engineering Test Reactor (HTTR). Prior to construction of a HTTR hydrogen production system, a mock-up test facility was constructed to investigate transient behavior of the hydrogen production system and to establish system controllability. The Mock-up test facility with a full-scale reaction tube is an approximately 1/30-scale model of the HTTR hydrogen production system and an electric heater is used as a heat source instead of a reactor. After its construction, a performance test of the test facility was carried out in the same pressure and temperature conditions as those of the HTTR hydrogen production system to investigate its performance such as hydrogen production ability, controllability and so on. It was confirmed that hydrogen was stably produced with a hot helium gas about 120m 3 /h, which satisfy the design value, and thermal disturbance of helium gas during the start-up could be mitigated within the design value by using a steam generator. The mock-up test of the HTTR hydrogen production system using this facility will continue until 2004. (author)

  19. Full scale mock-up tests for rod bundle thermal-hydraulics in Japan

    International Nuclear Information System (INIS)

    Sugawara, S.

    1995-01-01

    This poster describes tests aimed at development and validation of principal design methodology of rod bundle thermal-hydraulics correlations. The works are based on domestic data base using the full-scale mock-up test facilities. The scope of the tests comprises DNB heat flux, transient DNB heat flux, post DNB heat transfer, pressure drop and void distribution. The works have been performed under collaboration among electric facilities, NPP vendors, universities, governmental corporations. 1 tab., 14 figs

  20. Measurement and control system for the ITER remote handling mock-up test

    International Nuclear Information System (INIS)

    Oka, K.; Kakudate, S.; Takiguchi, Y.; Ako, K.; Taguchi, K.; Tada, E.; Ozaki, F.; Shibanuma, K.

    1998-01-01

    The mock-up test platforms composed of full-scale remote handling (RH) equipment were developed for demonstrating remote replacement of the ITER blanket and divertor. In parallel, the measurement and control system for operating these RH equipment were constructed on the basis of open architecture with object oriented feature, aiming at realization of fully-remoted automatic operation required for ITER. This paper describes the design concept of the measurement and control system for the remote handling equipment of ITER, and outlines the measured performances of the fabricated measurement system for the remote handling mock-up tests, which includes Data Acquisition System (DAS), Visual Monitoring System (VMS) and Virtual Reality System (VRS). (authors)

  1. Overview of the EU small scale mock-up tests for ITER high heat flux components

    International Nuclear Information System (INIS)

    Vieider, G.; Barabash, V.; Cardella, A.

    1998-01-01

    This task within the EU R and D for ITER was aimed at the development of basic manufacturing solutions for the high heat flux plasma facing components such as the divertor targets, the baffles and limiters. More than 50 representative small-scale mock-ups have been manufactured with beryllium, carbon and tungsten armour using various joining technologies. High heat flux testing of 20 of these mock-ups showed the carbon mono-blocks to be the most robust solution, surviving 2000 cycles at absorbed heat fluxes of up to 24 MW m -2 . With flat armour tiles rapid joint failures occurred at 5-16 MW m -2 depending on joining technology and armour material. These test results serve as a basis for the selection of manufacturing options and materials for the prototypes now being ordered. (orig.)

  2. Thermal fatigue tests with actively cooled divertor mock-ups for ITER

    International Nuclear Information System (INIS)

    Roedig, M.; Duwe, R.; Linke, J.; Schuster, A.; Wiechers, B.; Ibbott, C.; Jacobson, D.; Le Marois, G.; Lind, A.; Lorenzetto, P.; Vieider, G.; Peacock, A.; Ploechl, L.; Severi, Y.; Visca, E.

    1998-01-01

    Mock-ups for high heat flux components with beryllium and CFC armour materials have been tested by means of the electron beam facility JUDITH. The experiments concerned screening tests to evaluate heat removal efficiency and thermal fatigue tests. CFC monoblocks attached to DS-Cu (Glidcop Al25) and CuCrZr tubes by active metal casting and Ti brazing showed the best thermal fatigue behaviour. They survived more than 1000 cycles at heat loads up to 25 MW m -2 without any indication of failure. Operational limits are given only by the surface temperature on the CFC tiles. Most of the beryllium mock-ups were of the flat tile type. Joining techniques were brazing, hot isostatic pressing (HIP) and diffusion bonding. HIPed and diffusion bonded Be/Cu modules have not yet reached the standards for application in high heat flux components. The limit of this production method is reached for heat loads of approximately 5 MW m -2 . Brazing with and without silver seems to be a more robust solution. A flat tile mock-up with CuMnSnCe braze was loaded at 5.4 MW m -2 for 1000 cycles without damage The first test with a beryllium monoblock joined to a CuCrZr tube by means of Incusil brazing shows promising results; it survived 1000 cycles at 4.5 MW m -2 without failure. (orig.)

  3. Beryllium mock-ups development and ultrasonic testing for ITER divertor conditions

    International Nuclear Information System (INIS)

    Barabash, V.R.; Bykov, V.A.; Giniyatulin, R.N.; Gervash, A.A.; Gurieva, T.M.; Egorov, K.E.; Komarov, V.L.; Korolkov, M.D.; Mazul, I.V.; Gitarsky, L.S.; Strulia, I.L.; Sizenev, V.S.; Pronyakin, V.T.

    1995-01-01

    At the present time beryllium is considered as the most suitable armour material for the ITER divertor application. Different types of Be-divertor mock-up construction are compared in the report. Two different technologies of beryllium tiles joining to a heat sink body are analysed: high temperature brazing and thermodiffusion bonding. The comparative analysis of different constructions has been performed on the basis of 2-D finite element calculation for temperatures and stresses. The main parameters and diagnostic capabilities of electron beam facility for HHF testing of beryllium mock-ups are described. The first results of HHF tests of ''beryllium-copper saddle-MAGT tube'' and ''beryllium-copper plate-SS body'' mock-ups are presented. The reasons of the damages during the HHF are analysed. The technique of ultrasonic testing of the thermodifussion bonding and brazing quality for beryllium-copper joints is presented. The recorded results are prepared in the form of ultrasound grams. The testing results are compared with the metallographic analysis. (orig.)

  4. Mock-up test results of monoblock-type CFC divertor armor for JT-60SA

    Energy Technology Data Exchange (ETDEWEB)

    Higashijima, S. [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)], E-mail: higashijima.satoru@jaea.go.jp; Sakurai, S.; Suzuki, S.; Yokoyama, K.; Kashiwa, Y.; Masaki, K.; Shibama, Y.K.; Takechi, M.; Shibanuma, K.; Sakasai, A.; Matsukawa, M.; Kikuchi, M. [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

    2009-06-15

    The JT-60 Super Advanced (JT-60SA) tokamak project starts under both the Japanese domestic program and the international program 'Broader Approach'. The maximum heat flux to JT-60SA divertor is estimated to {approx}15 MW/m{sup 2} for 100 s. Japan Atomic Energy Agency (JAEA) has developed a divertor armor facing high heat flux in the engineering R and D for ITER, and it is concluded that monoblock-type CFC divertor armor is promising for JT-60SA. The JT-60SA armor consists of CFC monoblocks, a cooling CuCrZr screw-tube, and a thin oxygen-free high conductivity copper (OFHC-Cu) buffer layer between the CFC monoblock and the screw-tube. CFC/OFHC-Cu and OFHC-Cu/CuCrZr joints are essential for the armor, and these interfaces are brazed. Needed improvements from ITER engineering R and D are good CFC/OFHC-Cu and OFHC-Cu/CuCrZr interfaces and suppression of CFC cracking. For these purposes, metalization inside CFC monoblock is applied, and we confirmed again that the mock-up has heat removal capability in excess of ITER requirement. For optimization of the fabrication method and understanding of the production yield, the mock-ups corresponding to quantity produced in one furnace at the same time is also produced, and the half of the mock-ups could remove 15 MW/m{sup 2} as required. This paper summarizes the recent progress of design and mock-up test results for JT-60SA divertor armor.

  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. Analysis of free and forced excitation tests of 394 KN isolated structure mock-up

    International Nuclear Information System (INIS)

    Serino, G.; Martelli, A.; Bonacina, G.

    1993-01-01

    At the 1991 ASME-PVP Conference, some first experimental results obtained from static and dynamic tests on high damping steel laminated rubber bearings (Martelli et al., 1991) and from free and forced excitation tests on a 394 kN isolated structure mock-up were presented (Forni et al., 1991). In this paper, the most significant test data are reorganized and discussed in order to assess the suitability of single bearing test results to predict the dynamic response of an isolated structure. Three mathematical models of the single isolator having different levels of approximation are proposed, and their capability to estimate the experimental response of the mock-up is evaluated. It is shown that a non-linear hysteretic model, defined by three rubber parameters only, allows a very good complete simulation of the dynamic behavior of the isolated structure in both free and forced vibration tests. A simpler equivalent linear viscous model permits a good prediction of the peak absolute acceleration and relative displacement values if bearing stiffness and damping parameters are properly selected, and can be used in a response spectrum analysis, but reproduces less exactly the experimental behavior. An equivalent linear hysteretic model represents more correctly the actual rubber damping behavior, but gives results very similar to those obtained through the equivalent linear viscous model because of the practically mono-frequencial response of the isolated structure

  7. Manufacturing and testing of a copper/CFC divertor mock-up for JET

    International Nuclear Information System (INIS)

    Brossa, M.; Ciric, D.; Deksnis, E.; Falter, H.; Guerreschi, U.; Peacock, A.; Pick, M.; Rossi, M.; Shen, Y.; Zacchia, F.

    1995-01-01

    An actively cooled divertor is a possible option for future developments at The Joint European Torus (JET). A proof of principle actively cooled tile has been produced in order to qualify the relevant manufacturing technologies and the non destructive control processes. In this frame Ansaldo Ricerche (ARI) has been involved in the construction of a mock-up comprising 6 OFHC copper tubes for water cooling that are brazed to a plate made out of carbon fibre composite (CFC). The final objective was the high heat flux testing of the mock-up at JET in order to evaluate the general behaviour of the component under relevant operating conditions. The key point of the work was the realisation of a sound joint by adapting the expertise gained in ARI in previous R and D activities on brazing heterogeneous materials. Reliable methods for ultrasonic examinations of the pieces were also set up. For successful application to the JET pumped divertor a water-cooled CFC target plate must show surface temperatures of 2 . Furthermore, global hydraulic considerations specific to JET limit the system pressure to 0.7 MPa. In such a design, critical heat flux is not the key limit, rather the reliability of the CFC-copper joint in terms of extent of wetting. First tests in the neutral beam test bed at JET show an adequate response for fluxes up to 15 MW/m 2 . (orig.)

  8. TANK 18 AND 19-F TIER 1A EQUIPMENT FILL MOCK UP TEST SUMMARY

    Energy Technology Data Exchange (ETDEWEB)

    Stefanko, D.; Langton, C.

    2011-11-04

    The United States Department of Energy (US DOE) has determined that Tanks 18-F and 19-F have met the F-Tank Farm (FTF) General Closure Plan Requirements and are ready to be permanently closed. The high-level waste (HLW) tanks have been isolated from FTF facilities. To complete operational closure they will be filled with grout for the purpose of: (1) physically stabilizing the tanks, (2) limiting/eliminating vertical pathways to residual waste, (3) discouraging future intrusion, and (4) providing an alkaline, chemical reducing environment within the closure boundary to control speciation and solubility of select radionuclides. Bulk waste removal and heel removal equipment remain in Tanks 18-F and 19-F. This equipment includes the Advance Design Mixer Pump (ADMP), transfer pumps, transfer jets, standard slurry mixer pumps, equipment-support masts, sampling masts, dip tube assemblies and robotic crawlers. The present Tank 18 and 19-F closure strategy is to grout the equipment in place and eliminate vertical pathways by filling voids in the equipment to vertical fast pathways and water infiltration. The mock-up tests described in this report were intended to address placement issues identified for grouting the equipment that will be left in Tank 18-F and Tank 19-F. The Tank 18-F and 19-F closure strategy document states that one of the Performance Assessment (PA) requirements for a closed tank is that equipment remaining in the tank be filled to the extent practical and that vertical flow paths 1 inch and larger be grouted. The specific objectives of the Tier 1A equipment grout mock-up testing include: (1) Identifying the most limiting equipment configurations with respect to internal void space filling; (2) Specifying and constructing initial test geometries and forms that represent scaled boundary conditions; (3) Identifying a target grout rheology for evaluation in the scaled mock-up configurations; (4) Scaling-up production of a grout mix with the target rheology

  9. Preparation of W/CuCrZr monoblock test mock-up using vacuum brazing technique

    International Nuclear Information System (INIS)

    Singh, Kongkham Premjit; Khirwadkar, Samir S.; Bhope, Kedar; Patel, Nikunj; Mokaria, Prakash K.; Mehta, Mayur

    2015-01-01

    Development of the joining for W/CuCrZr monoblock PFC test mock-up is an interest area in Fusion R and D. W/Cu bimetallic material has prepared using OFHC copper casting approach on the radial surface of W monoblock tile surface. The W/Cu bimetallic material has been joined with CuCrZr tube (heat sink) material with the vacuum brazing route. Vacuum brazing of W/Cu-CuCrZr has been performed @ 970 °C for 10 mins using NiCuMn-37 filler material under deep vacuum environment (10 -6 mbar). Graphite fixtures were used for OFHC copper casting and vacuum brazing experiments. The joint integrity of W/Cu-CuCrZr monoblock mock-up on W/Cu and Cu-CuCrZr has been checked using ultrasonic immersion technique. Micro-structural examination and Spot-wise elemental analysis have been carried out using HR-SEM and EDAX. The results of the experimental work will be discussed in the paper. (author)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  11. High heat load tests on W/Cu mock-ups and evaluation of their application to EAST device

    Energy Technology Data Exchange (ETDEWEB)

    Li, H. [Institute of Plasma Physics, Chinese Academy of Science, Hefei, Anhui 230031 (China); Hefei Electronic Engineering Institute, Hefei, Anhui 230037 (China)], E-mail: lih72@hotmail.com; Chen, J.L.; Li, J.G. [Institute of Plasma Physics, Chinese Academy of Science, Hefei, Anhui 230031 (China); Sun, X.J. [Hefei Electronic Engineering Institute, Hefei, Anhui 230037 (China)

    2009-01-15

    Tungsten has been considered as the primary candidate plasma-facing materials (PFM) for the EAST device. Three actively cooled W/Cu mock-ups with an interlayer made of tungsten-copper alloy (1.5 mm) were designed and manufactured. The tungsten armors, pure sintered tungsten plate (1 mm) and plasma-sprayed tungsten coatings (0.3 and 0.9 mm), were bonded to the interlayer by brazing and depositing respectively. All mock-ups can withstand high heat flux up to 5 MW/m{sup 2} and no obvious failure was found after tests. The thermal performance experiments and microstructure analyses indicated the structure of mock-ups possess good thermal contact and high heat transfer capability. WCu alloy as an interlayer can largely reduce the stress due to the mismatch and improve the reliability. The mock-up with 0.9 mm coating had the highest surface temperature than the other two mock-ups, delaminations of this mock-up were found in the near surface by SEM. The primary results show that pure sintered tungsten brazed to WCu alloy is a possible way, and thick plasma-sprayed coating technique still need to be improved.

  12. DWPF Sample Vial Insert Study-Statistical Analysis of DWPF Mock-Up Test Data

    Energy Technology Data Exchange (ETDEWEB)

    Harris, S.P. [Westinghouse Savannah River Company, AIKEN, SC (United States)

    1997-09-18

    This report is prepared as part of Technical/QA Task Plan WSRC-RP-97-351 which was issued in response to Technical Task Request HLW/DWPF/TTR-970132 submitted by DWPF. Presented in this report is a statistical analysis of DWPF Mock-up test data for evaluation of two new analytical methods which use insert samples from the existing HydragardTM sampler. The first is a new hydrofluoric acid based method called the Cold Chemical Method (Cold Chem) and the second is a modified fusion method.Either new DWPF analytical method could result in a two to three fold improvement in sample analysis time.Both new methods use the existing HydragardTM sampler to collect a smaller insert sample from the process sampling system. The insert testing methodology applies to the DWPF Slurry Mix Evaporator (SME) and the Melter Feed Tank (MFT) samples.The insert sample is named after the initial trials which placed the container inside the sample (peanut) vials. Samples in small 3 ml containers (Inserts) are analyzed by either the cold chemical method or a modified fusion method. The current analytical method uses a HydragardTM sample station to obtain nearly full 15 ml peanut vials. The samples are prepared by a multi-step process for Inductively Coupled Plasma (ICP) analysis by drying, vitrification, grinding and finally dissolution by either mixed acid or fusion. In contrast, the insert sample is placed directly in the dissolution vessel, thus eliminating the drying, vitrification and grinding operations for the Cold chem method. Although the modified fusion still requires drying and calcine conversion, the process is rapid due to the decreased sample size and that no vitrification step is required.A slurry feed simulant material was acquired from the TNX pilot facility from the test run designated as PX-7.The Mock-up test data were gathered on the basis of a statistical design presented in SRT-SCS-97004 (Rev. 0). Simulant PX-7 samples were taken in the DWPF Analytical Cell Mock-up

  13. DWPF Sample Vial Insert Study-Statistical Analysis of DWPF Mock-Up Test Data

    International Nuclear Information System (INIS)

    Harris, S.P.

    1997-01-01

    This report is prepared as part of Technical/QA Task Plan WSRC-RP-97-351 which was issued in response to Technical Task Request HLW/DWPF/TTR-970132 submitted by DWPF. Presented in this report is a statistical analysis of DWPF Mock-up test data for evaluation of two new analytical methods which use insert samples from the existing HydragardTM sampler. The first is a new hydrofluoric acid based method called the Cold Chemical Method (Cold Chem) and the second is a modified fusion method.Both new methods use the existing HydragardTM sampler to collect a smaller insert sample from the process sampling system. The insert testing methodology applies to the DWPF Slurry Mix Evaporator (SME) and the Melter Feed Tank (MFT) samples. Samples in small 3 ml containers (Inserts) are analyzed by either the cold chemical method or a modified fusion method. The current analytical method uses a HydragardTM sample station to obtain nearly full 15 ml peanut vials. The samples are prepared by a multi-step process for Inductively Coupled Plasma (ICP) analysis by drying, vitrification, grinding and finally dissolution by either mixed acid or fusion. In contrast, the insert sample is placed directly in the dissolution vessel, thus eliminating the drying, vitrification and grinding operations for the Cold chem method. Although the modified fusion still requires drying and calcine conversion, the process is rapid due to the decreased sample size and that no vitrification step is required.A slurry feed simulant material was acquired from the TNX pilot facility from the test run designated as PX-7.The Mock-up test data were gathered on the basis of a statistical design presented in SRT-SCS-97004 (Rev. 0). Simulant PX-7 samples were taken in the DWPF Analytical Cell Mock-up Facility using 3 ml inserts and 15 ml peanut vials. A number of the insert samples were analyzed by Cold Chem and compared with full peanut vial samples analyzed by the current methods. The remaining inserts were analyzed by

  14. Mock-up test on key components of ITER blanket remote handling system

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Matsumoto, Yasuhiro; Taguchi, Koh; Kozaka, Hiroshi; Shibanuma, Kiyoshi; Tesini, Alessandro

    2009-01-01

    The maintenance operation of the ITER in-vessel component, such as a blanket and divertor, must be executed by the remote equipment because of the high gamma-ray environment. During the Engineering Design Activity (EDA), the Japan Atomic Energy Agency (then called as Japan Atomic Energy Research Institute) had been fabricated the prototype of the vehicle manipulator system for the blanket remote handling and confirmed feasibility of this system including automatic positioning of the blanket and rail deployment procedure of the articulated rail. The ITER agreement, which entered into force in the last year, formally decided that Japan will procure the blanket remote handling system and the JAEA, as the Japanese Domestic Agency, is continuing several R and Ds so that the system can be procured smoothly. The residual key issues after the EDA are rail connection and cable handling. The mock-ups of the rail connection mechanism and the cable handling system were fabricated from the last year and installed at the JAEA Naka Site in this March. The former was composed of the rail connecting mechanism, two rail segments and their handling systems. The latter one utilized a slip ring, which implemented 80 lines for power and 208 lines for signal, because there is an electrical contact between the rotating spool and the fixed base. The basic function of these systems was confirmed through the mock-up test. The rail connection mechanism, for example, could accept misalignment of 1.5-2 mm at least. The future test plan is also mentioned in the paper.

  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. Study on control characteristics for HTTR hydrogen production system with mock-up test facility

    International Nuclear Information System (INIS)

    Inaba, Yoshitomo; Ohashi, Hirofumi; Nishihara, Tetsuo; Sato, Hiroyuki; Inagaki, Yoshiyuki; Takeda, Tetsuaki; Hayashi, Koji; Takada, Shoji

    2005-01-01

    The Japan Atomic Energy Research Institute has a demonstration test plan of a hydrogen production system by steam reforming of methane coupling with the High-Temperature Engineering Test Reactor (HTTR). Prior to the coupling of a hydrogen production plant with the HTTR, simulation tests with a mock-up test facility of the HTTR hydrogen production system (HTTR-H2) is underway. The test facility is a 1/30-scale of the HTTR-H2 and simulates key components downstream from an intermediate heat exchanger of the HTTR. The main objective of the simulation tests is the establishment and demonstration of control technology, focusing on the mitigation of a thermal disturbance to the reactor by a steam generator (SG) and on the controllability of the pressure difference between the helium and process gases at the reaction tube in a steam reformer (SR). It was confirmed that the fluctuation of the outlet helium gas temperature at the SG and the pressure difference in the SR can be controlled within the allowable range for the HTTR-H2 in the case of the system controllability test for the fluctuation of chemical reaction. In addition, a dynamic simulation code for the HTTR-H2 was verified with the obtained test data

  17. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, Masataka; Takeda, Nobukazu; Onozuka, Masanori [Japan Atomic Energy Agency (Japan); Kakudate, Satoshi [Mitsubishi Heavy Industries, Ltd. (Japan)

    2007-07-01

    The structure and fabrication methods of the ITER vacuum vessel have been investigated and defined by the ITER international team. However, some of the current specifications are very difficult to be achieved from the manufacturing point of view and will lead to cost increase. In the mock-up fabrication, it is planned to conduct the following items: 1. Feasibility of the Japanese proposed VV structure and fabrication methods and the applicability to the ITER are to be confirmed; 2. Assembly procedure and inspection procedure are to be confirmed; 3. Manufacturing tolerances are to be assessed; 4. Manufacturing schedule is to be assessed. This report summarizes the Japanese proposed specification of the VV mock-up describing differences between the ITER supplied design. General scope of the mock-up fabrication and the detailed dimensions are also shown. In the VV fabrication, several types of weld joint configuration will be used. This report shows the joint configurations proposed by Japan to be used for the inner shell connection, the rib-to-shell connection and outer shell connection, and the housing-to-shell connection, respectively. Non-destructive testing considered to be applied to each joint configuration is also presented. A series of the fabrication and assembly procedures for the mock-up are presented in this report, together with candidates of welding configurations. Finally, the report summarizes the results of mock-up fabrication, including results of nondestructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. (orig.)

  18. Development of control technology for HTTR hydrogen production system with mock-up test facility

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Inaba, Yoshitomo; Nishihara, Tetsuo; Takeda, Tetsuaki; Hayashi, Koji; Takada, Shoji; Inagaki, Yoshiyuki

    2006-01-01

    The Japan Atomic Energy Agency has been planning the demonstration test of hydrogen production with the High Temperature Engineering Test Reactor (HTTR). In a HTTR hydrogen production system (HTTR-H2), it is required to control a primary helium temperature within an allowable value at a reactor inlet to prevent a reactor scram. A cooling system for a secondary helium with a steam generator (SG) and a radiator is installed at the downstream of a chemical rector in a secondary helium loop in order to mitigate the thermal disturbance caused by the hydrogen production system. Prior to HTTR-H2, the simulation test with a mock-up test facility has been carried out to establish the controllability on the helium temperature using the cooling system against the loss of chemical reaction. It was confirmed that the fluctuations of the helium temperature at chemical reactor outlet, more than 200 K, at the loss of chemical reaction could be successfully mitigated within the target of ±10 K at SG outlet. A dynamic simulation code of the cooling system for HTTR-H2 was verified with the obtained test data

  19. HTTR hydrogen production system. Structure and main specifications of mock-up test facility (Contract research)

    International Nuclear Information System (INIS)

    Kato, Michio; Aita, Hideki; Inagaki, Yoshiyuki; Hayashi, Koji; Ohashi, Hirofumi; Sato, Hiroyuki; Iwatsuki, Jin; Takada, Shoji; Inaba, Yoshitomo

    2007-03-01

    The mock-up test facility was fabricated to investigate performance of the steam generator for mitigation of the temperature fluctuation of helium gas and transient behavior of the hydrogen production system for HTTR and to obtain experimental data for verification of a dynamic analysis code. The test facility has an approximate hydrogen production capacity of 120Nm 3 /h and the steam reforming process of methane; CH 4 +H 2 O=3H 2 +CO, was used for hydrogen production of the test facility. An electric heater was used as a heat source instead of the reactor in order to heat helium gas up to 880degC (4MPa) at the chemical reactor inlet which is the same temperature as the HTTR hydrogen production system. Fabrication of the test facility was completed in February in 2002, and seven cycle operations were carried out from March in 2002 to December in 2004. This report describes the structure and main specifications of the test facility. (author)

  20. Thermo-siphon Mock-up Test for the HANARO-CNS

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jungwoon; Lee, Kye Hong; Kim, Hark Rho; Kim, Youngki; Kim, Myong Seop; Wu, Sang Ik; Kim, Bong Su

    2006-04-15

    In order to moderate thermal neutrons into cold neutrons, the liquid hydrogen is selected as a moderator for the HANARO CNS. By the non-nuclear heat load and nuclear heat load induced from collision of gamma-ray, beta-ray, and thermal neutrons, the liquid hydrogen in the moderator cell evaporates and flows into the heat exchanger. This evaporated hydrogen gas is liquefied by the cryogenic helium supplied from the helium refrigeration system,, then flows back to the moderator cell. This is so-called two-phase thermo-siphon. The most important point in the stable thermo-siphon is to have the good balance between the cooling capacity of the HRS and the heat load on the moderator cell so as to maintain the stable two-phase liquid level in the moderator cell. Accordingly, for not only the experience of the cryogenic two-phase thermo-siphon but also setup of the operation procedure, the full-scaled mock-up test has been performed using the liquid hydrogen. Through the test, the stable thermo-siphon establishment is confirmed at the cold normal operation; furthermore, the detail design parameter is validated. On top of the normal operation procedure setup, the abnormal operation procedure is settled based on the understanding the abnormal pressure and temperature transient dynamics in the hydrogen system.

  1. The HIE-ISOLDE alignment and monitoring system software and test mock up

    CERN Document Server

    Kautzmann, G; Kadi, Y; Leclercq, Y; Waniorek, S; Williams, L

    2012-01-01

    For the HIE Isolde project a superconducting linac will be built at CERN in the Isolde facility area. The linac will be based on the creation and installation of 2 high- β and 4 low- β cryomodules containing respectively 5 high-β superconducting cavities and 1 superconducting solenoid for the two first ones, 6 low-β superconducting cavities and 2 superconducting solenoids for the four other ones. An alignment and monitoring system of the RF cavities and solenoids placed inside the cryomodules is needed to reach the optimum linac working conditions. The alignment system is based on opto-electronics, optics and precise mechanical instrumentation. The geometrical frame configuration, the data acquisition and the 3D adjustment will be managed using a dedicated software application. In parallel to the software development, an alignment system test mock-up has been built for software validation and dimensional tests. This paper will present the software concept and the development status, and then will describe...

  2. Manufacturing, testing and post-test examination of ITER divertor vertical target W small scale mock-ups

    International Nuclear Information System (INIS)

    Visca, Eliseo; Cacciotti, Emanuele; Komarov, Anton; Libera, Stefano; Litunovsky, Nikolay; Makhankov, Alexey; Mancini, Andrea; Merola, Mario; Pizzuto, Aldo; Riccardi, Bruno; Roccella, Selanna

    2011-01-01

    ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R and D activities. During the last years ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP), suitable for the construction of high heat flux plasma-facing components, such as the divertor targets. In the frame of the EFDA contract six mock-ups were manufactured by HRP in the ENEA labs using W monoblocks supplied by the Efremov Institute in St. Petersburg, Russian Federation and IG CuCrZr tubes. According to the technical specifications the mock-ups were examined by ultrasonic technique and after their acceptance they were delivered to the Efremov Institute TSEFEY-M e-beam facility for the thermal fatigue testing. The test consisted in 3000 cycles of 15 s heating and 15 s cooling at 10 MW/m 2 and finally 1000 cycles at 20 MW/m 2 . After the testing the ultrasonic non-destructive examination was repeated and the results compared with the investigation performed before the testing. A microstructure modification of the W monoblock material due to the overheating of the surfaces and the copper interlayer structure modification were observed in the high heat flux area. The leakage points of the mock-ups that did not conclude the testing were localized in the middle of the monoblock while they were expected between two monoblocks. This paper reports the manufacturing route, the thermal fatigue testing, the pre and post non destructive examination and finally the results of the destructive examination performed on the monoblock small scale mock-ups.

  3. 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'.

  4. High heat flux testing of EU tungsten monoblock mock-ups for the ITER divertor

    International Nuclear Information System (INIS)

    Gavila, P.; Riccardi, B.; Pintsuk, G.; Ritz, G.; Kuznetsov, V.; Durocher, A.

    2015-01-01

    Highlights: • All the tested items sustained the ITER Full W divertor qualification program requirements. This confirms that the technology for the manufacturing of the first set of the ITER Divertor is available in Europe. • The surface roughening and local melting of the W surface under high heat flux was proven to be significantly reduced for an armour thickness lower or equal to 6 mm. • However, this campaign highlighted some specific areas of improvement to be implemented ideally before the upcoming ITER Divertor IVT serial production. • The issue of the self-castellation of the W monoblocks, which typically appears after a few tenths of cycles at 20 MW/m"2, is critical because it generates some uncontrolled defects at the amour to heat sink joints. Besides, they create a gap which exposure is almost perpendicular to the magnetic field lines and which might lead to local W melting in the strike point region. • This campaign also evidenced that the minimum IO requirements on the CuCrZr ductility could be revised to avoid the occurrence of rather early fatigue failures. Although the W material characterization program has been set up by the IO, the strategy on the CuCrZr still needs to be defined. - Abstract: With the aim to assess the option to start the ITER operation with a full tungsten divertor, an R&D program was launched in order to evaluate the performances of tungsten (W) armoured plasma facing components (PFCs) under high heat flux. The F4E program consisted in the manufacturing and high heat flux (HHF) testing of W monoblock mock-ups and medium scale prototypes up to 20 MW/m"2. During the test campaign, 26 W mock-ups and two medium scale prototypes manufactured by Plansee SE (Austria) and by Ansaldo Nucleare (Italy) have been tested at the FE200 (AREVA, Le Creusot, France) and ITER Divertor Test Facility (IDTF) (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facilities. The high heat flux (HHF) testing program

  5. High heat flux testing of EU tungsten monoblock mock-ups for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Gavila, P., E-mail: pierre.gavila@f4e.europa.eu [Fusion for Energy, 08019 Barcelona (Spain); Riccardi, B. [Fusion for Energy, 08019 Barcelona (Spain); Pintsuk, G. [Forschungszentrum Juelich, 52425 Juelich (Germany); Ritz, G. [AREVA NP, Centre Technique France, 71205 Le Creusot (France); Kuznetsov, V. [JCS “Efremov Institute”, Doroga na Metallostroy 3, Metallostroy, Saint-Petersburg 196641 (Russian Federation); Durocher, A. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 Saint Paul-lez-Durance (France)

    2015-10-15

    Highlights: • All the tested items sustained the ITER Full W divertor qualification program requirements. This confirms that the technology for the manufacturing of the first set of the ITER Divertor is available in Europe. • The surface roughening and local melting of the W surface under high heat flux was proven to be significantly reduced for an armour thickness lower or equal to 6 mm. • However, this campaign highlighted some specific areas of improvement to be implemented ideally before the upcoming ITER Divertor IVT serial production. • The issue of the self-castellation of the W monoblocks, which typically appears after a few tenths of cycles at 20 MW/m{sup 2}, is critical because it generates some uncontrolled defects at the amour to heat sink joints. Besides, they create a gap which exposure is almost perpendicular to the magnetic field lines and which might lead to local W melting in the strike point region. • This campaign also evidenced that the minimum IO requirements on the CuCrZr ductility could be revised to avoid the occurrence of rather early fatigue failures. Although the W material characterization program has been set up by the IO, the strategy on the CuCrZr still needs to be defined. - Abstract: With the aim to assess the option to start the ITER operation with a full tungsten divertor, an R&D program was launched in order to evaluate the performances of tungsten (W) armoured plasma facing components (PFCs) under high heat flux. The F4E program consisted in the manufacturing and high heat flux (HHF) testing of W monoblock mock-ups and medium scale prototypes up to 20 MW/m{sup 2}. During the test campaign, 26 W mock-ups and two medium scale prototypes manufactured by Plansee SE (Austria) and by Ansaldo Nucleare (Italy) have been tested at the FE200 (AREVA, Le Creusot, France) and ITER Divertor Test Facility (IDTF) (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facilities. The high heat flux (HHF) testing

  6. High heat flux testing of ITER ICH&CD antenna beryllium faraday screen bars mock-ups

    International Nuclear Information System (INIS)

    Courtois, X.; Meunier, L.; Kuznetsov, V.; Beaumont, B.; Lamalle, P.; Conchon, D.; Languille, P.

    2016-01-01

    Highlights: • ITER ICH&CD antenna beryllium faraday screen bars mock-ups were manufactured. • The mock-ups are submitted to high heat loads to test their heat exhaust capabilities. • The mock-ups withstand without damage the design limit load. • Lifetime is gradually reduced when the heat load is augmented beyond the design limit. • Thermal and mechanical behavior are reproducible, and coherent with the calculation. - Abstract: The Faraday Screen (FS) is the plasma facing component of ITER ion cyclotron heating antennas shielding. The requirement for the high heat exhaust, and the limitation of the temperatures to minimize strain and thus offer sufficient resistance to fatigue, imply the need for high conductivity materials and a high cooling flow rate. The FS bars are constructed by a hipping process involving beryllium tiles, a pure copper layer, a copper chrome zirconium alloy for the cooling channel and a stainless steel backing strip. Two FS bars small scale mock-ups were manufactured and tested under high heat flux. They endured 15,000 heating cycles without degradation under nominal heat flux, and revealed growing flaws when the heat flux was progressively augmented beyond. In this case, the ultrasonic test confirms a strong delamination of the Be tiles.

  7. High RF power test of a lower hybrid module mock-up in carbon fiber composite

    International Nuclear Information System (INIS)

    Goniche, M.; Bibet, P.; Brossaud, J.; Cano, V.; Froissard, P.; Kazarian, F.; Rey, G.; Maebara, S.; Kiyono, K.; Seki, M.; Suganuma, K.; Ikeda, Y.; Imai, T.

    1999-02-01

    A mock-up module of a Lower Hybrid Current Drive antenna module of a Carbon Fiber Composite (CFC) was fabricated for the development of heat resistive front facing the plasma. This module is made from CFC plates and rods which are copper coated to reduce the RF losses. The withstand-voltage, the RF properties and outgassing rates for long pulses and high RF power were tested at the Lower Hybrid test bed facility of Cadarache. After the short pulse conditioning, long pulses with a power density ranging between 50 and 150 MW/m 2 were performed with no breakdowns. During these tests, the module temperature was increasing from 100-200 deg. C to 400-500 deg. C. It was also checked that high power density, up to 150 MW/m 2 , could be transmitted when the waveguides are filled with H 2 at a pressure of 5 x 10 -2 Pa. No significant change in the reflection coefficient is measured after the long pulse operation. During a long pulse, the power reflection increases during the pulse typically from 0.8% to 1.3%. It is concluded that the outgassing rate of Cu-plated CFC is about 6 times larger than of Dispersion Strengthened Copper (DSC) module at the module temperature of 300 deg. C. No significant increase of the global outgassing of the CFC module was measured after hydrogen pre-filling. After the test, visual inspection revealed that peeling of the copper coating occurred at one end of the module only on a very small area (0.2 cm 2 ). It is assessed that a CFC module is an attractive candidate for the hardening of the tip of the LHCD antenna. (authors)

  8. High RF power test of a lower hybrid module mock-up in Carbon Fiber Composite

    International Nuclear Information System (INIS)

    Maebara, Sunao; Kiyono, Kimihiro; Seki, Masami

    1997-11-01

    A mock-up module of a Lower Hybrid Current Drive antenna module of a Carbon Fiber Composite (CFC) was fabricated for the development of heat resistive front facing the plasma. This module is made from CFC plates and rods which are copper coated to reduce the RF losses. The withstand-voltage, the RF properties and outgassing rates for long pulses and high RF power were tested at the Lower Hybrid test bed facility of Cadarache. After the short pulse conditioning, long pulses with a power density ranging between 50 and 150 MW/m 2 were performed with no breakdowns. During these tests, the module temperature was increasing from 100-200degC to 400-500degC. It was also checked that high power density, up to 150 MW/m 2 , could be transmitted when the waveguides are filled with H 2 at a pressure of 5 x 10 -2 Pa. No significant change in the reflection coefficient is measured after the long pulse operation. During a long pulse, the power reflection increases during the pulse typically from 0.8 % to 1.3 %. It is concluded that the outgassing rate of Cu-plated CFC is about 6-7 times larger than of Dispersion Strengthened Copper (DSC) module at the module temperature of 300degC. No significant increase of the global outgassing of the CFC module was measured after hydrogen prefilling. After the test, visual inspection revealed that peeling of the copper coating occurred at one end of the module only on a very small area (0.2 cm 2 ). It is assessed that a CFC module is an attractive candidate for the hardening of the tip of the LHCD antenna. (author)

  9. Visualization test using piping group mock up specimen for evaluation of wastage phenomena in steam generator for FBR

    International Nuclear Information System (INIS)

    Kato, Keisuke; Yoshida, Atsuro; Arae, Kunihiko; Narabayashi, Tadashi; Ohshima, Hiroyuki; Kurihara, Akikazu

    2012-01-01

    There is a need for quantitative evaluation of wastage phenomena in steam generator for FBR. We focused attention on liquid droplet impingement erosion (LDIE) in wastage phenomena and performed basic study with piping group mock up specimen for quantitative evaluation of LDIE. First, we did visualization test of high pressure and high speed jet into the water. Test section mock up the crack of heat exchanger tube and neighboring heat exchanger tubes. We did the test under the following test conditions. Upstream pressure is 0.3MPa, vapor temperature is 300K, crack width is 0.1mm, and crack length is 40mm. (crack diameter is 0.2mm) Second, we did pressure and temperature measurement test in the same test conditions as before. We evaluated jet behavior at test section by those two tests. In addition, we did two phase flow analysis of the jet with TRAC code. (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. 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).

  12. Inspection of heat transfer tubes after mock-up tests of miniaturized apparatus for the acid recovery evaporator. Contract research

    International Nuclear Information System (INIS)

    Hamada, Shozo; Fukaya, Kiyoshi; Kato, Chiaki; Yanagihara, Takao; Doi, Masamitu; Kiuchi, Kiyoshi

    2001-10-01

    The demonstration test for the acid recovery evaporator and the dissolver used in the major equipment of Rokkasho Reprocessing Plant (RRP), has been carried out. The mock-up miniature equipment has been employed to it. This test had been performed from April in 1998. The total time of demonstration test using the mock-up equipment is about two and half years, which corresponds to about 20,000 hours. After that, four of the seven heat transfer tubes used in the evaporator were drawn out and the corrosion level and the mechanical properties were evaluated for one of them. As a result, intergranular corrosion was recognized in the inner surface of the heat transfer tube and the corrosion depth at the grain boundary was statistically shown to be about one grain from the inner surface. Further, no change in mechanical properties was observed and growth of intergranular cracks in the inner surface of the specimen was found after flattering test. (author)

  13. Thermal cycling tests of 1st wall mock-ups with beryllium/CuCrZr bonding

    International Nuclear Information System (INIS)

    Uda, M.; Iwadachi, T.; Uchida, M.; Yamada, H.; Nakamichi, M.; Kawamura, H.

    2004-01-01

    The innovative bonding technology between beryllium and CuCrZr with Hot Isostatic Pressing (HIP) has been proposed for the manufacturing of the ITER first wall. In the next step, thermal cycling test of first wall mock-ups manufactured with the bonding technology, were carried out under the ITER heat load condition. The test condition is 1000 cycles of On and Off under 5 MW/m 2 , and two types of the mock-up were manufactured for evaluation of the effects on HIP temperature (520 degree C and 610 degree C). The tensile properties of the bonding were also evaluated in room temperature and 200 degree C. As for the results of the thermal cycling tests, the temperature near the bonding interface were scarcely any change up to 1000 cycles, and obvious damage of the mock-up was not detected under the tests. As for the results of the tensile tests in 200 degree C, the test pieces of the HIP bonding at 610 degree C were broken in parent CuCrZr material, not broken in the bonding interface. (author)

  14. Tests and measurements with a thermal VXD mock-up for BELLE II

    International Nuclear Information System (INIS)

    Huebner, Lars

    2015-03-01

    As part of the Belle detector upgrade, located at the KEK in Tsukuba, Japan, a CO 2 cooling system will be added. Using new detector components, which are easily damageable or influenced by heat, make this step necessary. Particularly the next to the beam pipe located PXD is strained by high thermal load and therefore requires cooling. The CDC needs a constant temperature for precise measurements, but it could be influenced by heat from the SVD. Knowledge about the heat generation and distribution is needed before assembling the full detector. A mock-up of the innermost parts of the detector and a CO 2 cooling system is under construction at DESY in Hamburg, Germany, to gather such knowledge. The mock-up should be able to emulate the thermal properties of the final detector. Within the scope of this bachelor's thesis, the outermost VXD Layer 6 was studied in a flat arrangement. Focus lay on the heat dissipation at the sensors and on pressure drop measurements of the cooling pipe. It was investigated whether the applied heat load can be sufficiently lead away and how large the pressure drop is along the experiment line. Despite cooling was applied, a remarkable rise in temperature was observed. However, the unfavorable position of the thermistors make reliable quantitative statements of the sensor dummies' temperatures impossible. The pressure drop was determined, but is of limited accuracy due to large uncertainties. Further investigations have to be made with a better set-up.

  15. Measurement and analysis of neutron flux spectra in a neutronics mock-up of the HCLL test blanket module

    International Nuclear Information System (INIS)

    Klix, A.; Batistoni, P.; Boettger, R.; Lebrun-Grandie, D.; Fischer, U.; Henniger, J.; Leichtle, D.; Villari, R.

    2010-01-01

    Fast neutron and gamma-ray flux spectra and time-of-arrival spectra of slow neutrons have been measured in a neutronics mock-up of the European Helium-Cooled Lithium-Lead Test Blanket Module with the aim to validate nuclear cross-section data. The mock-up was irradiated with fusion peak neutrons from the DT neutron generator of the Technical University of Dresden. A well characterized cylindrical NE-213 scintillator was inserted into two positions in the LiPb/EUROFER assembly. Pulse height spectra from neutrons and gamma-rays were recorded from the NE-213 output. The spectra were then unfolded with experimentally obtained response matrices of the NE-213 detector. Time-of-arrival spectra of slow neutrons were measured with a 3 He counter placed in the mock-up, and the neutron generator was operated in pulsed mode. Monte Carlo calculations using the MCNP code and nuclear cross-section data from the JEFF-3.1.1 and FENDL-2.1 libraries were performed and the results are compared with the experimental results. A good agreement of measurement and calculation was found with some deviations in certain energy intervals.

  16. Tests of load resilient matching procedures for the ITER ICRH system on a mock-up and layout proposals

    International Nuclear Information System (INIS)

    Dumortier, P.; Lamalle, P.; Messiaen, A.; Vervier, M.

    2006-01-01

    The ICRH antenna of ITER consists of an array of 24 radiating straps and must radiate 20 MW with resilience to load variations due to the ELMs. Because of its compactness the mutual coupling effects between the straps are far from negligible. Moreover they considerably increase the difficulty of matching and lead to coupling between the generators. Different external matching system layouts are under consideration. A reduced scale (1/5) mock-up loaded by a movable water tank is used for their experimental investigation. A first layout using full passive power distribution among the straps and a single matching circuit with one '' Conjugate-T '' (CT) or one hybrid has already been successfully tested. Its drawbacks are the difficulty of changing the toroidal phasing and the use of a single 20 MW feeding line section. In this paper we describe the mock-up tests of a second layout based on two 10 MW CT circuits, and allowing switching between heating or current drive phasings without any hardware modification. Two decouplers are used to minimize the effect of mutual coupling on matching. A robust four-parameter CT matching procedure has been developed based on adjusting the two first parameters - the positions of the line stretchers in the CT branches - of each CT in vacuum conditions (this is done once for all for each frequency). High load resilience, i.e. a VSWR remaining < 1.5 for an 8-fold increase of antenna resistance, can be obtained for the 4 toroidal phasing configurations considered: (0π/2π3π/2), (0-π/2-π-3π/2), (00ππ) and (0ππ0). The change of phasing only requires the adjustment of the phase difference between the two power sources and of the two last parameters (stub and line stretcher in the common line) of each of the two CT circuits. These properties have first been derived from the experimental scattering matrix of the antenna array and are verified by reflection measurements on the mock-up. Feedback control of the phasing and the last two

  17. Construction of PREMUX and preliminary experimental results, as preparation for the HCPB breeder unit mock-up testing

    Energy Technology Data Exchange (ETDEWEB)

    Hernández, F., E-mail: francisco.hernandez@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany); Kolb, M. [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-WPT) (Germany); Annabattula, R. [Indian Institute of Technology Madras (IITM), Department of Mechanical Engineering (India); Weth, A. von der [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany)

    2014-10-15

    Highlights: • PREMUX has been constructed as preparation for a future out-of-pile thermo-mechanical qualification of a HCPB breeder unit mock-up. • The rationale and constructive details of PREMUX are reported in this paper. • PREMUX serves as a test rig for the new heater system developed for the HCPB-BU mock-up. • PREMUX will be used as benchmark for the thermal and thermo-mechanical models developed in ANSYS for the pebble beds of the HCPB-BU. • Preliminary results show the functionality of PREMUX and the good agreement of the measured temperatures with the thermal model developed in ANSYS. - Abstract: One of the European blanket designs for ITER is the Helium Cooled Pebble Bed (HCPB) blanket. The core of the HCPB-TBM consists of so-called breeder units (BUs), which encloses beryllium as neutron multiplier and lithium orthosilicate (Li{sub 4}SiO{sub 4}) as tritium breeder in form of pebble beds. After the design phase of the HCPB-BU, a non-nuclear thermal and thermo-mechanical qualification program for this device is running at the Karlsruhe Institute of Technology. Before the complex full scale BU testing, a pre-test mock-up experiment (PREMUX) has been constructed, which consists of a slice of the BU containing the Li{sub 4}SiO{sub 4} pebble bed. PREMUX is going to be operated under highly ITER-relevant conditions and has the following goals: (1) as a testing rig of new heater concept based on a matrix of wire heaters, (2) as benchmark for the existing finite element method (FEM) codes used for the thermo-mechanical assessment of the Li{sub 4}SiO{sub 4} pebble bed, and (3) in situ measurement of thermal conductivity of the Li{sub 4}SiO{sub 4} pebble bed during the tests. This paper describes the construction of PREMUX, its rationale and the experimental campaign planned with the device. Preliminary results testing the algorithm used for the temperature reconstruction of the pebble bed are reported and compared qualitatively with first analyses

  18. Comparison between FEM and high heat flux thermal fatigue testing results of ITER divertor plasma facing mock-ups

    Energy Technology Data Exchange (ETDEWEB)

    Crescenzi, F., E-mail: fabio.crescenzi@enea.it; Roccella, S.; Visca, E.; Moriani, A.

    2014-10-15

    Highlights: • Divertor is an important part of the ITER machine. • Finite element analysis allows designers to explore multiple design options, reducing physical prototypes and optimizing design performance. • The hydraulic thermal-mechanical analysis performed by ANSYS and the test results on small-scale mock-ups manufactured by HRP were compared. • FEA results confirmed many experimental data, then it could be very useful for next design optimization. - Abstract: The divertor is one of the most challenging components of “DEMO” the next step ITER machine, so many tasks regarding modeling and experiments have been made in the past years to assess manufacturing processes, materials and thus the life-time of the components. In this context the finite element analysis (FEA) allows designers to explore multiple design options, to reduce physical prototypes and to optimize design performance. The comparison between the hydraulic thermal-mechanical analysis performed by ANSYS WORKBENCH 14.5 and the test results [1] on small-scale mock-ups manufactured with the Hot Radial Pressing (HRP) [2] technology is presented in this paper. During the thermal fatigue testing in the Efremov TSEFEY facility to assess the heat flux load-carrying capability of the mock-ups, only the surface temperature was measured, so the FEA was important because it allowed to know any other information (temperature inside the materials, local water temperature, local stress, etc.). FEA was performed coupling the thermal-hydraulic analysis, that calculated the temperature distributions on the components and the heat transfer coefficient (HTC) between water and heat sink tube, with the mechanical analysis. The comparison between analysis and testing results was based on the temperature maps of the loaded surface and on number of the cycles supported during the testing and those predicted by the mechanical analysis using the experimental fatigue curves for CuCrZr-IG, that is the structural

  19. Tests on a mock-up of the feedback controlled matching options of the ITER ICRH system

    International Nuclear Information System (INIS)

    Grine, D.; Vervier, M.; Messiaen, A.; Dumortier, P.

    2009-01-01

    Automatic control of the matching of the ITER ICRH antenna array on a reference load is presently developed and tested for optimization on a low-powered scaled (1:5) mock-up. Resilience to fast load variations is obtained either by 4 Conjugate-T (CT) or 4 quadrature hybrid circuits, the latter being the reference option. The main results are (i) for the CT option: successful implementation of the simultaneous feedback control of 11 actuators for the matching of the 4 CT and for the control of the array toroidal phasing; (ii) for the hybrid option: the matching and the array current control via feedback control of the decouplers and double stub tuners. This system is being progressively implemented and the simultaneous control of matching and antenna current has already been successfully tested on half of the array for heating and current drive phasings.

  20. Examination of C/C flat tile mock-ups with hypervapotron cooling after high heat flux testing

    International Nuclear Information System (INIS)

    Schedler, B.; Friedrich, T.; Traxler, H.; Eidenberger, E.; Scheu, C.; Clemens, H.; Pippan, R.; Escourbiac, F.

    2007-01-01

    Two C/C flat tile mock-ups with a hypervapotron cooling concept, have been successfully tested beyond ITER specification (3000 cycles at 15 MW/m 2 , 300 cycles at 20 MW/m 2 and 800-1000 cycles at 25 MW/m 2 ) in two electron beam testing facilities [F. Escourbiac, et al., Experimental simulation of cascade failure effect on tungsten and CFC flat tile armoured HHF components, Fusion Eng. Des., submitted for publication; F. Escourbiac, et al., A mature industrial solution for ITER divertor plasma facing components: hypervapotron cooling concept adapted to Tore Supra flat tile technology, Fusion Eng. Des. 75-79 (2005) 387-390]. Both mock-ups provide a SNECMA SEPCARB NS31 armour, which has been joined onto the CuCrZr heat sink by active metal casting (AMC) and electron beam welding (EBW). No tile detachment or sudden loss of single tiles has been observed; a cascade-like failure of flat tile armours was impossible to generate. At the maximum cyclic heat flux load of 25 MW/m 2 all tested tiles performed well except one, which revealed already a clear indication in the thermographic examination at the end of the manufacture. Visual examination and analysis of metallographic cuts of the remaining tiles demonstrated that the interface has not been altered. In addition, the shear strength of the C/C to copper joints measured after the high heat flux (HHF) test has been found to be still above the interlamellar shear strength of the used C/C material. The high resistance of the interface is explained by a modification of the C/C to copper joint interface due to silicon originating from the used C/C material

  1. Examination of C/C flat tile mock-ups with hypervapotron cooling after high heat flux testing

    Energy Technology Data Exchange (ETDEWEB)

    Schedler, B. [Technology Centre of PLANSEE SE, A-6600 Reutte (Austria)], E-mail: bertram.schedler@plansee.com; Friedrich, T.; Traxler, H. [Technology Centre of PLANSEE SE, A-6600 Reutte (Austria); Eidenberger, E.; Scheu, C.; Clemens, H. [Department of Physical Metallurgy and Materials Testing, University of Leoben, A-8700 Leoben (Austria); Pippan, R. [Austrian Academy of Sciences, Erich-Schmid-Institute of Material Science, A-8700 Leoben (Austria); Escourbiac, F. [Association EURATOM-CEA, DSM/DRFC, CEA Cadarache, F-13108 St. Paul Lez Durance (France)

    2007-04-15

    Two C/C flat tile mock-ups with a hypervapotron cooling concept, have been successfully tested beyond ITER specification (3000 cycles at 15 MW/m{sup 2}, 300 cycles at 20 MW/m{sup 2} and 800-1000 cycles at 25 MW/m{sup 2}) in two electron beam testing facilities [F. Escourbiac, et al., Experimental simulation of cascade failure effect on tungsten and CFC flat tile armoured HHF components, Fusion Eng. Des., submitted for publication; F. Escourbiac, et al., A mature industrial solution for ITER divertor plasma facing components: hypervapotron cooling concept adapted to Tore Supra flat tile technology, Fusion Eng. Des. 75-79 (2005) 387-390]. Both mock-ups provide a SNECMA SEPCARB NS31 armour, which has been joined onto the CuCrZr heat sink by active metal casting (AMC) and electron beam welding (EBW). No tile detachment or sudden loss of single tiles has been observed; a cascade-like failure of flat tile armours was impossible to generate. At the maximum cyclic heat flux load of 25 MW/m{sup 2} all tested tiles performed well except one, which revealed already a clear indication in the thermographic examination at the end of the manufacture. Visual examination and analysis of metallographic cuts of the remaining tiles demonstrated that the interface has not been altered. In addition, the shear strength of the C/C to copper joints measured after the high heat flux (HHF) test has been found to be still above the interlamellar shear strength of the used C/C material. The high resistance of the interface is explained by a modification of the C/C to copper joint interface due to silicon originating from the used C/C material.

  2. Behavior of underclad cracks in reactor pressure vessels - evaluation of mechanical analyses with tests on cladded mock-ups

    International Nuclear Information System (INIS)

    Moinereau, D.; Rousselier, G.; Bethmont, M.

    1993-01-01

    Innocuity of underclad flaws in the reactor pressure vessels must be demonstrated in the French safety analyses, particularly in the case of a severe transient at the end of the pressure vessel lifetime, because of the radiation embrittlement of the vessel material. Safety analyses are usually performed with elastic and elasto-plastic analyses taking into account the effect of the stainless steel cladding. EDF has started a program including experiments on large size cladded specimens and their interpretations. The purpose of this program is to evaluate the different methods of fracture analysis used in safety studies. Several specimens made of ferritic steel A508 C1 3 with stainless steel cladding, containing small artificial defects, are loaded in four-point bending. Experiments are performed at very low temperature to simulate radiation embrittlement and to obtain crack instability by cleavage fracture. Three tests have been performed on mock-ups containing a small underclad crack (with depth about 5 mn) and a fourth test has been performed on one mock-up with a larger crack (depth about 13 mn). In each case, crack instability occurred by cleavage fracture in the base metal, without crack arrest, at a temperature of about - 170 deg C. Each test is interpreted using linear elastic analysis and elastic-plastic analysis by two-dimensional finite element computations. The fracture are conservatively predicted: the stress intensity factors deduced from the computations (K cp or K j ) are always greater than the base metal toughness. The comparison between the elastic analyses (including two plasticity corrections) and the elastic-plastic analyses shows that the elastic analyses are often conservative. The beneficial effect of the cladding in the analyses is also shown : the analyses are too conservative if the cladding effects is not taken into account. (authors). 9 figs., 6 tabs., 10 refs

  3. HHF test with 80x80x1 Be/Cu/SS Mock-ups for verifying the joining technology of the ITER blanket First Wall

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Bae, Young Dug; Kim, Suk Kwon; Hong, Bong Guen; Jeong, Yong Hwan; Park, Jeong Yong; Choi, Byung Kwon; Jung, Hyun Kyu

    2008-11-15

    Through the fabrication of the Cu/SS and Be/Cu joint specimens, fabrication procedure such as material preparation, canning, degassing, HIP (Hot Isostatic Pressing), PHHT (Post HIP heat treatment) was established. The HIP conditions (1050 .deg. C, 100 MPa 2 hr for Cu/SS, 580 .deg. C 100 MPa 2 hr for Be/Cu) were developed through the investigation on joint specimen fabricated with the various HIP conditions; the destructive tests of joint include the microstructure observation of the interface with the examination of the elemental distribution, tension test, bend test, Charpy impact test and fracture toughness test. However, since the joint should be tested under the High Heat Flux (HHF) conditions like the ITER operation for verifying its joint integrity, several HHF tests were performed like the previous HHF test with the Cu/SS, Be/Cu, Be/Cu/SS Mock-ups. In the present study, the HHF test with Be/Cu/SS Mock-ups, which have 80 mm x 80 mm single Be tile and each material depths were kept to be the same as the ITER blanket FW. The Mock-ups fabricated with three kinds of interlayers such as Cr/Ti/Cu, Ti/Cr/Cu, Ti/Cu, which were different from the developed interlayer (Cr/Cu), total 6 Mock-ups were fabricated. Preliminary analysis were performed to decide the test conditions; they were tested with up to 2.5 MW/m2 of heat fluxes and 20 cycles for each Mock-up in a given heat flux. They were tested with JUDITH-1 at FZJ in Germany. During tests, all Mock-ups showed delamination or full detachment of Be tile and it can be concluded that the joints with these interlayers have a bad joining but it can be used as a good data for developing the Be/Cu joint with HIP.

  4. Cold Pump Test and Training and Mock-Up Facility Functions and Requirements

    International Nuclear Information System (INIS)

    BELLOMY, J.R.

    2000-01-01

    This document defines the functions and requirements (F and R) for a test facility to provide for pre-deployment, checkout, testing, and training for the underground storage tank retrieval equipment, systems, and crews that will be developed or deployed as part of Waste Feed Delivery. The F and R for a River Protection Project retrieval test facility, one that supports a production mode tank farm system, are identified

  5. Mock-up tests of rail-mounted vehicle type in-vessel transporter/manipulator

    International Nuclear Information System (INIS)

    Oka, K.; Kakaudate, S.; Fukatsu, S.

    1995-01-01

    A rail-mounted vehicle system has been developed for remote maintenance of in-vessel components for fusion experimental reactor. In this system, a rail deploying/storing system is installed at outside of the reactor core and used to deploy a rail transporter and vehicle/manipulator for the in-vessel maintenance. A prototype of the rail deploying/storing system has been fabricated for mockup tests. This paper describes structural design of the prototypical rail deploying/storing system and results of the performance tests such as payload capacity, position control and rail deployment/storage performance

  6. Ultrasonic non-destructive testing on CFC monoblock divertor mock-up

    International Nuclear Information System (INIS)

    Ezato, K.; Taniguchi, M.; Sato, K.; Araki, M.; Akiba, M.

    2001-01-01

    Non-destructive ultrasonic testing has been applied for the characterization of joints by means of a polymer transducer. One of the advantages of the polymer transducer is flexibility in its shape and the possibility to install multiple transducers in one probe, which can reduce the time for inspection. As a first step, the size effect of the transducer on the resolution and sensitivity was examined to detect the joint flaw. Transducers with circumferential angles of 5 , 10 and 30 were tested. For this test a small divertor element with a driller hole was prepared, which simulates a joint defect. The transducers with angles of 30 could not characterize the size of the artificial joint flaw. On the contrary, the size of the artificial defect was successfully detected with an accuracy of 90% by means of the transducers with angles of 5 and 10 . From the viewpoint of the sensitivity of the detection of the joint flaw, the transducer with the angle of 10 is appropriate because it could detect the largest intensity of the reflected signal caused by the same artificial defect of the joint interface. (orig.)

  7. Chemical-Cleaning Demonstration Test No. 2 in a mock-up steam generator

    International Nuclear Information System (INIS)

    Jevec, J.M.; Leedy, W.S.

    1983-04-01

    This report describes the results of the mockup demonstration test of the first modified baseline process under Contract S-127, Chemical Cleaning of Nuclear Steam Generators. The objective of this program is to determine the feasibility of cleaning the secondary side of nuclear steam generators with state-of-the-art chemical cleaning technology. The first step was to benchmark a baseline process. This process was then modified to attempt to eliminate the causes of unacceptable cleaning performance. The modified baseline process consists of an EDTA/H 2 O 2 -based copper solvent and a near-neutral, EDTA/N 2 H 4 -based magnetite and crevice solvent. This report also presents the results of three inhibitor evaluation mockup runs used in the evaluation of the modified baseline process

  8. Thermal fatigue testing of a diffusion-bonded beryllium divertor mock-up under ITER relevant conditions

    International Nuclear Information System (INIS)

    Youchison, D.L.; Guiniiatouline, R.; Watson, R.D.

    1994-01-01

    Thermal response and thermal fatigue tests of four 5 mm thick beryllium tiles on a Russian divertor mock-up were completed on the Electron Beam Test System at Sandia National Laboratories. The beryllium tiles were diffusion bonded onto an OFHC copper saddleblock and a DSCu (MAGT) tube containing a porous coating. Thermal response tests were performed on the tiles to an absorbed heat flux of 5 MW/m 2 and surface temperatures near 300 degrees C using 1.4 MPa water at 5.0 m/s flow velocity and an inlet temperature of 8-15 degrees C. One tile was exposed to incrementally increasing heat fluxes up to 9.5 MW/m 2 and surface temperatures up to 690 degrees C before debonding at 10 MW/m 2 . A third tile debonded after 9200 thermal fatigue cycles at 5 MW/m 2 , while another debonded after 6800 cycles. In all cases, fatigue failure occurred in the intermetallic layers between the beryllium and copper. No fatigue cracking of the bulk beryllium was observed. During thermal cycling, a gradual loss of porous coating produced increasing sample temperatures. These experiments indicate that diffusion-bonded beryllium tiles can survive several thousand thermal cycles under ITER relevant conditions without failure. However, the reliability of the diffusion bonded Joint remains a serious issue

  9. Remote handling of the blanket segments: Testing of 1/3 scale mock-ups on the ROBERTINO facility

    International Nuclear Information System (INIS)

    Maisonnier, D.; Amelotti, F.; Chiasera, A.

    1994-01-01

    The remotized replacement of the blanket segments inside the Vacuum Vessel of a fusion reactor is one of the critical tasks for reactor components design, operational procedures, and safety. This open-quotes hostile environmentclose quotes task must be accomplished by a specific Blanket Handling Device, with a grasping device acting as open-quotes end-effectorclose quotes, because of intervention complexity, of components dimensions and weights, and of consequences of possible accidents during the blanket segments handling operations. Therefore, specific support experimental studies in this field appear to be necessary in order to: select appropriate blanket handling devices and procedures; assess the design of all components involved in the handling operations; perform checks in all field related to the robotized handling control (kinematics and dynamics of the grasping device trajectory planning and motion control, sensing and intelligence of the blanket handling devices, etc.); improve reliability and safety for the replacement sequences; give a realistic estimation of the time duration of the replacement duration. During the test phase, handling operations were carried out on the blanket mock-ups by means of different gripping devices. The operations were driven in the control room by means of the Motion command computer and the real time sensing data display allowed operations' control. The results were analyzed by charting the sensors' data

  10. Manufacturing and testing of W/Cu mono-block small scale mock-up for EAST by HIP and HRP technologies

    Energy Technology Data Exchange (ETDEWEB)

    Li, Qiang [Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), Hefei, Anhui (China); Qin, Sigui [Advanced Technology and Materials Co., Ltd, Beijing (China); Wang, Wanjing; Qi, Pan [Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), Hefei, Anhui (China); Roccella, Selanna; Visca, Eliseo [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Liu, Guohui [Advanced Technology and Materials Co., Ltd, Beijing (China); Luo, Guang-Nan, E-mail: liqiang577@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), Hefei, Anhui (China)

    2013-10-15

    ITER-like W/Cu mono-block plasma-facing components (PFCs) will be used in vertical target regions of the experimental advanced superconducting tokamak (EAST) divertor. The first W/Cu mono-block small scale mock-up with five W mono-blocks has been manufactured successfully by technological combination of hot isostatic pressing (HIP) and hot radial pressing (HRP). The joining of a W mono-block and a pure copper interlayer was achieved by means of HIP technology and the bonding strength was over 150 MPa. The good bonding between the pure copper interlayer and a CuCrZr cooling tube was obtained by means of HRP technology. In order to understand deeply the process of HRP, the stress distribution of the mock-up during HRP process was simulated using ANSYS code. Ultrasonic Nondestructive Testing (NDT) of the W/Cu and Cu/CuCrZr interfaces was performed, showing that excellent bonding of the W/Cu and Cu/CuCrZr interfaces. The thermal cycle fatigue testing of the mock-up has been carried out by means of an e-beam device in Southwest Institute of Physics, Chengdu (SWIP) and the mock-up withstood 1000 cycles of heat loads up to 8.4 MW/m{sup 2} with the cooling water of 2 m/s, 20 °C, 0.2 MPa.

  11. Strength of Mock-up Trial Grout

    DEFF Research Database (Denmark)

    Sørensen, Eigil V.

    The present report describes tests carried out on samples taken and cast during the execution of a mock-up trial placement of the high performance grout MASTERFLOW 9500 on January 21, 2009.......The present report describes tests carried out on samples taken and cast during the execution of a mock-up trial placement of the high performance grout MASTERFLOW 9500 on January 21, 2009....

  12. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    International Nuclear Information System (INIS)

    Nakahira, M.; Takeda, N.; Kakudate, S.; Onozuka, M.

    2008-01-01

    The structure and fabrication methods of the ITER vacuum vessel (VV) have been investigated and defined by the ITER International Team (IT). However, some of the current technical specifications are difficult to be achieved from the manufacturing point of view. To solve such an issue, this paper proposes an alternative specification of the VV to the IT's design. A series of the fabrication and assembly procedures for the mock-up are presented, together with candidates of welding configurations. Finally, the paper summarizes the results of mock-up fabrication, such as non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. Based on the results, it is suggested that several issues such as clarification of conditions of repair welding, demonstration of welding distortion control and detectability/localization of internal defects should be solved before manufacturing the ITER VV

  13. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, M. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan)], E-mail: nakahira.masataka@jaea.go.jp; Takeda, N.; Kakudate, S. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Onozuka, M. [Mitsubishi Nuclear Energy Systems, Inc., 1700K Street NW, Suite 440, Washington, DC 20006 (United States)

    2008-12-15

    The structure and fabrication methods of the ITER vacuum vessel (VV) have been investigated and defined by the ITER International Team (IT). However, some of the current technical specifications are difficult to be achieved from the manufacturing point of view. To solve such an issue, this paper proposes an alternative specification of the VV to the IT's design. A series of the fabrication and assembly procedures for the mock-up are presented, together with candidates of welding configurations. Finally, the paper summarizes the results of mock-up fabrication, such as non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. Based on the results, it is suggested that several issues such as clarification of conditions of repair welding, demonstration of welding distortion control and detectability/localization of internal defects should be solved before manufacturing the ITER VV.

  14. Low drift type N thermocouples in out-of-pile advanced gas reactor mock-up test: metallurgical analysis

    International Nuclear Information System (INIS)

    Scervini, M.; Palmer, J.; Haggard, D.C.; Swank, W.D.

    2015-01-01

    Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. They are crucial for the control of current nuclear reactors and for the development of GEN IV reactors. In nuclear applications thermocouples are strongly affected by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. Previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of Nickel based thermocouples is limited to temperatures lower than 1000 deg. C due to drift related to phenomena other than nuclear irradiation. As part of a collaboration between Idaho National Laboratory (INL) and the University of Cambridge a variety of Type N thermocouples have been exposed at INL in an Advanced Gas Reactor mock-up test at 1150 deg. C for 2000 h, 1200 deg. C for 2000 h, 125 deg. C for 200 h and 1300 deg. C for 200 h, and later analysed metallurgically at the University of Cambridge. The use of electron microscopy allows to identify the metallurgical changes occurring in the thermocouples during high temperature exposure and correlate the time dependent thermocouple drift with the microscopic changes experienced by the thermoelements of different thermocouple designs. In this paper conventional Inconel 600 sheathed type N thermocouples and a type N using a customized sheath developed at the University of

  15. Low drift type N thermocouples in out-of-pile advanced gas reactor mock-up test: metallurgical analysis

    Energy Technology Data Exchange (ETDEWEB)

    Scervini, M. [University of Cambridge, Department of Materials Science and Metallurgy, 27 Charles Babbage Road, CB30FS Cambridge, (United Kingdom); Palmer, J.; Haggard, D.C.; Swank, W.D. [Idaho National Laboratory, Idaho Falls, ID 83415-3840, (United States)

    2015-07-01

    Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. They are crucial for the control of current nuclear reactors and for the development of GEN IV reactors. In nuclear applications thermocouples are strongly affected by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. Previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of Nickel based thermocouples is limited to temperatures lower than 1000 deg. C due to drift related to phenomena other than nuclear irradiation. As part of a collaboration between Idaho National Laboratory (INL) and the University of Cambridge a variety of Type N thermocouples have been exposed at INL in an Advanced Gas Reactor mock-up test at 1150 deg. C for 2000 h, 1200 deg. C for 2000 h, 125 deg. C for 200 h and 1300 deg. C for 200 h, and later analysed metallurgically at the University of Cambridge. The use of electron microscopy allows to identify the metallurgical changes occurring in the thermocouples during high temperature exposure and correlate the time dependent thermocouple drift with the microscopic changes experienced by the thermoelements of different thermocouple designs. In this paper conventional Inconel 600 sheathed type N thermocouples and a type N using a customized sheath developed at the University of

  16. 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.

  17. Design and validation of the THMC China-Mock-Up test on buffer material for HLW disposal

    Directory of Open Access Journals (Sweden)

    Yuemiao Liu

    2014-04-01

    Full Text Available According to the preliminary concept of the high-level radioactive waste (HLW repository in China, a large-scale mock-up facility, named China-Mock-Up was constructed in the laboratory of Beijing Research Institute of Uranium Geology (BRIUG. A heater, which simulates a container of radioactive waste, is placed inside the compacted Gaomiaozi (GMZ-Na-bentonite blocks and pellets. Water inflow through the barrier from its outer surface is used to simulate the intake of groundwater. The numbers of water injection pipes, injection pressure and the insulation layer were determined based on the numerical modeling simulations. The current experimental data of the facility are herein analyzed. The experiment is intended to evaluate the thermo-hydro-mechano-chemical (THMC processes occurring in the compacted bentonite-buffer during the early stage of HLW disposal and to provide a reliable database for numerical modeling and further investigation of engineered barrier system (EBS, and the design of HLW repository.

  18. FEBEX II Project Final report on thermo-hydro-mechanical laboratory tests

    Energy Technology Data Exchange (ETDEWEB)

    Lloret, A.; Romero, E.; Villar, M. V.

    2004-07-01

    The results of the thermo-hydro-mechanical (THM) study of the FEBEX bentonite performed during FEBEX II are presented. The laboratory test program continued in part with the works carried out during FEBEX I, particularly in activities related to tests aimed to the calibration of the models, the acquisition of parameters by back-analysis and the improvement of the knowledge on the behaviour of expansive clays. But the program has also included tests on new areas: investigations about the influence of the microstructure changes in bentonite, of temperature and of the solute concentration on the behaviour of clay. Besides, several tests were proposed in order to understand the unexpected behaviour observed in the mock-up test, towards the end of year 2. Temperature effects on water retention curves in confined and unconfined conditions were determined, and swelling pressure, hydraulic conductivity and swelling and consolidation strains as a function of temperature were successfully measured. Different experimental techniques and equipments were developed to study thermal induced changes under partially saturated states, covering a wide range of suctions. FEBEX bentonite remains suitable as a sealing material in HLW repositories (from the hydro- mechanical point of view) for temperatures of up to 80 C, as it keeps its high water retention capacity, low permeability and self-healing ability. The extrapolation of results points out to the preservation of properties for at least up to 100 C. Mercury intrusion porosimetry and environmental scanning electron microscopy provided promising results in order to characterise the bentonite microstructure and to give information about the mechanisms influencing pore size distribution changes on high active clays. The use of digital imaging techniques allowed verifying that at micro-scale level, where chemical phenomena prevail, strains are almost reversible as it is considered in the two-level elasto-plastic models. The swelling

  19. FEBEX II Project Final report on thermo-hydro-mechanical laboratory tests

    International Nuclear Information System (INIS)

    Lloret, A.; Romero, E.; Villar, M. V.

    2004-01-01

    The results of the thermo-hydro-mechanical (THM) study of the FEBEX bentonite performed during FEBEX II are presented. The laboratory test program continued in part with the works carried out during FEBEX I, particularly in activities related to tests aimed to the calibration of the models, the acquisition of parameters by back-analysis and the improvement of the knowledge on the behaviour of expansive clays. But the program has also included tests on new areas: investigations about the influence of the microstructure changes in bentonite, of temperature and of the solute concentration on the behaviour of clay. Besides, several tests were proposed in order to understand the unexpected behaviour observed in the mock-up test, towards the end of year 2. Temperature effects on water retention curves in confined and unconfined conditions were determined, and swelling pressure, hydraulic conductivity and swelling and consolidation strains as a function of temperature were successfully measured. Different experimental techniques and equipments were developed to study thermal induced changes under partially saturated states, covering a wide range of suctions. FEBEX bentonite remains suitable as a sealing material in HLW repositories (from the hydro- mechanical point of view) for temperatures of up to 80 C, as it keeps its high water retention capacity, low permeability and self-healing ability. The extrapolation of results points out to the preservation of properties for at least up to 100 C. Mercury intrusion porosimetry and environmental scanning electron microscopy provided promising results in order to characterise the bentonite microstructure and to give information about the mechanisms influencing pore size distribution changes on high active clays. The use of digital imaging techniques allowed verifying that at micro-scale level, where chemical phenomena prevail, strains are almost reversible as it is considered in the two-level elasto-plastic models. The swelling

  20. Siloette, Siloe mock-up

    International Nuclear Information System (INIS)

    Delcroix, V.; Jeanne, G.; Mitault, G.; Schulhof, P.

    1964-01-01

    Siloette is the Siloe mock-up. The main installations are described: various tanks, building, auxiliaries, control systems... Precis ions are given about precautions taken for using spent fuel elements. (authors) [fr

  1. Report on Thermo-Hydro-Mechanical Laboratory Tests Performed by CIEMAT on Febex Bentonite 2004-2008

    Energy Technology Data Exchange (ETDEWEB)

    Villar, M. V.; Gomez-Espina, R.

    2009-11-25

    The results of the laboratory studies performed by CIEMAT with the FEBEX bentonite in the context of WP3.2 of the NF-PRO Project and of the Agreement ENRESA-CIEMAT Anexo V are presented and analysed in this report. They refer to the effect of the hydraulic gradient on the permeability of bentonite, the effect of the thermal gradient on the hydration kinetics of bentonite, and the repercussion of temperature on the hydro-mechanical properties of bentonite (swelling, permeability and water retention capacity). In all the cases the bentonite has been used compacted to densities expected in the engineered barrier of a high-level radioactive waste repository. The existence of threshold and critical hydraulic gradients has been observed, both of them dependent on bentonite density and water pressures. After more than seven years of hydration, the 40-cm high bentonite columns are far from full saturation, the thermal gradient additionally delaying the process, which is very slow. Temperatures below 100 degree centigrade slightly decrease the swelling and the water retention capacity of the bentonite and increase its permeability. The information obtained improves the knowledge on the behaviour of expansive clay and will help the development of constitutive models and the interpretation of the results obtained in the mock-up and the in situ tests. (Author) 35 refs.

  2. Report on Thermo-Hydro-Mechanical Laboratory Tests Performed by CIEMAT on Febex Bentonite 2004-2008

    International Nuclear Information System (INIS)

    Villar, M. V.; Gomez-Espina, R.

    2009-01-01

    The results of the laboratory studies performed by CIEMAT with the FEBEX bentonite in the context of WP3.2 of the NF-PRO Project and of the Agreement ENRESA-CIEMAT Anexo V are presented and analysed in this report. They refer to the effect of the hydraulic gradient on the permeability of bentonite, the effect of the thermal gradient on the hydration kinetics of bentonite, and the repercussion of temperature on the hydro-mechanical properties of bentonite (swelling, permeability and water retention capacity). In all the cases the bentonite has been used compacted to densities expected in the engineered barrier of a high-level radioactive waste repository. The existence of threshold and critical hydraulic gradients has been observed, both of them dependent on bentonite density and water pressures. After more than seven years of hydration, the 40-cm high bentonite columns are far from full saturation, the thermal gradient additionally delaying the process, which is very slow. Temperatures below 100 degree centigrade slightly decrease the swelling and the water retention capacity of the bentonite and increase its permeability. The information obtained improves the knowledge on the behaviour of expansive clay and will help the development of constitutive models and the interpretation of the results obtained in the mock-up and the in situ tests. (Author) 35 refs

  3. Improvement works report on mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system (Contract research)

    International Nuclear Information System (INIS)

    Sakaki, Akihiro; Kato, Michio; Hayashi, Koji; Fujisaki, Katsuo; Aita, Hideki; Ohashi, Hirofumi; Takada, Shoji; Shimizu, Akira; Morisaki, Norihiro; Maeda, Yukimasa; Sato, Hiroyuki; Hanawa, Hiromi; Yonekawa, Hideo; Inagaki, Yoshiyuki

    2005-04-01

    In order to establish the system integration technology to connect a hydrogen production system to a high temperature gas cooled reactor; the mock-up test facility with a full-scale reaction tube for the steam reforming HTTR hydrogen production system was constructed in fiscal year 2001 and its functional test operation was performed in the year. Seven experimental test operations were performed from fiscal year 2001 to 2004. On a period of each test operation, there happened some troubles. For each trouble, the cause was investigated and the countermeasures and the improvement works were performed to succeed the experiments. The tests were successfully achieved according to plan. This report describes the improvement works on the test facility performed from fiscal year 2001 to 2004. (author)

  4. Realisation of a test facility for the ITER ICRH antenna plug-in by means of a mock-up with salted water load

    International Nuclear Information System (INIS)

    Messiaen, A.; Dumortier, P.; Koch, R.; Lamalle, P.; Louche, F.; Martini, J.L.; Vervier, M.

    2005-01-01

    By the use of a mock-up operated at higher frequency it is possible to measure with good accuracy the rf characteristics of an ICRH antenna, the plasma loading being simulated by a water tank in front of it. This concept has motivated the construction of the mock-up of the antenna array foreseen for ITER

  5. Performance test results of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Inagaki, Yoshiyuki; Hayashi, Koji; Kato, Michio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2003-03-01

    Research on a hydrogen production system by steam reforming of methane, chemical reaction; CH{sub 4} + H{sub 2}O {yields} 3H{sub 2}O + CO, has been carried out to couple with the HTTR for establishment of high-temperature nuclear heat utilization technology and contribution to hydrogen energy society in future. The mock-up test facility with a full-scale reaction tube test facility, a model simulating one reaction tube of a steam reformer of the HTTR hydrogen production system in full scale, was fabricated to perform tests on controllability, hydrogen production performance etc. under the same pressure and temperature conditions as those of the HTTR hydrogen production system. The design and fabrication of the test facility started from 1997, and the all components were installed until September in 2001. In a performance test conducted from October in 2001 to February in 2002, performance of each component was examined and hydrogen of 120m{sup 3}{sub N}/h was successfully produced with high-temperature helium gas. This report describes the performance test results on components performance, hydrogen production characteristics etc., and main troubles and countermeasures. (author)

  6. Performance test results of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system. Contract research

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Hayashi, Koji; Kato, Michio

    2003-03-01

    Research on a hydrogen production system by steam reforming of methane, chemical reaction; CH 4 + H 2 O → 3H 2 O + CO, has been carried out to couple with the HTTR for establishment of high-temperature nuclear heat utilization technology and contribution to hydrogen energy society in future. The mock-up test facility with a full-scale reaction tube test facility, a model simulating one reaction tube of a steam reformer of the HTTR hydrogen production system in full scale, was fabricated to perform tests on controllability, hydrogen production performance etc. under the same pressure and temperature conditions as those of the HTTR hydrogen production system. The design and fabrication of the test facility started from 1997, and the all components were installed until September in 2001. In a performance test conducted from October in 2001 to February in 2002, performance of each component was examined and hydrogen of 120m 3 N /h was successfully produced with high-temperature helium gas. This report describes the performance test results on components performance, hydrogen production characteristics etc., and main troubles and countermeasures. (author)

  7. ISSUES AND FEASIBILITY DEMONSTRATION OF CLIC SUPPORTING SYSTEM CHAIN ACTIVE PRE-ALIGNMENT USING A MULTI-MODULE TEST SETUP (MOCK-UP)

    CERN Document Server

    Sosin, Mateusz

    2016-01-01

    The implementation study of the CLIC (Compact LInear Collider) is under way at CERN with a focus on the challenging issues. The pre-alignment precision and accuracy requirements are part of these technical challenges: the permissible transverse position errors of the linac components are typically 14 micrometers over sliding windows of 200m. To validate the proposed methods and strategies, the Large Scale Metrology section at CERN has performed campaigns of measurements on the CLIC Two Beam Test Modules, focusing inter alia on the alignment performance of the CLIC “snake”- girders configuration and the Main Beam Quadrupoles supporting structures. This paper describes the activities and results of tests which were performed on the test mock-up for the qualification of the CLIC supporting system chain for active pre-alignment. The lessons learnt (“know how”), the issues encountered in the girder position determination as well as the behaviour of the mechanical components are presented.

  8. Validation of a Micrometric remotely controlled pre-alignment system for the CLIC Linear Collider using a test setup (Mock-Up) with 5 degrees of freedom

    CERN Document Server

    Mainaud Durand, H; Griffet, S; Kemppinen, J; Leuxe, R; Sosin, M

    2011-01-01

    The CLIC main beam quadrupoles need to be prealigned within 17 um rms with respect to a straight reference line along a sliding window of 200 m. A readjustment system based on eccentric cam movers, which will provide stiffness to the support assembly, is being studied. The cam movers were qualified on a 1 degree of freedom (DOF) test setup, where a repeatability of adjustment below 1um was measured along their whole range. This paper presents the 5 DOF mock-up, built for the validation of the eccentric cam movers, as well as the first results of tests carried out: resolution of displacement along the whole range, measurements of the support eigenfrequencies.

  9. Measurement and Analysis of the Neutron and Gamma-Ray Flux Spectra in a Neutronics Mock-Up of the HCPB Test Blanket Module

    International Nuclear Information System (INIS)

    Seidel, K.; Freiesleben, H.; Poenitz, E.; Klix, A.; Unholzer, S.; Batistoni, P.; Fischer, U.; Leichtle, D.

    2006-01-01

    The nuclear parameters of a breeding blanket, such as tritium production rate, nuclear heating, activation and dose rate, are calculated by integral folding of an energy dependent cross section (or coefficient) with the neutron (or gamma-ray) flux energy spectra. The uncertainties of the designed parameters are determined by the uncertainties of both the cross section data and the flux spectra obtained by transport calculations. Also the analysis of possible discrepancies between measured and calculated integral nuclear parameter represents a two-step procedure. First, the energy region and the amount of flux discrepancies has to be found out and second, the cross section data have to be checked. To this end, neutron and gamma-ray flux spectra in a mock-up of the EU Helium-Cooled Pebble Bed (HCPB) breeder Test Blanket Module (TBM), irradiated with 14 MeV neutrons, were measured and analysed by means of Monte Carlo transport calculations. The flux spectra were determined for the energy ranges that are relevant for the most important nuclear parameters of the TBM, which are the tritium production rate and the shielding capability. The fast neutron flux which determines the tritium production on 7 Li and dominates the shield design was measured by the pulse-height distribution obtained from an organic liquid scintillation detector. Simultaneously, the gamma-ray flux spectra were measured. The neutron flux at lower energies, down to thermal, which determines the tritium production on 6 Li, was measured with time-of-arrival spectroscopy. For this purpose, the TUD neutron generator was operated in pulsed mode (pulse width 10 μs, frequency 1 kHz) and the neutrons arriving at a 3 He proportional counter in the mock-up were recorded as a function of time after the source neutron pulse. The spectral distributions for the two positions in the mock-up, where measurements were carried out, were calculated with the Monte Carlo code MCNP, version 5, and nuclear data from the

  10. Issues and Feasibility Demonstration of Positioning Closed Loop Control for the CLIC Supporting System Using a Test Mock-up with Five Degrees of Freedom

    CERN Document Server

    Sosin, M; Chritin, N; Griffet, S; Kemppinen, J; Mainaud Durand, H; Rude, V; Sterbini, G

    2012-01-01

    Since several years, CERN is studying the feasibility of building a high energy e+ e- linear collider: the CLIC (Compact LInear Collider). One of the challenges of such a collider is the pre-alignment precision and accuracy requirement on the transverse positions of the linac components, which is typically 14 μm over a window of 200 m. To ensure the possibility of positioning within such tight constraints, CERN Beams Department’s Survey team has worked intensively at developing the methods and technology needed to achieve that objective. This paper describes activities which were performed on a test bench (mock-up) with five degrees of freedom (DOF) for the qualification of control algorithms for the CLIC supporting system active-pre-alignment. Present understanding, lessons learned (“know how”), issues of sensors noise and mechanical components nonlinearities are presented.

  11. Performance test results of helium gas circulator of mock-up test facility with full-scale reaction tube for HTTR hydrogen production system. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Akira; Kato, Michio; Hayashi, Koji [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2003-03-01

    Hydrogen production system by steam reforming of methane will be connected to the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) against development of nuclear heat utilization system. To obtain design and safety database of the HTTR hydrogen production system, mock-up test facility with full-scale reaction was constructed in FY 2001 and hydrogen of 120m{sup 3}N{sub /}h was successfully produced in overall performance test. This report describes performance test results of a helium gas circulator in this facility. The circulator performance curves regarding to pressure-rise, input power and adiabatic thermal efficiency at standard revolution number were made based on the measured flow-rate, temperature and pressure data in overall performance test. The circulator performance prediction code was made based on these performance curves. The code can calculate revolution number, electric power and temperature-rise of the circulator using flow-rate, inlet temperature, inlet pressure and pressure-rise data. The verification of the code was carried out with the test data in FY 2002. Total pressure loss of the helium gas circulation loop was also evaluated. The circulator should be operated in conditions such as pressure from 2.7MPa to 4.0MPa and flow-rate from 250g/s to 400g/s and at maximum pressure-rise of 250 kPa in test operation. It was confirmed in above verification and evaluations that the circulator had performance to satisfy above conditions within operation limitation of the circulator such as maximum input-power of 150 kW and maximum revolution number of 12,000 rpm. (author)

  12. Heat transfer characteristics evaluation of heat exchangers of mock-up test facility with full-scale reaction tube for HTTR hydrogen production system (Contract research)

    International Nuclear Information System (INIS)

    Shimizu, Akira; Ohashi, Hirofumi; Kato, Michio; Hayashi, Koji; Aita, Hideki; Nishihara, Tetsuo; Inaba, Yoshitomo; Takada, Shoji; Morisaki, Norihiro; Sakaki, Akihiro; Maeda, Yukimasa; Sato, Hiroyuki; Inagaki, Yoshiyuki; Hanawa, Hiromi; Fujisaki, Katsuo; Yonekawa, Hideo

    2005-06-01

    Connection of hydrogen production system by steam reforming of methane to the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) has been surveyed until now. Mock-up test facility of this steam reforming system with full-scale reaction tube was constructed in FY 2001 and hydrogen of 120 Nm 3 /h was successfully produced in overall performance test. Totally 7 times operational tests were performed from March 2002 to December 2004. A lot of operational test data on heat exchanges were obtained in these tests. In this report specifications and structures of steam reformer, steam superheater, steam generator, condenser, helium gas cooler, feed gas heater and feed gas superheater were described. Heat transfer correlation equations for inside and outside tube were chosen from references. Spreadsheet programs were newly made to evaluate heat transfer characteristics from measured test data such as inlet and outlet temperature pressure and flow-rate. Overall heat-transfer coefficients obtained from the experimental data were compared and evaluated with the calculated values with heat transfer correlation equation. As a result, actual measurement values of all heat exchangers gave close agreement with the calculated values with correlation equations. Thermal efficiencies of the heat exchangers were adequate as they were well accorded with design value. (author)

  13. Remote handling of the blanket segments: testing of 1/3 scale mock-ups at the Robertino facility

    International Nuclear Information System (INIS)

    Maisonnier, D.; Amelotti, F.; Chiasera, A.; Gaggini, P.; Damiani, C.; Degli Esposti, L.; Gatti, G.; Castillo, E.; Caravati, D.; Farfalletti-Casali, F.; Gritzmann, P.; Ruiz, E.

    1995-01-01

    The remote replacement of blanket segments inside the vacuum vessel of a fusion reactor is probably the most complex task from the maintenance standpoint. Its success will rely on the definition of appropriate handling concepts and equipment, but also on a ''maintenance friendly'' reactor layout and blanket design. The key difficulty is the lack of rigidity of the segments which results in considerable deformations since they cannot be gripped above their centre of gravity. These deformations may be up to five times greater than the assembly clearance and one order of magnitude larger than the required positioning accuracy. Experimental activities have been undertaken to select appropriate handling devices and procedures, to assess the design of the components handled, and to review specific technical issues such as kinematics and dynamics performance, trajectory planning and control and sensors requirement for the handling devices. Work was performed in the Robertino facility where two handling concepts have been tested at a 1/3 scale. (orig.)

  14. Non destructive testing of concrete nuclear containment plants with surface waves: Lab experiment on decimeter slabs and on the VeRCoRs mock-up

    Science.gov (United States)

    Abraham, Odile; Legland, Jean-Baptiste; Durand, Olivier; Hénault, Jean-Marie; Garnier, Vincent

    2018-04-01

    The maintenance and evaluation of concrete nuclear containment walls is a major concern as they must, in case of an accident, ensure the confinement of the nuclear radiations and resist to the loads. A homemade multi-receiver multi-source dry contact linear probe to record ultrasonic surface waves on concrete in the frequency range [60 kHz - 200 kHz] has been used in this context. The measurement protocol includes the summation of up to 50 spatially distributed seismograms and the determination of the surface waves phase velocity dispersion curve. The probe has been tested against several concrete states under no loading (water saturation level, temperature damage). Then, the same measurements have been performed on sound and fire damaged slabs submitted to uniaxial loading (stress up to 30 % of the concrete compression resistance). It is shown that the robustness and precision of the surface waves measurement protocol make it possible to follow the stress level. In March 2017 a first experiment with this surface wave probe has been conducted on a reduced 1:3 scale nuclear containment plant (EDF VeRCoRs mock-up) under loading conditions that replicates that of decennial inspection. The surface wave phase velocity dispersion curves of each state are compared and cross-validated with other NDT results.

  15. Damage and fatigue crack growth of Eurofer steel first wall mock-up under cyclic heat flux loads. Part 1: Electron beam irradiation tests

    Energy Technology Data Exchange (ETDEWEB)

    You, J.H., E-mail: you@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Euratom Association, Boltzmannstr. 2, 85748 Garching (Germany); Höschen, T. [Max-Planck-Institut für Plasmaphysik, Euratom Association, Boltzmannstr. 2, 85748 Garching (Germany); Pintsuk, G. [Forschungszentrum Jülich GmbH, IEK2, Euratom Association, 52425 Jülich (Germany)

    2014-04-15

    Highlights: • Clear evidence of microscopic damage and crack formation at the notch root in the early stage of the fatigue loading (50–100 load cycles). • Propagation of fatigue crack at the notch root in the course of subsequent cyclic heat-flux loading followed by saturation after roughly 600 load cycles. • No sign of damage on the notch-free surface up to 800 load cycles. • No obvious effect of the pulse time duration on the crack extension. • Slight change in the grain microstructure due to the formation of sub-grain boundaries by plastic deformation. - Abstract: Recently, the idea of bare steel first wall (FW) is drawing attention, where the surface of the steel is to be directly exposed to high heat flux loads. Hence, the thermo-mechanical impacts on the bare steel FW will be different from those of the tungsten-coated one. There are several previous works on the thermal fatigue tests of bare steel FW made of austenitic steel with regard to the ITER application. In the case of reduced-activation steel Eurofer97, a candidate structural material for the DEMO FW, there is no report on high heat flux tests yet. The aim of the present study is to investigate the thermal fatigue behavior of the Eurofer-based bare steel FW under cyclic heat flux loads relevant to DEMO operation. To this end, we conducted a series of electron beam irradiation tests with heat flux load of 3.5 MW/m{sup 2} on water-cooled mock-ups with an engraved thin notch on the surface. It was found that the notch root region exhibited a marked development of damage and fatigue cracks whereas the notch-free surface manifested no sign of crack formation up to 800 load cycles. Results of extensive microscopic investigation are reported.

  16. Experimental investigation of the IFMIF target mock-up

    International Nuclear Information System (INIS)

    Loginov, N.; Mikheyev, A.; Morozov, V.; Aksenov, Yu.; Arnol'dov, M.; Berensky, L.; Fedotovsky, V.; Chernov, V.; Nakamura, H.

    2009-01-01

    The international fusion materials irradiation facility (IFMIF) lithium neutron target mock-ups have been constructed and tested at water and lithium test facilities in the IPPE of Russia. Jet velocity in both mock-ups was up to 20 m/s. Calculations and experiments showed lithium flow instability at conjunction point of straight and concave sections of the mock-up back wall. Water velocity profile across the mock-up width, jet thickness, and wave height were measured. The significant increase of thickness of both water and lithium jets near the mock-up sidewalls was observed. The influence of shape of the nozzle outlet part on jet stability was investigated. Lithium evaporation from the jet free surface was investigated as well as lithium deposition on vacuum pipe walls of the target mock-up. It was shown that these phenomena are not very critical for the target efficiency. The possibility of lithium denitration down to 2 ppm (at 10 ppm requested) by means of aluminium getter was shown. Two types of cold traps and plug indicators of impurities were tested. The results are presented in the paper.

  17. FST-formation of cryogenic layer inside spherical shells of HiPER-class. Results of mathematical modeling and mock-ups testing

    International Nuclear Information System (INIS)

    Belolipetskiy, A.A.; Lalinina, E.A.; Panina, L.V.

    2010-01-01

    ; Layering stage optimization (computation and experiments): First step requires (a) choosing an optimal temperature of target input into the layering channel, (b) analyzing the 'liquid-vapor' interface behavior at different cooling rates, and (c) computation of the FST layering time; A preliminary concept of the FST-layering module for HiPER-class targets (including the interface unit for target-and-sabot assembly) is presented. The results of the mock-ups testing are discussed, namely: (a) FST layering channels (LC) of different geometry have been created and the time of target residence inside LC has been measured for 2 mm diameter targets of different weight and material, including a HiPER-class surrogate target. This allows determining the requirements on the LC manufacturing. (b) Mock-up of target positioning device has been constructed and tested. It was found that the device ensures a comprehensive look at the target for the time less than 1 sec. Acknowledgements. This work supported by the International Science and Technology Center (ISTC) under contract No. 3927 (partner of the project is Science and Technology Facilities Council (STFC), UK).

  18. Electron beam irradiation experiments of monoblock divertor mock-up

    International Nuclear Information System (INIS)

    Satoh, Kazuyoshi; Akiba, Masato; Araki, Masanori; Suzuki, Satoshi; Yokoyama, Kenji; Smid, I.; Cardella, A.; Duwe, R.; Di Pietro, E.

    1993-03-01

    It is one of the key issues for ITER to develop the divertor plate. Electron beam irradiation tests were carried out on a NET divertor mock-up using JEBIS at JAERI under a collaboration between The NET team, JAERI and KFA Juelich. Screening tests (maximum heat flux of 23 MW/m 2 ) and thermal cycling tests (18 MW/m 2 , 30s, 1000cycle) were carried out. As a result of the screening tests, the erosion caused by sublimation of C/C was observed on the surface of armor tile. No serious damage such as cracks or detachments, however, were found. As a result of the thermal cycling tests, no major damage was detected on the C/C surface. However cooling time constant of the divertor mock-up increased over 600cycle. Therefore it implies that some defects would occur at the brazing interface of the divertor mock-up. (author)

  19. FBR core mock-up RAPSODIE I - experimental analysis

    International Nuclear Information System (INIS)

    Brochard, D.; Buland, P.; Gantenbein, F.

    1990-01-01

    The main phenomena which influence the LMFBR core response to a seismic excitation are the fluid structure interaction and the impacts between subassemblies. To study the core behaviour, seismic tests have been performed on the core mock-up RAPSODIE with and without fluid and restraint ring and for different levels of excitation. This paper summarizes the results of these tests. (author)

  20. Annual report on experimental operation of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system in 2001 fiscal year (Contract research)

    International Nuclear Information System (INIS)

    Hayashi, Koji; Inagaki, Yoshiyuki; Kato, Michio; Fujisaki, Katsuo; Aita, Hideki; Takeda, Tetsuaki; Nishihara, Tetsuo; Inaba, Yoshitomo; Ohashi, Hirofumi; Katanishi, Shoji; Takada, Shoji; Shimizu, Akira; Morisaki, Norihiro; Sakaki, Akihiro; Maeda, Yukimasa; Sato, Hiroyuki

    2005-06-01

    This is an annual report on the experimental operation of the mock-up test facility with a full-scale reaction tube for the HTTR hydrogen production system in 2001 fiscal year. The first experimental operation was performed during two weeks from March 1, 2002 to March 13, 2002 to test on the thermal hydraulic performance of the steam reformer and also to train the operators. The thermal hydraulic performance test of the steam reformer was performed to evaluate the heat transfer characteristics between helium gas and process gas in the steam reformer. This report is summarized with an overview of the test, the results and its operation records. (author)

  1. Annual report on experimental operations and maintenances of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system in 2003 fiscal year (Contract research)

    International Nuclear Information System (INIS)

    Hayashi, Koji; Morisaki, Norihiro; Ohashi, Hirofumi; Kato, Michio; Aita, Hideki; Takeda, Tetsuaki; Nishihara, Tetsuo; Inaba, Yoshitomo; Takada, Shoji; Inagaki, Yoshiyuki

    2006-03-01

    This is a report on the experimental operations and maintenances of the mock-up test facility with a full-scale reaction tube for the HTTR hydrogen production system in 2003 fiscal year. The fourth and fifth experimental test operations were performed, from May to July and from October to December in 2003, for the following tests; (a) start-up and shutdown operation test, (b) process change test, (c) continuous hydrogen-production test and (d) chemical reaction shutdown test. From the results, a long time-range stability of the hydrogen production system was confirmed, a behavior of the helium-gas cooling system, consists of steam generator and radiator; during chemical reaction shutdown, was understood, and so on. Periodic inspections on boiler equipment and high-pressure gas production facilities were performed from end of July 2003. This report is summarized on outlines and results of the tests, outlines and results of the periodic inspections, and operation records of the mock-up test facility. (author)

  2. Role of irradiation reactor mock-ups

    International Nuclear Information System (INIS)

    Casali, F.; Cerles, J.M.; Debrue, J.

    1977-01-01

    A survey is given of the utilization of low power facilities in support to irradiation reactor experiments. The BRO2, ISIS and RB3 facilities are described as neutronic mock-ups of the BR2, OSIRIS and ESSOR reactors respectively

  3. The making of a mock-up

    DEFF Research Database (Denmark)

    Rosenqvist, Tanja Schultz; Heimdal, Elisabeth Jacobsen

    2011-01-01

    As part of a research project about user involvement in textile design we have carried out two Design:Labs (Binder & Brandt 2008) engaging different stakeholders in designing textile products for Danish hospital environments. In this paper we follow a mock-up session done as part of the second...

  4. The numeracy test workbook everything you need for a successful programme of self study including quick tests and full-length realistic mock-ups

    CERN Document Server

    Bryon, Mike

    2011-01-01

    One of the most common types of psychometric test used in assessment and selection procedures, The Numeracy Test Workbook provides practice questions and mock tests designed to build confidence and improve performance.

  5. Annual report on experimental operations and maintenance of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system in 2004 fiscal year (Contract research)

    International Nuclear Information System (INIS)

    Hayashi, Koji; Ohashi, Hirofumi; Morisaki, Norihiro; Kato, Michio; Aita, Hideki; Takeda, Tetsuaki; Nishihara, Tetsuo; Inaba, Yoshitomo; Takada, Shoji; Inagaki, Yoshiyuki

    2006-03-01

    This is annual report on the experimental test operations and maintenances of the mock-up test facility with a full-scale reaction tube for the HTTR hydrogen production system in 2004 fiscal year. The improvement work of catalyst dust filter in combustion system was performed in May 2004, and the performance was confirmed. The sixth experimental test operation was performed from June to July 2004. Periodic inspections on boiler equipment and high-pressure gas production facilities were performed from end of July to September 2004. The seventh experimental test operation was performed from October to December 2004 for chemical reaction shutdown test. From the results, a behavior of the helium-gas cooling system, consists of steam generator and radiator, during chemical reaction shutdown was confirmed. This report is summarized with the outline and the results of the test, maintenance works and inspections, and operation records in mentioned above. (author)

  6. Characterization of flaws in a tube bundle mock-up for reliability studies

    International Nuclear Information System (INIS)

    Kupperman, D.S.; Bakhtiari, S.

    1997-01-01

    As part of an assessment of in-service inspection of steam generator tubes, the authors will assemble a steam generator mock-up for round robin studies and use as a test bed in evaluating emerging technologies. Progress is reported on the characterization of flaws that will be part of the mock-up. Eddy current and ultrasonic techniques are being evaluated as a means to characterize the flaws in the mock-up tubes before final assembly. Twenty Inconel 600 tubes with laboratory-grown cracks, typical of those to be used in the mock-up, were provided by Pacific Northwest National Laboratory for laboratory testing. After the tubes were inspected with eddy current and ultrasonic techniques, they were destructively analyzed to establish the actual depths, lengths, and profiles of the cracks. The analysis of the results will allow the best techniques to be used for characterizing the flaws in the mock-up tubes

  7. Characterization of flaws in a tube bundle mock-up for reliability studies

    International Nuclear Information System (INIS)

    Kupperman, D.S.; Bakhtiari, S.

    1996-10-01

    As part of an assessment of in-service inspection of steam generator tubes, the authors will assemble a steam generator mock-up for round robin studies and use as a test bed in evaluating emerging technologies. Progress is reported on the characterization of flaws that will be part of the mock-up. Eddy current and ultrasonic techniques are being evaluated as a means to characterize the flaws in the mock-up tubes before final assembly. Twenty Inconel 600 tubes with laboratory-grown cracks, typical of those to be used in the mock-up, were provided by Pacific Northwest National Laboratory for laboratory testing. After the tubes were inspected with eddy current and ultrasonic techniques, they were destructively analyzed to establish the actual depths, lengths, and profiles of the cracks. The analysis of the results will allow the best techniques to be used for characterizing the flaws in the mock-up tubes

  8. How tangible mock-ups support design collaboration

    DEFF Research Database (Denmark)

    Brandt, Eva

    2007-01-01

    This paper is a contribution to a more conscious use of tangible mock-ups in collaborative design processes. It describes a design team's use of mock-ups in a series of workshops involving potential customers and users. Focus is primarily on the use of three-dimensional design mock-ups and how...... differences in these affected the dialogue. Reflective conversations were established by using tangible mock-ups as 'things-to-think with'. They served as boundary objects that spanned the gap between the different competencies and interests of participants in design. The design mock-ups evoked different...... things for different participants whereas the challenge for the design team was to find boundaries upon which everybody could agree. The level of details represented in a mock-up affected the communication so that a mock-up with few details evoked different issues whereas a very detailed mock-up evoked...

  9. FEBEX II Project THG Laboratory Experiments

    International Nuclear Information System (INIS)

    Missana, T.

    2004-01-01

    anionic exclusion process and determine the accessible porosity to diffusion at different clay densities. To complete the diffusion coefficients database for the tracers used in the mock-up and in situ tests. The aim of this report is to summarise the main results obtained by all the research groups involved in the THG Laboratory Experiments programme, during the second phase of the FEBEX project. The report will be organised in four main blocks in which the contributions of different institutions will be collected. CIEMAT (Spain), PSI (Switzerland), CSIC (Spain) and VTT (Finland) contributed to the study of the pore water in the clay barrier; CSIC (Spain) contributed to the study of the geochemical processes at the solution/ bentonite interface and to the study of the effects of the interlayer cations on the rheological properties of bentonite. Finally CIEMAT (Spain) and PSI (Switzerland) contributed to investigate radionuclides sorption and migration in bentonite. (Author)

  10. FEBEX II Project THG Laboratory Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Missana, T.

    2004-07-01

    anionic exclusion process and determine the accessible porosity to diffusion at different clay densities. To complete the diffusion coefficients database for the tracers used in the mock-up and in situ tests. The aim of this report is to summarise the main results obtained by all the research groups involved in the THG Laboratory Experiments programme, during the second phase of the FEBEX project. The report will be organised in four main blocks in which the contributions of different institutions will be collected. CIEMAT (Spain), PSI (Switzerland), CSIC (Spain) and VTT (Finland) contributed to the study of the pore water in the clay barrier; CSIC (Spain) contributed to the study of the geochemical processes at the solution/ bentonite interface and to the study of the effects of the interlayer cations on the rheological properties of bentonite. Finally CIEMAT (Spain) and PSI (Switzerland) contributed to investigate radionuclides sorption and migration in bentonite. (Author)

  11. Annual report on experimental operations and maintenances of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system in 2002 fiscal year (Contract research)

    International Nuclear Information System (INIS)

    Hayashi, Koji; Ohashi, Hirofumi; Inaba, Yoshitomo; Kato, Michio; Aita, Hideki; Morisaki, Norihiro; Takeda, Tetsuaki; Nishihara, Tetsuo; Takada, Shoji; Inagaki, Yoshiyuki

    2006-03-01

    This report describes 2002 fiscal-year experimental test operations of the mock-up test facility with a full-scale reaction tube for the HTTR hydrogen production system. The improvement works were performed in May 2002. The second experimental test operation was performed from June 2002 and the performances of the improved parts were confirmed. Periodic inspections on boiler equipment and high-pressure gas production facilities were performed from end of July 2002. The third experimental test operation was performed, from October 2002, for (a) start-up and shutdown test, (b) process change test, (c) chemical reaction shutdown test and (d) characteristics test on steam reformer. It was confirmed that the changes of helium gas temperature, caused at steam reformer, could be mitigated into the target range by the steam generator. Maintenance works of high-pressure gas production facilities were also performed in February 2003. This report is summarized with the outline and the results of the test, maintenance works and inspections, and operation records in mentioned above. (author)

  12. Mock-up tests on the combustion of hydrogen-air mixture in the vertical tube simulating the CNS channel of the CARR

    International Nuclear Information System (INIS)

    Yu Qingfeng; Feng Quanke; Kawai, Takeshi; Xu Jian

    2007-01-01

    A two-phase thermo-siphon loop for removing nuclear heating and maintaining the stable liquid level in the moderator cell was adopted for the cold neutron source (CNS) of the China advanced research reactor (CARR). The moderator is liquid hydrogen. The two-phase thermo-siphon loop consists of the crescent-shape moderator cell, the moderator transfer tube, and the condenser. The hydrogen is supplied from the buffer tank to the condenser. The main feature of the loop is that the moderator cell is covered by the helium sub-cooling system. The cold helium gas from the helium refrigerator is firstly introduced into the helium sub-cooling system and then flows up through the tube covering the moderator transfer tube into the condenser. The main part of this system is installed in the CNS vertical channel made of aluminum alloy 6061 T6 (Al-6061-T6) of 6 mm in thickness, 270 mm in outer diameter and about 6 m in height. For confirming the safety of the CNS channel, the combustion tests using a tube compatible with the CNS channel were carried out using the hydrogen-air mixture under which air is introduced into the tube at 1 atmosphere, and then hydrogen gas is supplied from the gas cylinder up to the test pressures. And maximum test pressure is 0.14 MPa G. This condition is involved with the maximum design basis accident of the CARR-CNS. The peak pressure due to combustion was 1.09 MPa, and the design pressure of the CNS channel is 3 MPa. The safety of the CNS was thus verified even if the maximum design basis accident occurs. The pressure and stress distributions along the axial direction and the displacement of the tube were also measured

  13. Thermal hydraulic considerations and mock-up tests for developing two-phase thermo-siphon loop of CARR-CNS

    International Nuclear Information System (INIS)

    Shejiao, Du; Qincheng, Bi; Tingkuan, Chen; Quanke, Feng

    2005-01-01

    The main component of the China Advanced Research Reactor Cold Neutron Source (CARR-CNS), which is under design, is a two-phase thermo-siphon loop of hydrogen. It consists of a condenser, a single tube with counter current flow avoiding flooding and a cylindrical-annulus moderator cell. The mockup tests were carried out using a full-scale loop with Freon-113, to validate the self-regulating characteristics of the loop, void fraction less than 20% in the liquid of the moderator cell and the requirements for establishing the condition under which the inner shell of the moderator cell has only vapor and the outer shell liquid. In the case of these mockup tests the density ratio of liquid to vapor and the volumetric vapor evaporation rate due to heat load are kept the same as those in normal operation of the CARR-CNS. The results show that the loop has the self-regulating characteristics and the inner shell of the moderator cell contains only vapor, the outer shell liquid. The average void fraction of the moderator cell was verified less than 20% under the volumetric vapor generation of 0.65 l/s corresponding to the nuclear heating of 800 W in the case of the liquid hydrogen. The local void fraction in the liquid hydrogen increases with the increase of the loop pressure under the condition of a constant volumetric evaporation

  14. Thermo-hydro-mechanical characterization of the Spanish reference clay material for engineered barrier for granite and clay HLW repository: laboratory and small mock up testing

    International Nuclear Information System (INIS)

    Villar, M.V.

    1995-01-01

    This report refers to the work carried out by Technic Geologic Division of CIEMAT (CIEMAT.DT.TG) coordinated by SCK/CEN (Belgium), participating besides UPC-DIT and University of Wales on the framework of CEC Contract F12W-CT91-0102 (DOEO). It presents the results obtained. The total results on the project will be published by CE in the EUR series. The role of CIEMAT in this project was to carry out tests in which the conditions of the clay barrier in the repository were simulated. The interaction of heat coming from the wastes and of water coming from the geological medium has been reproduced on compacted clay blocks. For the performance of tests on high density compacted clay blocks (Task 2.1) and for the cementation and chemical-mineralogical transformation studies two different cells were designed and constructed in stainless steel: a thermohydraulic cell and an alteration cell. The experiments performed in these cells have provided us with a better knowledge of the heat source, hydration system and sensors, as well as interesting data on heat and water diffusion. A revision of the experiments performed on the thermohydraulic cell was presented at the ''International Workshop on Thermomechanics of Clays and Clay Barriers'' held in Bergamo in October'93 (Villar et al. 1993)

  15. Experimental Investigation of the IFMIF Target Mock-up

    International Nuclear Information System (INIS)

    Loginov, N.; Mikheyev, A.; Morozov, V.; Aksenov, Y.; Arnoldov, M.; Berensky, L.; Fedotovsky, V.; Chernov, V.M.; Nakamura, H.

    2007-01-01

    Full text of publication follows: The IFMIF lithium neutron target mock-ups have been constructed and tested at the water and lithium test facilities. Description of the mock-ups and test facilities is presented in the paper, as well as the main results obtained. Reference geometry was used but the mockup flow cross-section was decreased. Velocity of water and lithium was up to reference value of 20 m/s. Features of lithium and water hydrodynamics were observed. The calculations and experiments showed that conjunction point of back wall straight and concave sections generated instability of lithium flow because of centrifugal force sudden change at this place. Therefore, it was proposed to use parabolic shape of the target back wall. Generation of wakes at the corners of cross-section of the Shima nozzle outlet was observed, and, as a result, surface waves appeared on the lithium jet. Observations of lithium and water jets and measurements of water jet thickness showed significant increasing the thickness near sidewalls of the mock-up concave section. It is because of absence of the centrifugal force at these places. Very large instability of the water jet surface was observed when outlet part of the Shima nozzle was divergent slightly (about 1 deg.), and vice versa very smooth jet surface occurred in confusing case (of about 0.5 deg.). So, nozzle outlet shape is very critical. Evaporation of lithium from the jet surface was investigated as well as deposition of vapor on vacuum pipe wall. It turned out to be not so critical. Significant part of the work concerned purification of lithium and monitoring impurities. The possibility of denitration of lithium down to 2 ppm by means of aluminum soluble getter was showed. Two types of both cold traps and plug indicators of impurities were tested. The results are presented in the paper. (authors)

  16. Modelling the cooling and partial dismantling of the Febex in-situ test

    International Nuclear Information System (INIS)

    Sanchez, M.; Gens, A.; Guimaraes, L.

    2010-01-01

    predictions from analysis. The operation related to the partial dismantling included the demolition of the concrete plug and the removal of the sections of the barrier corresponding to 'Heater 1'. The objective was to carry out the partial dismantling causing minimum disturbance to the sections of test corresponding to the second heater, which remained in operation at all times. A new concrete plug was constructed immediately after excavation. A detailed description of the work performed during the partial dismantling of the in-situ test can be found in Huertas et al. (2006). This contribution focuses on the modelling of the cooling and partly dismantling of the FEBEX in-situ experiment. The finite element computer program CODE-BRIGHT has been used for the numerical analysis. CODE-BRIGHT is a program developed to handle coupled Thermo-Hydro- Mechanical and Geochemical problems in geological media. It has been observed a very good performance of the model to reproduce the evolution of the main THM variables of the tests, during the cooling of the Heater No.1, concrete demolition and excavation of the clay barrier. It is worth mentioning that these are a kind of 'blind model predictions', as the constitutive laws and model parameters adopted at the beginning of the heating were used in this analysis. (authors)

  17. FEBEX. Investigations on gas generation, release and migration

    International Nuclear Information System (INIS)

    Jockwer, Norbert; Wieczorek, Klaus

    2008-06-01

    The FEBEX project is based on the Spanish reference concept for the disposal of radioactive waste in crystalline rock, which considers the emplacement of the canisters enclosing the conditioned waste surrounded by clay barriers constructed of high-compacted bentonite blocks in horizontal drifts /ENR 957. The whole project consisted of an experimental and a modelling part. The experimental part itself was divided into the in-situ test, a mock-up test performed at the CIEMAT laboratory, and various small-scale laboratory tests. In the modelling part it was expected to develop and validate the thermo-hydro-mechanical (THM) and the thermo-hydro-chemical (THC) processes for the performance assessment of the near-field behaviour. GRS was only involved in the in-situ test and some additional laboratory work with regard to gas generation, gas migration, and pore pressure build-up in the buffer constructed of high-compacted bentonite blocks around the electrical heaters simulating the waste containers. The following topics are covered: installation and dismantling of the heater pipes; methods of gas generation and release measurement, summary of results and discussion

  18. DECOVALEX III PROJECT. Modelling of FEBEX In-Situ Test. Task1 Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, E.E.; Alcoverro, J. [Univ. Politecnica de Catalunya, Barcelona (Spain)] (comps.)

    2005-02-15

    Task 1 of DECOVALEX III was conceived as a benchmark exercise supported by all field and laboratory data generated during the performance of the FEBEX experiment designed to study thermo-hydro-mechanical and thermo-hydro-geochemical processes of the buffer and rock in the near field. The task was defined as a series of three successive blind prediction exercises (Parts A, B and C), which cover the behaviour of both the rock and bentonite barrier. Research teams participating in the FEBEX task were given, for each of the three parts, a set of field and laboratory data theoretically sufficient to generate a proper model and were asked to submit predictions, at given locations and time, for some of the measured variables. The merits and limitations of different modeling approaches were therefore established. The teams could perform additional calculations, once the actual 'solution' was disclosed. Final calculations represented the best approximation that a given team could provide, always within the general time constraints imposed by the General DECOVALEX III Organization. This report presents the works performed for Task 1. It contains the case definitions and evaluations of modelling results for Part A, B and C, and the overall evaluation of the works performed. The report is completed by a CD-ROM containing a set of final reports provided by the modeling teams participating in each of the three parts defined. These reports provide the necessary details to better understand the nature of the blind or final predictions included in this report. The report closes with a set of conclusions, which provides a summary of the main findings and highlights the lessons learned, some of which were summarized below. The best predictions of the water inflow into the excavated tunnel are found when the hydro geological model is properly calibrated on the basis of other known flow measurements in the same area. The particular idealization of the rock mass (equivalent

  19. DECOVALEX III PROJECT. Modelling of FEBEX In-Situ Test. Task1 Final Report

    International Nuclear Information System (INIS)

    Alonso, E.E.; Alcoverro, J.

    2005-02-01

    Task 1 of DECOVALEX III was conceived as a benchmark exercise supported by all field and laboratory data generated during the performance of the FEBEX experiment designed to study thermo-hydro-mechanical and thermo-hydro-geochemical processes of the buffer and rock in the near field. The task was defined as a series of three successive blind prediction exercises (Parts A, B and C), which cover the behaviour of both the rock and bentonite barrier. Research teams participating in the FEBEX task were given, for each of the three parts, a set of field and laboratory data theoretically sufficient to generate a proper model and were asked to submit predictions, at given locations and time, for some of the measured variables. The merits and limitations of different modeling approaches were therefore established. The teams could perform additional calculations, once the actual 'solution' was disclosed. Final calculations represented the best approximation that a given team could provide, always within the general time constraints imposed by the General DECOVALEX III Organization. This report presents the works performed for Task 1. It contains the case definitions and evaluations of modelling results for Part A, B and C, and the overall evaluation of the works performed. The report is completed by a CD-ROM containing a set of final reports provided by the modeling teams participating in each of the three parts defined. These reports provide the necessary details to better understand the nature of the blind or final predictions included in this report. The report closes with a set of conclusions, which provides a summary of the main findings and highlights the lessons learned, some of which were summarized below. The best predictions of the water inflow into the excavated tunnel are found when the hydro geological model is properly calibrated on the basis of other known flow measurements in the same area. The particular idealization of the rock mass (equivalent porous media

  20. Manufacturing of a HCLL cooling plate mock up

    International Nuclear Information System (INIS)

    Rigal, E.; Dinechin, G. de; Rampal, G.; Laffont, G.; Cachon, L.

    2007-01-01

    The European DEMO blankets and associated Test Blanket Modules (TBM) are made of a set of components cooled by flowing helium at 80bar pressure. Hot Isostatic Pressing (HIP) is one of the very few processes that allow manufacturing such components exhibiting complex cooling channels. In HIP technology, the parts used to manufacture components with embedded channels are usually machined plates, blocks and tubes. Achievable geometries are limited in shape because it is not always possible to figure the channels by bent tubes. This occurs for example when channels present sharp turns, when the cross section of the channels is rectangular or when the rib between channels is so small that very thin tubes would be required. In these cases, bending is unpractical. The breeder unit cooling plates of the Helium Cooled Lithium Lead (HCLL) blanket have eight 4 x 4.5 mm parallel channels that run following a double U scheme. Turns are sharp and the wall thickness is small (1mm), so the manufacturing process described above cannot be used. An alternative process has been developed which has many advantages. It consists in machining grooves in a base plate, then closing the top of the grooves using thin welded strips, and finally adding a plate by HIP. There is then no need for the use of tubes with associated bending and deformation issues. The final component contains welds, but it must be stressed out that these potentially brittle zones do not connect the channels to the external surface because they are covered by the HIPed plate. Furthermore, the welds are homogenised during the HIP operation and further heat treatments. This paper describes the design of a simplified cooling plate mock up and its fabrication using this so-called weld+HIP process. The thermal fatigue testing of this mock up is presented somewhere else in this conference. (orig.)

  1. Mock-up qualification and prototype manufacture for ITER current leads

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Tingzhi, E-mail: tingszhou@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Lu, Kun; Ran, Qingxiang; Ding, Kaizhong; Feng, Hansheng; Wu, Huan; Liu, Chenglian; Song, Yuntao [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Niu, Erwu [CNDA, Ministry of Science and Technology, Beijing (China); Bauer, Pierre; Devred, Arnaud [Magnet Division, ITER Organization, Cadarache (France)

    2015-10-15

    Highlights: • Vacuum brazing and electron beam welding qualification. • Machine and assembly strategy of fin type heat exchanger. • Soldering and joint resistance test of superconducting joint. • Pre-preg technology with vacuum bag on insulation. - Abstract: Three types of high temperature superconducting current leads (HTSCL) are designed to carry 68 kA, 55 kA or 10 kA to the ITER magnets. Before the supply of the HTS current lead series, the design and manufacturing process is qualified through mock-ups and prototypes. Seven mock-ups, representing the critical technologies of the current leads, were built and tested successfully in the Institute of Plasma Physics of the Chinese Academy of Sciences (ASIPP) in 2013. After the qualification some design features of the HTS leads were updated. This paper summarizes the qualification through mock-ups. In 2014 ASIPP started the manufacture of the prototypes. The preparation and manufacturing process are also described.

  2. Fabrication of ITER first wall mock-ups with beryllium armour

    International Nuclear Information System (INIS)

    Mohri, K.; Nomoto, Y.; Uda, M.; Enoeda, M.; Akiba, M.

    2004-01-01

    This paper presents the fabric ability development for the ITER first wall through the fabrication of a real size first wall panel mock-up without beryllium armor and a partial mock-up of the first wall panel with beryllium armor. Microscopic observation and mechanical test of the hot isostatic pressed Be/Cu-alloy joints were also performed of which results showed good bond ability of the joints. Finally the fabrication procedure of the ITER first wall panel has been established. (author)

  3. F.B.R. Core mock-up RAPSODIE- I: Experimental analysis

    International Nuclear Information System (INIS)

    Brochard, D.; Buland, P.; Gantenbein, F.

    1990-01-01

    The main phenomena which influence the LMFBR core response to a seismic excitation are the fluid structure interaction and the impacts between subassemblies. To study the core behaviour, seismic tests have been performed on the core mock-up RAPSODIE with or without fluid and restraint ring and for different levels of excitation. This paper summarizes the results of these tests

  4. F.B.R. Core mock-up RAPSODIE - II - numerical models

    International Nuclear Information System (INIS)

    Brochard, D.; Hammami, L.; Gantenbein, F.

    1990-01-01

    To study the behaviour of LMFBR cores excited by a seism, tests have been performed on the RAPSODIE core mock-up. The aim of this paper is to present the numerical models used to interprete these tests and the comparisons between calculations and experimental results

  5. A coupled THC model of the FEBEX in situ test with bentonite swelling and chemical and thermal osmosis

    International Nuclear Information System (INIS)

    Zheng, L.; Samper, J.; Montenegro, L.

    2011-01-01

    The performance assessment of a geological repository for radioactive waste requires quantifying the geochemical evolution of the bentonite engineered barrier. This barrier will be exposed to coupled thermal (T), hydrodynamic (H), mechanical (M) and chemical (C) processes. This paper presents a coupled THC model of the FEBEX (Full-scale Engineered Barrier EXperiment) in situ test which accounts for bentonite swelling and chemical and thermal osmosis. Model results attest the relevance of thermal osmosis and bentonite swelling for the geochemical evolution of the bentonite barrier while chemical osmosis is found to be almost irrelevant. The model has been tested with data collected after the dismantling of heater 1 of the in situ test. The model reproduces reasonably well the measured temperature, relative humidity, water content and inferred geochemical data. However, it fails to mimic the solute concentrations at the heater-bentonite and bentonite-granite interfaces because the model does not account for the volume change of bentonite, the CO 2 (g) degassing and the transport of vapor from the bentonite into the granite. The inferred HCO 3 - and pH data cannot be explained solely by solute transport, calcite dissolution and protonation/deprotonation by surface complexation, suggesting that such data may be affected also by other reactions.

  6. A coupled THC model of the FEBEX in situ test with bentonite swelling and chemical and thermal osmosis

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, L.; Samper, J.; Montenegro, L.

    2011-04-01

    The performance assessment of a geological repository for radioactive waste requires quantifying the geochemical evolution of the bentonite engineered barrier. This barrier will be exposed to coupled thermal (T), hydrodynamic (H), mechanical (M) and chemical (C) processes. This paper presents a coupled THC model of the FEBEX (Full-scale Engineered Barrier EXperiment) in situ test which accounts for bentonite swelling and chemical and thermal osmosis. Model results attest the relevance of thermal osmosis and bentonite swelling for the geochemical evolution of the bentonite barrier while chemical osmosis is found to be almost irrelevant. The model has been tested with data collected after the dismantling of heater 1 of the in situ test. The model reproduces reasonably well the measured temperature, relative humidity, water content and inferred geochemical data. However, it fails to mimic the solute concentrations at the heater-bentonite and bentonite-granite interfaces because the model does not account for the volume change of bentonite, the CO{sub 2}(g) degassing and the transport of vapor from the bentonite into the granite. The inferred HCO{sub 3}{sup -} and pH data cannot be explained solely by solute transport, calcite dissolution and protonation/deprotonation by surface complexation, suggesting that such data may be affected also by other reactions.

  7. Neutronics Experiment on A HCPB Breeder Blanket Mock-Up

    International Nuclear Information System (INIS)

    Paola Batistoni, P.; Angelone, M.; Bettinali, L.

    2006-01-01

    A neutronics experiment has been performed in the frame of European Fusion Technology Program on a mock-up of the EU Test Blanket Module (TBM), Helium Cooled Pebble Bed (HCPB) concept, with the objective to validate the capability of nuclear data to predict nuclear responses, such as the tritium production rate (TPR), with qualified uncertainties. The experiment has been carried out at the FNG 14-MeV neutron source in collaboration between ENEA, Technische Universitaet Dresden, Forschungszentrum Karlsruhe, J. Stefan Institute Ljubljana and with the participation of JAEA. The mock-up, designed in such a way to replicate all relevant nuclear features of the TBM-HCPB, consisted of a steel box containing beryllium block and two intermediate steel cassettes, filled with of Li 2 CO 3 powder, replicating the breeder insert main characteristics: radial thickness, distance between ceramic layers, thickness of ceramic layers and of steel walls. In the experiment, the TPR has been measured using Li 2 CO 3 pellets at various depths at two symmetrical positions at each depth, one in the upper and one in the lower cassette. Twelve pellets were used at each position to determine the TPR profile through the cassette. Three independent measurements were performed by ENEA, TUD/VKTA and JAEA. The neutron flux in the beryllium layer was measured as well using activation foils. The measured tritium production in the TBM (E) was compared with the same quantity (C) calculated by the MCNP.4c using a very detailed model of the experimental set up, and using neutron cross sections from the European Fusion File (EFF ver.3.1) and from the Fusion Evaluated Nuclear Data Library (FENDL ver. 2.1, ITER reference neutron library). C/E ratios were obtained with a total uncertainty on the C/E comparison less than 9% (2 s). A sensitivity and uncertainty analysis has also been performed to evaluate the calculation uncertainty due to the uncertainty on neutron cross sections. The results of such

  8. FEBEX-DP. Dismantling the ''full-scale engineered barrier experiment'' after 18 years of operation at the Grimsel Test Site, Switzerland

    Energy Technology Data Exchange (ETDEWEB)

    Kober, Florian; Gaus, Irina [Nagra, Wettingen (Switzerland)

    2015-07-01

    The FEBEX experiment at the Grimsel Test Site (GTS) consists of an in-situ full-scale engineered barrier system (EBS) test for the disposal of high-level waste (HLW). It is performed under natural conditions in crystalline rock in which the canisters are placed horizontally in drifts and are surrounded by a clay barrier constructed of highly compacted bentonite blocks. A partial dismantling and sampling of the EBS was carried out during 2002. Heating of the FEBEX started in 1997 and since then a constant temperature of 100 deg C has been maintained, while the bentonite buffer has been slowly hydrating in a natural way. A total of 632 sensors in the bentonite barrier, the rock mass, the heaters and the service zone record temperature, water saturation, humidity, total pressure, displacement, and water pressure. The hydration pattern is relatively symmetric, with no major differences along the axis. Although the host rock is characterized by heterogeneities with zones of higher permeability, the resaturation process is driven by the suction of the bentonite rather than by the availability of water in the rock, especially in the early phase. After 17 years, the water content in the buffer close to the heater still continues to increase slowly. The hydraulic pore pressures in the buffer and the geosphere have practically stabilized. The total pressure in general continues to increase in most points into the buffer, where in some parts pressures of over 6 MPa are registered. The long monitoring phase and the partial dismantling in 2002 indicate that the EBS has largely performed as expected and the major processes and couplings affecting the buffer saturation during the initial thermal period identified prior to the start of the experiment have been confirmed. A comprehensive report documents and reviews the state of the FEBEX (Lanyon and Gaus, 2013). After 18 years of operation the experiment will be excavated and dismantled in 2015. The main objectives of the FEBEX

  9. 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)

  10. ABOUT DIGITAL MOCK-UP FOR MECHANICAL PRODUCTS

    Directory of Open Access Journals (Sweden)

    GHERGHINA George

    2015-06-01

    The digital mock-up of the product is built at a design stage, and is applicable to the whole life-cycle of the product, including design, manufacture, marketing and aftermarket. The digital mock-up could achieve interference check, motion analysis, simulation of performance and manufacturing, technical training, advertising and maintenance, planning etc. The DMU of mechanical products, as important engineering data in a company, is supposed to be able to support all the activities in the whole life-cycle of the product including design, manufacture, marketing and aftermarket

  11. Siloette, Siloe mock-up; Siloette, modele nucleaire de siloe

    Energy Technology Data Exchange (ETDEWEB)

    Delcroix, V; Jeanne, G; Mitault, G; Schulhof, P [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1964-07-01

    Siloette is the Siloe mock-up. The main installations are described: various tanks, building, auxiliaries, control systems... Precis ions are given about precautions taken for using spent fuel elements. (authors) [French] Siloette est le modele nucleaire de SILOE. On decrit ses diverses installations: bassins, batiments, auxiliaires, controle... Des precisions sont donnees sur les precautions prises pour y utiliser des elements uses. (auteurs)

  12. Demonstration of the LHC Safety Training Tunnel Mock-Up

    CERN Multimedia

    Brice, Maximilien

    2014-01-01

    Members of CERN's management visit the LHC tunnel mock-up at the Safety Training Centre on the Prévessin site. The facility is used to train personnel in emergency responses including the use of masks and safe evacuation.

  13. Mock-up-CZ: dismantling of the experiment - Geotechnical results

    International Nuclear Information System (INIS)

    Svoboda, J.; Vasicek, R.

    2010-01-01

    Document available in extended abstract form only. The issue of the disposal of radioactive waste is one of the most pressing challenges of our age, for which, in most countries, the deep repository concept is generally considered to be the most suitable final solution. In order to make such a repository both safe and reliable, intensive research is underway worldwide. The construction of physical models is one approach to the study of the engineered barriers for deep geological repositories; one such experiment, Mock-Up-CZ, has been performed at the Centre of Experimental Geotechnics, CTU in Prague. The Mock-Up-CZ experiment simulated the vertical placement of a container with radioactive waste, an approach that is in line with the Swedish KBS-3 system. The physical model consisted of a barrier made up of bentonite blocks, powdered bentonite backfill, a heater and hydration and monitoring systems. The whole experiment was enclosed in a cylindrical box, whose construction was able to withstand high pressure due to bentonite swelling. A number of sensors (monitoring changes in temperature, pressure and moisture) were placed inside the bentonite barrier. The basic material used in the experiment consisted of a mixture of Czech bentonite from the Rokle deposit (85%), quartz sand (10%) and graphite (5%). The first phase of the experiment commenced on 7 May 2002, during which the heater was switched on, with no water input. After 6 months the second phase commenced in which water was introduced through the hydration system. This phase ended on 2nd January 2006 when the heater was switched off. After allowing time for cooling, the dismantling phase commenced (30 January 2006). After a further one and a half months (17 March 2006) the dismantling of the experimental vessel was completed. Post-decommissioning analysis continued until the end of 2007. Dismantling and post-decommissioning analysis were carried out according to a very detailed plan which included not only

  14. Fabrication of small mock-ups for the KO HCCR TBM

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Jae Sung; Kim, Suk Kwon; Lee, Eo Hwak; Jin, Hyung Gon; Lee, Dong Won [KAERI, Daejeon (Korea, Republic of); Cho, Seung Yon [NFRI, Daejeon (Korea, Republic of)

    2016-05-15

    A fabrication procedure for the manufacturing of the HCCR TBM sub-module was performed and small mock-ups were fabricated using an E-beam and laser beam weld to verify the manufacturing procedure and method of the HCCR TBM sub-module. To establish and optimize the welding procedure in an E-beam weld from ARAA material, the distortion and radiographic tests were carried out from the E-beam weld results. It could be noted that a small amount of distortion occurred, but the values are small enough to neglect for the fabrication. In addition, a helium leak test and water pressure test will be performed for verification of the fabricated small mock-ups.

  15. A new magnet for the LHC mock-up

    CERN Multimedia

    HSE Unit

    2013-01-01

    This year, the safety training centre on the Prévessin site acquired a mock-up of the LHC, which simulates the work and safety conditions in the tunnel.   Photo: Christoph Balle. A new dummy quadrupole has just been added to the magnet chain, making the mock-up even more realistic. The new facility, which was a joint endeavour by the TE, GS, BE and EN Departments, will significantly improve the quality of the various training courses held at the centre, particularly the course on the use of self-rescue masks. To consult the safety training catalogue and/or sign up for radiation protection training, please go to: https://cta.cern.ch. For further information, please contact the Safety Training and Awareness service by telephone on 73811 or 79935 or by e-mail to safety-training@cern.ch.  

  16. ASTP crewmen in Soyuz orbital module mock-up during training session at JSC

    Science.gov (United States)

    1975-01-01

    An interior view of the Soyuz orbital module mock-up in bldg 35 during Apollo Soyuz Test Project (ASTP) joint crew training at JSC. The ASTP crewmen are Astronaut Vance D. Brand (on left), command module pilot of the American ASTP prime crew; and Cosmonaut Valeriy N. Kubasov, engineer on the Soviet ASTP first (prime) crew. The training session simulated activities on the second day in Earth orbit.

  17. FEBEX project: full-scale engineered barriers experiment for a deep geological repository for high level radioactive waste in crystalline host rock. Final report

    International Nuclear Information System (INIS)

    Alberdi, J.; Barcala, J. M.; Campos, R.; Cuevas, A. M.; Fernandez, E.

    2000-01-01

    FEBEX has the multiple objective of demonstrating the feasibility of manufacturing, handling and constructing the engineered barriers and of developing codes for the thermo-hydro-mechanical and thermo-hydro-geochemical performance assessment of a deep geological repository for high level radioactive wastes. These objectives require integrated theoretical and experimental development work. The experimental work consists of three parts: an in situ test, a mock-up test and a series of laboratory tests. The experiments is based on the Spanish reference concept for crystalline rock, in which the waste capsules are placed horizontally in drifts surround by high density compacted bentonite blocks. In the two large-scale tests, the thermal effects of the wastes were simulated by means of heaters; hydration was natural in the in situ test and controlled in the mock-up test. The large-scale tests, with their monitoring systems, have been in operation for more than two years. the demonstration has been achieved in the in situ test and there are great expectation that numerical models sufficiently validated for the near-field performance assessment will be achieved. (Author)

  18. FEBEX project: full-scale engineered barriers experiment for a deep geological repository for high level radioactive waste in crystalline host rock

    Energy Technology Data Exchange (ETDEWEB)

    Alberid, J; Barcala, J M; Campos, R; Cuevas, A M; Fernandez, E [Ciemat. Madrid (Spain)

    2000-07-01

    FEBEX has the multiple objective of demonstrating the feasibility of manufacturing, handling and constructing the engineered barriers and of developing codes for the thermo-hydro-mechanical and thermo-hydro-geochemical performance assessment of a deep geological repository for high level radioactive wastes. These objectives require integrated theoretical and experimental development work. The experimental work consists of three parts: an in situ test, a mock-up test and a series of laboratory tests. The experiments is based on the Spanish reference concept for crystalline rock, in which the waste capsules are placed horizontally in drifts surround by high density compacted bentonite blocks. In the two large-scale tests, the thermal effects of the wastes were simulated by means of heaters; hydration was natural in the in situ test and controlled in the mock-up test. The large-scale tests, with their monitoring systems, have been in operation for more than two years. the demonstration has been achieved in the in situ test and there are great expectation that numerical models sufficiently validated for the near-field performance assessment will be achieved. (Author)

  19. The supply of small scale mock-ups of the primary wall module concepts for ITER

    International Nuclear Information System (INIS)

    Walsh, G.; Cheyne, K.; Lorenzetto, P.

    1998-01-01

    The present design of Blanket Shield and Primary Wall for ITER envisages construction of the wall with a water cooled, stainless steel outer layer and a water cooled, copper liner on the inside plasma facing surface. Protection of the inner copper surface with an armour layer is necessary to cope with plasma to wall interaction. There are a number of armour materials under consideration, for this project beryllium was used. The scope of work was to produce a series of mock-ups, each consisting of a different combination of materials, which included Dispersion Strengthened Copper, Copper-Chrome-Zirconium alloy, Beryllium and Stainless Steel. Hot Isostatic Pressing (HIP) was the method used to ensure that a fully diffused bonded joint was achieved giving the necessary strength and thermal conductivity. The first five of the mock ups have been successfully completed and are being tested at the various laboratories in Europe. The remaining mock ups are awaiting the results of this test work prior to being completed. (authors)

  20. Mock-up critical experiments for prototype fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    Zukeran, Atsushi; Inoue, Teruji; Suzuki, Takeo; Kawashima, Kanau

    1976-01-01

    The mock-up criticality experiments for Monju are roughly divided into the full mock-up test using the ZEBRA of Winfrith Institute, UK AEA, and the partial mock-up experiment with FCA of JAERI. The former test has been carried out over 18 months from September 1971 as the Japan-UK cooperative research project MOZART. With the FCA, the experiment complementing the MOZART has been carried out, focusing on the nuclear characteristics of Monju which can be simulated with a relatively small core, and the experiment on highly enriched control rods and shielding is being continued now with the FCA 7 core. The experimental data of the MOZART and the ZPPR series in USA were exchanged at the international symposium in Tokyo, thus the prediction and the accuracy evaluation of the nuclear characteristics of Monju became possible, and the highly reliable core design was able to be accomplished. The simulated criticality experiment is necessary for directly grasping the reliability of calculated values in comparison with the experimental values, and also for the experimental prediction of the nuclear characteristics. The outline and the analysis of the simulated criticality experiment such as reactivity factor, control rod value, reaction rate distribution and sodium void reactivity are described, and the reflection of the results to the design of the core of Monju is explained. (Kako, I.)

  1. Destructive analysis on the ITER FW small scale mock-ups

    International Nuclear Information System (INIS)

    Wang, Pinghuai; Chen, Jiming; Liu, Danhua; Jin, Fanya; Yang, Bo

    2015-01-01

    As one of the core components of ITER, the first wall (FW) panel of shield blanket defines a physical boundary for the plasma transients and exhausts the majority of the plasma heat flux. China will undertake 12.64% of FW manufacturing tasks, and all of them are enhanced heat flux (EHF) components which will suffer surface heat flux of 4 - 5 MW/m 2 . The FW will be manufactured by a combination technology of explosion bonding CuCrZr alloy/316L (N) stainless steel plate and hot iso-static pressing (HIP) joining of beryllium tiles/CuCrZr alloy. The Be/Cu joint qualities is the key issue for the manufacturing of the FW panels. Several small scale mock-ups were manufactured for the qualification of the HIP technology for the FW. To avoid the brittle Be-Cu phase formed during the HIPing process, different thick Ti and pure Cu were coated on the beryllium tiles before HIPing to CuCrZr alloy. Ultrasonic testing was conducted on the mock-ups and destructive analysis was carried out on the mock-ups. For the failed ones, the results show that in the UT indication area brittle fracture occurs at the Be/Ti interface and then Ti/Cu interface in other areas. Based on these results, the manufacturing technology was improved mainly on the beryllium tiles quality, coating process and canister design. (author)

  2. Blanket Cooling Plates Mock-ups Manufactured in different Diffusion Weld Setup

    International Nuclear Information System (INIS)

    Von Der Weth, A.; Aktaa, J.

    2007-01-01

    Full text of publication follows: The breeding blanket box is considered as one of the most important components of a future fusion power plant. It will be assembled by so called cooling plates (CP) with a system of internal cooling channels. Such a CP is produced by two symmetric half pieces with half milled-in channels. Both pieces will be joined by a diffusion weld (DW) process. Within recent years a two step DW process for different EUROFER batches has been developed. It has been first applied to small laboratory scaled samples with dimensions of 25 mm x 30 mm x 40 mm. Then the DW process had then been successfully transferred to so called compact mock ups which are small CPs with dimensions of 67 mm x 70 mm x 50 mm. As third step this process has been used to manufacture a CP (465 mm x 205 mm x 50 mm) of a breeder unit in an industrial uniaxial diffusion weld setup. This paper treats the manufacturing sequence of a cooling plate and a first wall mock up in an industrial hot isostatic pressing (HIP) setup. The firstly laboratory specimens scaled diffusion weld process has been adjusted to different cooling channel dimensions and a different DW setup. The weld quality is investigated by tensile and Charpy impact testing. This allows comparison of the weld quality of mock ups welded in different DW setups. (authors)

  3. Analysis of the impacts of the J-TEXT TBM mock-up on the equilibrium magnetic field

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Zhengqing [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, and College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Rao, Bo, E-mail: borao@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, and College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Zhang, Ming; Zhang, Jun [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, and College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Wang, Weihua [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); New Star Institute of Applied Technology, Hefei 230031 (China); Liu, Sumei [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); School of Engineering,Anhui Agricultural University, Hefei 230036 (China)

    2016-11-01

    Highlights: • J-TEXT TBM mock-up was designed and fabricated to test and study the distribution of eddy current, electromagnetic and thermal load on the TBM during plasma disruption. • This paper focuses on evaluating the influence of the TBM structural material (RAMF steel) to tokamak discharge and security. The simulation data presents a relatively complete assessment of impacts of the J-TEXT TBM mock-up on the equilibrium magnetic field. • The conclusion of the simulation will offer the guidance for installation interface design of the TBM mock-up. - Abstract: The Test Blanket Module (TBM) will be used in the test port of ITER to demonstrate tritium self-sufficiency and the extraction of high grade heat for electricity production. J-TEXT TBM mock-up using reduced activation ferritic/martensitic (RAFM) steel as structural material was designed and fabricated to perform and validate relevant electromagnetic and thermal technologies of the China Helium-Cooled Ceramic Breeder Test Blanket Module (CN HCCB-TBM) on the J-TEXT. Its size is one third of the CN HCCB-TBM. By using the finite element analysis technology, this paper analyzed the impacts on the equilibrium magnetic field over the plasma region after introducing the structure material RAFM steel. The distribution of toroidal field (TF) ripple and the magnitude of the error field with the mock-up at different positions were given. Simulation shows the distribution of the null field region formed by poloidal field (PF). The influence to tokamak discharge has been evaluated by drawing the magnetic field lines. Based on the results above, we have optimized and finished the installation of the mock-up to J-TEXT which meets the needs of the experiments and to ensure the normal discharge.

  4. Digital mock-up for the spent fuel disassembly processes

    International Nuclear Information System (INIS)

    Lee, J. Y.; Kim, S. H.; Song, T. G.; Kim, Y. H.; Hong, D. H.; Yoon, J. S.

    2000-12-01

    In this study, the graphical design system is developed and the digital mock-up is implemented for designing the spent fuel handling and disassembly processes. The system consists of a 3D graphical modeling system, a devices assembling system, and a motion simulation system. This system is used throughout the design stages from the conceptual design to the motion analysis. By using this system, all the process involved in the spent fuel handling and disassembly processes are analyzed and optimized. Also, this system is used in developing the on-line graphic simulator which synchronously simulates the motion of the equipment in a real time basis by connecting the device controllers with the graphic server through the TCP/IP network. This simulator can be effectively used for detecting the malfunctions of the process equipment which is remotely operated. Thus, the simulator enhances the reliability and safety of the spent fuel handling process by providing the remote monitoring function of the process. The graphical design system and the digital mock-up system can be effectively used for designing the process equipment, as well as the optimized process and maintenance process. And the on-line graphic simulator can be an alternative of the conventional process monitoring system which is a hardware based system

  5. Conceptual design of a First Wall mock-up experiment in preparation for the qualification of breeding blanket technologies in the Helium Loop Karlsruhe (HELOKA) facility

    Energy Technology Data Exchange (ETDEWEB)

    Zeile, C., E-mail: christian.zeile@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Abou-Sena, A.; Boccaccini, L.V.; Ghidersa, B.E.; Kang, Q.; Kunze, A. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Lamberti, L. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Dipartimento Energia, Politecnico di Torino (Italy); Maione, I.A.; Rey, J.; Weth, A. von der [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2016-11-01

    Highlights: • Experiment in preparation for the qualification of Breeding Blanket technologies in HELOKA facility is proposed. • Experimental capabilities, instrumentation of the mock-up and experimental program are presented. • Design and manufacturing of the mock-up is described. • Design of modular attachment system to obtain different stress levels and distributions on the mock-up is discussed. - Abstract: An experimental program based on a First Wall mock-up is presented as preparation for the qualification of breeding blanket mock-ups at high heat flux in the Helium Loop Karlsruhe (HELOKA) facility. Two objectives of the experimental program have been defined: testing of the experimental setup and a first validation of FE models. The design and manufacturing of mock-up representing about 1/3 of the heated zone of an ITER Test Blanket Module (TBM) First Wall is discussed. A modular attachment system concept has been developed for the fixation of the mock-up in order to be able to generate different stress distributions and levels on the plate, which is confirmed by thermo-mechanical analyses. The HELOKA facility is able to provide a TBM relevant helium cooling system and to generate the required surface heat flux by an electron beam gun. An installed IR camera can be used to measure the temperature distribution on the surface.

  6. Lessons learned from the Febex in situ test: geochemical processes associated to the microbial degradation and gas generation

    International Nuclear Information System (INIS)

    Fernandez, A. M.; Sanchez, D.M.; Melon, A.; Mingarro, M.; Wieczorek, K.

    2012-01-01

    existence of gaps between the bentonite blocks, which favour the development and growth of inactive and dormant cells or spores belonging to the original bentonite. In this work, the observed geochemical and corrosion processes influenced both by organic matter degradation and micro-organisms in the 1:1 scale FEBEX in situ test (Grimsel, Switzerland) are described. This test consists of two heaters, simulating radioactive waste containers, emplaced in a horizontal gallery and surrounded by a highly compacted bentonite barrier. Samples from pore water, gases and bentonite (SHSDI-01: clay in contact with AISI 316L metal; S29 and BSBI-26: clay in contact with carbon steel) have been analysed. The samples were obtained during the test and the dismantling of the heater 1 after six years of experiment. The solid samples were analysed by XRD, SEM, XPS, FTIR, ATD-TG and chemical analysis; the water samples by IC and ICP-OES, and the gases by gas chromatography. Different geochemical processes have been detected as a function of the temperature and water content of the samples. When the water content is high, there are aerobic respiration and fermentation processes, anaerobic respiration with SO 4 2- as electron acceptor, and anaerobic production of methane with CO 2 as electron acceptor. In a first phase, both oxygen consumption and an increase of CH 4 and CO 2 is observed. Afterwards, there is a reduction of sulfates by SRB bacteria, which provokes corrosion processes. As a consequence, a precipitation of sulphurs, iron oxy-hydroxides and carbonates occurs, as well as H 2 generation. There is an increase of the iron content in the smectite and the neo-formation of zeolites. However this alteration is punctual and localized. The redox potential of the bentonite pore water was of -284 mV. When the temperature is high and water content is low, other processes take place

  7. Non destructive examination of primary wall small scale mock-up DS-1F

    International Nuclear Information System (INIS)

    Jeskanen, H.; Lahdenperae, K.; Kauppinen, P.; Taehtinen, S.

    1998-06-01

    Ultrasonic examination of primary wall small scale mock up DS-1F before thermal testing showed no major defects on studied interfaces. However, some small indications were found on copper to copper and copper to steel interfaces and surface roughness of the outer surface of copper layer gave clear indications on ultrasonic images. After thermal test a curved 50 mm long crack along the Y- direction in the middle of the heated surface of the mock up and a 220 mm long crack along the copper to copper interface on the side surface of the mock up were detected. Small cracks, less than 60-80 μm in depth, were observed on copper surface. After thermal test the corresponding ultrasonic examination showed a strong effect on ultrasonic attenuation properties and on leaky Rayleigh waves on outer surface of copper layer. A major indication was found on copper to copper interface. About 50% of the copper to copper interface was delaminated. However, some small indications found already before thermal test were also found after thermal test and they were not grown in size. No indications were observed on copper to stainless steel interfaces. Additionally, major indications were found on stainless steel tube to copper interfaces. Tubes No. 1 and 2 were almost completely whereas tube No. 3 only partly separated from copper. No indications were found on stainless steel tube to copper interface on tube No. 4. Eddy current measurements showed no volumetric or crack like flaws in the stainless steel tubes, however, delamination of the copper to copper interface along the tubes No. 1, 2 and 3 was observed. (orig.)

  8. Fabrication of a full-size mock-up for inboard 10o section of ITER vacuum vessel thermal shield

    International Nuclear Information System (INIS)

    Chung, W.; Nam, K.; Noh, C.H.; Kang, D.K.; Kang, S.M.; Oh, Y.G.; Choi, S.W.; Kang, S.H.; Utin, Y.; Ioki, K.; Her, N.; Yu, J.

    2011-01-01

    A full-scale mock-up of VVTS inboard section was made in order to validate its manufacturing processes before manufacturing the vacuum vessel thermal shield (VVTS) for ITER tokamak. VVTS inboard 10 o section consists of 20 mm shells on which cooling tubes are welded and flange joints that connect adjacent thermal shield sectors. The whole VVTS inboard is divided into two by bisectional flange joint located at the center. All the manufacturing processes except silver coating were tested and verified in the fabrication of mock-up. For the forming and the welding, pre-qualification tests were conducted to find proper process conditions. Shell thickness change was measured after bending, forming and buffing processes. Shell distortion was adjusted after the welding. Welding was validated by non-destructive examination. Bisectional flange joint was successfully assembled by inserting pins and tightening with bolt/nut. Bolt hole margin of 2 mm for sector flange was revealed to be sufficient by successful sector assembly of upper and lower parts of mock-up. Handling jig was found to be essential because the inboard section was flexible. Dimensional inspection of the fabricated mock-up was performed with a 3D laser scanner.

  9. The natural and artificial hydration of a bentonite engineered barrier system in a full-scale KBS-3V mock-up; results from the first 7 years of the large scale gas injection test (LASGIT)

    International Nuclear Information System (INIS)

    Cuss, R.J.; Harrington, J.F.; Noy, D.J.; Bennett, D.P.; Sellin, P.

    2012-01-01

    Document available in extended abstract form only. The Large scale gas injection test is a full-scale in situ canister test designed to answer specific questions regarding the movement of gas through bentonite in a mock KBS-3v deposition hole. The test is located at 420 m depth within SKB's Aespoe Hard Rock Laboratory (HRL) in Sweden. The objective of Lasgit is to provide quantitative data to improve process understanding and test/validate modelling approaches which might be used in performance assessment. The deposition hole has a depth of 8.5 m and a diameter of around 1.75 m. A full scale KBS-3 canister has been modified for the Lasgit experiment with thirteen circular filters of varying dimensions located on its surface to provide point sources for gas injection, mimicking potential canister defects. These filters can also be used to inject water during the hydration stage, with hydration also conducted through 4 filter mats within the buffer. The deposition hole, buffer and canister are equipped with instrumentation to measure the total stress, pore water pressure and relative humidity in 32, 26 and 7 positions respectively. Additional instrumentation continually monitors variations in temperature, relative displacement of the lid and the restraining forces on the rock anchors. Groundwater inflow through a number of highly-conductive discrete fractures quickly resulted in elevated pore water pressures in sections of the borehole. This lead to the formation of conductive channels, the extrusion of bentonite from the deposition hole, and the discharge of groundwater to the gallery floor. Artificial hydration began after 106 days of testing. Up until the first gas injection test (day 843), the pressures in all of the canister filters and hydration mats were used to hydrate the clay. Initial attempts to raise pore water pressure in the artificial hydration arrays occasionally resulted in the formation of preferential pathways resulting in localized increases in

  10. Interactive virtual mock-ups for Remote Handling compatibility assessment of heavy components

    Energy Technology Data Exchange (ETDEWEB)

    Oosterhout, J. van, E-mail: j.vanoosterhout@differ.nl [FOM Institute DIFFER (Dutch Institute for Fundamental Energy Research), Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Heemskerk, C.J.M.; Koning, J.F. [Heemskerk Innovative Technology B.V., Jonckerweg 12, 2201 DZ Noordwijk 6 (Netherlands); Ronden, D.M.S.; Baar, M. de [FOM Institute DIFFER (Dutch Institute for Fundamental Energy Research), Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster, PO Box 1207, 3430 BE Nieuwegein (Netherlands)

    2014-10-15

    Highlights: •Specific ITER components require RHCA on hardware mock-ups. •Hardware mock-ups are expensive and have a long lead time. •Interactive Virtual Reality mock-ups are readily available and easily adapted. •This paper analysis and proposes improvements to simulator capabilities. -- Abstract: ITER standards Tesini (2009) require hardware mock-ups to validate the Remote Handling (RH) compatibility of RH class 1- and critical class 2-components. Full-scale mock-ups of large ITER components are expensive, have a long lead time and lose their relevance in case of design changes. Interactive Virtual Reality simulations with real time rigid body dynamics and contact interaction allow for RH Compatibility Assessment during the design iterations. This paper explores the use of interactive virtual mock-ups to analyze the RH compatibility of heavy component handling and maintenance. It infers generic maintenance operations from the analysis and proposes improvements to the simulator capabilities.

  11. Interactive virtual mock-ups for Remote Handling compatibility assessment of heavy components

    International Nuclear Information System (INIS)

    Oosterhout, J. van; Heemskerk, C.J.M.; Koning, J.F.; Ronden, D.M.S.; Baar, M. de

    2014-01-01

    Highlights: •Specific ITER components require RHCA on hardware mock-ups. •Hardware mock-ups are expensive and have a long lead time. •Interactive Virtual Reality mock-ups are readily available and easily adapted. •This paper analysis and proposes improvements to simulator capabilities. -- Abstract: ITER standards Tesini (2009) require hardware mock-ups to validate the Remote Handling (RH) compatibility of RH class 1- and critical class 2-components. Full-scale mock-ups of large ITER components are expensive, have a long lead time and lose their relevance in case of design changes. Interactive Virtual Reality simulations with real time rigid body dynamics and contact interaction allow for RH Compatibility Assessment during the design iterations. This paper explores the use of interactive virtual mock-ups to analyze the RH compatibility of heavy component handling and maintenance. It infers generic maintenance operations from the analysis and proposes improvements to the simulator capabilities

  12. Experimental investigation of MHD pressure losses in a mock-up of a liquid metal blanket

    Science.gov (United States)

    Mistrangelo, C.; Bühler, L.; Brinkmann, H.-J.

    2018-03-01

    Experiments have been performed to investigate the influence of a magnetic field on liquid metal flows in a scaled mock-up of a helium cooled lead lithium (HCLL) blanket. During the experiments pressure differences between points on the mock-up have been recorded for various values of flow rate and magnitude of the imposed magnetic field. The main contributions to the total pressure drop in the test-section have been identified as a function of characteristic flow parameters. For sufficiently strong magnetic fields the non-dimensional pressure losses are practically independent on the flow rate, namely inertia forces become negligible. Previous experiments on MHD flows in a simplified test-section for a HCLL blanket showed that the main contributions to the total pressure drop in a blanket module originate from the flow in the distributing and collecting manifolds. The new experiments confirm that the largest pressure drops occur along manifolds and near the first wall of the blanket module, where the liquid metal passes through small openings in the stiffening plates separating two breeder units. Moreover, the experimental data shows that with the present manifold design the flow does not distribute homogeneously among the 8 stacked boxes that form the breeding zone.

  13. Parallel Execution of Functional Mock-up Units in Buildings Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Ozmen, Ozgur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Nutaro, James J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); New, Joshua Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-06-30

    A Functional Mock-up Interface (FMI) defines a standardized interface to be used in computer simulations to develop complex cyber-physical systems. FMI implementation by a software modeling tool enables the creation of a simulation model that can be interconnected, or the creation of a software library called a Functional Mock-up Unit (FMU). This report describes an FMU wrapper implementation that imports FMUs into a C++ environment and uses an Euler solver that executes FMUs in parallel using Open Multi-Processing (OpenMP). The purpose of this report is to elucidate the runtime performance of the solver when a multi-component system is imported as a single FMU (for the whole system) or as multiple FMUs (for different groups of components as sub-systems). This performance comparison is conducted using two test cases: (1) a simple, multi-tank problem; and (2) a more realistic use case based on the Modelica Buildings Library. In both test cases, the performance gains are promising when each FMU consists of a large number of states and state events that are wrapped in a single FMU. Load balancing is demonstrated to be a critical factor in speeding up parallel execution of multiple FMUs.

  14. Sensitivity and uncertainty analyses of the HCLL mock-up experiment

    International Nuclear Information System (INIS)

    Leichtle, D.; Fischer, U.; Kodeli, I.; Perel, R.L.; Klix, A.; Batistoni, P.; Villari, R.

    2010-01-01

    Within the European Fusion Technology Programme dedicated computational methods, tools and data have been developed and validated for sensitivity and uncertainty analyses of fusion neutronics experiments. The present paper is devoted to this kind of analyses on the recent neutronics experiment on a mock-up of the Helium-Cooled Lithium Lead Test Blanket Module for ITER at the Frascati neutron generator. They comprise both probabilistic and deterministic methodologies for the assessment of uncertainties of nuclear responses due to nuclear data uncertainties and their sensitivities to the involved reaction cross-section data. We have used MCNP and MCSEN codes in the Monte Carlo approach and DORT and SUSD3D in the deterministic approach for transport and sensitivity calculations, respectively. In both cases JEFF-3.1 and FENDL-2.1 libraries for the transport data and mainly ENDF/B-VI.8 and SCALE6.0 libraries for the relevant covariance data have been used. With a few exceptions, the two different methodological approaches were shown to provide consistent results. A total nuclear data related uncertainty in the range of 1-2% (1σ confidence level) was assessed for the tritium production in the HCLL mock-up experiment.

  15. Fabrication of a 1/6-scale mock-up and manifolds for the Korea first wall in the ITER

    International Nuclear Information System (INIS)

    Yoon, Jae Sung; Kim, Suk Kwon; Lee, Eo Hwak; Lee, Dong Won

    2012-01-01

    Korea has developed and participated in the Test Blanket Module (TBM) program of the International Thermo-nuclear Experimental Reactor (ITER). The first wall (FW) of the TBM is an important component that faces the plasma directly and therefore it is subjected to high heat and neutron loads. To fabricate the TBM FW, the Hot Isostatic Pressing (HIP) bonding method has been investigated. In the present study, the manufacturing method of the TBM FW is introduced through the fabrication and testing of a 1/6-scale mockup. To distribute fluid uniformly in the mock-up, a manifold was designed and fabricated using the ANSYS-CFX analysis. After the mock-up was fabricated and its fluid distribution tests performed, we compared the results of tests with the simulated results

  16. Progress on pebble bed experimental activity for the HE-FUS3 mock-ups

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Sansone, L.; Simoncini, M.; Zito, D.

    2002-01-01

    The EU Long Term for DEMO Programme foresees the qualification of the reference design of the helium cooled pebble bed (HCPB) - test blanket module (TBM) to be tested in ITER Reactor. In this frame, FZK and ENEA have launched many experimental activities for the evaluation of the interactions between the Tritium breeder and neutron multiplier pebble beds and the steel containment walls. Main aim of these activities is the measuring the pebble bed effective thermal conductivity, the wall heat transfer coefficient as well as their dependency from the mechanical constraints. The paper presents the progress of the testing activity and results of the tests on two mock-up, called Tazza and Helichetta, carried out on the HE-FUS3 facility at ENEA Brasimone. (orig.)

  17. Simulation in full-scale mock-ups: an ergonomics evaluation method?

    DEFF Research Database (Denmark)

    Andersen, Simone Nyholm; Broberg, Ole

    2014-01-01

    This paper presents and exploratory study of four simulation sessions in full-scale mock-ups of future hospital facilities.......This paper presents and exploratory study of four simulation sessions in full-scale mock-ups of future hospital facilities....

  18. Thermo-Hydro-Mechanical Characterisation of the Bentonite of a Simulated HLW Repository after Five Years Operation ( In Situ Test of the FEBEX Project)

    Energy Technology Data Exchange (ETDEWEB)

    Villar, M. V.

    2004-07-01

    After five years of operation, heater 1 of the FEBEX experimental at the Grimsel Test Sites was switched off in February 2002. Following cooling of the system, the bentonite barrier was dismantled and the heater extracted. During dismantling many bentonite samples were taken. Several determinations were carried out in these samples with the aim of: (1) characterise the actual state of the bentonite and (2) determine the possible changes in its properties occurred during the experiment. The results of the thermo-hydro-mechanical characterisation performed at CIEMAT are reported and analysed. The distribution of water content and dry density of the bentonite in vertical sections presents axial symmetry. The construction gaps of the barrier have been filled by the expansion of the bentonite. The water retention capacity, the hydraulic conductivity and the swelling capacity of the samples from Grimsel have not irreversible changed. The pre consolidation pressure of the Grimsel samples has decreased due to the microstructural changes associated to the volume increase experienced during hydration. The thermal conductivity is higher for the bentonite blocks of the external ring of the barrier. (Author)

  19. Thermo-Hydro-Mechanical Characterisation of the Bentonite of a Simulated HLW Repository after Five Years Operation ( In Situ Test of the FEBEX Project)

    International Nuclear Information System (INIS)

    Villar, M. V.

    2004-01-01

    After five years of operation, heater 1 of the FEBEX experimental at the Grimsel Test Sites was switched off in February 2002. Following cooling of the system, the bentonite barrier was dismantled and the heater extracted. During dismantling many bentonite samples were taken. Several determinations were carried out in these samples with the aim of: (1) characterise the actual state of the bentonite and (2) determine the possible changes in its properties occurred during the experiment. The results of the thermo-hydro-mechanical characterisation performed at CIEMAT are reported and analysed. The distribution of water content and dry density of the bentonite in vertical sections presents axial symmetry. The construction gaps of the barrier have been filled by the expansion of the bentonite. The water retention capaciaty, the hydraulic conductivity and the swelling capacity of the samples from Grimsel have not irreversible changed. The preconsolidation pressure of the Grimsel samples has decreased due to the microstructural changes asswociated to the volume increase experienced during hydration. The thermal conductivity is higher for the bentonite blocks of the external ring of the barrier. (Author)

  20. Re-analysis of HCPB/HCLL Blanket Mock-up Experiments Using Recent Nuclear Data Libraries

    International Nuclear Information System (INIS)

    Kondo, K.; Fischer, U.; Klix, A.; Pereslavtsev, P.; Serikov, A.; Villari, R.

    2014-01-01

    We have re-analysed the two breeding blankets experiments performed previously in the frame of the European fusion program on two mock-ups of the European Helium-Cooled-Lithiium Lead (HCLL) and Helium-Cooled-Pebble-Bed (HCPB) test blanket modules for ITER. The tritium production rate and the neutron and photon spectra measured in these mock-ups were compared with calculations using FENDL-3 Starter Library, release 4 and state-of-the-art nuclear data evaluations, JEFF-3.1.2, JENDL-4.0 and ENDF/B-VII.0. The tritium production calculated for the HCPB mock-up underestimates the experimental result by about 10%. The result calculated with FENDL-3/SLIB4 gives slightly smaller tritium production by 2% than the one with FENDL-2.1. The difference attributes to the slight modification of the total and elastic scattering cross section of Be. For the HCLL experiment, all libraries reproduce the experimental results well. FENDL-3/SLIB4 gives better result both for the measured spectra and the tritium production compared to FENDL-2.1

  1. FEBEX Full-Scalle Engineered barriers experiment in crystalline host rock Preoperational thermo-hydro-mechanical (THM) modelling of the in situ test

    International Nuclear Information System (INIS)

    1998-01-01

    This report contains the results of a set of 1-D and 2-D coupled thermo-hydro-mechanical (THM) analyses carried out during the preoperational stage simulating the in situ FEBEX test. The analyses incorporate available information concerning rock and bentonite properties as well as the final test layout and conditions. The main goals are: -To provide the best estimate of test performance given current models and information - To define a basis for future model improvements. The theoretical bases of the analyses and the computer code used are reviewed. Special reference is made to the process of parameter estimation that tries to incorporate available information on material behaviour obtained in the characterisation work carried out both in the laboratory and in the field. Data obtained in the characterisation stage is also used to define initial and boundary conditions. The results of the 1-D THM Base Case analysis are used to gain a good understanding of expected test behaviour concerning thermal, hydraulic and mechanical problems. A quite extensive programme of sensitivity analyses is also reported in which the effect of a number of parameters and boundary conditions are examined. The results of the sensitivity analyses place an appropriate context the information obtained from the Base Case showing, for instance, that rock desaturation and degree of buffer hydration depend on some critical parameters in a complex way. Two-dimensional effects are discussed on the basis of the results of 2-D axisymmetric THM analysis performed using a longitudinal section that provides a better representation of real test geometry. Quantitative but not qualitative differences are found with respect to the 1-D results. Finally, a 2-D THM cross section analysis has been performed under plane strain conditions. No specific 2-D effects are observed in this case as quasi-axisymmetric conditions have been prescribed. The models employed in the analyses included in this report have not

  2. Adequacy of the analysis of mock-up control rod experiment with FCA

    International Nuclear Information System (INIS)

    Mizoo, Nobutatsu; Nakano, Masafumi

    1977-07-01

    A method of numerical analysis has been investigated for the mock-up control rod experiment of FCA VII-1 assembly constructed as the engineering mock-up of prototype fast breeder reactor MONJU. The results of criticality and B 4 C mock-up control rod worths analysis for the assembly are described in comparison with the experimental ones. The tendency of the C/E value with 10 B enrichment and the interaction effect of the multiple rods array was also examined. Reactivities and the mock-up rods worths were obtained with the X-Y geometry six groups diffusion theory. Twelve kinds of the mock-up rods with different 10 B contents and/or enrichments were used in the experiment; effective cross-sections are provided for each rod by calculation using the collision probability method. Criticality of VII-1 90Z assembly is underestimated for 3 reference critical configurations, ranging from -0.65%Δk/k to -0.77%Δk/k. The C/E values at core center for 12 kinds of B 4 C mock-up rods range from 1.03 to 1.09. The overestimate of the rod worth increases with macroscopic absorption cross-section of the rod region. The C/E values for 24 different arrays of the mock-up rods ranging from single rod to five rods lie between 1.04 and 1.08. The C/E value tends to decrease with increase in the number of rods inserted, the values for five rods arrays being about 4% lower than those for single rod arrays. The calculated interaction effects of the multiple rods arrays are slightly more negative than the experimental ones. (auth.)

  3. Mock-up experiments for the project of high dose irradiation on the RPV concrete

    International Nuclear Information System (INIS)

    Zdarek, J.; Brabec, P.; Frybort, O.; Lahodova, Z.; Vit, J.; Stemberk, P.

    2015-01-01

    Aging of NPP's concrete structures comes into growing interest in connection with solution of life extension programmes of operated units. Securing continued safe operation of NPPs calls for additional proofs of suitable long term behaviour of loaded reinforced concrete structures. An irradiation test of concrete samples was performed in the core of the LVR-15 reactor. The irradiation capsule was hung in the irradiation channel and the cooling of the capsule was ensured through direct contact of the capsule wall with the primary circuit water. Cylindrical, serpentine concrete samples (50 mm in diameter and 100 mm in length), representing composition of WWER RPV cavity, was chosen as a compromise of mechanical properties testing needs and dimension limitations of reactor irradiation channel. Heating during irradiation test was maintained under 93 Celsius degrees by cooling and was controlled by embedded thermocouple. Design of the cooling management was supported by computational analysis. The dependencies of heated concrete samples to the neutron fluence and the gamma heating were obtained by changing the thermal power of the reactor and by changing the vertical position of the sample in the irradiation channel. The irradiation capsule was filled with inert gas (helium) to allow the measurement of generated gas. The determination of concrete samples activity for long-term irradiation was performed on the principles of the Neutron Activation Analysis. Preliminary mock-up tests have proved the ability to fulfill technical needs for planned high dose irradiation experiment

  4. The state of art of the manufacturing technology of FW blanket and the development of mock-up for fusion reactor in Russia

    International Nuclear Information System (INIS)

    Baek, Jong Hyuk; Jeong, Y. H.; Park, J. Y.; Kim, J. H.; Kim, H. G.

    2004-08-01

    In early 1990s, Russia had carried out the performance tests to verify the optimization of Be tile geometry and the bonding integrity of small mock-up using a HHF (High Heat Flux) test and an in-pile test in a research reactor. They had obtained the reliability of the brazing technologies for the Be/Cu bonding. And they had manufactured the near real-size large mock-up (about 0.8 mm in length) to find the bonding integrity by a fast brazing technique. They had a satisfied results from the HHF test for the large mock-up. Additionally, an alternative FW mock-ups, which were manufactured by both casting and fast brazing techniques to reduce the joining parts, showed a good joining performance from the HHF test. Therefore, it was concluded that the fast brazing techniques could be strongly recommended as a one of the preferable joining techniques and be possible to apply to joining for the Be/Cu joining of FW blanket

  5. Development of the ITER IOIS assembly tool and mock-up

    International Nuclear Information System (INIS)

    Nam, Kyoungo; Kim, Dongjin; Park, Hyunki; Ahn, Heejae; Kim, Kyoungkyu; Yoo, Yongsoo; Watson, Emma; Shaw, Robert

    2014-01-01

    The ITER toroidal field coils (TFCs) are connected by 3 different connecting structures as follows; Outer Intercoil Structure (OIS), Inner Intercoil Structure (IIS), Intermediate Outer Intercoil Structure (IOIS). In assessing the assembly, requirements and environmental conditions of each Intercoil structure, the IOIS and IIS assembly were thought to be the most challenging compared to the OIS assembly due to the very limited assembly space available and the strict requirements requested by IO, especially the IOIS assembly, which has particularly difficult installation requirements including complicated shear pin assemblies. A conceptual and preliminary design has been developed by the Korean domestic agency (KODA) for the sub assembly and final assembly phase; the tool includes the ability to control both IOIS plates simultaneously. For design verification of the IOIS assembly tool mentioned above, structural analysis has been carried out considering seismic event. Also, a half sized mock-up has been fabricated and tested according to assembly procedures. In this paper, a description of tool design and the results of analysis and mock-test will be introduced

  6. Characterization of ITER tungsten qualification mock-ups exposed to high cyclic thermal loads

    Energy Technology Data Exchange (ETDEWEB)

    Pintsuk, Gerald, E-mail: g.pintsuk@fz-juelich.de [Forschungszentrum Jülich GmbH, D-52425 Jülich (Germany); Bednarek, Maja; Gavila, Pierre [Fusion for Energy, E-08019 Barcelona (Spain); Gerzoskovitz, Stefan [Plansee SE, Innovation Services, 6600 Reutte (Austria); Linke, Jochen [Forschungszentrum Jülich GmbH, D-52425 Jülich (Germany); Lorenzetto, Patrick; Riccardi, Bruno [Fusion for Energy, E-08019 Barcelona (Spain); Escourbiac, Frederic [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 Saint Paul lez Durance (France)

    2015-10-15

    Highlights: • Mechanical deformation of CuCrZr in case a thermal barrier layer has been formed due to impurity content in the cooling water. • Crack formation at the W/Cu interface starting at the block edge. • Porosity formation in the pure Cu interlayer. • Microstructural changes in tungsten down to the W/Cu interface, which indicates also high temperatures for the pure Cu interlayer. • Macrocrack formation in tungsten which is assumed to be ductile at the initiation point and brittle when proceeding toward the cooling tube. - Abstract: High heat flux tested small-scale tungsten monoblock mock-ups (5000 cycles at 10 MW/m{sup 2} and up to 1000 cycles at 20 MW/m{sup 2}) manufactured by Plansee and Ansaldo were characterized by metallographic means. Therein, the macrocrack formation and propagation in tungsten, its recrystallization behavior and the surface response to different heat load facilities were investigated. Furthermore, debonding at the W/Cu interface, void formation in the soft copper interlayer and microcrack formation at the inner surface of the CuCrZr cooling tube were found.

  7. Water Mock-up for the Sodium Waste Treatment Process

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Ho Yun; Kim, Jong Man; Kim, Byung Ho; Lee, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    complex reacting phenomena in the system to observe with the naked eye. Therefore, a water mockup was carried out for the practical use of the data in the waste sodium treatment test

  8. Pre-brazed casting and hot radial pressing: A reliable process for the manufacturing of CFC and W monoblock mock-ups

    Energy Technology Data Exchange (ETDEWEB)

    Visca, Eliseo [Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, Via E. Fermi, 45, IT-00044 Frascati, RM (Italy)], E-mail: visca@frascati.enea.it; Libera, S.; Mancini, A.; Mazzone, G.; Pizzuto, A. [Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, Via E. Fermi, 45, IT-00044 Frascati, RM (Italy); Testani, C. [CSM S.p.A., IT-00128 Castel Romano, RM (Italy)

    2007-10-15

    ENEA is involved in the European International Thermonuclear Experimental Reactor (ITER) R and D activities and, in particular, for the manufacturing of high heat flux plasma-facing components (HHFC), such as the divertor targets, the baffles and the limiters. During last years, ENEA has manufactured actively cooled mock-ups by using different technologies, namely brazing, diffusion bonding and hot isostatic pressing (HIPping). A new manufacturing process has been set up and tested. It was successfully applied for the manufacturing of W armoured monoblock mock-ups. This technique is the HRP (hot radial pressing) based on performing a radial diffusion bonding between the cooling tube and the armour tile by pressurizing only internal tube and by keeping the joining zone in vacuum at the required bonding temperature. The heating is obtained by a standard air furnace. The HRP technique is now used for the manufacturing of CFC armoured monoblock components. For this purpose, some issues have to be faced, like the low CFC tensile strength, the pure copper interlayer between the heat sink and the armour necessary to mitigate the stress at the joint interface, and the low wettability of the pure copper on the CFC matrix. This paper reports the research path followed to manufacture a medium scale vertical target CFC and W armoured mock-up by HRP. A casting of a soft copper interlayer between the tube and the tile was obtained by a new technique: the pre-brazed casting (PBC, ENEA patent). Some preliminary mock-ups with three NB31 CFC tiles were successfully manufactured and tested to thermal fatigue using electron beam facilities. They all reached at least 1000 cycles at 20 MW/m{sup 2} without suffering any damage. The manufactured medium scale vertical target mock-up is now under testing at the FE2000 (France) facility. These activities were performed in the frame of ITER-EFDA contracts.

  9. Pre-brazed casting and hot radial pressing: A reliable process for the manufacturing of CFC and W monoblock mock-ups

    International Nuclear Information System (INIS)

    Visca, Eliseo; Libera, S.; Mancini, A.; Mazzone, G.; Pizzuto, A.; Testani, C.

    2007-01-01

    ENEA is involved in the European International Thermonuclear Experimental Reactor (ITER) R and D activities and, in particular, for the manufacturing of high heat flux plasma-facing components (HHFC), such as the divertor targets, the baffles and the limiters. During last years, ENEA has manufactured actively cooled mock-ups by using different technologies, namely brazing, diffusion bonding and hot isostatic pressing (HIPping). A new manufacturing process has been set up and tested. It was successfully applied for the manufacturing of W armoured monoblock mock-ups. This technique is the HRP (hot radial pressing) based on performing a radial diffusion bonding between the cooling tube and the armour tile by pressurizing only internal tube and by keeping the joining zone in vacuum at the required bonding temperature. The heating is obtained by a standard air furnace. The HRP technique is now used for the manufacturing of CFC armoured monoblock components. For this purpose, some issues have to be faced, like the low CFC tensile strength, the pure copper interlayer between the heat sink and the armour necessary to mitigate the stress at the joint interface, and the low wettability of the pure copper on the CFC matrix. This paper reports the research path followed to manufacture a medium scale vertical target CFC and W armoured mock-up by HRP. A casting of a soft copper interlayer between the tube and the tile was obtained by a new technique: the pre-brazed casting (PBC, ENEA patent). Some preliminary mock-ups with three NB31 CFC tiles were successfully manufactured and tested to thermal fatigue using electron beam facilities. They all reached at least 1000 cycles at 20 MW/m 2 without suffering any damage. The manufactured medium scale vertical target mock-up is now under testing at the FE2000 (France) facility. These activities were performed in the frame of ITER-EFDA contracts

  10. Experiment and analysis of hypervapotron mock-ups for preparing the 2nd qualification of the ITER blanket first wall

    International Nuclear Information System (INIS)

    Lee, Dong Won; Bae, Young Dug; Kim, Suk Kwon; Bang, In Cheol

    2010-01-01

    According to the increased heat flux condition up to 5 MW/m 2 in the International Thermonuclear Experimental Reactor (ITER), new design of the blanket first wall (FW) has been considered and the analysis was performed with ANSYS-CFX for checking its temperature with the ITER operation conditions. And a semi-prototype of the FW was proposed to be tested with the similar heat flux conditions under the second qualification for the FW procurement. In order to investigate the fabrication procedure and analysis capability of the code, two types of mock-up were fabricated according to the current semi-prototype design except for bending shape; one with hypervapotron and another without it. They were tested with KoHLT-2 (Korea Heat Load Test) facility and the results were compared with the ones by CFX code. The mass flow rate of inlet coolant was the same as the ITER condition and heat flux was loaded up to 0.48 MW/m 2 heat flux. The results show that the temperature of the mock-up can be predicted using the CFX code even with the complex geometry and the hypervapotron shows its function to increase the cooling.

  11. Weld distortion prediction and control of the ITER vacuum vessel manufacturing mock-ups

    International Nuclear Information System (INIS)

    Ottolini, Marco; Barbensi, Andrea

    2014-01-01

    The fabrication of the ITER Vacuum Vessel Sectors is an unprecedented challenge, due to their dimensions, the close tolerances, the complex 'D' shape. The technological issues were faced by the production of full scale mock ups to confirm the manufacturing feasibility to achieve very tight tolerances and qualify the main manufacturing processes, by a step by step welding distortion control, by the qualification of not conventional NDT inspection techniques and by innovative 3D dimensional inspections. The Supplier is required to fabricate at least two mock ups, inboard and outboard, related to the manufacturing method of the VV Sectors, to demonstrate the control of the welding distortions to achieve tolerances, optimizing welding sequences and calibrating of welding distortions computer simulations. The stages of this preparatory activity are: prediction of welding distortion for fabrication mock ups representative of selected segments; demonstration that distortion predictions are consistent with experimental results from 3D dimensional inspection; understanding of reasons of possible deviations between numerical and experimental results and definition of action to solve these issues; demonstration that possible calculation simplifications, adopted to speed up the analysis process, do not affect significantly the welding distortion prediction. This paper describes the weld distortion prediction and control on the manufacturing mock-ups of ITER Vacuum Vessel Sectors, with particular emphasis to the lessons learned. (authors)

  12. Team training using full-scale reactor coolant pump seal mock-ups

    International Nuclear Information System (INIS)

    McDonald, T.J.; Hamill, R.W.

    1987-01-01

    The use of full-scale reactor coolant pump (RCP) seal mock-ups has greatly enhanced Northeast Utilities' ability to effectively utilize the team training approach to technical training. With the advent of the Institute of Nuclear Power Operations accreditation come a new emphasis and standards for the integrated training of plant engineering personnel, maintenance mechanics, quality control personnel, and health physics personnel. The results of purchasing full-scale RCP mock-ups to pilot the concept of team training have far exceeded expectations and cost-limiting factors. The initial training program analysis identified RCP seal maintenance as a task that required training for maintenance department personnel. Due to radiation exposure considerations and the unavailability of actual plant equipment for training purposes, the decision was made to procure a mock-up of an RCP seal assembly and housing. This mock-up was designed to facilitate seal cartridge removal, disassembly, assembly, and installation, duplicating all internal components of the seal cartridge and housing area in exact detail

  13. Advanced Spacesuit Portable Life Support System Packaging Concept Mock-Up Design & Development

    Science.gov (United States)

    O''Connell, Mary K.; Slade, Howard G.; Stinson, Richard G.

    1998-01-01

    A concentrated development effort was begun at NASA Johnson Space Center to create an advanced Portable Life Support System (PLSS) packaging concept. Ease of maintenance, technological flexibility, low weight, and minimal volume are targeted in the design of future micro-gravity and planetary PLSS configurations. Three main design concepts emerged from conceptual design techniques and were carried forth into detailed design, then full scale mock-up creation. "Foam", "Motherboard", and "LEGOtm" packaging design concepts are described in detail. Results of the evaluation process targeted maintenance, robustness, mass properties, and flexibility as key aspects to a new PLSS packaging configuration. The various design tools used to evolve concepts into high fidelity mock ups revealed that no single tool was all encompassing, several combinations were complimentary, the devil is in the details, and, despite efforts, many lessons were learned only after working with hardware.

  14. Status report on active stabilisation of a linear collider final focus quadrupole mock-up

    International Nuclear Information System (INIS)

    Lottin, J.; Brunetti, L.; Formosa, F.; Adloff, C.; Bastian, Y.; Bolzon, B.; Cadoux, F.; Geffroy, N.; Girard, C.; Jeremie, A.; Karyotakis, Y.; Peltier, F.

    2006-01-01

    The measurements done with the sensors available in our laboratories used for ground motion analysis are presented. The first sensors studied are seismic sensors measuring ground velocity, other sensors are accelerometers available for measuring ground acceleration. The first step has been to characterize the sensors, the second step has been to model and simulate the acceleration in order to identify Eigen frequencies and to display mode shapes. The third step has been to assess the performances of a new algorithm for disturbance rejection. In order to facilitate the analysis, a reduced-size mock-up has been used. The goal was to eliminate or at least to reduce as much as possible the main frequencies of the disturbance. A new mock-up is currently being developed that will have a geometry closer to a final focus quadrupole. Measurements will be done to validate the whole system in view of active stabilization for a future linear collider

  15. Status report on active stabilisation of a linear collider final focus quadrupole mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Lottin, J.; Brunetti, L.; Formosa, F. [Universite de Savoie, ESIA, 74 - Annecy (France); Adloff, C.; Bastian, Y.; Bolzon, B.; Cadoux, F.; Geffroy, N.; Girard, C.; Jeremie, A.; Karyotakis, Y.; Peltier, F. [LAPP-IN2P3-CNRS, 74 - Annecy-le-Vieux (France)

    2006-07-01

    The measurements done with the sensors available in our laboratories used for ground motion analysis are presented. The first sensors studied are seismic sensors measuring ground velocity, other sensors are accelerometers available for measuring ground acceleration. The first step has been to characterize the sensors, the second step has been to model and simulate the acceleration in order to identify Eigen frequencies and to display mode shapes. The third step has been to assess the performances of a new algorithm for disturbance rejection. In order to facilitate the analysis, a reduced-size mock-up has been used. The goal was to eliminate or at least to reduce as much as possible the main frequencies of the disturbance. A new mock-up is currently being developed that will have a geometry closer to a final focus quadrupole. Measurements will be done to validate the whole system in view of active stabilization for a future linear collider.

  16. Numerical studies on helium cooled divertor finger mock up with sectorial extended surfaces

    International Nuclear Information System (INIS)

    Rimza, Sandeep; Satpathy, Kamalakanta; Khirwadkar, Samir; Velusamy, Karupanna

    2014-01-01

    Highlights: • Studies on heat transfer enhancement for divertor finger mock-up. • Heat transfer characteristics of jet impingement with extended surfaces have been investigated. • Effect of critical parameters that influence the thermal performance of the finger mock-up by CFD approach. • Effect of extended surface in enhancing heat removal potential with pumping power assessed. • Practicability of the chosen design is verified by structural analysis. - Abstract: Jet impinging technique is an advance divertor concept for the design of future fusion power plants. This technique is extensively used due to its high heat removal capability with reasonable pumping power and for safe operation. In this design, plasma-facing components are fabricated with numerous fingers cooled by helium jets to reduce the thermal stresses. The present study is focused towards finding an optimum performance of one such finger mock-up through systematic computational fluid dynamics (CFD) studies. Heat transfer characteristics of jet impingement have been numerically investigated with sectorial extended surfaces (SES). The result shows that addition of SES enhances heat removal potential with minimum pumping power. Detailed parametric studies on critical parameters that influence thermal performance of the finger mock-up have been analyzed. Thermo-mechanical analysis has been carried out through finite element based approach to know the state of stress in the assembly as a result of large temperature gradients. It is seen that the stresses are within the permissible limits for the present design. The whole numerical simulation has been carried out using general-purpose CFD software (ANSYS FLUENT, Release 14.0, User Guide, Ansys, Inc., 2011). Benchmark validation studies have been performed against high-heat flux experiments (B. Končar, P. Norajitra, K. Oblak, Appl. Therm. Eng., 30, 697–705, 2010) and a good agreement is noticed between the present simulation and the reported

  17. Fabrication of full-size mock-up for 10° section of ITER vacuum vessel thermal shield

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dong Kwon [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Nam, Kwanwoo, E-mail: kwnam@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Kang, Kyoung-O; Noh, Chang Hyun; Chung, Wooho [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Lim, Kisuk; Kang, Youngkil [SFA Engineering Corp., Asan-si, Chungcheongnam-do 336-873 (Korea, Republic of); Hamlyn-Harris, Craig; Her, Namil; Robby, Hicks [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2015-10-15

    In this paper, a full-scale prototype fabrication for vacuum vessel thermal shield (VVTS) of ITER tokamak is described and test results are reported. All the manufacturing processes except for silver coating were performed in the fabrication of 10° section of VVTS. Pre-qualification test was conducted to compare the vertical and the horizontal welding positions. After shell welding, shell distortion was measured and adjusted. Shell thickness change was also measured after buffing process. Specially, VVTS ports need large bending and complex welding of shell and flange. Bending method for the complex and long cooling tube layout especially for the VVTS ports was developed in detail. Dimensional inspection of the fabricated mock-up was performed with a 3D laser scanner and the scanning data was analyzed.

  18. Thermal transient and the temperature profile in a HELICA mock-up simulated by a new finite element homogenous model

    International Nuclear Information System (INIS)

    Zaccari, Nicola; Aquaro, Donato

    2013-01-01

    Highlights: • We have developed a numerical model of the pebble beds is based on the results of a theoretical and experimental research activity performed. • The model has been used to simulate the experimental tests performed on HELICA mock-up (ENEA Italy). • Moreover the numerical results are compared with the experimental ones. Finally, a discussion on results obtained by other authors involved in the benchmark is reported. -- Abstract: This paper deals with a numerical approach for simulating the thermal and mechanical behaviour of pebble beds used as breeder and neutron multiplier in breeding blanket of nuclear fusion reactor. The model of the pebble beds is based on the results of a theoretical and experimental research activity performed by the Authors on ceramic pebble beds (lithium ortosilicate and lithium metatitanate). The results of this activity permitted to determine the effective thermal conductivity of the beds, versus the temperature and the axial pressure and to implement a homogenous model of pebble bed in a FEM code. This paper illustrates an application of the implemented model, considering pebble beds under several cycles of heating and cooling. The examined geometry corresponds to the HELICA mock-up tested by ENEA in the research centre Brasimone. The experimental tests performed on HELICA have been used as a benchmark problem in order to assess the different approaches for simulating pebble beds. In this paper, the simulations performed with two-dimensional models are illustrated. Moreover the numerical results are compared with the experimental ones. Finally, a discussion on results obtained by other authors involved in the benchmark is reported

  19. Fabrication of divertor mock-up with ODS-Cu and W by the improved brazing technique

    Science.gov (United States)

    Tokitani, M.; Hamaji, Y.; Hiraoka, Y.; Masuzaki, S.; Tamura, H.; Noto, H.; Tanaka, T.; Muroga, T.; Sagara, A.; FFHR Design Group

    2017-07-01

    Copper alloy has been considered as a divertor cooling tube or heat sink not only in the helical reactor FFHR-d1 but also in the tokamak DEMO reactor, because it has a high thermal conductivity. This work focused on applying an oxide dispersion strengthened copper alloy (ODS-Cu), GlidCop® (Cu-0.3 wt%Al2O3) as the divertor heat sink material of FFHR-d1. This alloy has superior high temperature yield strength exceeding 300 MPa at room temperature even after annealing up to ~1000 °C. The change in material properties of Pure-Cu, GlidCop® and CuCrZr by neutron irradiation are summarized in this paper. A primary dose limit is the radiation-induced hardening/softening (~0.2 dpa/1-2 dpa) which has a temperature dependence. According to such an evaluation, the GlidCop® can be selected as the current best candidate material in the commercial base of the divertor heat sink, and its temperature should be maintained as close as possible to 300 °C during operation. Bonding between the W armour and the GlidCop® heat sink was successfully performed by using an improved brazing technique with BNi-6 (Ni-11%P) filler material. The bonding strength was measured by a three-point bending test and reached up to approximately 200 MPa. Surprisingly, several specimens showed an obvious yield point. This means that the BNi-6 brazing (bonding) layer caused relaxation of the applied stress. The small-scale divertor mock-up of the W/BNi-6/GlidCop® was successfully fabricated by using the improved brazing technique. The heat loading test was carried out by the electron beam device ACT2 in NIFS. The mock-up showed an excellent heat removal capability for use in the FFHR-d1 divertor.

  20. First steps in designing an all-in-one ICT-based device for persons with cognitive impairment: evaluation of the first mock-up.

    Science.gov (United States)

    Boman, Inga-Lill; Persson, Ann-Christine; Bartfai, Aniko

    2016-03-07

    This project Smart Assisted Living involving Informal careGivers++ (SALIG) intends to develop an ICT-based device for persons with cognitive impairment combined with remote support possibilities for significant others and formal caregivers. This paper presents the identification of the target groups' needs and requirements of such device and the evaluation of the first mock-up, demonstrated in a tablet. The inclusive design method that includes end-users in the design process was chosen. First, a scoping review was conducted in order to examine the target group's need of an ICT-based device, and to gather recommendations regarding its design and functionalities. In order to capture the users' requirements of the design and functionalities of the device three targeted focus groups were conducted. Based on the findings from the publications and the focus groups a user requirement specification was developed. After that a design concept and a first mock-up was developed in an iterative process. The mock-up was evaluated through interviews with persons with cognitive impairment, health care professionals and significant others. Data were analysed using content analysis. Several useful recommendations of the design and functionalities of the SALIG device for persons with cognitive impairment were identified. The main benefit of the mock-up was that it was a single device with a set of functionalities installed on a tablet and designed for persons with cognitive impairment. An additional benefit was that it could be used remotely by significant others and formal caregivers. The SALIG device has the potentials to facilitate everyday life for persons with cognitive impairment, their significant others and the work situation for formal caregivers. The results may provide guidance in the development of different types of technologies for the target population and for people with diverse disabilities. Further work will focus on developing a prototype to be empirically tested

  1. A SCALE-UP Mock-Up: Comparison of Student Learning Gains in High- and Low-Tech Active-Learning Environments.

    Science.gov (United States)

    Soneral, Paula A G; Wyse, Sara A

    2017-01-01

    Student-centered learning environments with upside-down pedagogies (SCALE-UP) are widely implemented at institutions across the country, and learning gains from these classrooms have been well documented. This study investigates the specific design feature(s) of the SCALE-UP classroom most conducive to teaching and learning. Using pilot survey data from instructors and students to prioritize the most salient SCALE-UP classroom features, we created a low-tech "Mock-up" version of this classroom and tested the impact of these features on student learning, attitudes, and satisfaction using a quasi--experimental setup. The same instructor taught two sections of an introductory biology course in the SCALE-UP and Mock-up rooms. Although students in both sections were equivalent in terms of gender, grade point average, incoming ACT, and drop/fail/withdraw rate, the Mock-up classroom enrolled significantly more freshmen. Controlling for class standing, multiple regression modeling revealed no significant differences in exam, in-class, preclass, and Introduction to Molecular and Cellular Biology Concept Inventory scores between the SCALE-UP and Mock-up classrooms. Thematic analysis of student comments highlighted that collaboration and whiteboards enhanced the learning experience, but technology was not important. Student satisfaction and attitudes were comparable. These results suggest that the benefits of a SCALE-UP experience can be achieved at lower cost without technology features. © 2017 P. A. G. Soneral and S. A. Wyse. CBE—Life Sciences Education © 2017 The American Society for Cell Biology. This article is distributed by The American Society for Cell Biology under license from the author(s). It is available to the public under an Attribution–Noncommercial–Share Alike 3.0 Unported Creative Commons License (http://creativecommons.org/licenses/by-nc-sa/3.0).

  2. Use of nuclear data sensitivity and uncertainty analysis for the design preparation of the HCLL breeder blanket mock-up experiment for ITER

    International Nuclear Information System (INIS)

    Kodeli, I.

    2007-01-01

    An experiment on a mock-up of the Test Blanket module based on Helium Cooled Lithium Lead (HCLL) concept will be performed in 2007 in the FNG utility in Frascati in order to study neutronics characteristics of the module and the performance of the computational tools in the accurate prediction of the neutron transport. With the objective to prepare and optimise the design of the mock-up in the sense to provide maximum information on the state-of-the-art of the cross section data the mock-up was pre-analysed using the deterministic codes for the sensitivity/uncertainty analysis. The neutron fluxes and tritium production rate (TPR), their sensitivity to the underlying basic cross sections, as well as the corresponding uncertainty estimations were calculated using the deterministic transport codes (DOORS package), the sensitivity/uncertainty code package SUSD3D and the VITAMIN-J/COVA covariance matrix libraries. The cross section reactions with largest contribution to the uncertainty in the calculation of the TPR were identified to be (n,2n) and (n,3n) reactions on plumb. The conclusions of this work support the main benchmark design and suggest some modifications and improvements. In particular this study recommends the use, as far as possible, of both natural and enriched lithium pellets for the TRP measurements. The combined use is expected to provide additional and complementary information on the sensitive cross sections. (author)

  3. A SCALE-UP Mock-Up: Comparison of Student Learning Gains in High- and Low-Tech Active-Learning Environments

    Science.gov (United States)

    Soneral, Paula A. G.; Wyse, Sara A.

    2017-01-01

    Student-centered learning environments with upside-down pedagogies (SCALE-UP) are widely implemented at institutions across the country, and learning gains from these classrooms have been well documented. This study investigates the specific design feature(s) of the SCALE-UP classroom most conducive to teaching and learning. Using pilot survey data from instructors and students to prioritize the most salient SCALE-UP classroom features, we created a low-tech “Mock-up” version of this classroom and tested the impact of these features on student learning, attitudes, and satisfaction using a quasi-­experimental setup. The same instructor taught two sections of an introductory biology course in the SCALE-UP and Mock-up rooms. Although students in both sections were equivalent in terms of gender, grade point average, incoming ACT, and drop/fail/withdraw rate, the Mock-up classroom enrolled significantly more freshmen. Controlling for class standing, multiple regression modeling revealed no significant differences in exam, in-class, preclass, and Introduction to Molecular and Cellular Biology Concept Inventory scores between the SCALE-UP and Mock-up classrooms. Thematic analysis of student comments highlighted that collaboration and whiteboards enhanced the learning experience, but technology was not important. Student satisfaction and attitudes were comparable. These results suggest that the benefits of a SCALE-UP experience can be achieved at lower cost without technology features. PMID:28213582

  4. Containment Evaluation under Severe Accidents (CESA): synthesis of the predictive calculations and analysis of the first experimental results obtained on the Civaux mock-up

    International Nuclear Information System (INIS)

    Granger, L.; Rieg, C.Y.; Touret, J.P.; Fleury, F.; Nahas, G.; Danisch, R.; Brusa, L.; Millard, A.; Laborderie, C.; Ulm, F.; Contri, P.; Schimmelpfennig, K.; Barre, F.; Firnhaber, M.; Gauvain, J; Coulon, N.; Dutton, L.M.C.; Tuson, A.

    2001-01-01

    In 1996, EDF decided to build a containment model at the scale 1:3, the MAEVA mock-up, in order to check and study the behaviour of a pre-stressed concrete containment vessel without a liner in terms of mechanical strength and leaktightness, for loadings corresponding to its design and beyond design conditions. In parallel with the construction and testing of the mock-up, a cost-shared R and D action supported by the European Union, the CESA project, is dealing with quantification of leak rates through concrete cracks and porosity, predictive calculations of the behaviour of the mock-up and analysis of the experimental results. In this paper, we propose a synthesis of the main theoretical and experimental results, obtained after 2.5 years. It should however be noted that, due to some unexpected delays in the experimental programme, quite natural with such a huge and unique experimental set-up, only the design-basis accident sequences, already performed, have been reported in this paper. The first results are nevertheless very interesting, both from a scientific and nuclear utility point of view

  5. Status of the Digital Mock-up System for the dismantling of the nuclear facilities

    International Nuclear Information System (INIS)

    Park, Hee Seoung; Kim, S. K.; Lee, K. W.; Oh, W. J.

    2004-12-01

    The database system have already developed is impossible to solve a quantitative evaluation about a various situation from the dismantle activities of the reactor had contaminated with radioactivity. To satisfy the requirements for safety and economical efficiency among a major decommissioning technologies, it need a system that can evaluate and estimate dismantling scheduling, amount of radioactive waste being dismantled, and decommissioning cost. We have review and analyzed status of the digital mock-up system to get a technical guide because we have no experience establishment of one relation to dismantling of research reactor and nuclear power plant

  6. NDE of explosion welded copper stainless steel first wall mock-up

    International Nuclear Information System (INIS)

    Taehtinen, S.; Kauppinen, P.; Jeskanen, H.; Lahdenperae, K.; Ehrnsten, U.

    1997-04-01

    The study showed that reflection type C-mode scanning acoustic microscope (C-SAM) and internal ultrasonic inspection (IRIS) equipment can be applied for ultrasonic examination of copper stainless steel compound structures of ITER first wall mock-ups. Explosive welding can be applied to manufacture fully bonded copper stainless steel compound plates. However, explosives can be applied only for mechanical tightening of stainless steel cooling tubes within copper plate. If metallurgical bonding between stainless steel tubes and copper plate is required Hot Isostatic Pressing (HIP) method can be applied. (orig.)

  7. Plasma cleaning of ITER edge Thomson scattering mock-up mirror in the EAST tokamak

    Science.gov (United States)

    Yan, Rong; Moser, Lucas; Wang, Baoguo; Peng, Jiao; Vorpahl, Christian; Leipold, Frank; Reichle, Roger; Ding, Rui; Chen, Junling; Mu, Lei; Steiner, Roland; Meyer, Ernst; Zhao, Mingzhong; Wu, Jinhua; Marot, Laurent

    2018-02-01

    First mirrors are the key element of all optical and laser diagnostics in ITER. Facing the plasma directly, the surface of the first mirrors could be sputtered by energetic particles or deposited with contaminants eroded from the first wall (tungsten and beryllium), which would result in the degradation of the reflectivity. The impurity deposits emphasize the necessity of the first mirror in situ cleaning for ITER. The mock-up first mirror system for ITER edge Thomson scattering diagnostics has been cleaned in EAST for the first time in a tokamak using radio frequency capacitively coupled plasma. The cleaning properties, namely the removal of contaminants and homogeneity of cleaning were investigated with molybdenum mirror insets (25 mm diameter) located at five positions over the mock-up plate (center to edge) on which 10 nm of aluminum oxide, used as beryllium proxy, were deposited. The cleaning efficiency was evaluated using energy dispersive x-ray spectroscopy, reflectivity measurements and x-ray photoelectron spectroscopy. Using argon or neon plasma without magnetic field in the laboratory and with a 1.7 T magnetic field in the EAST tokamak, the aluminum oxide films were homogeneously removed. The full recovery of the mirrors’ reflectivity was attained after cleaning in EAST with the magnetic field, and the cleaning efficiency was about 40 times higher than that without the magnetic field. All these results are promising for the plasma cleaning baseline scenario of ITER.

  8. Development of a digital mock-up system for selecting a decommissioning scenario

    International Nuclear Information System (INIS)

    Kim, Sung-Kyun; Park, Hee-Sung; Lee, Kune-Woo; Jung, Chong-Hun

    2006-01-01

    The evaluation of decommissioning scenarios is critical to the successful development and execution of a decommissioning project. In the past, many experts have used a physical mock-up system to find the exact work processes and the working positions. Nowadays, these jobs are being done by a Digital Mock-Up (DMU) system. The DMU, which is a technology to realize an effective work process by using virtual environments through representing the physical and logical schema and the behavior of a real decommissioning work, can save on the cost and time, reduce the risk of making later changes, and develop various decommissioning scenarios. In this research, a decommissioning DMU system was developed for simulating the relevant dismantling processes. Decommissioning data-computing modules which can calculate a dismantling schedule, quantify a radioactive waste, visualize a radioactive inventory, estimate a decommissioning cost, and estimate a worker's exposure were also developed to qualitatively assess the decommissioning information. And an analytic hierarchy process (AHP) model was developed to evaluate the decommissioning scenarios which reflected the quantitative and qualitative considerations. To establish the proper scenario for the thermal column in KRR-1, the developed decommissioning DMU system was applied to evaluate the two candidate scenarios of it

  9. FEBEX: An example of a major international collaborative project

    International Nuclear Information System (INIS)

    Ulibarri, A.M.; Olmo, C. del; Huertas, F.

    1996-01-01

    There are many similarities in the high-level waste (HLW) disposal programmes in Switzerland and Spain. In both cases, alternative crystalline and sedimentary host rocks are currently under investigation, options for disposal of both vitrified reprocessing waste and spent fuel are considered and repository designs utilize massive engineered barriers. For the case of HLW disposal in a granite rock, the reference engineered barrier system (EBS) concepts are almost identical. The waste, in its steel fabrication container, is sealed in a massive steel canister which is emplaced horizontally in drilled tunnels. The canister is surrounded by a highly compacted bentonite backfill. Individual components of this waste package have been tested in isolation or on a small scale, but the aim of the full-scale engineered barrier experiment is to examine some properties of a real size system in a realistic natural environment. FEBEX was proposed by ENRESA and the experimental studies at Grimsel are run as an ENRESA/NAGRA collaboration. The field experiments are, however, only one component of a project which includes a large-scale laboratory 'mockup' and supporting materials tests and modelling. FEBEX, as a whole, is sponsored by the European Union as part of the 'Nuclear Fission Safety' research programme (the Swiss component being supported by the Bundesamt fur Bildung und Wissenschaft)

  10. Validation results of satellite mock-up capturing experiment using nets

    Science.gov (United States)

    Medina, Alberto; Cercós, Lorenzo; Stefanescu, Raluca M.; Benvenuto, Riccardo; Pesce, Vincenzo; Marcon, Marco; Lavagna, Michèle; González, Iván; Rodríguez López, Nuria; Wormnes, Kjetil

    2017-05-01

    The PATENDER activity (Net parametric characterization and parabolic flight), funded by the European Space Agency (ESA) via its Clean Space initiative, was aiming to validate a simulation tool for designing nets for capturing space debris. This validation has been performed through a set of different experiments under microgravity conditions where a net was launched capturing and wrapping a satellite mock-up. This paper presents the architecture of the thrown-net dynamics simulator together with the set-up of the deployment experiment and its trajectory reconstruction results on a parabolic flight (Novespace A-310, June 2015). The simulator has been implemented within the Blender framework in order to provide a highly configurable tool, able to reproduce different scenarios for Active Debris Removal missions. The experiment has been performed over thirty parabolas offering around 22 s of zero-g conditions. Flexible meshed fabric structure (the net) ejected from a container and propelled by corner masses (the bullets) arranged around its circumference have been launched at different initial velocities and launching angles using a pneumatic-based dedicated mechanism (representing the chaser satellite) against a target mock-up (the target satellite). High-speed motion cameras were recording the experiment allowing 3D reconstruction of the net motion. The net knots have been coloured to allow the images post-process using colour segmentation, stereo matching and iterative closest point (ICP) for knots tracking. The final objective of the activity was the validation of the net deployment and wrapping simulator using images recorded during the parabolic flight. The high-resolution images acquired have been post-processed to determine accurately the initial conditions and generate the reference data (position and velocity of all knots of the net along its deployment and wrapping of the target mock-up) for the simulator validation. The simulator has been properly

  11. Design and Analysis of the Korean Small Semi-prototype Mock-up for the 2nd Qualification of the ITER Blanket First Wall

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Suk Kwon; Yoon, Jae Sung; Lee, Eo Hwak; Lee, Seung Jae; Choi, Bo Guen; Park, Jeong Yong; Jung, Yang Il; Choi, Byung Kwon; Kim, Byoung Yoon

    2011-01-01

    Since the blanket First Wall (FW) of the International Thermonuclear Experimental Reactor (ITER) is subjected to a high heat and high neutron loads, it is one of the most important components. It composed of a beryllium (Be) layer as a plasma facing material, a copper alloy (CuCrZr) layer as a heat sink and type 316L authentic stainless steel (SS316L) as a structure material. The joining of the three different metals is the key issue to be solved. And more, the peak heat load was assumed to be 0.5 MW/m 2 in the initial design of the FW, but it was changed to be up to 5 MW/m 2 In Korea, the joining method has developed and it was proved through the several mock-up fabrication and high heat flux tests for confirming the joining integrity. Some of them were tested in the foreign facilities such as JEBIS at JAEA in Japan, TSEFEY at Efremov in Russia, and JUDITH at FZJ in Germany, and others were tested in our own facilities such as KoHLT-1 and -2. And finally, the 1 st , recently. Therefore, the FW panel design has been changed for enhancing the cooling and ITER Organization will provide the proposed design. Qualification was passed, in which two 80x80x3 Be/Cu/SS mock-ups were tested under 0.625 and 0.875 MW/m 2 heat fluxes for 12,000 cycles and then tested under 1.75 and 1.40 MW/m 2 Currently, the 2 heat fluxes for 1,000 cycles at FZJ and SNL, respectively. Currently, the 2 nd qualification program was started and the semi-prototype should be fabricated by the end of 2011 for testing under 5.0 MW/m 2 heat flux for certain number of cycles. In order to prepare the semi-prototype, several fabrication methods should be developed through the fabrication and test with the several mock-ups. In the present study, small Be mock-up was fabricated as the first step for the preparation. It was fabricated according to the designs considering the currently modified design of the FW. In the present paper, the fabrication objectives, methods, results and related tests were

  12. XML-based assembly visualization for a multi-CAD digital mock-up system

    International Nuclear Information System (INIS)

    Song, In Ho; Chung, Sung Chong

    2007-01-01

    Using a virtual assembly tool, engineers are able to design accurate and interference free parts without making physical mock-ups. Instead of a single CAD source, several CAD systems are used to design a complex product in a distributed design environment. In this paper, a multi-CAD assembly method is proposed through an XML and the lightweight CAD file. XML data contains a hierarchy of the multi-CAD assembly. The lightweight CAD file produced from various CAD files through the ACIS kemel and InterOp includes not only mesh and B-Rep data, but also topological data. It is used to visualize CAD data and to verify dimensions of the parts. The developed system is executed on desktop computers. It does not require commercial CAD systems to visualize 3D assembly data. Multi-CAD models have been assembled to verify the effectiveness of the developed DMU system on the Internet

  13. Mock-up experiment and analysis for the primary shield of the N.S. MUTSU

    International Nuclear Information System (INIS)

    Miyasaka, S.; Asaoka, T.; Taji, Y.; Ise, T.; Koyama, K.; Tsutsui, T.; Takeuchi, M.; Fuse, T.; Miura, T.; Yamaji, Y.

    1977-01-01

    A series of shielding mock-up experiments was performed at JRR-4, a swimming pool type reactor, of Japan Atomic Energy Research Institute (JAERI) to obtain the necessary experimental data and the sufficiently accurate method of calculation adopted for the modification of the MUTSU primary shield. Analyses for the experiments were carried out by using of the Ssub(n) codes, ANISN and TWOTRAN. The two dimensional calculations were performed with the P 1 -S 8 approximation. The neutron streaming through the annular gap between the pressure vessel and the primary shield has been confirmed to be estimated from the present method of calculation. The agreement between the calculated and the measured values is generally in about a factor of 2 to 4. (orig.) [de

  14. A generalized approach for historical mock-up acquisition and data modelling: Towards historically enriched 3D city models

    Science.gov (United States)

    Hervy, B.; Billen, R.; Laroche, F.; Carré, C.; Servières, M.; Van Ruymbeke, M.; Tourre, V.; Delfosse, V.; Kerouanton, J.-L.

    2012-10-01

    Museums are filled with hidden secrets. One of those secrets lies behind historical mock-ups whose signification goes far behind a simple representation of a city. We face the challenge of designing, storing and showing knowledge related to these mock-ups in order to explain their historical value. Over the last few years, several mock-up digitalisation projects have been realised. Two of them, Nantes 1900 and Virtual Leodium, propose innovative approaches that present a lot of similarities. This paper presents a framework to go one step further by analysing their data modelling processes and extracting what could be a generalized approach to build a numerical mock-up and the knowledge database associated. Geometry modelling and knowledge modelling influence each other and are conducted in a parallel process. Our generalized approach describes a global overview of what can be a data modelling process. Our next goal is obviously to apply this global approach on other historical mock-up, but we also think about applying it to other 3D objects that need to embed semantic data, and approaching historically enriched 3D city models.

  15. Validation of CLIC Re-Adjustment System Based on Eccentric Cam Movers One Degree of Freedom Mock-Up

    CERN Document Server

    Kemppinen, J; Lackner, F

    2011-01-01

    Compact Linear Collider (CLIC) is a 48 km long linear accelerator currently studied at CERN. It is a high luminosity electron-positron collider with an energy range of 0.5-3 TeV. CLIC is based on a two-beam technology in which a high current drive beam transfers RF power to the main beam accelerating structures. The main beam is steered with quadrupole magnets. To reach CLIC target luminosity, the main beam quadrupoles have to be actively pre-aligned within 17 µm in 5 degrees of freedom and actively stabilised at 1 nm in vertical above 1 Hz. To reach the pre-alignment requirement as well as the rigidity required by nano-stabilisation, a system based on eccentric cam movers is proposed for the re-adjustment of the main beam quadrupoles. Validation of the technique to the stringent CLIC requirements was started with tests in one degree of freedom on an eccentric cam mover. This paper describes the dedicated mock-up as well as the tests and measurements carried out with it. Finally, the test results are present...

  16. A constitutive model for the thermo-mechanical behaviour of fusion-relevant pebble beds and its application to the simulation of HELICA mock-up experimental results

    International Nuclear Information System (INIS)

    Vella, G.; Maio, P.A. Di; Giammusso, R.; Tincani, A.; Orco, G. Dell

    2006-01-01

    Within the framework of the activities promoted by European Fusion Development Agreement on the technology of the Helium Cooled Pebble Bed Test Blanket Module to be irradiated in one of the ITER equatorial ports, attention has been focused on the theoretical modelling of the thermo-mechanical constitutive behaviour of both beryllium and lithiated ceramics pebble beds, that are envisaged to act respectively as neutron multiplier and tritium breeder. The thermo-mechanical behaviour of the pebble beds and their nuclear performances in terms of tritium production depend on the reactor relevant conditions (heat flux and neutron wall load), the pebble sizes and the breeder cell geometries (bed thickness, pebble packing factor, bed overall thermal conductivity). ENEA-Brasimone and the Department of Nuclear Engineering (DIN) of the Palermo University have performed intense research activities intended to investigate fusion-relevant pebble bed thermo-mechanical behaviour by adopting both experimental and theoretical approaches. In particular, ENEA has carried out several experimental campaigns on small scale mock-ups tested in out-of-pile conditions, while DIN has developed a proper constitutive model that has been implemented on commercial FEM code, for the prediction of the thermal and mechanical performances of fusion-relevant pebble beds and for the comparison with the experimental results of the ENEA tests. In that framework, HELICA mock-up has been set-up and tested to investigate the behaviour of pebble bed in reactor-relevant geometries, providing useful data sets to be numerically reproduced by means of the DIN constitutive model, contributing to its assessment. The paper presents the constitutive model developed and the main experimental results of two test campaigns on HELICA mock-up carried out at HE-FUS 3 facility of ENEA Brasimone, the geometry of the mock-up, the adopted thermal and mechanical boundary conditions and the test operating conditions. The most

  17. Benchmarking of MCNPX Results with Measured Tritium Production Rate and Neutron Flux at the Mock-up of EU TBM (HCPB concept)

    Energy Technology Data Exchange (ETDEWEB)

    Tore, C.; Ortego, P.

    2013-07-01

    In order to reassesses the available design results of Test Breeder Modules (TBMs) a framework contract agreement between F4E and IDOM-Spain has been signed. SEA SL-Spain and UNED-Spain participate as sub-contractors of IDOM. In this study, a qualification of MCNPX code and nuclear data libraries are performed with benchmarking of measured tritium production and neutron flux at the mock-up of the EU TBM, HCPB concept. The irradiation and measurements had been performed in the frame of European Fusion Technology Program by ENEA-Italy, TUD-Germany and JAERI -Japan.

  18. Bouyancy effects on sodium coolant temperature profiles measured in an electrically heated mock-up of a 61-rod breeder reactor blanket assembly

    International Nuclear Information System (INIS)

    Engel, F.C.; Markley, R.A.; Minushkin, B.

    1978-01-01

    The paper describes test results selected to demonstrate the effect of buoyancy on the temperature profiles in a 61-rod electrically heated mock-up of an LMFBR radial blanket assembly. In these assemblies, heat transfer occurs over a wide range of complex operating conditions. The range and complexity of conditions are the result of the steep flux and power gradients which are an inherent feature of the blanket region and the power generation level in an assembly which can vary from 20 to 1100 kW

  19. Preparing ITER ICRF: development and analysis of the load resilient matching systems based on antenna mock-up measurements

    International Nuclear Information System (INIS)

    Messiaen, A.; Vervier, M.; Dumortier, P.; Grine, D.; Lamalle, P.U.; Durodie, F.; Koch, R.; Louche, F.; Weynants, R.

    2009-01-01

    The reference design for the ICRF antenna of ITER is constituted by a tight array of 24 straps grouped in eight triplets. The matching network must be load resilient for operation in ELMy discharges and must have antenna spectrum control for heating or current drive operation. The load resilience is based on the use of either hybrid couplers or conjugate-T circuits. However, the mutual coupling between the triplets at the low expected loading strongly counteracts the load resilience and the spectrum control. Using a mock-up of the ITER antenna array with adjustable water load matching solutions are designed. These solutions are derived from transmission line modelling based on the measured scattering matrix and are finally tested. We show that the array current spectrum can be controlled by the anti-node voltage distribution and that suitable decoupler circuits can not only neutralize the adverse mutual coupling effects but also monitor this anti-node voltage distribution. A matching solution using four 3 dB hybrids and the antenna current spectrum feedback control by the decouplers provides outstanding performance if each pair of poloidal triplets undergoes a same load variation. Finally, it is verified by modelling that this matching scenario has the same antenna spectrum and load resilience performances as the antenna array loaded by plasma as described by the TOPICA simulation. This is true for any phasing and frequency in the ITER frequency band. The conjugate-T solution is presently considered as a back-up option.

  20. Numerical evaluation of weld overlay applied to a pressurized water reactor nozzle mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Rabello, Emerson G.; Silva, Luiz L.; Gomes, Paulo T.V., E-mail: egr@cdtn.b, E-mail: silvall@cdtn.b, E-mail: gomespt@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Integridade Estrutural

    2011-07-01

    The primary water stress corrosion cracking (PWSCC) is a major mechanism of failure in the primary circuit of PWR type nuclear power plants. The PWSCC is associated with the presence of corrosive environment, the susceptibility to corrosion cracking of the materials involved and the tensile stresses presence. Residual stresses generated during dissimilar materials welding can contribute to PWSCC. An alternative to the PWSCC mitigation is the application of external weld layers in the regions of greatest susceptibility to corrosion cracking. This process, called Weld Overlay (WOL), has been widely used in regions of dissimilar weld (low alloy steel and stainless steel with nickel alloy addition) of nozzles and pipes on the primary circuit in order to promote internal compressive stresses on the wall of these components. This paper presents the steps required to the numerical stress evaluation (by finite element method) during the dissimilar materials welding as well as application of Weld Overlay process in a nozzle mock-up. Thus, one can evaluate the effectiveness of the application of weld overlay process to internal compressive stress generation on the wall nozzle. (author)

  1. Implementation of a Digital Mock-up for Remote Hot cell Operations

    International Nuclear Information System (INIS)

    Park, Hee Seong; Park, Byung Suk; Kim, Sung Hyun; Kim, Ki Ho; Kim, Ho Dong

    2010-01-01

    A remote manipulation environment that a human operator has to observe is the inner side of a hotcell through a lead grass window which has many obstacles due to many existing 'blind-spots' where are several cameras installed. The lack of visual information when operating in a cluttered environment makes manoeuvering a manipulator very difficult and when this situation is exacerbated by strict time limits for a task completion, then a manipulator and environmental collisions and resultant damage can occur. To cope with these problems, there has been efforts to develop a virtual simulator to validate control programs visually and to establish maintainability-engineering tools that automate generation assembly/disassembly procedures by using Computer Aided Design(CAD) visualization systems with human figure models to virtual reality systems where engineers can interact with the system using virtual input devices. This article introduces a system that can simulate a deployment analysis on a digital mock-up effectively and proposes a scheme to enable an operator to improve a remote manipulation by using a haptic device

  2. Numerical evaluation of weld overlay applied to a pressurized water reactor nozzle mock-up

    International Nuclear Information System (INIS)

    Rabello, Emerson G.; Silva, Luiz L.; Gomes, Paulo T.V.

    2011-01-01

    The primary water stress corrosion cracking (PWSCC) is a major mechanism of failure in the primary circuit of PWR type nuclear power plants. The PWSCC is associated with the presence of corrosive environment, the susceptibility to corrosion cracking of the materials involved and the tensile stresses presence. Residual stresses generated during dissimilar materials welding can contribute to PWSCC. An alternative to the PWSCC mitigation is the application of external weld layers in the regions of greatest susceptibility to corrosion cracking. This process, called Weld Overlay (WOL), has been widely used in regions of dissimilar weld (low alloy steel and stainless steel with nickel alloy addition) of nozzles and pipes on the primary circuit in order to promote internal compressive stresses on the wall of these components. This paper presents the steps required to the numerical stress evaluation (by finite element method) during the dissimilar materials welding as well as application of Weld Overlay process in a nozzle mock-up. Thus, one can evaluate the effectiveness of the application of weld overlay process to internal compressive stress generation on the wall nozzle. (author)

  3. Use of hydraulic and aerial mock up to study atmospheric pollution

    International Nuclear Information System (INIS)

    Facy, L.; Perrin De Brichambaut, C.; Doury, A.; Le Quinio, R.

    1962-01-01

    Fundamental studies on turbulent atmospheric diffusion of finely divided particles, cannot remain on a purely theoretical basis. Further experimental studies must be considered. - In full scale, from accidental and induced releases. - On a reduced scale, in aerodynamic wind tunnels or hydraulic water tunnels. A first set of studies on reduced scale models has been worked out according to a contract between French 'Meteorologie Nationale' and French 'Commissariat a l'Energie Atomique' and with the Collaboration of Saint-Cyr 'Institut Aerotechnique'. Essentially two kinds of results have been obtained: - The mathematical model of SUTTON for the turbulent diffusion in the atmosphere, deduced from the SUTTON theory, generally used by us, has been correctly verified, qualitatively and quantitatively whenever experiments were consistent with the theory conditions. - The quantitative assays of photographic and cinematographic visualization have given precise details on the phenomena inaccessible to calculations, due to the influence of obstacles and release conditions. - Generally, it can be asserted, that the atmospheric pollution studies are worked out by mock up experimentations and that, in some cases these experiments never can be replaced by mathematically pure models. (authors) [fr

  4. Fabrication of small mock-ups reflecting the design features of the ITER semi-prototype

    International Nuclear Information System (INIS)

    Jung, Yang-Il; Choi, Byoung-Kwon; Park, Jeong-Yong; Kim, Suk-Kwon; Lee, Dong Won; Kim, Byoung Yoon

    2012-01-01

    The ITER semi-prototype was designed to qualify the manufacturing technology for the ITER blanket first wall. However, its fabrication is expected to face great difficulty due to a design complexity. Even though joining technology for different materials such as beryllium, CuCrZr, and stainless steel (SS) was developed during the first stage of qualification, the joining is still a key issue for the fabrication of the semi-prototype. In this study, small mock-ups (SMU) were fabricated to realize and verify the manufacturing of the semi-prototype reflecting the described design features. The joining of multiple beryllium tiles on the angled CuCrZr surface was confirmed with SMU no. 1. Six beryllium tiles were joined using hot isostatic pressing (HIP), and slitting was then performed to form multiple tiles. In SMU no. 2, HIP was performed two times in order to facilitate the cooling channels at the CuCrZr/SS interface, and to join the beryllium tiles on CuCrZr/SS. The method used to form a pressure boundary for the complex cooling channels was also developed by fabricating the SMU no. 3. The SMUs confirmed the applicability of the HIP for the manufacturing of the semi-prototype.

  5. SCC behavior of alloy 690 from a CDRM mock-up

    International Nuclear Information System (INIS)

    Lapena, J.; Sol Garcia-Redondo, M. del; Perosanz, F.J.; Saez, A.; Gomez-Briceno, D.; Castelao, C.

    2015-01-01

    Stress corrosion cracking (SCC) response of Alloy 690 when the material has been subjected to nonuniform cold working is of interest to understand the behavior of the weld heat affected zone (HAZ) of Alloy 690 in which localised plastic strain exists due to weld shrinkage. This has a special interest in the case of control-rod-drive mechanisms (CRDM) of vessel head. To simulate these conditions during last years many crack growth rate (CGR) data were obtained in deformed material by cold work (rolling, forging or tensile straining), up to 40% of cold working. However, it is unclear to what extent this simulation procedure reproduces the conditions of the material in a CRDM. A research project is being carried out in order to obtain CGR data in realistic situations existing in operating power plants, by the use of CT specimens extracted from CRDMs. This presentation shows the characterization and some results of crack growth rate data on Alloy 690 TT base metal/HAZ/weld metal using specimens made from a CRDM mock-up. It has been fabricated following the usual procedures used for the RPV head fabrication for the Spanish PWR NPP. (authors)

  6. Dismantling Experiment of Mock-up Tube Bundle of Steam Generator

    International Nuclear Information System (INIS)

    Kim, Sung Kyun; Lee, Kune Woo

    2010-01-01

    A SG (steam generator) is one of the biggest decommissioning components in nuclear power plants and one has been replaced 2∼6 times during the whole operation of a nuclear power plant. The old SG should be decommissioned for the purpose of the volume reduction of radioactive waste. Among the components of SG, the tube bundle is one of the most difficult items to be dismantled due to the fact that it is very hard to cut since it is made of Inconel 600 which has high resistance of corrosion and abrasion. Moreover, All cutting process should be performed by remotely since radioactive contamination of the internal surface of SG tubes is very high (about 150,000∼300,000 Bq/cm 2 ). Therefore, it is necessary to choose the appropriate cutting methods by the pros and cons analysis for candidate dismantling technologies and to do experiment study for the validation. In this study, the results of cutting experiment for a mock-up bundle by using band saw cutting method are described herein

  7. Determination of power density in VVER-1000 Mock-Up in LR-0 reactor

    Directory of Open Access Journals (Sweden)

    Košál Michal

    2017-01-01

    Full Text Available The pin power density is an important quantity which has to be monitored during the reactor operation, for two main reasons. Firstly, it is part of the limits and conditions of safe operation and, secondly, it is source term in neutron transport calculations used for the adequate assessing of the state of core structures and pressure vessel material. It is often calculated using deterministic codes which may have problems with an adequate definition of boundary conditions in subcritical regions. This may lead to overestimation of real situation, and therefore the validation of the utility codes contributes not only to better fuel utilization, but also to more precise description of radiation situation in structural components of core. Current paper presents methods developed at LR-0 reactor, as well as selected results for pin power density measurement in peripheral regions of VVER-1000 mock-up. The presented data show that the results of a utility diffusion code at core boundary overestimate the measurement. This situation, however satisfactory safe, may lead to unduly conservative approach in the determination of radiation damage of core structures.

  8. Mock-up development of new warship protective armor structure and feasibility analysis of ship installation

    Directory of Open Access Journals (Sweden)

    ZHENG Pan

    2017-05-01

    Full Text Available To ensure the installation of the new design of protective armor structure on larger warships,a study into the installation process of the structure of this armor is carried out to improve installation efficiency and ensure the protective effect. This paper proposes a typical composite armor structure design which is composed of ‘silicate aerogel/ballistic ceramic/high-strength polyethylene/silicate aerogel’. The study analyzes the modeling design,down-selection of materials and equipment,and real ship mock-up technical development. The reliability and application of high strength polyethylene in response to high temperatures in the real ship installation process is discussed. The results show that high-temperatures during welding have no negative impact on the high strength polyethylene of the armored structure. The design demonstrates that this installation process is feasible and can be provided as an alternative solution by virtues of its good maneuverability,controllable precision,checkable quality and high reliability.

  9. Using an integrative mock-up simulation approach for evidence-based evaluation of operating room design prototypes.

    Science.gov (United States)

    Bayramzadeh, Sara; Joseph, Anjali; Allison, David; Shultz, Jonas; Abernathy, James

    2018-07-01

    This paper describes the process and tools developed as part of a multidisciplinary collaborative simulation-based approach for iterative design and evaluation of operating room (OR) prototypes. Full-scale physical mock-ups of healthcare spaces offer an opportunity to actively communicate with and to engage multidisciplinary stakeholders in the design process. While mock-ups are increasingly being used in healthcare facility design projects, they are rarely evaluated in a manner to support active user feedback and engagement. Researchers and architecture students worked closely with clinicians and architects to develop OR design prototypes and engaged clinical end-users in simulated scenarios. An evaluation toolkit was developed to compare design prototypes. The mock-up evaluation helped the team make key decisions about room size, location of OR table, intra-room zoning, and doors location. Structured simulation based mock-up evaluations conducted in the design process can help stakeholders visualize their future workspace and provide active feedback. Copyright © 2018 Elsevier Ltd. All rights reserved.

  10. Structural analysis, design and evaluation of mock-up platform, monorail, and tank plate cut-out

    International Nuclear Information System (INIS)

    Hundal, T.S.

    1995-01-01

    Platform - Structural analyses were performed for design seismic, live and dead load combinations for the freestanding platform over the partial DST mock-up section. The platform is to be used for Robotic ultrasonic inspection of the tank wall. It is a free standing structure anchored to floor slab with Hilti Kwik bolts

  11. Neutronics experiments for uncertainty assessment of tritium breeding in HCPB and HCLL blanket mock-ups irradiated with 14 MeV neutrons

    International Nuclear Information System (INIS)

    Batistoni, P.; Angelone, M.; Pillon, M.; Villari, R.; Fischer, U.; Klix, A.; Leichtle, D.; Kodeli, I.; Pohorecki, W.

    2012-01-01

    Two neutronics experiments have been carried out at 14 MeV neutron sources on mock-ups of the helium cooled pebble bed (HCBP) and the helium cooled lithium lead (HCLL) variants of ITER test blanket modules (TBMs). These experiments have provided an experimental validation of the calculations of the tritium production rate (TPR) in the two blanket concepts and an assessment of the uncertainties due to the uncertainties on nuclear data. This paper provides a brief summary of the HCPB experiment and then focuses in particular on the final results of the HCLL experiment. The TPR has been measured in the HCLL mock-up irradiated for long times at the Frascati 14 MeV Neutron Generator (FNG). Redundant and well-assessed experimental techniques have been used to measure the TPR by different teams for inter-comparison. Measurements of the neutron and gamma-ray spectra have also been performed. The analysis of the experiment, carried out by the MCNP code with FENDL-2.1 and JEFF-3.1.1 nuclear data libraries, and also including sensitivity/uncertainty analysis, shows good agreement between measurements and calculations, within the total uncertainty of 5.9% at 1σ level. (paper)

  12. Chemically deposited tungsten fibre-reinforced tungsten – The way to a mock-up for divertor applications

    Directory of Open Access Journals (Sweden)

    J. Riesch

    2016-12-01

    Full Text Available The development of advanced materials is essential for sophisticated energy systems like a future fusion reactor. Tungsten fibre-reinforced tungsten composites (Wf/W utilize extrinsic toughening mechanisms and therefore overcome the intrinsic brittleness of tungsten at low temperature and its sensitivity to operational embrittlement. This material has been successfully produced and tested during the last years and the focus is now put on the technological realisation for the use in plasma facing components of fusion devices. In this contribution, we present a way to utilize Wf/W composites for divertor applications by a fabrication route based on the chemical vapour deposition (CVD of tungsten. Mock-ups based on the ITER typical design can be realized by the implementation of Wf/W tiles. A concept based on a layered deposition approach allows the production of such tiles in the required geometry. One fibre layer after the other is positioned and ingrown into the W-matrix until the final sample size is reached. Charpy impact tests on these samples showed an increased fracture energy mainly due to the ductile deformation of the tungsten fibres. The use of Wf/W could broaden the operation temperature window of tungsten significantly and mitigate problems of deep cracking occurring typically in cyclic high heat flux loading. Textile techniques are utilized to optimise the tungsten wire positioning and process speed of preform production. A new device dedicated to the chemical deposition of W enhances significantly, the available machine time for processing and optimisation. Modelling shows that good deposition results are achievable by the use of a convectional flow and a directed temperature profile in an infiltration process.

  13. CFD prediction of mixing in a steam generator mock-up: Comparison between full geometry and porous medium approaches

    International Nuclear Information System (INIS)

    Dehbi, A.; Badreddine, H.

    2013-01-01

    Highlights: • CFD is used to simulate single phase mixing in a model steam generator. • Motive of the work is to compare porous media approach with full geometry representation of tubes. • Porous media approach is found to compare favorably with full representation in steady states. - Abstract: In CFD simulations of single phase flow mixing in a steam generator (SG) during a station blackout severe accident, one is faced with the problem of representing the thousands of SG U-tubes. Typically simplifications are made to render the problem computationally tractable. In particular, one or a number of tubes are lumped in one volume that is treated as a single porous medium which replicates the pressure loss and heat transfer characteristics of the real tube. This approach significantly reduces the computational size of the problem and hence simulation time. In this work, we endeavor to investigate the adequacy of this approach by performing a series of simulations. We first validate the porous medium approach against results of the 1/7th scale Westinghouse SG-S3 test. In a second step, we make two separate simulations of flow in the PSI SG mock-up, i.e. one in which the porous medium model is used for the tube bundle, and another in which the full geometry is represented. In all simulations, the Reynolds Stress (RSM) model of turbulence is used. We show that in steady state conditions, the porous medium treatment yields results which are comparable to those of the full geometry representation (temperature distribution, recirculation ratio, hot plume spread, etc.). Hence, the porous medium approach can be extended with a good degree of confidence to model single phase mixing in the full scale SG

  14. Seismic analysis methods for LMFBR core and verification with mock-up vibration tests

    International Nuclear Information System (INIS)

    Sasaki, Y.; Kobayashi, T.; Fujimoto, S.

    1988-01-01

    This paper deals with the vibration behaviors of a cluster of core elements with the hexagonal cross section in a barrel under the dynamic excitation due to seismic events. When a strong earthquake excitation is applied to the core support, the cluster of core elements displace to a geometrical limit determined by restraint rings in the barrel, and collisions could occur between adjacent elements as a result of their relative motion. For these reasons, seismic analysis on LMFBR core elements is a complicated non-linear vibration problem, which includes collisions and fluid interactions. In an actual core design, it is hard to include hundreds of elements in the numerical calculations. In order to study the seismic behaviors of core elements, experiments with single row 29 elements (17 core fuel assemblies, 4 radial blanket assemblies, and 8 neutron shield assemblies) simulated all elements in MONJU core central row, and experiments with 7 cluster rows of 37 core fuel assemblies in the core center were performed in a fluid filled tank, using a large-sized shaking table. Moreover, the numerical analyses of these experiments were performed for the validation of simplified and detailed analytical methods. 4 refs, 18 figs

  15. Ultrasonic testing results of fatigue cracks in PWR mock-up

    International Nuclear Information System (INIS)

    Gondard, C.

    1990-01-01

    The Ispra Joint Research Center has entered, since many years a study on fatigue crack propagation in PWR reactor vessels. The objective of this study is to establish a relation between the size and the location of defects and the lifetime of the vessel. For verifying the theoretical models validity a mockup has been built. This document gives the results of CEA for 6 in service inspection during 5 years [fr

  16. Ion Beam Analysis methods applied to the examination of Be//Cu joints in hipped Be tiles for ITER first wall mock- ups

    International Nuclear Information System (INIS)

    Vito, E. de; Cayron, C.; Hicham Khodja; Lorenzetto, P.

    2006-01-01

    complementary techniques like SEM and EBSD. Next steps to this work will be to apply the analysis procedure described in this study on the soon coming tested mock-ups, in order to characterize the impact of the different tests (high heat flux tests and thermal fatigue tests) on the evolution and performance of the Be/Cu junctions. (author)

  17. Residual stress measurement inside a dissimilar metal weld mock-up of the pressurizer safety and relief nozzle

    Energy Technology Data Exchange (ETDEWEB)

    Campos, Wagner R.C.; Rabello, Emerson G.; Silva, Luiz L.; Mansur, Tanius R., E-mail: wrcc@cdtn.br, E-mail: egr@cdtn.br, E-mail: silvall@cdtn.br, E-mail: tanius@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte (Brazil). Servico de Integridade Estrutural; Martins, Ketsia S., E-mail: ketshinoda@hotmail.com [Universidade Federal de Minas Gerais (UFMG), Nelo Horizonte (Brazil). Departamento de Engenharia Metalurgica

    2015-07-01

    Residual stresses are present in materials or structural component in the absence of external loads or changes in temperatures. The most common causes of residual stresses being present are the manufacturing or assembling processes. All manufacturing processes, such as casting, welding, machining, molding, heat treatment, among others, introduces residual stresses into the manufactured object. The residual stresses effects could be beneficial or detrimental, depending on its distribution related to the component or structure, its load service and if it is compressive or tensile. In this work, the residual strains and stresses inside a mock-up that simulates the safety and relief nozzle of Angra 1 Nuclear Power Plant pressurizer were studied. The current paper presents a blind hole-drilling method residual stress measurements both at the inner surface of dissimilar metal welds of dissimilar metal weld nozzle mock-up. (author)

  18. Residual stress measurement inside a dissimilar metal weld mock-up of the pressurizer safety and relief nozzle

    International Nuclear Information System (INIS)

    Campos, Wagner R.C.; Rabello, Emerson G.; Silva, Luiz L.; Mansur, Tanius R.; Martins, Ketsia S.

    2015-01-01

    Residual stresses are present in materials or structural component in the absence of external loads or changes in temperatures. The most common causes of residual stresses being present are the manufacturing or assembling processes. All manufacturing processes, such as casting, welding, machining, molding, heat treatment, among others, introduces residual stresses into the manufactured object. The residual stresses effects could be beneficial or detrimental, depending on its distribution related to the component or structure, its load service and if it is compressive or tensile. In this work, the residual strains and stresses inside a mock-up that simulates the safety and relief nozzle of Angra 1 Nuclear Power Plant pressurizer were studied. The current paper presents a blind hole-drilling method residual stress measurements both at the inner surface of dissimilar metal welds of dissimilar metal weld nozzle mock-up. (author)

  19. Mock-up experiment at Birmingham University for BNCT project of Osaka University – Neutron flux measurement with gold foil

    International Nuclear Information System (INIS)

    Tamaki, S.; Sakai, M.; Yoshihashi, S.; Manabe, M.; Zushi, N.; Murata, I.; Hoashi, E.; Kato, I.; Kuri, S.; Oshiro, S.; Nagasaki, M.; Horiike, H.

    2015-01-01

    Mock-up experiment for development of accelerator based neutron source for Osaka University BNCT project was carried out at Birmingham University, UK. In this paper, spatial distribution of neutron flux intensity was evaluated by foil activation method. Validity of the design code system was confirmed by comparing measured gold foil activities with calculations. As a result, it was found that the epi-thermal neutron beam was well collimated by our neutron moderator assembly. Also, the design accuracy was evaluated to have less than 20% error. - Highlights: • Accelerator based neutron source for BNCT is being developed in Osaka University. • Mock-up experiment was carried out at Birmingham University, UK. • Neutronics performance of our assembly was evaluated from gold foil activation. • Gold foil activation was determined by using HPGe detectors. • Validity of the neutronics design code system was confirmed.

  20. Original Research. Statistical Study Regarding Differences Between the Wax-Up, Mock-Up, and Final Restoration

    Directory of Open Access Journals (Sweden)

    Jánosi Kinga

    2017-03-01

    Full Text Available The aesthetic rehabilitation of patients remains a challenge for practicians. To facilitate the clinicians’ and technicians’ task, several innovative methods were developed, like the diagnostic wax-up and mock-up. The width-to-length ratio of the maxillary frontal teeth can be used to evaluate dentofacial aesthetics. Our study presents the variations between the teeth size measured on casts obtained during the prosthodontic treatment.

  1. Mock-up facilities for the development of an advanced spent fuel management process using molten salt technology

    International Nuclear Information System (INIS)

    Young-Joon Shin; Ik-Soo Kim; Seung-Chul Oh; Soo-Haeng Cho; Yo-Taik Song; Hyun-Soo Park

    2000-01-01

    The Korea Atomic Energy Research Institute (KAERI) has investigated a new approach to spent fuel storage technology that would reduce the total storage volume and the amount of decay heat. The technology utilizes the reduction of oxide fuel to a metal to reduce the volume and preferentially removing the fission products to reduce the decay heat. The uranium oxide is reduced to uranium metal by Li metal in a molten LiCl salt bath. During the reduction process, fission products are dissolved into the LiCl bath and some of the highly radioactive elements, such as Sr and Cs, are preferentially removed from the bath. The reduced uranium metal is cast into an ingot, put into a storage capsule, and stored using conventional storage methods. The fission products are treated as high level radioactive wastes. Each process of the technology has been studied and analyzed for technical feasibility, and has come to the point for designing and constructing of the mock-up for a demonstration of the technology. This paper presents the detailed design of the mock-up of the system and operational characteristics, along with all the details of the equipment for the system. KAERI plans to use the mock-up for the demonstration using an in-active spent fuel specimen. (authors)

  2. FEBEX bentonite colloid stability in ground water

    Energy Technology Data Exchange (ETDEWEB)

    Seher, H.; Schaefer, T.; Geckeis, H. [Inst. fuer Nukleare Entsorgung (INE), Forschungszentrum Karlsruhe, 76021 Karlsruhe (Germany)]. e-mail: holger.seher@ine.fzk .de; Fanghaenel, T. [Ruprecht-Karls-Univ. Heidelberg, Physikalisch-Chemisches In st., D-69120 Heidelberg (Germany)

    2007-06-15

    Coagulation experiments are accomplished to identify the geochemical conditions for the stability of Febex bentonite colloids in granite ground water. The experiments are carried out by varying pH, ionic strength and type of electrolyte. The dynamic light scattering technique (photon correlation spectroscopy) is used to measure the size evolution of the colloids with time. Agglomeration rates are higher in MgCl{sub 2} and CaCl{sub 2} than in NaCl solution. Relative agglomeration rates follow approximately the Schulze-Hardy rule. Increasing agglomeration rates at pH>8 are observed in experiments with MgCl{sub 2} and CaCl{sub 2} which are, however, caused by coprecipitation phenomena. Bentonite colloid stability fields derived from the colloid agglomeration experiments predict low colloid stabilization in granite ground water taken from Aespoe, Sweden, and relatively high colloid stability in Grimsel ground water (Switzerland)

  3. FEBEX II Project Post-mortem analysis EDZ assessment

    International Nuclear Information System (INIS)

    Bazargan Sabet, B.; Shao, H.; Autio, J.; Elorza, F. J.

    2004-01-01

    Within the framework of the FEBEX II project a multidisciplinary team studied the mechanisms of creation of the potential damaged zone around the test drift. The research program includes laboratory and in situ investigations as well as the numerical modelling of the observed phenomena. Where laboratory investigations are concerned, the 14C-PMMA technique was applied to study the spatial distribution of porosity in the samples taken from the test drift wall. In addition complementary microscopy and scanning electron microscopy (SEM) studies were performed to make qualitative investigations on the pore apertures and minerals in porous regions. The results obtained with the PMMA method have not shown any clear increased porosity zone adjacent to the tunnel wall. The total porosity of the samples varied between 0.6-1.2%. The samples of unplugged region did not differ from the samples of plugged region. A clear increase in porosity to depths of 10-15 mm from the tunnel wall was detected in lamprophyre samples. According to the SEM/EDX analyses the excavation-disturbed zone in the granite matrix extended to depths of 1-3 mm from the wall surface. A few quartz grains were crushed and some micro fractures were found. Gas permeability tests were carried out on two hollow cylinder samples of about 1m long each taken on the granite wall perpendicular to the drift axis. The first sample was cored in the service area far from the heated zone and the second one at the level of the heater. The tests were performed at constant gas pressure by setting a steady state radial flow through a section of 1cm wide isolated by means of four mini-packers. The profile of the gas permeability according to the core length has been established. The results obtained for both considered samples have shown permeability ranging between 3.5 10-18 and 8.4 10-19m2, pointing out the absence of a marked damage. Acoustic investigations have been carried out with the objective of quantifying the

  4. FEBEX II Project Post-mortem analysis EDZ assessment

    Energy Technology Data Exchange (ETDEWEB)

    Bazargan Sabet, B.; Shao, H.; Autio, J.; Elorza, F. J.

    2004-07-01

    Within the framework of the FEBEX II project a multidisciplinary team studied the mechanisms of creation of the potential damaged zone around the test drift. The research program includes laboratory and in situ investigations as well as the numerical modelling of the observed phenomena. Where laboratory investigations are concerned, the 14C-PMMA technique was applied to study the spatial distribution of porosity in the samples taken from the test drift wall. In addition complementary microscopy and scanning electron microscopy (SEM) studies were performed to make qualitative investigations on the pore apertures and minerals in porous regions. The results obtained with the PMMA method have not shown any clear increased porosity zone adjacent to the tunnel wall. The total porosity of the samples varied between 0.6-1.2%. The samples of unplugged region did not differ from the samples of plugged region. A clear increase in porosity to depths of 10-15 mm from the tunnel wall was detected in lamprophyre samples. According to the SEM/EDX analyses the excavation-disturbed zone in the granite matrix extended to depths of 1-3 mm from the wall surface. A few quartz grains were crushed and some micro fractures were found. Gas permeability tests were carried out on two hollow cylinder samples of about 1m long each taken on the granite wall perpendicular to the drift axis. The first sample was cored in the service area far from the heated zone and the second one at the level of the heater. The tests were performed at constant gas pressure by setting a steady state radial flow through a section of 1cm wide isolated by means of four mini-packers. The profile of the gas permeability according to the core length has been established. The results obtained for both considered samples have shown permeability ranging between 3.5 10-18 and 8.4 10-19m2, pointing out the absence of a marked damage. Acoustic investigations have been carried out with the objective of quantifying the

  5. Weld defects analysis of 60 mm thick SS316L mock-ups of TIG and EB welds by ultrasonic inspection for fusion reactor vacuum vessel applications

    International Nuclear Information System (INIS)

    Buddu, Ramesh Kumar; Shaikh, Shamsuddin; Raole, P.M.; Sarkar, B.

    2015-01-01

    The present paper reports the weld quality inspections carried with 60 mm thick AISI welds of SS316L. The high thickness steel plates requirement is due to the specific applications in case of advanced fusion reactor structural components like vacuum vessel and others. Different kind welds are proposed for the thick plate joints like Tungsten Inert Gas (TIG) welding, Electron beam welding as per stringent conditions (like very low distortions and residual stresses) for the vacuum vessel fabrication. Mock-ups of laboratory scale welds are fabricated by TIG (multi-pass) and EB (double pass) process techniques and different weld quality inspections are carried by different NDT tests. The welds are examined with Liquid penetrant examination to check sub surface cracks/discontinuities towards the defects observation

  6. Qualification and post-mortem characterization of tungsten mock-ups exposed to cyclic high heat flux loading

    Energy Technology Data Exchange (ETDEWEB)

    Pintsuk, G., E-mail: g.pintsuk@fz-juelich.de [Forschungszentrum Jülich GmbH, Euratom Association, D-52425 Jülich (Germany); Bobin-Vastra, I.; Constans, S. [AREVA NP PTCMI-F, Centre Technique, Fusion, F-71200 Le Creusot (France); Gavila, P. [Fusion for Energy, E-08019 Barcelona (Spain); Rödig, M. [Forschungszentrum Jülich GmbH, Euratom Association, D-52425 Jülich (Germany); Riccardi, B. [Fusion for Energy, E-08019 Barcelona (Spain)

    2013-10-15

    Highlights: • We characterize tungsten mono-block components after exposure to ITER relevant heat loads. • We qualify the manufacturing technology, i.e., hot isostatic pressing and hot radial pressing, and repair technologies. • We determine the microstructural influences, i.e., rod vs. plate material, on the damage evolution. • Needle like microstructures increase the risk of deep crack formation due to a limited fracture strength. -- Abstract: In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, high heat flux tests were performed in the electron beam facility FE200, Le Creusot, France. Thereby, in total eight small-scale and three medium-scale monoblock mock-ups produced with different manufacturing technologies and different tungsten grades were exposed to cyclic steady state heat loads. The applied power density ranges from 10 to 20 MW/m{sup 2} with a maximum of 1000 cycles at each particular loading step. Finally, on a reduced number of tiles, critical heat flux tests in the range of 30 MW/m{sup 2} were performed. Besides macroscopic and microscopic images of the loaded surface areas, detailed metallographic analyses were performed in order to characterize the occurring damages, i.e., crack formation, recrystallization, and melting. Thereby, the different joining technologies, i.e., hot radial pressing (HRP) vs. hot isostatic pressing (HIP) of tungsten to the Cu-based cooling tube, were qualified showing a higher stability and reproducibility of the HIP technology also as repair technology. Finally, the material response at the loaded top surface was found to be depending on the material grade, microstructural orientation, and recrystallization state of the material. These damages might be triggered by the application of thermal shock loads during electron beam surface scanning and not by the steady state heat load only. However, the superposition of thermal fatigue loads and thermal shocks as also expected

  7. FEBEX Project Post-mortem Analysis: Corrosion Study

    International Nuclear Information System (INIS)

    Madina, V.; Azkarate, I.

    2004-01-01

    The partial dismantling of the FEBEX in situ test was carried out during de summer of 2002, following 5 years of continuous heating. The operation included the demolition of the concrete plug and the removal of the section of the test corresponding to the first heater. A large number of samples from all types of materials have been taken during the dismantling for subsequent analysis. Part of the samples collected were devoted to the analysis of the corrosion processes occurred during the first operational phase of the test. These samples comprised corrosion coupons from different metals installed for that purpose, sensors retrieved during the dismantling that were found severely corroded and bentonite in contact with those sensors. In addition, a corrosion study was performed on the heater extracted and on one section of liner surrounding it. All the analyses were carried out by the Fundacion INASMET (Spain). This report describes, in detail the studies carried out the different materials and the obtained results, as well as the drawn conclusions. (Author)

  8. FEBEX Project Post-mortem Analysis: Corrosion Study

    Energy Technology Data Exchange (ETDEWEB)

    Madina, V.; Azkarate, I.

    2004-07-01

    The partial dismantling of the FEBEX in situ test was carried out during de summer of 2002, following 5 years of continuous heating. The operation included the demolition of the concrete plug and the removal of the section of the test corresponding to the first heater. A large number of samples from all types of materials have been taken during the dismantling for subsequent analysis. Part of the samples collected were devoted to the analysis of the corrosion processes occurred during the first operational phase of the test. These samples comprised corrosion coupons from different metals installed for that purpose, sensors retrieved during the dismantling that were found severely corroded and bentonite in contact with those sensors. In addition, a corrosion study was performed on the heater extracted and on one section of liner surrounding it. All the analyses were carried out by the Fundacion INASMET (Spain). This report describes, in detail the studies carried out the different materials and the obtained results, as well as the drawn conclusions. (Author)

  9. The Mock-up of the "Ratto Delle Sabine" by Giambologna: Making and Utilization of a 3D Model

    Directory of Open Access Journals (Sweden)

    Grazia Tucci

    2015-02-01

    Full Text Available Within a project for the knowledge and preservation of the mock-up of Giambologna's Ratto delle Sabine housed in the Galleria dell'Accademia in Florence, the GeCO laboratory has made laser scanner acquisitions to create surface models at different resolutions for structural analysis, on which to check the coverage of the photographic campaign and to create a three-dimensional thematic mapping of data relating to investigations and restoration works. The PDF3D file format has been used to easily manage data on a platform immediately available to all operators.

  10. Fabrication data package for HEDL dosimetry in the ORNL Poolside Facility: LWR Pressure Vessel Mock-up irradiation

    International Nuclear Information System (INIS)

    Lippincott, E.P.; McElroy, W.N.; Kellogg, L.S.; Gold, R.; Guthrie, G.L.; Ruddy, F.H.; Ulseth, J.A.

    1981-09-01

    This document provides a complete description of the HEDL dosimetry inserted in the metallurgical specimen irradiation in the LWR Pressure Vessel Mock-up at the Oak Ridge Reactor Poolside Facility (PSF). This experiment is being conducted under the Nuclear Regulatory Commission sponsored program on Surveillance Dosimetry Improvement. The irradiation started April 1980 with recovery of the 2 x 10 19 (nominal fluence with E > 1 MeV) capsule in September 1980, the 4 x 10 19 surveillance capsule in November 1981 and the pressure vessel and void box capaules about August 1982

  11. Fabrication of mock-up with Be armour tiles diffusion bonded to the CuCrZr heat sink

    International Nuclear Information System (INIS)

    Moreschi, L.F.; Pizzuto, A.; Alessandrini, I.; Agostini, M.; Visca, E.; Merola, M.

    2001-01-01

    The aim of this work is the manufacture of high heat flux mock-ups with Be armour tiles on a CuCrZr heat sink for fabricating the beryllium section of the divertor vertical target (DVT) in the ITER reactor. Diffusion bonding between the CuCrZr bar and the beryllium tiles was obtained by inserting an aluminium interlayer to accommodate surface irregularities as well as to provide a compliant layer for accommodating thermal mismatches during both manufacturing and operation and cycles

  12. Sorption-desorption of samarium in Febex bentonite

    International Nuclear Information System (INIS)

    Ramirez-Guinart, O.; Rigol, A.; Vidal, M.; Fernandez-Poyatos, P.; Alba, M. D.

    2012-01-01

    Document available in extended abstract form only. The chemical and physical nature of the clay is a key issue in the design of engineered barriers. The FEBEX bentonite is one of the clays candidates to be used in engineered barriers in deep geology repositories (DGR). Here, its performance was tested with respect to the sorption-desorption of samarium, which is a lanthanide that, besides being considered as a natural analogue of actinides, may also be present in high level radioactive waste in the form of the radioactive isotope 151 Sm. FEBEX bentonite was used in this study. This is a di-octahedral smectite, with isomorphic substitutions in tetrahedral and octahedral sheets. Its theoretical cation exchange capacity value is 1500 meq kg -1 . Sorption isotherms were obtained for Sm in the range of initial concentrations of 0.01 and 9 meq l -1 . Tests were carried out in deionized water and in a medium simulating the composition of interstitial water. Sorption tests were performed equilibrating 30 ml of the Sm solution with 0.2 g of clay. After a contact time of 24 hours, supernatants were decanted off after centrifugation. The quantification of the concentration of Sm in the initial and final solutions allowed us to quantify the Sm equilibrium concentration (C eq ), the fraction sorbed in the FEBEX bentonite (C sorb ) and to derive the sorption K d data. Desorption tests were applied to determine the desorption K d and the percentage of Sm reversibly sorbed. Desorption tests were performed with the bentonite residue from the sorption step, under the same experimental conditions, but without Sm. Powder X-ray diffractograms were obtained from 3 to 70 deg. 2θ with a step of 0.05 deg. and a counting time of 3 s. The crystalline phases were identified using the computer program X'Pert HighScore. The morphology of the samples was analyzed by SEM at 20 kV. An EDX system was fitted to the SEM equipment to perform chemical analyses of the samples using a Si/Li detector

  13. Effective use of plant simulators and mock-up facilities for cultivation and training of younger regulators

    International Nuclear Information System (INIS)

    Tsuruga, Keisuke

    2010-01-01

    In order to achieve effective safety regulation, the staff members of a regulatory body who are engaged in regulatory work are requested to be well familiar with the characteristics, operations and maintenances of nuclear power plants at a practical level as far as possible. Although the regulators are not always required to have the same level of skills as those of plant designers or operators, the skills of the regulatory staff are essential elements to achieve high quality of the national nuclear safety regulation. Especially understanding of fundamentals such as operations, transient behaviors, trouble responses and plant inspections is indispensable not only to practical regulatory work but also to the establishment of the trust and confidence in safety regulation. To acquire these skills, the use of facilities such as plant simulators and inspection mock-up facilities is very effective to back up classroom lectures on theories and procedures. Practical training using these facilities under the guidance of well-experienced instructors inspires motivations and enhances capabilities of younger regulators. To support the countries newly embarking on nuclear power programs, JNES will continue to cooperate with those countries in cultivating and training younger regulators, by focusing on the training by veteran instructors using full-scale plant simulators and inspection mock-up facilities to give the trainees more practical skills and knowledge difficult to obtain through classroom lectures or textbooks. (author)

  14. Characterization by laser velocity of the flow in a ramjet chamber mock-up; Caracterisation par velocimetrie laser de l'ecoulement dans un foyer maquette de statoreacteur

    Energy Technology Data Exchange (ETDEWEB)

    Brossard, C.; Gicquel, P.; Barat, M.; Ristori, A.

    2002-07-01

    A three dimensional mock-up has been realized at the ONERA, to study the combustion in conditions of pressure, speed and temperatures similar as real temperatures in ramjet. The first step of the tests program allows to study the cold and non reactive flows, by an hydraulic simulation. In the second step, the hot reactive and non reactive flows have been studied in more realistic tests. This paper presents the results obtained in non reactive flow and on high speed reactive flow. (A.L.B.)

  15. Damage and fatigue crack growth of Eurofer steel first wall mock-up under cyclic heat flux loads. Part 2: Finite element analysis of damage evolution

    International Nuclear Information System (INIS)

    You, Jeong-Ha

    2014-01-01

    Highlights: • The surface heat flux load of 3.5 MW/m 2 produced substantial stresses and inelastic strains in the heat-loaded surface region, especially at the notch root. • The notch root exhibited a typical notch effect such as stress concentration and localized inelastic yield leading to a preferred damage development. • The predicted damage evolution feature agrees well with the experimental observation. • The smooth surface also experiences considerable stresses and inelastic strains. However, the stress intensity and the amount of inelastic deformation are not high enough to cause any serious damage. • The level of maximum inelastic strain is higher at the notch root than at the smooth surface. On the other hand, the amplitude of inelastic strain variation is comparable at both positions. • The amount of inelastic deformation is significantly affected by the length of pulse duration time indicating the important role of creep. - Abstract: In the preceding companion article (part 1), the experimental results of the high-heat-flux (3.5 MW/m 2 ) fatigue tests of a Eurofer bare steel first wall mock-up was presented. The aim was to investigate the damage evolution and crack initiation feature. The mock-up used there was a simplified model having only basic and generic structural feature of an actively cooled steel FW component for DEMO reactor. In that study, it was found that microscopic damage was formed at the notch root already in the early stage of the fatigue loading. On the contrary, the heat-loaded smooth surface exhibited no damage up to 800 load cycles. In this paper, the high-heat-flux fatigue behavior is investigated with a finite element analysis to provide a theoretical interpretation. The thermal fatigue test was simulated using the coupled damage-viscoplastic constitutive model developed by Aktaa. The stresses, inelastic deformation and damage evolution at the notch groove and at the smooth surface are compared. The different damage

  16. Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor

    Directory of Open Access Journals (Sweden)

    Košťál Michal

    2016-01-01

    Full Text Available A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1–10 MeV and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1. Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.

  17. Study of pressure drop in a mock-up of fuel element cluster

    International Nuclear Information System (INIS)

    Barros Filho, J.A.

    1987-01-01

    Results of single-phase tests performed in a 3 x 3 rod bundle arranged in square array are presented and analysed. The tests were performed in adiabatic conditions and with heat transfer, covering the following range of parameters: Reynolds no.: 1,5 to 20 x 10 4 ; inlet temperature [ 0 C]: 30 to 150; pressure [bar]: 1 to 15; heat flux (kW/cm 2 ]: 0 to 1000. Correlations were determined for the friction factor, isothermal and under conditions of heat transfer, spacer grids pressure drop coefficient and average heat transfer coefficient. The experimental data were compared with published data obtained by other researchers and with some theoretical models selected in the literature. (Author) [pt

  18. Thermal and mechanical behaviour of an experimental mock-up of a nuclear containment

    International Nuclear Information System (INIS)

    Chauvel, D.; Barre, F.

    2007-01-01

    In order to better understand the behaviour of a reactor containment submitted to combined pressure and temperature loads by means of studies of the concrete permeability and the state of cracking evolution, EDF and its French partners have built a prestressed concrete test model which represents a PWR containment typical section. The monitoring system was designed to follow the evolution of strains, temperature and state of cracking of the concrete wall from construction stage to air and steam tests. The measurements results as well as their comparison with theoretical laws or data and calculated values, allow to determine the main thermal and mechanical characteristics of the concrete, to analyse the thermo-mechanical behaviour of the structure and also to check the design criteria of prestressed concrete containments. (authors)

  19. Diffusion of strongly sorbing cations (60Co and 152Eu) in compacted FEBEX bentonite

    International Nuclear Information System (INIS)

    Garcia-Gutierrez, M.; Cormenzana, J. L.; Missana, T.; Alonso, U.; Mingarro, M.

    2011-01-01

    Diffusion experiments in compacted FEBEX bentonite were performed with strongly sorbing radionuclides, 60 Co and 152 Eu. Diffusion experiments with these radionuclides present several difficulties: first of all these tests are very time consuming because of the high sorption on the clays, secondly these elements not only present high sorption onto clays but also on diffusion cells, tubing, filters and reservoirs, typically used in the classical through-diffusion or in-diffusion methods, which makes difficult the interpretation of the results. In this study, the experiments were performed using the instantaneous planar source method, where a paper filter tagged with a tracer is placed between two tablets of compacted bentonite. The apparent diffusion coefficient (D a ) is obtained analysing the tracer concentration profile in the samples at the end of the experiment, both with an analytical and a numerical approach. The ranges of D a values obtained from these experiments in the FEBEX clay compacted at 1.65 g/cm 3 are (0.5-2.3) x 10 -13 m 2 /s for Co and (0.8-2.5) x 10 -14 m 2 /s for Eu. Results showed that the analytical solution is able to fit reasonably well the Eu concentration profiles, whereas Co concentration profiles show a different behavior, not straightforward to explain, which was also analyzed by numerical methods. (authors)

  20. Uncertainty analysis and flow measurements in an experimental mock-up of a molten salt reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Bogdan; Aszodi, Attila [Budapest University of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-09-15

    In the paper measurement results from the experimental modelling of a molten salt reactor concept will be presented along with detailed uncertainty analysis of the experimental system. Non-intrusive flow measurements are carried out on the scaled and segmented mock-up of a homogeneous, single region molten salt fast reactor concept. Uncertainty assessment of the particle image velocimetry (PIV) measurement system applied with the scaled and segmented model is presented in detail. The analysis covers the error sources of the measurement system (laser, recording camera, etc.) and the specific conditions (de-warping of measurement planes) originating in the geometry of the investigated domain. Effect of sample size in the ensemble averaged PIV measurements is discussed as well. An additional two-loop-operation mode is also presented and the analysis of the measurement results confirm that without enhancement nominal and other operation conditions will lead to strong unfavourable separation in the core flow. It implies that use of internal flow distribution structures will be necessary for the optimisation of the core coolant flow. Preliminary CFD calculations are presented to help the design of a perforated plate located above the inlet region. The purpose of the perforated plate is to reduce recirculation near the cylindrical wall and enhance the uniformity of the core flow distribution.

  1. Direct tritium measurement in lithium titanate for breeding blanket mock-up experiments with D-T neutrons

    International Nuclear Information System (INIS)

    Klix, A.; Ochiai, K.; Nishitani, T.; Takahashi, A.

    2004-01-01

    At Fusion Neutronics Source (FNS) of JAERI, tritium breeding experiments with blanket mock-ups consisting of advanced fusion reactor materials are in progress. The breeding zones are thin layers of lithium titanate which is one of the candidate tritium breeder materials for the DEMO fusion power reactor. It is anticipated that the application of small pellet-shaped lithium titanate detectors manufactured from the same material as the breeding layer will reduce experimental uncertainties arising from necessary corrections due to different isotopic lithium volume concentrations in breeding material and detector. Therefore, a method was developed to measure the local tritium production by means of lithium titanate pellet detectors and a liquid scintillation counting technique. The lithium titanate pellets were dissolved in concentrated hydrochloric acid solution and the resulting acidic solution was neutralized. Two ways of further processing were followed: direct incorporation into a liquid scintillation cocktail and distillation of the solution followed by mixing with liquid scintillator. Two types of lithium titanate pellets were investigated with different 6 Li enrichment and manufacturing procedure. It was found that lithium titanate is suitable for tritium production measurements. However some discrepancies in the measurement accuracy remained with one of the investigated pellet detectors when compared with a well-established lithium carbonate measurement technique and this issue needs further investigation

  2. Enhancement of the use of digital mock-ups in the verification and validation process for ITER remote handling systems

    Energy Technology Data Exchange (ETDEWEB)

    Sibois, R., E-mail: romain.sibois@vtt.fi [VTT Technical Research Centre of Finland, P.O. Box 1300, 33101 Tampere (Finland); Salminen, K.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, 33101 Tampere (Finland); Mattila, J. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T. [VTT Technical Research Centre of Finland, P.O. Box 1300, 33101 Tampere (Finland)

    2013-10-15

    Highlights: • Verification and validation process for ITER remote handling system. • Verification and validation framework for complex engineering systems. • Verification and validation roadmap for digital modelling phase. • Importance of the product life-cycle management in the verification and validation framework. -- Abstract: The paper is part of the EFDA's programme of European Goal Oriented Training programme on remote handling (RH) “GOT-RH”. The programme aims to train engineers for activities supporting the ITER project and the long-term fusion programme. This paper is written based on the results of a project “verification and validation (V and V) of ITER RH system using digital mock-ups (DMUs)”. The purpose of this project is to study efficient approach of using DMU for the V and V of the ITER RH system design utilizing a system engineering (SE) framework. This paper reviews the definitions of DMU and virtual prototype and overviews the current trends of using virtual prototyping in the industry during the early design phase. Based on the survey of best industrial practices, this paper proposes ways to improve the V and V process for ITER RH system utilizing DMUs.

  3. Investigation on water content in fresco mock-ups in the microwave and near-IR spectral regions

    International Nuclear Information System (INIS)

    Magrini, Donata; Riminesi, Cristiano; Cucci, Costanza; Olmi, Roberto; Picollo, Marcello

    2017-01-01

    Water diffusion inside masonry is responsible for the majority of the decay phenomena observed in wall paintings and frescos. Thus, the diagnostics of moisture and water content and their monitoring represent a key issue. In order to preserve the integrity of surfaces of artistic interest, investigations by means of non-destructive techniques (NDT) are preferred over others. The aim of this research is to determine methodologies to quantify the moisture content (MC) of frescos by means of the integrated use of two non-invasive techniques, namely fiber optic reflectance spectroscopy (FORS) in the near-IR region and evanescent field dielectrometry (EFD) in the microwave range. The FORS technique has been employed in order to assess the amount of water adsorbed from the surface by means of an analysis of the reflectance spectra in the Vis–NIR (350-2200 nm) range. This technique investigates the electronic and vibrational transitions that are characteristic of each compound and enables their identification. The water content is evaluated on the basis of the 1920 nm and 1450 nm absorption bands. The EFD system consists of a resonant probe connected to a network analyzer. The resonance frequency of the cavity under different moisture-content conditions of frescos is in the 1.0–1.5 GHz range. The device makes it possible to compute, in real time, the MC from a measurement of the transmission coefficient (amplitude versus frequency) through the probe. Fresco mock-ups have been prepared in collaboration with the Opificio delle Pietre Dure in order to recreate most of the possible chromatic shades obtained by mixing iron oxides and hydroxide-based pigments. Measurements were performed by employing both techniques on fresco models after wet-dry cycles obtained by means of poultices with a known water content. The results obtained with these two techniques were compared, and cross relationships between the EFD and FORS data were defined. (paper)

  4. FEBEX: Full-Scale engineered barriers experiment in crystalline host-rock: preoperational phase. Synthesized report

    International Nuclear Information System (INIS)

    1997-01-01

    The FEBEX project is being cofinanced by the EC under contract F 14WCT950006. In addition to the EC, seven partners from three countries of the EU. (France, Germany, and Spain) as well as one from EFTA (Switzerland) are participating in the project. ENRESA is the coordinating partner with NAGRA assisting in coordinating some aspects. The project consists of two large-scale tests and a series of complimentary laboratory tests. The work is being executed by the following organizations: CIEMAT, AITEMIN, UP-DIT (CIMNE), ULC, CSIC-Zaidin, and UPM (SPAIN) ANDRA and G.3S (FRANCE) GRS (GERMANY). This report includes a synthesized description of the experiment from its conception through the installation of the two large-scale tests (from the middle of 1994 to the beginning of 1997, preoperation stage). The experiment is described in detail in a series of specific reports. (Author)

  5. The state of the art report on the fabrication of FW blanket for the fusion reactor and mock-up development in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Whan; Baek, Jong Hyuk; Park, Jeong Yong; Kim, Hyun Gil; Jeong, Yong Hwan

    2004-08-01

    Blanket-shield system in ITER is the component where it directly is faced with high-heat plasma. Function of blanket is to sustain extremely high temperature environment as well as to remove heat flux generated its surface. It mainly consists of plasma facing part, heat sinking part and structural part. Plasma facing part is made of armour materials such as beryllium, tungsten and carbon fiber composite. Heat sinking part is made of copper alloy to maximize heat transfer into flowing coolant inside of blanket. Structural material is used in 316LN stainless steel. As joining such dissimilar materials emerged as an issue, many developed countries have spurred the development of joint technology. This technical report was focused on the activities of EU regarding joining beryllium, copper and stainless steel. EU have adopted to Hot Isostatic Pressing (HIP) to join beryllium, copper and stainless steel. Although brazing process is not actively investigated compared as HIP, it still investigated in some countries to support HIP. Fabrication of mock-up is accomplished by CEA in France to finish small scale mock-up in 1996, medium and large scale mock-up in 1997. In recent, FRAMATOME in EU has focused on manufacturing prototype used for ITER.

  6. Tool coupling for the design and operation of building energy and control systems based on the Functional Mock-up Interface standard

    Energy Technology Data Exchange (ETDEWEB)

    Nouidui, Thierry Stephane; Wetter, Michael

    2014-03-01

    This paper describes software tools developed at the Lawrence Berkeley National Laboratory (LBNL) that can be coupled through the Functional Mock-up Interface standard in support of the design and operation of building energy and control systems. These tools have been developed to address the gaps and limitations encountered in legacy simulation tools. These tools were originally designed for the analysis of individual domains of buildings, and have been difficult to integrate with other tools for runtime data exchange. The coupling has been realized by use of the Functional Mock-up Interface for co-simulation, which standardizes an application programming interface for simulator interoperability that has been adopted in a variety of industrial domains. As a variety of coupling scenarios are possible, this paper provides users with guidance on what coupling may be best suited for their application. Furthermore, the paper illustrates how tools can be integrated into a building management system to support the operation of buildings. These tools may be a design model that is used for real-time performance monitoring, a fault detection and diagnostics algorithm, or a control sequence, each of which may be exported as a Functional Mock-up Unit and made available in a building management system as an input/output block. We anticipate that this capability can contribute to bridging the observed performance gap between design and operational energy use of buildings.

  7. Manufacturing of small-scale mock-ups and of a semi-prototype of the ITER Normal Heat Flux First Wall

    International Nuclear Information System (INIS)

    Banetta, S.; Zacchia, F.; Lorenzetto, P.; Bobin-Vastra, I.; Boireau, B.; Cottin, A.; Mitteau, R.; Eaton, R.; Raffray, R.

    2014-01-01

    This paper describes the manufacturing development and fabrication of reduced scale ITER First Wall (FW) mock-ups of the Normal Heat Flux (NHF) design, including a “semi-prototype” with a dimension of 305 mm × 660 mm, corresponding to about 1/6 of a full-scale panel. The activity was carried out in the framework of the pre-qualification of the European Domestic Agency (EU-DA or F4E) for the supply of the European share of the ITER First Wall. The hardware consists of three Upgraded (2 MW/m 2 ) Normal Heat Flux (U-NHF) small-scale mock-ups, bearing 3 beryllium tiles each, and of one Semi-Prototype, representing six full-scale fingers and bearing a total of 84 beryllium tiles. The manufacturing process makes extensive use of Hot Isostatic Pressing, which was developed over more than a decade during ITER Engineering Design Activity phase. The main manufacturing steps for the semi-prototype are described, with special reference to the lessons learned and the implications impacting the future fabrication of the full-scale prototype and the series which consists of 218 panels plus spares. In addition, a “tile-size” mock-up was manufactured in order to assess the performance of larger tiles. The use of larger tiles would be highly beneficial since it would allow a significant reduction of the panel assembly time

  8. The state of the art report on the fabrication of FW blanket for the fusion reactor and mock-up development in Europe

    International Nuclear Information System (INIS)

    Kim, Jun Whan; Baek, Jong Hyuk; Park, Jeong Yong; Kim, Hyun Gil; Jeong, Yong Hwan

    2004-08-01

    Blanket-shield system in ITER is the component where it directly is faced with high-heat plasma. Function of blanket is to sustain extremely high temperature environment as well as to remove heat flux generated its surface. It mainly consists of plasma facing part, heat sinking part and structural part. Plasma facing part is made of armour materials such as beryllium, tungsten and carbon fiber composite. Heat sinking part is made of copper alloy to maximize heat transfer into flowing coolant inside of blanket. Structural material is used in 316LN stainless steel. As joining such dissimilar materials emerged as an issue, many developed countries have spurred the development of joint technology. This technical report was focused on the activities of EU regarding joining beryllium, copper and stainless steel. EU have adopted to Hot Isostatic Pressing (HIP) to join beryllium, copper and stainless steel. Although brazing process is not actively investigated compared as HIP, it still investigated in some countries to support HIP. Fabrication of mock-up is accomplished by CEA in France to finish small scale mock-up in 1996, medium and large scale mock-up in 1997. In recent, FRAMATOME in EU has focused on manufacturing prototype used for ITER

  9. The surface mock-up KENTEX: on the thermal-hydro-mechanical behaviors in the buffer of a Korean HLW repository

    International Nuclear Information System (INIS)

    Lee, Jae Owan; Cho, Won Jin; Choi, Jong Won

    2008-01-01

    The concept for a disposal of high-level wastes (HLW) in Korea is based upon a multi barrier system composed of engineered barriers and its surrounding plutonic rock (Kang et. al., 2002). A repository is constructed in a bedrock of several hundred meters in depth below the ground surface. The engineered barrier system (EBS), which is similar to the configuration considered by many other countries, consists of the HLW-encapsulating disposal container, the buffer between the container and the wall of a borehole, and the backfill in the inside space of the emplacement room, to isolate the HLW from the surrounding rock masses. The engineering performance of a HLW repository is dependent, to a large extent, upon the thermal-hydro-mechanical (THM) behaviors in the buffer which are complicated by the processes such as the decay heat generated from the HLW, the ground water flowing in from the surrounding host rock, and the swelling pressure exerted by compacted bentonite. For this reason, the Korea Atomic Energy Research Institute (KAERI), to investigate the THM behaviors in the buffer of the Korean reference disposal system (KRS), planned large-scale tests to be conducted in two stages: a surface mock-up and then a full-scale 'in situ' test. This paper deals with the surface mock-up called as 'KENTEX' and presents the THM behaviors in the buffer which have been investigated from the KENTEX test. The KENTEX is a third scale of the KRS. It consists of five major components: a heating system, a confining cylinder, a hydration tank, bentonite blocks, and sensors and instruments. The heating system measures 0.41 m in diameter and 0.68 m in length, which includes three heating elements in its inside, capable of supplying a thermal power of 1 kW each. The confining cylinder, which plays a role of the wall of a borehole excavated in the host rock, is a steel body with a length of 1.36 m and an inner diameter of 0.75 m, the inside wall of which is lined with layers of geotextile

  10. Geochemical Processes and compacted bentonite FEBEX with a thermohydraulic gradient with a thermohydraulic gradient; Procesos geoquimicos y modificaciones texturales en bentonita FEBEX compactada sometida a un gradiente termohidraulico

    Energy Technology Data Exchange (ETDEWEB)

    Leguey Jimenez, S; Cuevas Rodriguez, J; Martin Barca, M; Vigil de la Villa Mencia, R.; Ramirez Martin, S; Garcia Gimenez, R [Universidad Autonoma de Madrid (Spain)

    2002-07-01

    At present, the main source of High Level radioactive Waste (HLW) is the electrical energy production during all sep of developing. In almost all the countries with nuclear programs, the option for the final management of HLW is the Deep Geological Repository (DGR), based on the concept of multi barrier. According to this concept, the wastes is isolated from biosphere by the interposition of confinement barrier. In the context of an investigation of the near field for a repository of HLW, the FEBEX Project, a set of laboratory test has been designed to give a better understanding of the thermo-hydro-mechanical and geochemical behaviour of the compacted bentonite as a confinement barrier. The object of these work is to analyse the properties of the bentonite and its behaviour under conditions that will be found in a repository. The precipitation of mineral phases, due to local changes in the chemical equilibrium and the hydration itself, can produce changes in the salinity of the interstitial water and in the microstructural organisation of the clay particles. the hydraulic and mechanical properties of the bentonite can be modified by the special conditions of the barrier. (Author)

  11. Calculations of fission rate distribution in the core of WWER-1000 mock-up on the LR-0 reactor using alternative methods and comparison with results of measurements

    International Nuclear Information System (INIS)

    Zaritskiy, S.; Kovalishin, A.; Tsvetkov, T.; Rypar, V.; Svadlenkova, M.

    2011-01-01

    General review of experimental and calculation researches on WWER-440 and WWER-1000 mock-ups on the reactor LR-0 was introduced on the twentieth Symposium AER. The experimental core fission rate distribution was obtained by means of gamma-scanning of the fuel pins - 140La single peak (1596 keV) measurements and wide energy range (approximately 600-900 keV) measurements. Altogether from 260 to 500 fuel pins were scanned in different experiments. The measurements were arranged in the middle of the fuel (the active part of pin). Pin-to-pin calculations of the WWER-1000 mock-up core fission rate distribution were performed with several codes: Monte Carlo codes MCU-REA/2 and MCNPX with different nuclear data libraries, diffusion code RADAR (63 energy groups library) and code SVL based on Surface Harmonics Method (69 energy groups). Calculated data are compared with experimental ones. The obtained results allow developing the benchmark for core calculations methodologies, evaluating and validating source reliability for the out-of-core (inside and outside pressure vessel) neutron transport calculations. (Authors)

  12. FEBEX: Full-Scale engineered barriers experiment in crystalline host-rock: preoperational phase. Synthesized report; FEBEX: experimento de barreras de ingenieria a gran escala en rocas cristalinas: etapa preoperacional. Informe de sintesis

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    The FEBEX project is being cofinanced by the EC under contract F 14WCT950006. In addition to the EC, seven partners from three countries of the EU. (France, Germany, and Spain) as well as one from EFTA (Switzerland) are participating in the project. ENRESA is the coordinating partner with NAGRA assisting in coordinating some aspects. The project consists of two large-scale tests and a series of complimentary laboratory tests. The work is being executed by the following organizations: CIEMAT, AITEMIN, UP-DIT (CIMNE), ULC, CSIC-Zaidin, and UPM (SPAIN) ANDRA and G.3S (FRANCE) GRS (GERMANY). This report includes a synthesized description of the experiment from its conception through the installation of the two large-scale tests (from the middle of 1994 to the beginning of 1997, preoperation stage). The experiment is described in detail in a series of specific reports. (Author)

  13. Adsorption behaviour of bivalent ions onto Febex bentonite

    Energy Technology Data Exchange (ETDEWEB)

    Missana, T.; Garcia-Gutierrez, M. [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT), Dpt. de Impacto Ambiental de la Energia Madrid (Spain)

    2005-07-01

    The sorption and transport properties of radionuclides in the near and far field barriers of a deep geological radioactive waste repository are amongst the principal aspects to be evaluated for the performance assessment (PA) of such a kind of disposal. The study of the clayey materials is crucial because the backfill material is constituted by compacted clay in most countries design; in addition, argillaceous formations are particularly suitable as host rock formations. It is widely recognised that, to acquire predictive modelling capability, a theoretical effort is needed for a mechanistic understanding of sorption processes, as they greatly influence the transport of radionuclides in clay porous structures. In this work, an exhaustive experimental study of the Co(II), Sr (II) and Ca(II) sorption behaviour on a Spanish bentonite was carried out. The clay used for these experiments is the FEBEX bentonite, which is basically formed by smectite (93 {+-} 2%) with small percentages of quartz (2 {+-} 1 %), plagioclase (3 {+-} 1 %), cristobalite (2 {+-} 1 %) and traces of minerals such as K-feldspar and calcite. (authors)

  14. Geochemical processes and compacted bentonite FEBEX with a thermohydraulic gradient with a thermohydraulic gradient

    International Nuclear Information System (INIS)

    Leguey Jimenez, S.; Cuevas Rodriguez, J.; Martin Barca, M.; Vigil de la Villa Mencia, R.; Ramirez Martin, S.; Garcia Gimenez, R.

    2002-01-01

    At present, the main source of High Level radioactive Waste (HLW) is the electrical energy production during all sep of developing. In almost all the countries with nuclear programs, the option for the final management of HLW is the Deep Geological Repository (DGR), based on the concept of multi barrier. According to this concept, the wastes is isolated from biosphere by the interposition of confinement barrier. In the context of an investigation of the near field for a repository of HLW, the FEBEX Project, a set of laboratory test has been designed to give a better understanding of the thermo-hydro-mechanical and geochemical behaviour of the compacted bentonite as a confinement barrier. The object of these work is to analyse the properties of the bentonite and its behaviour under conditions that will be found in a repository. The precipitation of mineral phases, due to local changes in the chemical equilibrium and the hydration itself, can produce changes in the salinity of the interstitial water and in the microstructural organisation of the clay particles. the hydraulic and mechanical properties of the bentonite can be modified by the special conditions of the barrier. (Author)

  15. Microstructure and mechanical properties of reactor pressure vessel mock-up material treated by intercritical heat treatment

    International Nuclear Information System (INIS)

    Kim, M. C.; Lee, B. S.; Hong, J. H.; Lee, H. J.; Park, S. D.; Kim, K. B.; Yoon, J. H.; Kim, J. S.; Oh, J. M.

    2003-12-01

    The mechanical properties and microstructures of base metal and weld HAZ (Heat-Affected Zone) of a Mn-Mo-Ni low alloy steels treated by intercritical heat treatment were investigated to evaluate effects of intercritical heat treatment on mechanical properties. In order to clarify the effects of intercritical heat treatment, two types of specimen were prepared by CHT(Conventional Heat Treatment) and IHT(CHT+Intercritical Heat Treatment). Tensile test, charpy impact test and vickers hardness test were carried out to evaluate the mechanical properties. It is found that impact toughness and hardness were improved by intercritical heat treatment. Mean size of precipitates and effective grain were quantitatively analysed as microstructural factors. It is found that precipitate size was decreased and shape of precipitate was spherodized by intercritical heat treatment and grain size was also decreased. So, it is thought that these microstructural changes cause the improvement of mechanical properties by intercritical heat treatment. The simulated specimen using a Gleeble thermal simulator system was used to evaluate the mechanical properties of HAZ. It is well known that IRHAZ and SRHAZ have lower toughness than base metal. However, in the case of IHT, impact toughness of IRHAZ and SRHAZ were slightly higher than that of base metal. It is obvious that this improvement of fracture toughness in IRHAZ and SRHAZ region was closely related to the microstructural changes, such as spheroidization of precipitate and decreases of precipitate size and grain size

  16. Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept

    Science.gov (United States)

    Chen, Lei; Liu, Xiang; Lian, Youyun; Cai, Laizhong

    2015-09-01

    The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal-mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor configuration is under construction in SWIP, where ITER-like flat-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2011GB110001 and 2011GB110004)

  17. Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept

    International Nuclear Information System (INIS)

    Chen Lei; Liu Xiang; Lian Youyun; Cai Laizhong

    2015-01-01

    The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal–mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor configuration is under construction in SWIP, where ITER-like flat-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak. (paper)

  18. Simulation of Water Percolation in a FEBEX Bentonite Block using TOUGH2 Program

    International Nuclear Information System (INIS)

    Bru, A.

    2001-01-01

    We use Tough2 program to simulate the water percolation in a Febex bentonite Block. From obtained results, we conclude that mean field approximation does not describe this process because the heterogeneity of the medium it is not include in mathematical formalism. (Author) 17 refs

  19. Deterministic 3D transport, sensitivity and uncertainty analysis of TPR and reaction rate measurements in HCPB Breeder Blanket mock-up benchmark

    International Nuclear Information System (INIS)

    Kodeli, I.

    2006-01-01

    The Helium-Cooled Pebble Bed (HCPB) Breeder Blanket mock-up benchmark experiment was analysed using the deterministic transport, sensitivity and uncertainty code system in order to determine the Tritium Production Rate (TPR) in the ceramic breeder and the neutron reaction rates in beryllium, both nominal values and the corresponding uncertainties. The experiment, performed in 2005 to validate the HCPB concept, consists of a metallic beryllium set-up with two double layers of breeder material (Li 2 CO 3 powder). The reaction rate measurements include the Li 2 CO 3 pellets for the tritium breeding monitoring and activation foils, inserted at several axial and lateral locations in the block. In addition to the well established and validated procedure based on the 2-dimensional (2D) code DORT, a new approach for the 3D modelling was validated based on the TORT/GRTUNCL3D transport codes. The SUSD3D code, also in 3D geometry, was used for the cross-section sensitivity and uncertainty calculations. These studies are useful for the interpretation of the experimental measurements, in particular to assess the uncertainties linked to the basic nuclear data. The TPR, the neutron activation rates and the associated uncertainties were determined using the EFF-3.0 9 Be nuclear cross section and covariance data, and compared with those from other evaluations, like FENDL-2.1. Sensitivity profiles and nuclear data uncertainties of the TPR and detector reaction rates with respect to the cross-sections of 9 Be, 6 Li, 7 Li, O and C were determined at different positions in the experimental block. (author)

  20. Use of hydraulic and aerial mock up to study atmospheric pollution; L'utilisation des maquettes aeriennes et hydrauliques pour l'etude de la pollution atmospherique

    Energy Technology Data Exchange (ETDEWEB)

    Facy, L; Perrin De Brichambaut, C; Doury, A; Le Quinio, R

    1962-07-01

    Fundamental studies on turbulent atmospheric diffusion of finely divided particles, cannot remain on a purely theoretical basis. Further experimental studies must be considered. - In full scale, from accidental and induced releases. - On a reduced scale, in aerodynamic wind tunnels or hydraulic water tunnels. A first set of studies on reduced scale models has been worked out according to a contract between French 'Meteorologie Nationale' and French 'Commissariat a l'Energie Atomique' and with the Collaboration of Saint-Cyr 'Institut Aerotechnique'. Essentially two kinds of results have been obtained: - The mathematical model of SUTTON for the turbulent diffusion in the atmosphere, deduced from the SUTTON theory, generally used by us, has been correctly verified, qualitatively and quantitatively whenever experiments were consistent with the theory conditions. - The quantitative assays of photographic and cinematographic visualization have given precise details on the phenomena inaccessible to calculations, due to the influence of obstacles and release conditions. - Generally, it can be asserted, that the atmospheric pollution studies are worked out by mock up experimentations and that, in some cases these experiments never can be replaced by mathematically pure models. (authors) [French] Les etudes fondamentales portant sur la diffusion turbulente dans l'atmosphere de quantites de matieres finement divisees, ne peuvent se maintenir sur un plan exclusivement theorique. C'est pourquoi des etudes experimentales complementaires doivent etre obligatoirement envisagees: - en vraie grandeur, a partir d'emissions accidentelles ou provoquees; - sur modeles reduits, en soufflerie aerodynamique ou en veine hydraulique. En ce qui concerne les modeles reduits, une premiere serie d'etudes a pu etre menee a bien, dans le cadre d'un contrat passe entre la Meteorologie Nationale et le Commissariat a l'Energie Atomique, et avec la collaboration de l'Institut Aerotechnique de Saint

  1. Estudos etnoictiológicos sobre o pirarucu Arapaima gigas na Amazônia Central Ethnoictiology studies on Pirarucu (Arapaima mock-ups in Central Amazon

    Directory of Open Access Journals (Sweden)

    Liane Galvão de Lima

    2012-09-01

    Full Text Available O presente estudo visou identificar saberes comuns entre o conhecimento científico e o conhecimento local sobre a ecologia e biologia do pirarucu (Arapaima gigas, contribuindo com informações úteis para a implementação e consolidação de projetos de manejo participativo pesqueiro na região. Foram realizadas 57 entrevistas semi-estruturadas, com pescadores profissionais de Manaus e pescadores de subsistência de Manacapuru durante o período de junho a dezembro do ano de 2002. Foi observado que os pescadores profissionais possuem informações igualmente precisas e abrangentes em relação aos saberes dos pescadores ribeirinhos de subsistência nos aspectos de reprodução, predação, migração, crescimento e mortalidade. Os aspectos que não são equivalentes entre os pescadores profissionais comerciais citadinos e ribeirinhos de subsistência são nos aspectos de tipo de alimentação e no tamanho de recrutamento pesqueiro. Concluímos que os pescadores da Amazônia central possuem os conhecimentos necessários que possibilitam o manejo participativo do pirarucu, como um profundo saber nos aspectos comportamentais, biológicos e ecológicos desta espécie, podendo assim contribuir de fato com a participação de gestão nos recursos pesqueiros locais.Present study it aimed at to identify to know common between scientific knowledge and local knowledge on ecology and biology of pirarucu (Arapaima mock-ups, contributing with useful information for implementation and consolidation of projects of participative handling fishing boat in region. 57 half-structuralized interviews had been carried through, with fishing of Manaus and Manacapuru during period of June to December of year 2002. It was observed that professional fishermen also have accurate and comprehensive information in relation to knowledge of subsistence fishermen in coastal aspects of reproduction, predation, migration, growth and mortality. Aspects that are not equivalent

  2. Synthesis and characterisation of SiC{sub f}/Cu matrix composites and their application in a divertor flat-tile mock-up; Synthese und Charakterisierung von SiC{sub f}/Cu-Matrix-Verbundwerkstoffen und ihre Anwendung in einem Modell einer Divertor-Komponente

    Energy Technology Data Exchange (ETDEWEB)

    Paffenholz, Verena

    2010-06-30

    ) interlayers. Tensile tests of MMC specimens gave information about the tensile strength at different temperatures. A higher fibre volume fraction leads to an increased tensile strength, whereas higher temperatures reduce the tensile strength. The required tensile strength of 300 MPa was obtained with a fibre volume fraction of ∝10%. At temperatures above ∝300 C tensile tests, thermo mechanical fatigue tests, and the determination of the CTE showed an increased ductility of Cu. Contrary to tensile tests at room temperature - the weak point is the layered structure of the fibre - the failure occurred at the interface between fibre and matrix at temperatures of 550 C. Thermal investigations showed the dependence of the thermal diffusivity and thermal conductivity on the fibre volume fraction, the fibre arrangement, and the fibre distance. Narrow fibre distances as well as pores and cracks due to oxide layers within the MMC, will cause thermal barriers and reduce the heat transport. High consolidation temperatures (800 C) lead to a more uniform fibre arrangement and an increased thermal conductivity. However, a high pressure at high temperatures during consolidation can cause fibre rupture, in particular for the 0 /90 fibre orientation. MMC interlayers were synthesised at a pressure of ∝40 MPa and a temperature of 650 C to assemble three flat-tile mock-ups for high heat flux tests at the facility GLADIS at IPP. The MMC consisting of 4 and 5 layers, a fibre volume fraction of 14% and a heat conductivity of ∝180 Wm{sup -1}K{sup -1} were loaded with heat loads of 0.2-10.5 MW/m{sup 2} for a duration of 30 s. Additional cyclic tests at 10.5 MW/m{sup 2} for 20 s every minute showed a high potential of SiC{sub f}/Cu up to a temperature of ∝550 C and as heat sink material in future fusion reactors. For metallographic investigations the heat flux tests were stopped after 80 cycles at 10.5 MW/m{sup 2}. Microscopic investigations showed a good bonding, few cracks and no crack

  3. Design, calculation and testing on mock-up of B(U) f type LR 56 packaging for radioactive liquid effluent transport

    International Nuclear Information System (INIS)

    Belaud; Leconnetable; Daspet; Tombini; Tanguy

    1986-06-01

    Transport of radioactive acid liquid effluents are effected on tank truck inside nuclear center of the CEA. The cylindrical packaging type B(U) f has a capacity of 4,000l, a maximum permissible activity of 110 T Bq (3x10 4 Ci) and comprises a central element for liquid effluent containment to prevent contamination of environment and peripheral elements for mechanical, biological and thermal protection. This packaging is fixed on a trailer associated with a control box. Design and equipment of the packaging are studied for a maximum safety and in accordance with regulations [fr

  4. Simulation of Water Percolation in a FEBEX Bentonite Block Using TOUGH2 Program; Simulacion de la Percolacion de Agua en un Bloque de Bentonite Febex Utilizando el Programa TOUGH2

    Energy Technology Data Exchange (ETDEWEB)

    Bru, A.

    2001-07-01

    We use Tough2 program to simulate the water percolation in a Febex bentonite Block. From obtained results, we conclude that mean field approximation does not describe this process because the heterogeneity of the medium it is not include in mathematical formalism. (Author) 17 refs.

  5. CATSIUS CLAY PROJECT: Calculation and testing of behaviour of unsaturated clay as barrier in radioactive waste repositories: stage 3: validation exercises at a large in situ scale

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, E E; Alcoverro, J

    1999-07-01

    Stage 3 of CATSIUS CLAY Project: Validation Exercises at a Large in situ Scale includes two Benchmarks: Benchmark 3.1: In situ Hydration of Boom Clay Pellets (BACCHUS 2) and Benchmark 3.2: FEBEX Mock-up Test. The BACCHUS 2 in situ test was performed in the HADES underground laboratory (Mol, Belgium) to demonstrate and optimize an installation procedure for a clay based material and to study its hydration process. After drilling a vertical shaft (540 mm in diameter, 3.0 m in length) in the host Boom clay, a central filter (90 mm in diameter) was placed, the remaining space was filled with a mixture of clay pellets and clay powder and the assembly was sealed at the upper end by a resin plug (0.20 m in thickness) over which concrete was poured. The test was instrumented so that it could be used as a validation experiment. Total stress, pore water pressure and water content measurements were performed both in the backfill material and in the surrounding clay massif. Once the installation was complete, the natural hydration of the backfill material began (day 0). To accelerate the hydration process, on day 516 water was injected through the central filter. On day 624, after the saturation of the backfill was reached, the hydraulic circuit was closed and the undrained response of the system backfill-host clay was monitored until an overall steady state was reached. Partners were asked to provide predictions for the evolution of the pore water pressure and total pressure of various points where appropriate sensors are installed. This benchmark addresses the Hydro-Mechanical response of an unsaturated low density clay barrier under natural and artificial hydration. (Author)

  6. CATSIUS CLAY PROJECT: Calculation and testing of behaviour of unsaturated clay as barrier in radioactive waste repositories: stage 3: validation exercises at a large in situ scale

    International Nuclear Information System (INIS)

    Alonso, E. E.; Alcoverro, J.

    1999-01-01

    Stage 3 of CATSIUS CLAY Project: Validation Exercises at a Large in situ Scale includes two Benchmarks: Benchmark 3.1: In situ Hydration of Boom Clay Pellets (BACCHUS 2) and Benchmark 3.2: FEBEX Mock-up Test. The BACCHUS 2 in situ test was performed in the HADES underground laboratory (Mol, Belgium) to demonstrate and optimize an installation procedure for a clay based material and to study its hydration process. After drilling a vertical shaft (540 mm in diameter, 3.0 m in length) in the host Boom clay, a central filter (90 mm in diameter) was placed, the remaining space was filled with a mixture of clay pellets and clay powder and the assembly was sealed at the upper end by a resin plug (0.20 m in thickness) over which concrete was poured. The test was instrumented so that it could be used as a validation experiment. Total stress, pore water pressure and water content measurements were performed both in the backfill material and in the surrounding clay massif. Once the installation was complete, the natural hydration of the backfill material began (day 0). To accelerate the hydration process, on day 516 water was injected through the central filter. On day 624, after the saturation of the backfill was reached, the hydraulic circuit was closed and the undrained response of the system backfill-host clay was monitored until an overall steady state was reached. Partners were asked to provide predictions for the evolution of the pore water pressure and total pressure of various points where appropriate sensors are installed. This benchmark addresses the Hydro-Mechanical response of an unsaturated low density clay barrier under natural and artificial hydration. (Author)

  7. A coupled THMC model of a heating and hydration laboratory experiment in unsaturated compacted FEBEX bentonite

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, L.; Samper, J.; Montenegro, L.; Fernandez, A.M.

    2010-05-01

    Unsaturated compacted bentonite is foreseen by several countries as a backfill and sealing material in high-level radioactive waste repositories. The strong interplays between thermal (T), hydrodynamic (H), mechanical (M) and chemical (C) processes during the hydration stage of a repository call for fully coupled THMC models. Validation of such THMC models is prevented by the lack of comprehensive THMC experiments and the difficulties of experimental methods to measure accurately the chemical composition of bentonite porewater. We present here a non-isothermal multiphase flow and multicomponent reactive solute transport model for a deformable medium of a heating and hydration experiment performed on a sample of compacted FEBEX bentonite. Besides standard solute transport and geochemical processes, the model accounts for solute cross diffusion and thermal and chemical osmosis. Bentonite swelling is solved with a state-surface approach. The THM model is calibrated with transient temperature, water content and porosity data measured at the end of the experiment. The reactive transport model is calibrated with porewater chemical data derived from aqueous extract data. Model results confirm that thermal osmosis is relevant for the hydration of FEBEX bentonite while chemical osmosis can be safely neglected. Dilution and evaporation are the main processes controlling the concentration of conservative species. Dissolved cations are mostly affected by calcite dissolution-precipitation and cation exchange reactions. Dissolved sulphate is controlled by gypsum/anhydrite dissolution-precipitation. pH is mostly buffered by protonation/deprotonation via surface complexation. Computed concentrations agree well with inferred aqueous extract data at all sections except near the hydration boundary where cation data are affected by a sampling artifact. The fit of Cl{sup -} data is excellent except for the data near the heater. The largest deviations of the model from inferred aqueous

  8. The state of the art report on the development of manufacturing technology of fusion reactor FW blanket and mock-up in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. Y.; Jeong, Y. H.; Baek, J. H.; Kim, J. H.; Kim, H. G

    2004-08-15

    The joining technology of first wall blanket has been developed by JAERI in collaboration with Kawasaki Heavy Industry, Isuau Motors and University of Tsukuba in Japan. A variety of joining technologies including HIP, brazing, casing and friction welding was applied to the manufacturing of SS/SS and Cu/SS joint. In Be/Cu joining, it was emphasized to find the optimal HIP temperature lower than 650 .deg. C in order to avoid excessive SS sensitization because the joining of Be tile to Cu heat sink is a final processing step in the manufacturing of FW blanket. The selected HIP condition were 620 .deg. C, 150MPa and 2hr with Cu interlayer. Sample tests for joints was completed by 1995. The small scale mockup was manufactured and its performance was qualified by end of 2000. From 2001, the manufacturing and the characterization has been carried out for the larger scale mockup.

  9. The state of the art report on the development of manufacturing technology of fusion reactor FW blanket and mock-up in Japan

    International Nuclear Information System (INIS)

    Park, J. Y.; Jeong, Y. H.; Baek, J. H.; Kim, J. H.; Kim, H. G.

    2004-08-01

    The joining technology of first wall blanket has been developed by JAERI in collaboration with Kawasaki Heavy Industry, Isuau Motors and University of Tsukuba in Japan. A variety of joining technologies including HIP, brazing, casing and friction welding was applied to the manufacturing of SS/SS and Cu/SS joint. In Be/Cu joining, it was emphasized to find the optimal HIP temperature lower than 650 .deg. C in order to avoid excessive SS sensitization because the joining of Be tile to Cu heat sink is a final processing step in the manufacturing of FW blanket. The selected HIP condition were 620 .deg. C, 150MPa and 2hr with Cu interlayer. Sample tests for joints was completed by 1995. The small scale mockup was manufactured and its performance was qualified by end of 2000. From 2001, the manufacturing and the characterization has been carried out for the larger scale mockup

  10. Drilling Experiments of Dummy Fuel Rods Using a Mock-up Drilling Device and Detail Design of Device for Drilling of Irradiated Nuclear Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Yong; Lee, H. K.; Chun, Y. B.; Park, S. J.; Kim, B. G

    2007-07-15

    KAERI are developing the safety evaluation method and the analysis technology for high burn-up nuclear fuel rod that is the project, re-irradiation for re-instrumented fuel rod. That project includes insertion of a thermocouple in the center hole of PWR nuclear fuel rod with standard burn-up, 3,500{approx}4,000MWD/tU and then inspection of the nuclear fuel rod's heat performance during re-irradiation. To re-fabricate fuel rod, two devices are needed such as a drilling machine and a welding machine. The drilling machine performs grinding a center hole, 2.5 mm in diameter and 50 mm in depth, for inserting a thermocouple. And the welding machine is used to fasten a end plug on a fuel rod. Because these two equipment handle irradiated fuel rods, they are operated in hot cell blocked radioactive rays. Before inserting any device into hot cell, many tests with that machine have to be conducted. This report shows preliminary experiments for drilling a center hole on dummy of fuel rods and optimized drilling parameters to lessen operation time and damage of diamond dills. And the design method of a drilling machine for irradiated nuclear fuel rods and detail design drawings are attached.

  11. Development and trial manufacturing of 1/2-scale partial mock-up of blanket box structure for fusion experimental reactor

    International Nuclear Information System (INIS)

    Hashimoto, Toshiyuki; Takatsu, Hideyuki; Sato, Satoshi

    1994-07-01

    Conceptual design of breeding blanket has been discussed during the CDA (Conceptual Design Activities) of ITER (International Thermonuclear Experimental Reactor). Structural concept of breeding blanket is based on box structure integrated with first wall and shield, which consists of three coolant manifolds for first wall, breeding and shield regions. The first wall must have cooling channels to remove surface heat flux and nuclear heating. The box structure includes plates to form the manifolds and stiffening ribs to withstand enormous electromagnetic load, coolant pressure and blanket internal (purge gas) pressure. A 1/2-scale partial model of the blanket box structure for the outboard side module near midplane is manufactured to estimate the fabrication technology, i.e. diffusion bonding by HIP (Hot Isostatic Pressing) and EBW (Electron Beam Welding) procedure. Fabrication accuracy is a key issue to manufacture first wall panel because bending deformation during HIP may not be small for a large size structure. Data on bending deformation during HIP was obtained by preliminary manufacturing of HIP elements. For the shield structure, it is necessary to reduce the welding strain and residual stress of the weldment to establish the fabrication procedure. Optimal shape of the parts forming the manifolds, welding locations and welding sequence have been investigated. In addition, preliminary EBW tests have been performed in order to select the EBW conditions, and fundamental data on built-up shield have been obtained. Especially, welding deformation by joining the first wall panel to the shield has been measured, and total deformation to build-up shield by EBW has been found to be smaller than 2 mm. Consequently, the feasibility of fabrication technologies has been successfully demonstrated for a 1m-scaled box structure including the first wall with cooling channels by means of HIP, EBW and TIG (Tungsten Inert Gas arc)-welding. (author)

  12. The Plasma-Facing Components Transporter (PFCT) : a Prototype System for PFC Replacement on the new ITER 2001 Cassette Mock-up

    International Nuclear Information System (INIS)

    Micciche, G.; Lorenzelli, L.; Muro, L.; Irving, M.

    2006-01-01

    The remote maintainability of the early ITER divertor cassette (based on the ITER 1998 design) was successfully proved during test campaigns carried out in the Divertor Refurbishment Platform (DRP) at the ENEA research centre at Brasimone over the period 1999-2003. Due to subsequent major modifications in the ITER divertor cassette design, the main focus over the past few years has been on the design and manufacture of the various components, devices and tools needed for refurbishment of the new ITER 2001 Divertor Cassette. The design of this new cassette differs substantially from the earlier version: in particular the shape, weight and attachment system of the Plasma Facing Components (PFC's) has been completely revised, and this also entailed a review of the procedures adopted for its refurbishment. One of the major requirements of the cassette refurbishment process is removal and replacement of the three PFC's. In the old cassette concept, target replacement was performed by means of a purpose-built '' C '' frame slung from a standard bridge crane. The 2001 cassette design precludes such handling methods for a number of reasons, notably because of the extremely tight inter-PFC clearances, and the need for controlled inclination of the target in addition to normal translational movements, both impossible with a simple Cartesian crane. To demonstrate the refurbishment feasibility operations for the new ITER Divertor 2001 cassettes, an experimental machine known as the Plasma-Facing Component Transporter (PFCT) has been designed, fabricated and commissioned in the years 2004-5. This full six degree-of-freedom system has been designed to handle payloads of up to 5 tonnes with good positional accuracy, and axes capable of very low joint velocities, including inclination of the PFC's over the range of ± 10 o in both horizontal axes, and controlled rotation about the vertical axis. Preliminary trials carried out during the commissioning phase have proved its

  13. Diffusion of strongly sorbing cations (60Co and 152Eu) in compacted Febex bentonite

    International Nuclear Information System (INIS)

    Garcia-Gutierrez, M.; Missana, T.; Alonso, U.; Mingarro, M.; Cormenzana, J.L.

    2010-01-01

    Document available in extended abstract form only. Compacted bentonite is used as an engineered barrier in high-level radioactive waste (HLRW) repositories because is a swelling clay of very low permeability and high sorption capability for many solutes. The transport of radionuclides through compacted bentonite is a diffusion-controlled process retarded by sorption. Performance assessment calculations of a repository need diffusion coefficients data of relevant radionuclides. Several studies on diffusion behaviour of neutral, anionic and weakly sorbing elements on clay exist while very few studies are available for moderately sorbing elements, and almost no studies for Eu, a highly sorbing element are reported. In this study, diffusion experiments with strongly sorbing radionuclides, as 60 Co and 152 Eu, have been performed through compacted FEBEX bentonite. Diffusion essays with these strongly sorbing radionuclides are not straightforward to carry out because they are very time consuming essays, but also because sorption on the diffusion cells, tubing, filters and reservoirs, typically used in the classical through-diffusion or in-diffusion methods make hard the interpretation of the experimental results and the calculation of the diffusion coefficients. FEBEX bentonite was selected as Spanish reference buffer materials, and used in many national and international projects. The clay comes from the Cortijo de Archidona deposit (Almeria, Spain), and has a smectite content greater than 90% (93 ± 2%), with quartz (2 ± 1%), plagioclase (3 ± 1%), cristobalite (2 ± 1%), potassic feldspar, calcite, and trydimite as accessory minerals. The specific weight of the FEBEX bentonite is 2.7 g/cm 3 . Diffusion experiments were performed using the instantaneous plane source method. In this setup, a paper filter tagged with a tracer is introduced between two compacted tablets, avoiding contact between the tracer and the experimental vessels. The tracer can diffuse into both

  14. Eddy current testing system for bottom mounted instrumentation welds - 15206

    International Nuclear Information System (INIS)

    Kobayashi, N.; Ueno, S.; Suganuma, N.; Oodake, T.; Maehara, T.; Kasuya, T.; Ichikawa, H.

    2015-01-01

    We have demonstrated the scanning of eddy current testing (ECT) probe on the welds area including the nozzle, the J-welds and the buildup welds of the Bottom Mounted Instrumentation (BMI) mock-up using the developed ECT system and procedure. It is difficult to scan the probe on the BMI welds area because the area has a complex curved surface shape and narrow spaces. We made the space coordinates and the normal vectors on the scanning points as the scanning trajectory of probe on the welds area based on the measured results of welds surface shape on the mock-up. The multi-axis robot was used to scan the probe on the welds surface. Each motion axis position of the robot corresponding to each scanning point was calculated by the inverse kinematic algorithm. The BMI mock-up test was performed using the cross coil probe in the differential mode. The artificial stress corrosion cracking and the electrical discharge machining slits given on the mock-up surface were detected. The results show that the ECT can detect a defect of approximately 2.3 mm in length, 0.5 mm in depth and 0.2 mm in width for the BMI welds. From the output voltage of single coil, we estimated that the average and the maximum probe tilt angles on the mock-up surface under scanning were 2.6 degrees and 8.5 degrees, respectively

  15. Behaviour of Ti-doped CFCs under thermal fatigue tests

    Energy Technology Data Exchange (ETDEWEB)

    Centeno, A. [Instituto Nacional del Carbon (CSIC), Apdo. 73, 33080 Oviedo (Spain); Pintsuk, G.; Linke, J. [Forschungszentrum Juelich GmbH, EURATOM Association, 52425 Juelich (Germany); Gualco, C. [Ansaldo Energia, I-16152 Genoa (Italy); Blanco, C., E-mail: clara@incar.csic.es [Instituto Nacional del Carbon (CSIC), Apdo. 73, 33080 Oviedo (Spain); Santamaria, R.; Granda, M.; Menendez, R. [Instituto Nacional del Carbon (CSIC), Apdo. 73, 33080 Oviedo (Spain)

    2011-01-15

    In spite of the remarkable progress in the design of in-vessel components for the divertor of the first International Thermonuclear Experimental Reactor (ITER), a great effort is still put into the development of manufacturing technologies for carbon armour with improved properties. Newly developed 3D titanium-doped carbon fibre reinforced composites and their corresponding undoped counterparts were brazed to a CuCrZr heat sink to produce actively cooled flat tile mock-ups. By exposing the mock-ups to thermal fatigue tests in an electron beam test facility, the material behaviour and the brazing between the individual constituents in the mock-up was qualified. The mock-ups with titanium-doped CFCs exhibited a significantly improved thermal fatigue resistance compared with those undoped materials. The comparison of these mock-ups with those produced using pristine NB31, one of the reference materials as plasma facing material for ITER, showed almost identical results, indicating the high potential of Ti-doped CFCs due to their improved thermal shock resistance.

  16. Data Acquisition System Based in MDAS Unit for the FEBEX Project; Sistema de Toma de Datos Basado en el Equipo MDAS para el Proyecto FEBEX

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, J; Barcala, J M; Navarrete, J J; Martin, P L [CIEMAT. Madrid (Spain)

    1999-12-31

    The engineered barrier concept for the storage of radioactive wastes is being tested at CIEMAT facilities. This document describes the data acquisition system of one of that experiments. (Author) 2 refs.

  17. Long-term durability test of acid recovery evaporators made of Ti-5% Ta alloy and zirconium

    International Nuclear Information System (INIS)

    Takeuchi, Masayuki; Koizumi, Tsutomu; Koyama, Tomozo

    2001-05-01

    Mock-ups of acid recovery evaporators which are made of Ti-5% Ta alloy and Zr were tested under inactive condition for forty thousands hours to improve a corrosion resistance of acid recovery evaporator in Tokai reprocessing plant (TRP). The mock-up unit was designed and produced referring to the specification of acid recovery evaporator in TRP and the evaporation performance of the mock-up was 1/27 of TRP. A long-term durability of both evaporators was demonstrated by results of operation data, evaporation performance and corrosion resistance. The mock-up unit did not suffer from any trouble during the running test and the operation data such as temperature, flow, concentrations of nitric acid and metal ions were fairly stable within standard condition. As for the corrosion resistance, cracks and local corrosion such as intergranular attack were not observed on both evaporators after the running test, and a corrosion of weld was not selective. The average corrosion rates at measuring points were less than 0.1 mm/yr, respectively, however, thickness of the Ti-5% Ta alloy evaporator was slightly reduced at all points of vapor phase region. In addition, from the result by test coupon, it is found that both materials have low susceptibility to stress corrosion cracking in this environment. The destructive inspection showed that the mechanical properties of both materials were not degraded during the running test. Finally, the total running time of the mock-up unit is much more than a maximum running time of acid recovery evaporator made of stainless steel in TRP (nearly 15,000 hours). On the basis of the test results, an excellent durability of Ti-5% Ta alloy and Zr evaporators under was successfully demonstrated throughout the mock-up test from an engineering perspective. (author)

  18. Ultrasonic Guided Waves-Based Monitoring of Rail Head: Laboratory and Field Tests

    Directory of Open Access Journals (Sweden)

    Piervincenzo Rizzo

    2010-01-01

    The first part of the paper shows the prototype in action on a railroad track mock-up built at the University of California, San Diego. The mock-up contained surface and internal defects. The results from three experiments are presented. The importance of feature selection to maximize the sensitivity of the inspection system is demonstrated here. The second part of the paper shows the results of field testing conducted in south east Pennsylvania under the auspices of the U.S. Federal Railroad Administration.

  19. Thermal and mechanical behaviour of an experimental mock-up of a nuclear containment; Comportement thermique et mecanique d'une maquette experimentale d'enceinte de confinement nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Chauvel, D. [Electricite de France (EDF/SEPTEN), 69 - Villeurbanne (France); Barre, F. [Coyne et Bellier, 92 - Gennevilliers (France)

    2007-07-01

    In order to better understand the behaviour of a reactor containment submitted to combined pressure and temperature loads by means of studies of the concrete permeability and the state of cracking evolution, EDF and its French partners have built a prestressed concrete test model which represents a PWR containment typical section. The monitoring system was designed to follow the evolution of strains, temperature and state of cracking of the concrete wall from construction stage to air and steam tests. The measurements results as well as their comparison with theoretical laws or data and calculated values, allow to determine the main thermal and mechanical characteristics of the concrete, to analyse the thermo-mechanical behaviour of the structure and also to check the design criteria of prestressed concrete containments. (authors)

  20. Data Acquisition System Based in MDAS Unit for the FEBEX Project

    International Nuclear Information System (INIS)

    Alberdi, J.; Barcala, J.M.; Navarrete, J.J.; Martin, P.L.

    1998-01-01

    The engineered barrier concept for the storage of radioactive wastes is being tested at CIEMAT facilities. This document describes the data acquisition system of one of that experiments. (Author) 2 refs

  1. Results of the mock-up experiment on partial LOCA

    International Nuclear Information System (INIS)

    Dreier, J.; Winkler, H.

    1985-01-01

    A mockup experiment has been performed to verify the heat transfer model for a partial loss of coolant accident in the swimming pool reactor SAPHIR. Three coolant channels with the same dimensions as in a SAPHIR fuel element were simulated using four electrically heated plates. For a water level such that the heated plates are partially submerged, plate temperatures remain below 160 deg. C for plate powers of up to 650 W. For water levels low enough to just block the channels, plate temperatures of 400 deg. C are reached for plate powers as low as 60 W. Details of the experiment and further results are discussed. (author)

  2. PSF blind test SSC, SPVC, and SVBC physics-dosimetry-metallurgy data packages

    International Nuclear Information System (INIS)

    1984-01-01

    Information is presented concerning the final PSF radiometric data; calculated spectral fluences and dosimeter activities for the metallurgical blind test irradiations at the ORR-PSF; fabrication data package for HEDL dosimetry in the ORNL Poolside Facility LWR pressure vessel mock-up irradiation; SSC-1; NUREG-CR-3457; and NUREG-CR-3295

  3. Status of Italian test data on isolated structures and comparison with computer predictions

    Energy Technology Data Exchange (ETDEWEB)

    Bettinali, F; Dusi, A [ENEL S.p.A. - CRIS, Milan (Italy); Martelli, A; Forni, M [ENEA, Bologna (Italy)

    1993-07-01

    This paper presents the main features of the numerical and experimental studies that are in progress in Italy on isolated structures. Particular attention is paid to the MISS mock-up, for which test data will be provided by Italy to the participants in this Research Programme. (author)

  4. Status of Italian test data on isolated structures and comparison with computer predictions

    International Nuclear Information System (INIS)

    Bettinali, F.; Dusi, A.; Martelli, A.; Forni, M.

    1993-01-01

    This paper presents the main features of the numerical and experimental studies that are in progress in Italy on isolated structures. Particular attention is paid to the MISS mock-up, for which test data will be provided by Italy to the participants in this Research Programme. (author)

  5. Remote-handling demonstration tests for the Fusion Materials Irradiation Test (FMIT) Facility

    International Nuclear Information System (INIS)

    Shen, E.J.; Hussey, M.W.; Kelly, V.P.; Yount, J.A.

    1982-01-01

    The mission of the Fusion Materials Irradiation Test (FMIT) Facility is to create a fusion-like environment for fusion materials development. Crucial to the success of FMIT is the development and testing of remote handling systems required to handle materials specimens and maintenance of the facility. The use of full scale mock-ups for demonstration tests provides the means for proving these systems

  6. SP-100 reactor disassembly remote handling test program

    International Nuclear Information System (INIS)

    Wilson, C.E.; Potter, J.D.; Maiden, G.E.; Vader, D.P.

    1991-01-01

    This paper is presented as an overview of the remote handling equipment validation testing, which will be conducted before installation and use in the ground engineering test facility. This equipment will be used to defuel the SP-100 reactor core after removing it from the Test Assembly following nuclear testing. A series of full scale mock-up operational tests will be conducted at a Hanford Site facility to verify equipment design, operation, and capabilities

  7. Heating facility for blanket and performance test

    Energy Technology Data Exchange (ETDEWEB)

    Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Sato, Satoshi; Hatano, Toshihisa; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Hara, Shigemitsu

    1999-03-01

    A design and a fabrication of heating test facility for a mock-up of the blanket module to be installed in International Thermonuclear Experimental Reactor (ITER) have been conducted to evaluate/demonstrate its heat removal performance and structural soundness under cyclic heat loads. To simulate surface heat flux to the blanket module, infrared heating method is adopted so as to heat large surface area uniformly. The infrared heater is used in vacuum environment (10{sup -4} Torr{approx}), and the lamps are cooled by air flowing through an annulus between the lamp and a cover tube made of quartz glass. Elastomer O rings (available to be used up to {approx}300degC) and used for vacuum seal at outer surface of the cover tube. To prevent excessive heating of the O ring, the end part of the cover tube is specially designed including the tube shape, flow path of air and gold coating on the surface of the cover tube to protect the O ring against thermal radiation from glowing tungsten filament. To examine the performance of the facility, steady state and cyclic operation of the infrared heater were conducted using a small-scaled shielding blanket mock-up as a test specimen. The important results are as follows: (1) Heat flux at the surface of the small-scaled mock-up measured by a calorimeter was {approx}0.2 MW/m{sup 2}. (2) A comparison of thermal analysis results and measured temperature responses showed that the small-scaled mock-up had good heat removal performance. (3) Steady state operation and cyclic operation with step response between the rated and zero powers of the infrared heater were successfully performed, and it was confirmed that this heating facility was well-prepared and available for the thermal cyclic test of a blanket module. (author)

  8. Overview of the testing activities on ITER sub-scale pre-compression rings

    Energy Technology Data Exchange (ETDEWEB)

    Rossi, Paolo, E-mail: paolo.rossi@enea.it [Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, C.P. 65, 00044 Frascati, Rome (Italy); Capobianchi, Mario; Crescenzi, Fabio; Massimi, Alberto; Mugnaini, Giampiero; Pizzuto, Aldo [Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, C.P. 65, 00044 Frascati, Rome (Italy); Knaster, Juan [ITER Organisation, Route de Vinon sur Verdon, 13115, St. Paul lez Durance (France); Rajainmaki, Hannu [FUSION FOR ENERGY, Josep Pla no. 2, Torres Diagonal Litoral Edificio B3, 08019 Barcelona (Spain)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer ENEA developed a high strength glass fiber-epoxy composite for ITER pre-compression rings. Black-Right-Pointing-Pointer High UTS values were obtained at RT on linear specimens (2200 MPa) and on scaled ring mock-ups (1550 MPa). Black-Right-Pointing-Pointer Creep tests showed very low creep strain and creep rates. Black-Right-Pointing-Pointer Long term tests showed no significant stress relaxation on the ring mock-ups. - Abstract: After a first R and D and testing activity to develop and characterize by tensile and creep tests a high strength glass fiber-epoxy composite as reference material for the manufacture of ITER pre-compression rings, ENEA designed and manufactured a dedicated testing facility and different sub-scale composite ring mock-ups in order to characterize their mechanical properties. The paper reports the results of the overall testing activities performed during the last years on a total number of eleven sub-scale pre-compression ring mock-ups manufactured by winding S2 glass fibers on a diameter of 1 m (1/5 of the full scale) both by vacuum pressure epoxy impregnation (VPI) and filament wet winding techniques (WW). The first three rings were manufactured by ENEA Frascati thanks to a particular VPI technique; one of them was used as base composite material to manufacture different sets of specimens for shear, compression and non destructive tests (NDT). Then, five other mock-ups were manufactured following ENEA VPI process and three using WW technique by two different industrial companies. The rings were tested in ENEA Frascati in a dedicated hydraulic testing machine consisting of 18 radial actuators working in position control with a total load capability of 1000 tons. The complete testing campaign consisted of six ultimate tensile strength (UTS) tests and four stress relaxation (SR) tests. The tests demonstrated that the composite (S2 glass-epoxy) is a valid and viable solution for the ITER pre

  9. Effect of Heating/Hydratation on Compacted Bentonite: Tests in 60-cm Long Cells

    Energy Technology Data Exchange (ETDEWEB)

    Villar, M. V.; Fernandez, A. M.; Martin, P. L.; Barcala, J. M.; Gomez-Espina, R.; Rivas, P.

    2008-07-01

    The conditions of the bentonite in an engineered barrier for high-level radioactive waste disposal have been simulated in a series of tests. Cylindrical cells with an inner length of 60 cm and a diameter of 7 cm were constructed. Inside the cells, blocks of compacted FEBEX bentonite were put one on top of the other. the bottom surface of the material was heated at 100 degree centigree and the top surface was injected with granitic water. the duration of the tests was 0.5, 1,2 and 7,6 years. The temperatures and water intake were measured during the tests and, at the end, the cells were dismounted and the dry density, water content, mineralogy, geochemistry and some hydro-mechanical properties of the clay (permeability, swelling) were measured at different positions. the values obtained are compared among them and to those of the untreated FEBEX bentonite. The study has run over for 10 years in the context of the projects FEBEX I and II and NF-PRO. (Author) 50 refs.

  10. Effect of Heating/Hydratation on Compacted Bentonite: Tests in 60-cm Long Cells

    International Nuclear Information System (INIS)

    Villar, M. V.; Fernandez, A. M.; Martin, P. L.; Barcala, J. M.; Gomez-Espina, R.; Rivas, P.

    2008-01-01

    The conditions of the bentonite in an engineered barrier for high-level radioactive waste disposal have been simulated in a series of tests. Cylindrical cells with an inner length of 60 cm and a diameter of 7 cm were constructed. Inside the cells, blocks of compacted FEBEX bentonite were put one on top of the other. the bottom surface of the material was heated at 100 degree centigree and the top surface was injected with granitic water. the duration of the tests was 0.5, 1,2 and 7,6 years. The temperatures and water intake were measured during the tests and, at the end, the cells were dismounted and the dry density, water content, mineralogy, geochemistry and some hydro-mechanical properties of the clay (permeability, swelling) were measured at different positions. the values obtained are compared among them and to those of the untreated FEBEX bentonite. The study has run over for 10 years in the context of the projects FEBEX I and II and NF-PRO. (Author) 50 refs

  11. Test of fuel handling machine for Monju in sodium

    International Nuclear Information System (INIS)

    Ishii, Yoichiro; Masuda, Yoichi; Kataoka, Hajime

    1980-01-01

    Various types of fuel handling machines were studied, and under-the-plug method of fuel exchange and the fuel handling machine of single turning plug, fixed arm type were selected for the prototype reactor ''Monju'', because the turning plug is relatively small, and the rate of operation, safety, operational ability, maintainability and reliability required for the reactor are satisfied, moreover, the extrapolation to the demonstration reactor was considered. Attention must be paid to the points that the fuel handling machine is very long and invisible from outside, and the smooth operation and endurance in sodium are required for it. The full mock-up testing facility of single turning plug, fixed arm type was installed in 1974, and the full mock-up test has been carried out since 1975 in Oarai. Fuel exchange is carried out at about 6 months intervals in Monju, and about 20 to 30% of core and blanket fuels are exchanged for about one month period. The functions required for the fuel handling machine for Monju, the outline of the testing facility, the schedule of the testing, the items of testing and the results, and the matters to be specially written are described. The full mock-up test in sodium has been carried out for 5 years, and the functions and the endurance have been proved sufficiently. (Kako, I.)

  12. Surface inspection technique with an eddy current testing array probe

    International Nuclear Information System (INIS)

    Nishimizu, Akira; Endo, Hisashi; Tooma, Masahiro; Otani, Kenichi; Ouchi, Hirofumi; Yoshida, Isao; Nonaka, Yoshio

    2010-01-01

    An eddy current testing (ECT) system has been developed for inspecting weld surfaces of components in the reactor pressure vessel of nuclear plants. The system can be applied to curved surfaces with an ECT array probe, it can discriminate flaws from other signal factors by using a combination of arrayed coils signal-phase. The system is applied to a mock-up of core internal components and the signal discrimination using the signal-phase clearly separated flaw and noise signals. (author)

  13. Visualization of flaws within heavy section ultrasonic test blocks using high energy computed tomography

    International Nuclear Information System (INIS)

    House, M.B.; Ross, D.M.; Janucik, F.X.; Friedman, W.D.; Yancey, R.N.

    1996-05-01

    The feasibility of high energy computed tomography (9 MeV) to detect volumetric and planar discontinuities in large pressure vessel mock-up blocks was studied. The data supplied by the manufacturer of the test blocks on the intended flaw geometry were compared to manual, contact ultrasonic test and computed tomography test data. Subsequently, a visualization program was used to construct fully three-dimensional morphological information enabling interactive data analysis on the detected flaws. Density isosurfaces show the relative shape and location of the volumetric defects within the mock-up blocks. Such a technique may be used to qualify personnel or newly developed ultrasonic test methods without the associated high cost of destructive evaluation. Data is presented showing the capability of the volumetric data analysis program to overlay the computed tomography and destructive evaluation (serial metallography) data for a direct, three-dimensional comparison

  14. Eddy current testing system for bottom mounted instrumentation welds

    Directory of Open Access Journals (Sweden)

    Kobayashi Noriyasu

    2015-01-01

    Full Text Available The capability of eddy current testing (ECT for the bottom mounted instrumentation (BMI weld area of reactor vessel in a pressurized water reactor was demonstrated by the developed ECT system and procedure. It is difficult to position and move the probe on the BMI weld area because the area has complexly curved surfaces. The space coordinates and the normal vectors at the scanning points were calculated as the scanning trajectory of probe based on the measured results of surface shape on the BMI mock-up. The multi-axis robot was used to move the probe on the mock-up. Each motion-axis position of the robot corresponding to each scanning point was calculated by the inverse kinematic algorithm. In the mock-up test, the probe was properly contacted with most of the weld surfaces. The artificial stress corrosion cracking of approximately 6 mm in length and the electrical-discharge machining slit of 0.5 mm in length, 1 mm in depth and 0.2 mm in width given on the weld surface were detected. From the probe output voltage, it was estimated that the average probe tilt angle on the surface under scanning was 2.6°.

  15. Manufacturing and testing of ITER divertor gas box liners

    International Nuclear Information System (INIS)

    Mazul, I.; Giniatulin, R.; Komarov, V.L.; Krylov, V.; Kuzmin, Ye.; Makhankov, A.; Odintsov, V.; Zhuk, A.

    1998-01-01

    Among a variety of R and D works performed by different ITER parties there are seven large projects which deal with the development, manufacturing and testing of most important complex reactor components. One of the projects is directed to produce a prototype of divertor cassette. In according with integration plan two full size liners with dummy armour are manufactured by RF Home Team. Except for liners with dummy armors the large - scale mock-up with real armour have to be manufactured in order to demonstrate the semi-industrial possibilities for joining of Be and W to CuCrZr heat - sink structure. The design of this liners, technological approaches to their manufacturing are presented. The description of brazing facility and joining technology which use a fast ohmic heating by 15 kA current is made. A mock-up of 800 mm in length and 90 mm in width was armored by 18 Be tiles (44 x 44 mm 2 in plane, 10 mm - thick) and 16 W-Cu tiles (44 x 44 mm 2 in plane, 3 mm - thick W). The preliminary results of high heat flux testing of the armored mock-ups are also presented. (author)

  16. Development and testing of CFC-copper high heat flux elements

    International Nuclear Information System (INIS)

    Mitteau, R.; Chappuis, P.; Deschamps, P.; Schlosser, J.; Viallet, H.; Vieider, G.

    1994-01-01

    In the frame of high heat flux development for plasma facing components, CEA has designed, fabricated and tested over twenty specimens, with some of them for the NET divertor application. Several Carbon Fibre Composites (CFC) and copper grades have been used with flat tile or macro bloc configuration. All the mock-ups were tested in the electron beam facility EB200, for steady-state flux and fatigue up to 1000 cycles. The best four are presented. (author) 3 refs.; 11 figs

  17. Mechanical behaviour of the reactor vessel support of a pressurized water reactor: tests and analysis

    International Nuclear Information System (INIS)

    Bolvin, M.; L'huby, Y.; Quillico, J.J.; Humbert, J.M.; Thomas, J.P.; Hugenschmitt, R.

    1985-08-01

    The PWR reactor vessel is supported by a steel ring laying on the reactor pit. This support has to ensure a good behaviour of the vessel in the event of accidental conditions (earthquake and pipe rupture). A new evolution of the evaluation methods of the applied forces has shown a significant increase in the design loads used until now. In order to take into account these new forces, we carried out a test on a representative mock-up of the vessel support (scale 1/6). This test was performed by CEA, EDF and FRAMATOME. Several static equivalent forces were applied on the experimental mock-up. Displacements and strains were simultaneously recorded. The results of the test have enabled to justify the design of the pit and the ring, to show up a wide safety margin until the collapse of the structures and to check our hypothesis about the transmission of the forces between the ring and the pit

  18. Large scale reflood test

    International Nuclear Information System (INIS)

    Hirano, Kemmei; Murao, Yoshio

    1980-01-01

    The large-scale reflood test with a view to ensuring the safety of light water reactors was started in fiscal 1976 based on the special account act for power source development promotion measures by the entrustment from the Science and Technology Agency. Thereafter, to establish the safety of PWRs in loss-of-coolant accidents by joint international efforts, the Japan-West Germany-U.S. research cooperation program was started in April, 1980. Thereupon, the large-scale reflood test is now included in this program. It consists of two tests using a cylindrical core testing apparatus for examining the overall system effect and a plate core testing apparatus for testing individual effects. Each apparatus is composed of the mock-ups of pressure vessel, primary loop, containment vessel and ECCS. The testing method, the test results and the research cooperation program are described. (J.P.N.)

  19. MASURCA, a Fast-Neutron Critical Mock-Up: Operation and Uses; MASURCA. Maquette Critique a Neutrons Rapides. Description Fonctionnelle et Obiectifs; ''MAZURKA'' - kriticheskaya model' na bystrykh nejtronakh. funktsional'noe opisanie i tseli; Descripcion Funcional y Objetivos de la Maqueta Critica de Neutrones Rapidos 'MASURCA '

    Energy Technology Data Exchange (ETDEWEB)

    Schmitt, A. P.; Storrer, F.; Vendryes, G. [Association CEA-EURATOM, Cadarache (France); Tavernier, G.; Van Dievoet, J. [Societe Belgo-Nucleaire, Bruxelles (Belgium)

    1964-02-15

    Under the EURATOMCEA Association project a fast-neutron critical mock-up, Masurca, is now being built at the Cadarache Nuclear Research Centre. The main purpose of this extremely versatile facility is the study of non-moderated, plutonium critical assemblies of large volume and hence having a relatively soft neutron spectrum. The paper explains what these studies are for. The facility must satisfy certain conditions and, in essence, combine great versatility with almost absolute operational safety. The safety problem was dealt with by: (1) Seeking inherent safety: with simulated fuel elements it was possible to obtain (a) a negative reactivity coefficient from the cumulative longitudinal expansion of these elements: (b) a negative Doppler coefficient; (2) Using a set of shim-safety rods which can be placed in a square lattice with spacings of about 30cm; (3) A pressure vessel, containing reserves of argon in case of fire: and (4) Strict administrative supervision. A U-Pu-Fe metallic alloy being chosen as the basic element in the fuel simulation, provision for cooling large-volume critical assemblies must be incorporated in the facility. Sodium, the coolant used in simulated reactors, will be represented by sodium strips clad in stainless steel. The facility is designed as a vertical single-block unit in view of the maximum volume of the cores to be simulated (about 5000 1). The simulated elements are shaped like a right prism with a square base (except in the case of fuel elements which have a circular base) with an outer side (or diameter) of 12.7 mm and a height of 102 mm. They are placed in tubes having an over- all length of about 4 m and square sections whose outer side is 10.6 mm. These tubes are placed side by side and suspended. Smaller tubes can be placed in the central area of the suspension plate so that smaller cores can be made. A special heating loop can also be placed in the central part of the facility to measure the Doppler coefficient. The paper

  20. Pseudodynamic tests on a full-scale 3-storey precast concrete building: Global response

    OpenAIRE

    Negro, Paolo; Bournas, Dionysios A.; Molina, Francisco J.

    2013-01-01

    In the framework of the SAFECAST Project, a full-scale three-storey precast building was subjected to a series of pseudodynamic (PsD) tests in the European Laboratory for Structural Assessment (ELSA). The mock-up was constructed in such a way that four different structural configurations could be investigated experimentally. Therefore, the behaviour of various parameters like the types of mechanical connections (traditional as well as innovative) and the presence or absence of shear walls alo...

  1. MP98, an innovative reactivity control system for LWR and VHTR, tests results on prototypes

    International Nuclear Information System (INIS)

    Emin, M.; Gagne, J.F.

    2005-01-01

    MP98' Reactivity Control System is now a technology tested on a real scale prototypes. Demonstrator is working at PWR real conditions (155 bars, 320 degree C), and many accidents configurations could be simulated. This paper presents the technology, describes the demonstrator design and presents some results. Mock up and demonstrator have permitted to optimize the design of MP98' devices to prepare industrialization. (authors)

  2. Package testing capabilities at the Pacific Northwest Laboratory

    International Nuclear Information System (INIS)

    Taylor, J.M.

    1993-01-01

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

  3. Reactor group constants and benchmark test

    Energy Technology Data Exchange (ETDEWEB)

    Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-08-01

    The evaluated nuclear data files such as JENDL, ENDF/B-VI and JEF-2 are validated by analyzing critical mock-up experiments for various type reactors and assessing applicability for nuclear characteristics such as criticality, reaction rates, reactivities, etc. This is called Benchmark Testing. In the nuclear calculations, the diffusion and transport codes use the group constant library which is generated by processing the nuclear data files. In this paper, the calculation methods of the reactor group constants and benchmark test are described. Finally, a new group constants scheme is proposed. (author)

  4. RITD – Wind tunnel testing

    Science.gov (United States)

    Haukka, Harri; Harri, Ari-Matti; Aleksashkin, Sergei; Koryanov, Valeri; Schmidt, Walter; Heilimo, Jyri; Finchenko, Valeri; Martynov, Maxim; Ponomarenko, Andrey; Kazakovtsev, Victor; Arruego, Ignazio

    2015-04-01

    An atmospheric re-entry and descent and landing system (EDLS) concept based on inflatable hypersonic decelerator techniques is highly promising for the Earth re-entry missions. We developed such EDLS for the Earth re-entry utilizing a concept that was originally developed for Mars. This EU-funded project is called RITD - Re-entry: Inflatable Technology Development - and it was to assess the bene¬fits of this technology when deploying small payloads from low Earth orbits to the surface of the Earth with modest costs. The principal goal was to assess and develope a preliminary EDLS design for the entire relevant range of aerodynamic regimes expected to be encountered in Earth's atmosphere during entry, descent and landing. The RITD entry and descent system utilizes an inflatable hypersonic decelerator. Development of such system requires a combination of wind tunnel tests and numerical simulations. This included wind tunnel tests both in transsonic and subsonic regimes. The principal aim of the wind tunnel tests was the determination of the RITD damping factors in the Earth atmosphere and recalculation of the results for the case of the vehicle descent in the Mars atmosphere. The RITD mock-up model used in the tests was in scale of 1:15 of the real-size vehicle as the dimensions were (midsection) diameter of 74.2 mm and length of 42 mm. For wind tunnel testing purposes the frontal part of the mock-up model body was manufactured by using a PolyJet 3D printing technology based on the light curing of liquid resin. The tail part of the mock-up model body was manufactured of M1 grade copper. The structure of the mock-up model placed th center of gravity in the same position as that of the real-size RITD. The wind tunnel test program included the defining of the damping factor at seven values of Mach numbers 0.85; 0.95; 1.10; 1.20; 1.25; 1.30 and 1.55 with the angle of attack ranging from 0 degree to 40 degrees with the step of 5 degrees. The damping characteristics of

  5. Feasibly study of gas-cooled test cell for material testing in IFMIF

    International Nuclear Information System (INIS)

    Yonemoto, Yukihiro; Maki, Eiji; Ebara, Shinji; Yokomine, Takehiko; Shimizu, Akihiko; Korenaga, Tadashi

    2002-01-01

    Temperature control performance of test pieces enclosed in IFMIF capsule by using single phase gas was estimated experimentally. The key issue of this study is to obtain the definite value of dimension of test facility and flow conditions of coolant and to clarify the temperature response of test piece to the beam-off scenario. Firstly, we have examined the cooling performance of the test cell originally proposed in IFMIF-KEP and from results of this calculation performed in three dimensional system by using brand-new turbulence model for flow and thermal fields, it is concluded that the drastical change of design of test cell is needed in order to obtain the unformity of temperature of test piece, to improve the responsibility of temperature measurement of test piece, and to relieve the coolant flow condition, especially for inlet pressure value. Thus, we have proposed new design of test cell and test piece arrangement. A mock-up experimental facility was made based on our design and preliminary experiments for temperature control were performed. As a result, we have verified the cooling performance at the case that corresponds to two beam-off scenario by using mock-up facility

  6. SiC-CMC-Zircaloy-4 Nuclear Fuel Cladding Performance during 4-Point Tubular Bend Testing

    Energy Technology Data Exchange (ETDEWEB)

    IJ van Rooyen; WR Lloyd; TL Trowbridge; SR Novascone; KM Wendt; SM Bragg-Sitton

    2013-09-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE NE) established the Light Water Reactor Sustainability (LWRS) program to develop technologies and other solutions to improve the reliability, sustain the safety, and extend the life of current reactors. The Advanced LWR Nuclear Fuel Development Pathway in the LWRS program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. Recent investigations of potential options for “accident tolerant” nuclear fuel systems point to the potential benefits of silicon carbide (SiC) cladding. One of the proposed SiC-based fuel cladding designs being investigated incorporates a SiC ceramic matrix composite (CMC) as a structural material supplementing an internal Zircaloy-4 (Zr-4) liner tube, referred to as the hybrid clad design. Characterization of the advanced cladding designs will include a number of out-of-pile (nonnuclear) tests, followed by in-pile irradiation testing of the most promising designs. One of the out-of-pile characterization tests provides measurement of the mechanical properties of the cladding tube using four point bend testing. Although the material properties of the different subsystems (materials) will be determined separately, in this paper we present results of 4-point bending tests performed on fully assembled hybrid cladding tube mock-ups, an assembled Zr-4 cladding tube mock-up as a standard and initial testing results on bare SiC-CMC sleeves to assist in defining design parameters. The hybrid mock-up samples incorporated SiC-CMC sleeves fabricated with 7 polymer impregnation and pyrolysis (PIP) cycles. To provide comparative information; both 1- and 2-ply braided SiC-CMC sleeves were used in this development study. Preliminary stress simulations were performed using the BISON nuclear fuel performance code to show the stress distribution differences for varying lengths between loading points

  7. ORNL Pre-test Analyses of A Large-scale Experiment in STYLE

    International Nuclear Information System (INIS)

    Williams, Paul T.; Yin, Shengjun; Klasky, Hilda B.; Bass, Bennett Richard

    2011-01-01

    Oak Ridge National Laboratory (ORNL) is conducting a series of numerical analyses to simulate a large scale mock-up experiment planned within the European Network for Structural Integrity for Lifetime Management non-RPV Components (STYLE). STYLE is a European cooperative effort to assess the structural integrity of (non-reactor pressure vessel) reactor coolant pressure boundary components relevant to ageing and life-time management and to integrate the knowledge created in the project into mainstream nuclear industry assessment codes. ORNL contributes work-in-kind support to STYLE Work Package 2 (Numerical Analysis/Advanced Tools) and Work Package 3 (Engineering Assessment Methods/LBB Analyses). This paper summarizes the current status of ORNL analyses of the STYLE Mock-Up3 large-scale experiment to simulate and evaluate crack growth in a cladded ferritic pipe. The analyses are being performed in two parts. In the first part, advanced fracture mechanics models are being developed and performed to evaluate several experiment designs taking into account the capabilities of the test facility while satisfying the test objectives. Then these advanced fracture mechanics models will be utilized to simulate the crack growth in the large scale mock-up test. For the second part, the recently developed ORNL SIAM-PFM open-source, cross-platform, probabilistic computational tool will be used to generate an alternative assessment for comparison with the advanced fracture mechanics model results. The SIAM-PFM probabilistic analysis of the Mock-Up3 experiment will utilize fracture modules that are installed into a general probabilistic framework. The probabilistic results of the Mock-Up3 experiment obtained from SIAM-PFM will be compared to those results generated using the deterministic 3D nonlinear finite-element modeling approach. The objective of the probabilistic analysis is to provide uncertainty bounds that will assist in assessing the more detailed 3D finite

  8. Preparation of 3D Printed Divertor Mock-up Design and Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Park, Sung Dae; Kim, Dong Jun; Kim, Suk Kwon; Lee, Eo Hwak [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The divertor for fusion reactor is known to be able to remove the extreme heat flux up to 10 MW/m2 and the various type of divertors have been developed for enhancing the heat transfer such as hypervapotron, twisted tape insertion, screwed tube, and so on. In order to overcome this limitation, 3D printing method is considered to be used in the fusion reactor divertor design in present study. With the advantages of the 3D printing, the various shapes of the inner divertor cooling tube are investigated to enhance the turbulence of coolant and to reduce the pressure drop. The metallic powder of the fusion reactor candidate material is produced as the preliminary step for using in 3D printer. The material is a reduced activation ferritic-matensitic steel named as ARAA (Advanced Reduced Activation Alloy) which have been independently developed in Korea. Gas atomization method was used to make the spherical particles with average diameter of 100 μm. Several candidates were presented to achieve the excellent heat removal capacity and the low pressure drop. Thermal-hydraulic analysis was performed to confirm the effects of the inner cooling tube geometry with a conventional CFD code, ANSYS-CFX v14.5. The modified screw type called as a rail type twisted tube was presented through the optimization process. This complicated tube could be made by 3D printing technology. (metallic powder). Thermal-hydraulic analysis was conducted to compare the 3 type geometric divertor. A rail type twisted tube has good heat transfer performance in comparison with a conventional twisted tube. The pressure drop of a rail type twisted tube was reduced about 36% compared with a conventional twisted tube.

  9. Temperature field downstream of an heated bundle mock-up results for different power distribution

    International Nuclear Information System (INIS)

    Girard, J.P.; Buravand, Y.

    1982-10-01

    The aim of these peculiar experiments performed on the ML4 loop in ISPRA is to evaluate the characteristics of the temperature field over a length of 20 to 30 dias downstream of a rod bundle for different temperatures profiles at the bundle outlet. The final purpose of this work will be to establish either directly or through models whether it is possible or not to detect subassembly failures using suitable of the subassembly outlet temperature signal. 15 hours of digital and analog recording were taped for five different power distributions in the bundle. The total power dissipation remained constant during the whole run. Two flow rates and seven axial location were investigated. It is shown that the different temperature profiles produce slight differences in the variance and skewness of the temperature signal measured along the axis of the pipe over 20 dias

  10. Preparation of 3D Printed Divertor Mock-up Design and Fabrication

    International Nuclear Information System (INIS)

    Lee, Dong Won; Park, Sung Dae; Kim, Dong Jun; Kim, Suk Kwon; Lee, Eo Hwak

    2016-01-01

    The divertor for fusion reactor is known to be able to remove the extreme heat flux up to 10 MW/m2 and the various type of divertors have been developed for enhancing the heat transfer such as hypervapotron, twisted tape insertion, screwed tube, and so on. In order to overcome this limitation, 3D printing method is considered to be used in the fusion reactor divertor design in present study. With the advantages of the 3D printing, the various shapes of the inner divertor cooling tube are investigated to enhance the turbulence of coolant and to reduce the pressure drop. The metallic powder of the fusion reactor candidate material is produced as the preliminary step for using in 3D printer. The material is a reduced activation ferritic-matensitic steel named as ARAA (Advanced Reduced Activation Alloy) which have been independently developed in Korea. Gas atomization method was used to make the spherical particles with average diameter of 100 μm. Several candidates were presented to achieve the excellent heat removal capacity and the low pressure drop. Thermal-hydraulic analysis was performed to confirm the effects of the inner cooling tube geometry with a conventional CFD code, ANSYS-CFX v14.5. The modified screw type called as a rail type twisted tube was presented through the optimization process. This complicated tube could be made by 3D printing technology. (metallic powder). Thermal-hydraulic analysis was conducted to compare the 3 type geometric divertor. A rail type twisted tube has good heat transfer performance in comparison with a conventional twisted tube. The pressure drop of a rail type twisted tube was reduced about 36% compared with a conventional twisted tube

  11. Development of Digital Mock-Up for the Assessment of Dismantling Scenarios

    International Nuclear Information System (INIS)

    Kim, Sung-Kyun; Park, Hee-Sung; Lee, Kune-Woo; Jung, Chong-Hun

    2008-01-01

    As the number of superannuated research reactors and nuclear power plants increase, dismantling nuclear power facilities has become a big issue. However, decommissioning a nuclear facility is still a costly and possibly hazardous task. So prior to an actual decommission, what should be done foremost is to establish a proper procedure. Due to the fact that a significant difference in cost, exposure to a radiation, and safety might occur, a proper procedure is imperative for the entire engineering process. The purpose of this paper is to develop a system for evaluating the decommissioning scenarios logically and systematically. So a digital mockup system with functions such as a dismantling schedule, decommissioning costs, wastes, worker's exposure dose, and a radiation distribution was developed. Also on the basis of the quantitative information calculated from a DMU system and the data evaluated by decommissioning experts about qualitatively evaluating the items, the best decommissioning scenarios were established by using the analytic hierarchy process (AHP) method. Finally, the DMU was implemented in the thermal column of KRR-1 and adequate scenarios were provided after comparing and analyzing the two scenarios. In this paper, we developed the virtual environment of KRR-1 by using computer graphic technology and simulating the dismantling processes. The data-computing modules were also developed for quantitatively comparing the decommissioning scenarios. The decommissioning DMU system was integrated with both the VE system and the data-computing modules. In addition, we presented a decision-making method for selecting the best decommissioning scenario through the AHP. So the scenarios can be evaluated logically and quantitatively through the decommissioning DMU. As an implementation of the AHP, the plasma cutting scenario and the nibbler cutting scenario of the thermal column were prioritized. The fact that the plasma cutting scenario ranked the better than the nibbler cutting scenario is that the plasma scenario mostly got the higher scores than the nibbler scenario in the decommissioning cost and safety sections that have high weighting factors. Finally we decided that the plasma cutting scenario is appropriate to dismantle the thermal column. This study has a great meaning in that it can present a reliable scenario through the decommissioning DMU system while this work had only been done through a subjective evaluation in the past. The DMU system will be applied to the KRR-1 decommissioning project to obtain the best scenarios. We believe it will be a useful engineering tool for other nuclear facility decommissioning

  12. Experimental results of passive vibro-acoustic leak detection in SFR steam generator mock-up

    International Nuclear Information System (INIS)

    Moriot, J.; Gastaldi, O.; Maxit, L.; Guyader, J-L.; Perisse, J.; Migot, B.

    2013-06-01

    Regarding to GEN 4 context, it is necessary to fulfil the high safety standards for sodium fast reactors (SFR), particularly against water-sodium reaction which may occur in the steam generator units (SGU) in case of leak. This reaction can cause severe damages in the component in a short time. Detecting such a leak by visual in-sodium inspection is impossible because of sodium opacity. Hydrogen detection is then used but the time response of this method can be high in certain operating conditions. Active and passive acoustic leak detection methods were studied before SUPERPHENIX plant shutdown in 1997 to detect a water-into-sodium leak with a short time response. In the context of the new R and D studies for SFR, an innovative passive vibro-acoustic method is developed in the framework of a Ph.D. thesis to match with GEN 4 safety requirements. The method consists in assuming that a small leak emits spherical acoustic waves in a broadband frequency domain, which propagate in the liquid sodium and excite the SGU cylindrical shell. These spatially coherent waves are supposed to be buried by a spatially incoherent background noise. The radial velocities of the shell is measured by an array of accelerometers positioned on the external envelop of the SGU and a beam forming treatment is applied to increase the signal-to-noise ratio (SNR) and to detect and localize the acoustic source. Previous numerical experiments were achieved and promising results were obtained. In this paper, experimental results of the proposed passive vibro-acoustic leak detection are presented. The experiment consists in a cylindrical water-filled steel pipe representing a model of SGU shell without tube bundle. A hydro-phone emitting an acoustic signal is used to simulate an acoustic monopole. Spatially uncorrelated noise or water-flow induced shell vibrations are considered as the background noise. The beam-forming method is applied to vibration signals measured by a linear array of accelerometers on the shell. Satisfying results are obtained regarding to detection and localization of the source smothered by the background noise. (authors)

  13. De CAD/CAM-brug vervaardigd met behulp van een composiet mock-up

    NARCIS (Netherlands)

    Denissen, H.; Đozic, A.

    2010-01-01

    Het esthetische aspect van een brugconstructie in het front is niet altijd voorspelbaar. Dit geldt voor zowel conventionele bruggen als bruggen die computerondersteund worden ontworpen (computer-aided design, cad ) en geproduceerd (computer-aided manufacturing, cam ). Een cruciale factor is het

  14. Data management for digital mock-up visualization in immersive virtuel environment: application to car design

    OpenAIRE

    Paillot , Damien

    2004-01-01

    Virtual Reality techniques constitute new tools for the trades of the design of manufactured products and offer profits in term of innovation, time and of money. The industrial stake is the integration of these techniques in the product design process. These techniques must thus be interfaced with the various trade associations and be able to be in phase with the flow of the digital models updates. The technological problems relate to the management of the geometrical data produced by the eng...

  15. AMMU Automotive Mixed Mock-Up: Konzeption einer neuen Entwicklungsplattform für die Automobilindustrie

    OpenAIRE

    Geißel, Oliver

    2012-01-01

    Vor dem Hintergrund der großen Herausforderungen der Automobilindustrie zu Beginn des 21. Jahrhunderts beschäftigt sich diese Arbeit mit den Potentialen der Informationstechnologie, im speziellen der Mixed Reality, hinsichtlich der Anpassung von Methoden und Prozessen innerhalb der Produktentwicklung an die aktuellen Herausforderungen. Mit der Einführung von Mixed Reality in den Produktentwicklungsprozess wurde im Rahmen dieser Arbeit den Herausforderungen innerhalb der automobilen Entwick...

  16. Testing of high heat flux components manufactured by ENEA for ITER divertor

    International Nuclear Information System (INIS)

    Visca, Eliseo; Escourbiac, F.; Libera, S.; Mancini, A.; Mazzone, G.; Merola, M.; Pizzuto, A.

    2009-01-01

    ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R and D activities and in particular in the manufacturing of high heat flux plasma-facing components, such as the divertor targets. During the last years ENEA has manufactured actively cooled mock-ups by using different technologies, namely brazing, diffusion bonding and HIPping. A new manufacturing process that combines two main techniques PBC (Pre-Brazed Casting) and the HRP (Hot Radial Pressing) has been set up and widely tested. A full monoblock medium scale vertical target, having a straight CFC armoured part and a curved W armoured part, was manufactured using this process. The ultrasonic method was used for the non-destructive examinations performed during the manufacturing of the component, from the monoblock preparation up to the final mock-up assembling. The component was also examined by thermography on SATIR facility (CEA, France), afterwards it was thermal fatigue tested at FE200 (200 kW electron beam facility, CEA/AREVA France). The successful results of the thermal fatigue testing performed according the ITER requirements (10 MW/m 2 , 3000 cycles of 10 s on both CFC and W part, then 20/15 MW/m 2 , 2000 cycles of 10 s on CFC/W part, respectively) have confirmed that the developed process can be considerate a candidate for the manufacturing of monoblock divertor components. Furthermore, a 35-MW/m 2 Critical Heat Flux was measured at relevant thermal-hydraulics conditions at the end of the testing campaign. This paper reports the manufacturing route, the thermal fatigue testing results, the pre and post non-destructive examination and the destructive examination performed on the ITER vertical target medium scale mock-up. These activities were performed in the frame of EFDA contracts (04-1218 with CEA, 93-851 JN with AREVA and 03-1054 with ENEA).

  17. Investigation on cause of malfunction of wide range monitor (WRM) in high temperature engineering test reactor (HTTR). Sample tests and destructive tests

    International Nuclear Information System (INIS)

    Shinohara, Masanori; Saito, Kenji; Haga, Hiroyuki; Sasaki, Shinji; Katsuyama, Kozo; Motegi, Toshihiro; Takada, Kiyoshi; Higashimura, Keisuke; Fujii, Junichi; Ukai, Takayuki; Moriguchi, Yusuke

    2012-11-01

    An event, in which one of WRMs were disabled to detect the neutron flux in the reactor core, occurred during the period of reactor shut down of HTTR in March, 2010. The actual life time of WRM was unexpectedly shorter than the past developed life time. Investigation of the cause of the outage of WRM toward the recovery of the life time up to the past developed life is one of the issues to develop the technology basis of High Temperature Gas cooled Reactor (HTGR). Then, two experimental investigations were carried out to reveal the cause of the malfunction by specifying the damaged part causing the event in the WRM. One is an experiment using a mock-up sample test which strength degradation at assembly process and heat cycle to specify the damaged part in the WRM. The other is a destructive test in Fuels Monitoring Facility (FMF) to specify the damaged part in the WRM. This report summarized the results of the destructive test and the experimental investigation using the mock-up to reveal the cause of malfunction of WRM. (author)

  18. Installation Test of Cold Neutron Soruce In-pool Assembly

    International Nuclear Information System (INIS)

    Lee, Kye Hong; Choi, J.; Wu, S. I.; Kim, Y. K.; Cho, Y. G.; Lee, C. H.; Kim, K. R.

    2006-04-01

    Before installation of the final cold neutron source in-pool assembly (IPA) in the vertical CN hole at the HANARO, the research reactor, the installation test of IPA has been conducted in the CN hole of the reactor using a full-scaled mock-up in-pool assembly. The well-known cold neutron sources, being safely operated or being now constructed, had been constructed together with each research reactor; therefore, there was little limitation to obtain the optimal cold neutron source since a cold neutron source had been decided to be installed in the reactor from the beginning of the design for the reactor construction. Unlikely, the HANARO has been operated for 10 years so that we have got lots of design limitation in terms of the decisions in the optimal shape, size, minimal light-water gap, and adhesion degree to the CN beam tube, IPA installation tools, etc. for the construction of the CNS. Accordingly, the main objective of this test is to understand any potential problem or interference happened inside the reactor by installing the mock-up IPA and installation bracket. The outcomes from this test is reflected on the finalizing process of the IPA detail design

  19. Pseudodynamic tests on a full-scale 3-storey precast concrete building: Behavior of the mechanical connections and floor diaphragms

    OpenAIRE

    Bournas, Dionysios A.; Negro, Paolo; Molina, Francisco J.

    2013-01-01

    A full-scale three-storey precast building was tested under seismic conditions at the European Laboratory for Structural Assessment in the framework of the SAFECAST project. The unique research opportunity of testing a complete structural system was exploited to the maximum extent by subjecting the structure to a series of pseudodynamic (PsD) tests and by using four different structural layouts of the same mock-up, while 160 sensors were used to monitor the global and local response of each l...

  20. Modelling of bentonite-granite solutes transfer from an in situ full-scale experiment to simulate a deep geological repository (Grimsel Test Site, Switzerland)

    International Nuclear Information System (INIS)

    Buil, B.; Gomez, P.; Pena, J.; Garralon, A.; Turrero, M.J.; Escribano, A.; Sanchez, L.; Duran, J.M.

    2010-01-01

    Research highlights: → The FEBEX experiment is a 1:1 simulation of a high level waste disposal facility in crystalline rock according to the Spanish radwaste disposal concept. → Solute transfer processes occurrs at the bentonite-granite interface. → An increase of Cl and Na is observed in granitic water of the surrounding of the experiment. → Solute transfer does not affect the sealing and thermo-hydromechanical properties of the engineered barriers. → A diffusive transport of Cl and Na simulated by 1D transport modeling with an effective diffusion coefficient of D e ≅ 5.0 E-11 m 2 /s. - Abstract: The FEBEX experiment is a 1:1 simulation of a high level waste disposal facility in crystalline rock according to the Spanish radwaste disposal concept. This experiment has been performed in a gallery drilled in the underground laboratory Grimsel Test Site (Switzerland). Two boreholes parallel to the FEBEX drift were drilled 20 and 60 cm away from the granite-bentonite interface to provide data on potential bentonite-granite solutes transfer. Periodic sampling and analysis of the major ions showed: (a) the existence of solutes transfer from the bentonite porewater towards the granite groundwater, explaining the Cl - and Na + contents of the latter; (b) that the concentration of the natural tracers coming into the granite groundwater from the bentonite porewater increased over time. This bentonite-granite solutes transfer was modelled in order to predict the increase in the Cl - and Na + concentrations of the granite groundwater. The modelled results seem to confirm that the mechanism of solute migration in this scenario is that of diffusive transport. An effective diffusion coefficient of D e = 5 x 10 -11 m 2 /s was that which best fitted the data obtained.

  1. High-heat-flux testing of helium-cooled heat exchangers for fusion applications

    International Nuclear Information System (INIS)

    Youchison, D.L.; Izenson, M.G.; Baxi, C.B.; Rosenfeld, J.H.

    1996-01-01

    High-heat-flux experiments on three types of helium-cooled divertor mock-ups were performed on the 30-kW electron beam test system and its associated helium flow loop at Sandia National Laboratories. A dispersion-strengthened copper alloy (DSCu) was used in the manufacture of all the mock-ups. The first heat exchanger provides for enhanced heat transfer at relatively low flow rates and much reduced pumping requirements. The Creare sample was tested to a maximum absorbed heat flux of 5.8 MW/m 2 . The second used low pressure drops and high mass flow rates to achieve good heat removal. The GA specimen was tested to a maximum absorbed heat flux of 9 MW/m 2 while maintaining a surface temperature below 400 degree C. A second experiment resulted in a maximum absorbed heat flux of 34 MW/m 2 and surface temperatures near 533 degree C. The third specimen was a DSCu, axial flow, helium-cooled divertor mock-up filled with a porous metal wick which effectively increases the available heat transfer area. Low mass flow and high pressure drop operation at 4.0 MPa were characteristic of this divertor module. It survived a maximum absorbed heat flux of 16 MW/m 2 and reached a surface temperature of 740 degree C. Thermacore also manufactured a follow-on, dual channel porous metal-type heat exchanger, which survived a maximum absorbed heat flux of 14 MW/m 2 and reached a maximum surface temperature of 690 degree C. 11refs., 20 figs., 3 tabs

  2. 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)

  3. Testing external surface of fuel element tubes for power nuclear reactors

    International Nuclear Information System (INIS)

    Naugol'nykh, O.G.; Nelyubin, Yu.V.

    1987-01-01

    Optical methods are regarded perspective for discovery and detection of flaws of external surfaces of fuel element tubes. The TV method has highest information content among them. Two mock-ups of facilities based on the TV method using a ''dissector'' type TV device and a TV tube with charge accumulation (vidikon) have been developed. It is concluded that complex testing - combination of ultrasonic, photoelectric and TV methods in a facility is necessary for discovery and analysis of the whole variety of flaws, though sensitivity of the TV method is enough for disclosure of all the main defects

  4. Qualification test for the flexible receiver

    International Nuclear Information System (INIS)

    Keller, C.M.

    1994-01-01

    This document provides the test plan and procedures to certify and design verify the 42 inch and 4--6 inch Flexible Receiver (FR) is a safety class 3 system. Verification of the design will be handled in two parts. The first part will be to show that it meets design requirements set forth by documents and the second part will perform test(s) to verify its operational aspects. To qualify the design of the FR systems for field use this test will demonstrate environmentally safe removal of a Tank Farm pump mock-up from a Tank Farm riser mock-up. Testing will also demonstrate the performance of supporting equipment. The FR and the Secondary Bagging (SB) equipment shall be tested to verify successful operation of the equipment to the following criteria: The FR can be placed on a riser and connections made to the supporting equipment; The FR bag can accept equipment and be successfully sealed; The SB system encases the seal of the primary FR bag; The flexible bag(s) do not tear and maintain integrity during the entire test; The FR control system operates in the fail safe forced sequence mode; The FR control system will operate in the manual override mode (out of sequence operations); The CCTV Video system monitors and records the removal of the test item; The spray wash system operates without leaks and effectively provides coverage; The item being removed can be reinserted to a depth of 8 feet and the bag reinstalled onto the vertical bag supports; and The system prohibits momentary mechanical fluctuations due to the application of system power, including power interruptions

  5. Buffer Construction Methodology in Demonstration Test For Cavern Type Disposal Facility

    International Nuclear Information System (INIS)

    Yoshihiro, Akiyama; Takahiro, Nakajima; Katsuhide, Matsumura; Kenji, Terada; Takao, Tsuboya; Kazuhiro, Onuma; Tadafumi, Fujiwara

    2009-01-01

    A number of studies concerning a cavern type disposal facility have been carried out for disposal of low level radioactive waste mainly generated by power plant decommissioning in Japan. The disposal facility is composed of an engineered barrier system with concrete pit and bentonite buffer, and planed to be constructed in sub-surface 50 - 100 meters depth. Though the previous studies have mainly used laboratory and mock-up tests, we conducted a demonstration test in a full-size cavern. The main objectives of the test were to study the construction methodology and to confirm the quality of the engineered barrier system. The demonstration test was planned as the construction of full scale mock-up. It was focused on a buffer construction test to evaluate the construction methodology and quality control in this paper. Bentonite material was compacted to 1.6 Mg/m 3 in-site by large vibrating roller in this test. Through the construction of the buffer part, a 1.6 Mg/m 3 of the density was accomplished, and the data of workability and quality is collected. (authors)

  6. 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).

  7. Full scale test platform for European TBM systems integration and maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Vála, Ladislav, E-mail: ladislav.vala@cvrez.cz; Reungoat, Mathieu; Vician, Martin

    2016-11-01

    Highlights: • A platform for EU-TBS maintenance and integration tests is described. • Its modular design allows adaptation to non-EU TBSs. • Assembling of the facility will be followed by initial tests in 2016. - Abstract: This article deals with description and current status of a project of a non-nuclear, full size (1:1 scale) test platform dedicated to tests, optimization and validation of integration and maintenance operations for the European TBM systems in the ITER port cell #16. The facility called TBM platform reproduces the ITER port cell #16 and port interspace with all the relevant interfaces and mock-ups of the corresponding main components. Thanks to the modular design of the platform, it is possible to adapt or change completely the interfaces in the future if needed or required according to the updated configuration of TBSs. In the same way, based on customer requirements, it will be possible to adapt the interfaces and piping inside the mock-ups in order to represent also the other, non-EU configurations of TBM systems designed for port cells #02 and #18. Construction of this test platform is realized and funded within the scope of the SUSEN project.

  8. Final Design and Installation of the ''In Situ'' test at GRIMSEL

    International Nuclear Information System (INIS)

    Fuentes-Cantillana, J. L.; Garcia-Sineriz, J. L.

    1998-01-01

    The aim of the FEBEX project (Full-Scale Engineered Barriers Experiment) is the study of the near-field for a repository of high-level radioactive waste (HLW) in crystalline rock. The experiment has three major parts: 1) an in situ test, in natural conditions and at full scale; 2) a mack-up test, at almost full scale, and 3) a set of experimental laboratory tests, to complement the information from the two large-scale tests. The experiment is based on the Spanish reference concept for crystalline rock, in which the waste canisters are placed in horizontal drifts surrounded by a clay formed from highly-compacted bentonite blocks. The complete project, with about seven years of duration (1994-2001), has been divided into four sequential stages, defined by the main features of each stage of the two large-scale tests. This report is part of the pre-operational stage (1994-1996). (Author)

  9. High heat flux tests on beryllium and beryllium-copper joints

    International Nuclear Information System (INIS)

    Roedig, M.; Duwe, R.; Linke, J.; Schuster, A.

    1997-01-01

    A large test program has been set up to evaluate the performance of beryllium as a plasma facing material for the divertor in thermonuclear fusion devices. Simulation of steady state heat loads of 5 MWm -2 and above on actively cooled divertor modules, and off-normal plasma conditions with energy densities in the range 1-7 MJm -2 , have been investigated. Thermal shock tests were carried out with the ITER reference grade S65-C and several Russian grades of beryllium. At incident energies up to 7 MJm -2 the best erosion behaviour is observed for S65-C and for TGP-56. Steady state heating tests with actively cooled Be/Cu mock-ups were performed at incident powers of up to 5.8 MWm -2 . All samples investigated in these tests did not show any indications of failure. A Be/Cu mock-ups with Incusil braze was loaded in thermal fatigue up to 500 cycles at an incident power of 4.8 MWm -2 . Up to the end of the experiment no temperature increase of the surface and no indication of failure was observed. (orig.)

  10. Thermal fatigue equipment to test joints of materials for high heat flux components

    International Nuclear Information System (INIS)

    Visca, E.; Libera, S.; Orsini, A.; Riccardi, B.; Sacchetti, M.

    2000-01-01

    The activity, carried out in the framework of an ITER divertor task, was aimed at defining a suitable method in order to qualify junctions between armour materials and heat sink of plasma-facing components (PFCs) mock-ups. An equipment able to perform thermal fatigue testing by electrical heating and active water-cooling was constructed and a standard for the sample was defined. In this equipment, during operation cycles, two samples are heated by thermal contact up to a relevant temperature value (350 deg. C) and then the water flow is switched on, thus producing fast cooling with time constants and gradients close to the real operating conditions. The equipment works with a test cycle of about 60 s and is suitable for continuous operation. A complete test consists of about 10000 cycles. After the assembling, the equipment and the control software were optimized to obtain a good reliability. Preliminary tests on mock-ups with flat CFC tiles joined to copper heat sink were performed. Finite-elements calculations were carried out in order to estimate the value of the thermal stresses arising close to the joint under the transient conditions that are characteristic of this equipment

  11. Pretreatment Engineering Platform Phase 1 Final Test Report

    International Nuclear Information System (INIS)

    Kurath, Dean E.; Hanson, Brady D.; Minette, Michael J.; Baldwin, David L.; Rapko, Brian M.; Mahoney, Lenna A.; Schonewill, Philip P.; Daniel, Richard C.; Eslinger, Paul W.; Huckaby, James L.; Billing, Justin M.; Sundar, Parameshwaran S.; Josephson, Gary B.; Toth, James J.; Yokuda, Satoru T.; Baer, Ellen B.K.; Barnes, Steven M.; Golovich, Elizabeth C.; Rassat, Scot D.; Brown, Christopher F.; Geeting, John G.H.; Sevigny, Gary J.; Casella, Amanda J.; Bontha, Jagannadha R.; Aaberg, Rosanne L.; Aker, Pamela M.; Guzman-Leong, Consuelo E.; Kimura, Marcia L.; Sundaram, S.K.; Pires, Richard P.; Wells, Beric E.; Bredt, Ofelia P.

    2009-01-01

    Pacific Northwest National Laboratory (PNNL) was tasked by Bechtel National Inc. (BNI) on the River Protection Project, Hanford Tank Waste Treatment and Immobilization Plant (RPP-WTP) project to conduct testing to demonstrate the performance of the WTP Pretreatment Facility (PTF) leaching and ultrafiltration processes at an engineering-scale. In addition to the demonstration, the testing was to address specific technical issues identified in Issue Response Plan for Implementation of External Flowsheet Review Team (EFRT) Recommendations - M12, Undemonstrated Leaching Processes. Testing was conducted in a 1/4.5-scale mock-up of the PTF ultrafiltration system, the Pretreatment Engineering Platform (PEP). Parallel laboratory testing was conducted in various PNNL laboratories to allow direct comparison of process performance at an engineering-scale and a laboratory-scale. This report presents and discusses the results of those tests.

  12. Pretreatment Engineering Platform Phase 1 Final Test Report

    Energy Technology Data Exchange (ETDEWEB)

    Kurath, Dean E.; Hanson, Brady D.; Minette, Michael J.; Baldwin, David L.; Rapko, Brian M.; Mahoney, Lenna A.; Schonewill, Philip P.; Daniel, Richard C.; Eslinger, Paul W.; Huckaby, James L.; Billing, Justin M.; Sundar, Parameshwaran S.; Josephson, Gary B.; Toth, James J.; Yokuda, Satoru T.; Baer, Ellen BK; Barnes, Steven M.; Golovich, Elizabeth C.; Rassat, Scot D.; Brown, Christopher F.; Geeting, John GH; Sevigny, Gary J.; Casella, Amanda J.; Bontha, Jagannadha R.; Aaberg, Rosanne L.; Aker, Pamela M.; Guzman-Leong, Consuelo E.; Kimura, Marcia L.; Sundaram, S. K.; Pires, Richard P.; Wells, Beric E.; Bredt, Ofelia P.

    2009-12-23

    Pacific Northwest National Laboratory (PNNL) was tasked by Bechtel National Inc. (BNI) on the River Protection Project, Hanford Tank Waste Treatment and Immobilization Plant (RPP-WTP) project to conduct testing to demonstrate the performance of the WTP Pretreatment Facility (PTF) leaching and ultrafiltration processes at an engineering-scale. In addition to the demonstration, the testing was to address specific technical issues identified in Issue Response Plan for Implementation of External Flowsheet Review Team (EFRT) Recommendations - M12, Undemonstrated Leaching Processes.( ) Testing was conducted in a 1/4.5-scale mock-up of the PTF ultrafiltration system, the Pretreatment Engineering Platform (PEP). Parallel laboratory testing was conducted in various PNNL laboratories to allow direct comparison of process performance at an engineering-scale and a laboratory-scale. This report presents and discusses the results of those tests.

  13. A framework for the testing and validation of the I and C system based on a simulator

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Jun; Kwon, Kee Choon; Lee, Jang Soo [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The I and C system for a nuclear power plant should be developed as a prototype or mock-up from the concept phase of the development process, and the function and performance of the computer system also have to be tested and validated. If possible, the developed prototype or mock-up could receive the signals of a normal or abnormal operation status of a nuclear power plant and generate the proper requirement output signal. Using these processes, it can be verified that the status of a plant is changed to the design state or the state needed by the plant operator. A simulation-based conformity evaluation platform is an environment that can automate the testing and validation actions. A traditional testing and validation method defines the static test requirements and extracts the input data from the defined requirement using IO signal generation devices. On the contrary, a simulation-based test method can generate the real calculated input data from a simulator and send the signals to the test devices directly. In this paper, we developed a framework that can conduct a conformity evaluation based on a simulator and implement the communication and monitoring program.

  14. Experimental results and validation of a method to reconstruct forces on the ITER test blanket modules

    International Nuclear Information System (INIS)

    Zeile, Christian; Maione, Ivan A.

    2015-01-01

    Highlights: • An in operation force measurement system for the ITER EU HCPB TBM has been developed. • The force reconstruction methods are based on strain measurements on the attachment system. • An experimental setup and a corresponding mock-up have been built. • A set of test cases representing ITER relevant excitations has been used for validation. • The influence of modeling errors on the force reconstruction has been investigated. - Abstract: In order to reconstruct forces on the test blanket modules in ITER, two force reconstruction methods, the augmented Kalman filter and a model predictive controller, have been selected and developed to estimate the forces based on strain measurements on the attachment system. A dedicated experimental setup with a corresponding mock-up has been designed and built to validate these methods. A set of test cases has been defined to represent possible excitation of the system. It has been shown that the errors in the estimated forces mainly depend on the accuracy of the identified model used by the algorithms. Furthermore, it has been found that a minimum of 10 strain gauges is necessary to allow for a low error in the reconstructed forces.

  15. DSCu/SS joining techniques development and testing

    International Nuclear Information System (INIS)

    Sato, Satoshi; Hatano, Toshihisa; Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Takatsu, Hideyuki

    1998-01-01

    Joining techniques of alumina dispersion strengthened copper alloy (DSCu) and type 316L stainless steel (SS) has been investigated aiming at applying to the fabrication of the ITER first wall/blanket. As the joining method, Hot Isostatic Pressing (HIP) of solid plates and/or blocks has been pursued. By a screening test including HIP temperatures of 980-1050degC, it was concluded that the HIP temperature of 1050degC would be optimum for the simultaneous HIPping of DSCu/DSCu, DSCu/SS and SS/SS. With DSCu/SS joint specimens HIPped at 1050degC, tensile, impact, fatigue, crack propagation, and fracture toughness tests were performed as well as mechanical test of structural model with one SS circular tube embedded. Typically, the properties of the joints were almost the same as those of DSCu or SS base metal with the same heat treatment of the HIP process, thus good joints were obtained, though parts of properties were decreased at elevated test temperature. Typical results of the mechanical test of structural mode indicated that a crack initiated at the inner surface of the SS tube under cyclic operation, and the lifetime of the first wall structure could be evaluated by existing SS fatigue data. Two HIPped first wall panel mock-ups were successfully fabricated with built-in coolant tubes: one was 300 mm long and the other 800 mm long. The former was thermo-mechanically tested with high heat fluxes corresponding to the ITER operation conditions. The mock-up showed good heat removal performance during the high heat flux tests. In addition, there were no cracks and delaminations found at HIPped interfaces by microscopic observation after all tests. Ultrasonic testing have been tried as a non-destructive examination method, and detectable defect size at SS/SS, DSCu/DSCu and DSCu/SS joint interfaces were estimated. (author)

  16. Testing remotely operated module technique for Wackersdorf reprocessing plant at Lahde test rig

    International Nuclear Information System (INIS)

    Leister, P.; Schroeder, G.; Boehme, G.

    1986-01-01

    The FEMO technique represents a plant concept which makes it possible to carry out the repair of high and medium activity wet chemical stages of the process by remote handling without direct access by staff. For this purpose, the apparatus of this step of the process is arranged modularly in large cells, so that movable large handling devices such as cranes and manipulator systems can replace process components subject to wear via the process modules. The machine room of the former coal-fired power station Heyden I at Lahde was, after removal of the turbines and generators, converted to a hall in which the following test areas were accommodated: FEMO cell section with 10 positions for module, cell wall mock-up and wall penetration, module mounting area, module measuring position, workplace for service area, training position, welding position and FEMO control position. (orig./HP) [de

  17. Operability test procedure [Tank] 241-SY-101 equipment removal system

    International Nuclear Information System (INIS)

    Mast, J.C.

    1994-01-01

    The 241-SY-101 equipment removal system (ERS) consists of components, equipment, instrumentation and procedures that will provide the means to disconnect, retrieve, contain, load and transport the Mitigation Pump Assembly (MPA) from waste Tank 241-SY-101 to the Central Waste Complex (CWC). The Operability Test Procedure (OTP) will test the interfaces between ERS components and will rehearse the procedure for MPA removal and transportation to the extent they can be mocked-up at the CTF (Cold Test Facility). At the conclusion of the OTP, the ERS components and equipment will be removed from the CTF, entered into the Component Based Recall System (CBRS), and stored until needed for actual MPA removal and transportation

  18. Design manufacturing and thermo-mechanical testing of a relevant size mono block divertor prototype

    International Nuclear Information System (INIS)

    Cardella, A.; Vieider, G.; Di Pietro, E.; Orsini, A.; Febvre, M.; Guerreschi, U.; Reheis, N.; Bruno, L.

    1994-01-01

    Following a technological development of joining techniques between carbon fibre composite tiles and metallic tubes, and the manufacturing and testing of small size actively cooled mock-ups, a relevant size divertor prototype has been designed, manufactured and tested. The prototype consisted of a series of metallic tubes surrounded by CFC tiles, cooling collectors and a supporting system representative of a divertor dump plate for high power reactors. The tubes have been preliminary tested at the CEA 200 kW electron beam facility with uniform fluxes up to 5 MW/m 2 to select the best five tubes, which together with a sixth non tested tube have been then assembled to form the prototype. This has been tested at the JET high power neutral beam injector test facility. After screening tests the prototype has been subjected to thermal cycling at more than 15 MW/m 2 . (author) 12 refs.; 4 figs

  19. Core Seismic Tests for a Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, J. H

    2007-01-15

    This report describes the results of the comparison of the core seismic responses between the test and the analysis for the reduced core mock-up of a sodium-cooled fast reactor to verify the FAMD (Fluid Added Mass and Damping) code and SAC-CORE (Seismic Analysis Code for CORE) code, which implement the application algorithm of a consistent fluid added mass matrix including the coupling terms. It was verified that the narrow fluid gaps between the duct assemblies significantly affect the dynamic characteristics of the core duct assemblies and it becomes stronger as a number of duct increases within a certain level. As conclusion, from the comparison of the results between the tests and the analyses, it is verified that the FAMD code and the SAC-CORE code can give an accurate prediction of a complex core seismic behavior of the sodium-cooled fast reactor.

  20. Development and testing of 140 GHz absorber coatings for the water baffle of W7-X cryopumps

    International Nuclear Information System (INIS)

    Floristan, Miriam; Mueller, Philipp; Gebhardt, Andreas; Killinger, Andreas; Gadow, Rainer; Cardella, Antonio; Li, Chuanfei; Stadler, Reinhold; Zangl, Guenter; Hirsch, Matthias; Laqua, Heinrich P.; Kasparek, Walter

    2011-01-01

    Due to the relatively high strayfield radiation (140 GHz) from the electron cyclotron radio frequency heating system to which the W7-X cryopumps are expected to be subjected, coating systems acting as an efficient absorber for 140 GHz radiation have been developed for the water-cooled baffle shield in order to reduce the thermal load on the liquid N shield and the liquid He cryopanel. Several types of oxide ceramic coatings were applied on planar copper substrates by Atmospheric Plasma Spraying. The influence of the process parameters on the coating properties and microwave absorbing capability was analysed. It was found that film thickness and microstructure of the sprayed coatings have a significant influence on microwave absorption behaviour. For Al 2 O 3 /TiO 2 coatings, absorption values over 90% were obtained for the 140 GHz probing beam. After optimisation of the coating structure for maximum microwave absorption, the coating procedure was adapted by special robot trajectories to the complex water baffle geometry. The selected spray parameters and kinematics were then used for the complete coating of four mock-ups, which have been tested in the W7-X strayfield test facility Mistral. The mock-ups showed absorption values of 75%.

  1. Development and testing of 140 GHz absorber coatings for the water baffle of W7-X cryopumps

    Energy Technology Data Exchange (ETDEWEB)

    Floristan, Miriam, E-mail: miriam.floristan@gsame.uni-stuttgart.de [Graduate School for advanced Manufacturing Engineering (GSaME), Universitaet Stuttgart (Germany); Institute for Manufacturing Technologies of Ceramic Components and Composites (IMTCCC), Universitaet Stuttgart, Allmandring 7 b, D-70569 Stuttgart (Germany); Mueller, Philipp; Gebhardt, Andreas; Killinger, Andreas; Gadow, Rainer [Institute for Manufacturing Technologies of Ceramic Components and Composites (IMTCCC), Universitaet Stuttgart, Allmandring 7 b, D-70569 Stuttgart (Germany); Cardella, Antonio [European Commission c/o Wendelstein 7X, Boltzmannstasse 2, D-85748 Garching (Germany); Li, Chuanfei; Stadler, Reinhold; Zangl, Guenter; Hirsch, Matthias; Laqua, Heinrich P. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Assoc., Wendelsteinstrasse 1, D-17491 Greifswald (Germany); Kasparek, Walter [Institut fuer Plasmaforschung, Universitaet Stuttgart, Pfaffenwaldring 31, D-70569 Stuttgart (Germany)

    2011-10-15

    Due to the relatively high strayfield radiation (140 GHz) from the electron cyclotron radio frequency heating system to which the W7-X cryopumps are expected to be subjected, coating systems acting as an efficient absorber for 140 GHz radiation have been developed for the water-cooled baffle shield in order to reduce the thermal load on the liquid N shield and the liquid He cryopanel. Several types of oxide ceramic coatings were applied on planar copper substrates by Atmospheric Plasma Spraying. The influence of the process parameters on the coating properties and microwave absorbing capability was analysed. It was found that film thickness and microstructure of the sprayed coatings have a significant influence on microwave absorption behaviour. For Al{sub 2}O{sub 3}/TiO{sub 2} coatings, absorption values over 90% were obtained for the 140 GHz probing beam. After optimisation of the coating structure for maximum microwave absorption, the coating procedure was adapted by special robot trajectories to the complex water baffle geometry. The selected spray parameters and kinematics were then used for the complete coating of four mock-ups, which have been tested in the W7-X strayfield test facility Mistral. The mock-ups showed absorption values of 75%.

  2. Achievements in the development of the Water Cooled Solid Breeder Test Blanket Module of Japan to the milestones for installation in ITER

    International Nuclear Information System (INIS)

    Tsuru, Daigo; Tanigawa, Hisashi; Hirose, Takanori; Mohri, Kensuke; Seki, Yohji; Enoeda, Mikio; Ezato, Koichiro; Suzuki, Satoshi; Nishi, Hiroshi; Akiba, Masato

    2009-01-01

    As the primary candidate of ITER Test Blanket Module (TBM) to be tested under the leadership of Japan, a water cooled solid breeder (WCSB) TBM is being developed. This paper shows the recent achievements towards the milestones of ITER TBMs prior to the installation, which consist of design integration in ITER, module qualification and safety assessment. With respect to the design integration, targeting the detailed design final report in 2012, structure designs of the WCSB TBM and the interfacing components (common frame and backside shielding) that are placed in a test port of ITER and the layout of the cooling system are presented. As for the module qualification, a real-scale first wall mock-up fabricated by using the hot isostatic pressing method by structural material of reduced activation martensitic ferritic steel, F82H, and flow and irradiation test of the mock-up are presented. As for safety milestones, the contents of the preliminary safety report in 2008 consisting of source term identification, failure mode and effect analysis (FMEA) and identification of postulated initiating events (PIEs) and safety analyses are presented.

  3. Selection of dissolution process for spent fuels and preparation of corrosion test solution simulated to dissolver (contract research)

    International Nuclear Information System (INIS)

    Motooka, Takafumi; Terakado, Shogo; Koya, Toshio; Hamada, Shozo; Kiuchi, Kiyoshi

    2001-03-01

    In order to evaluate the reliability of reprocessing equipment materials used in the Rokkasho Reprocessing Plant, we have proceeded a mock-up test and laboratory tests for getting corrosion parameters. In a dissolver made of zirconium, the simulation of test solutions to the practical solution which includes the high concentration of radioactive elements such as FP and TRU is one of the important issues with respect to the life prediction. On this experiment, the dissolution process of spent fuels and the preparation of test solution for evaluating the corrosion resistance of dissolver materials were selected. These processes were tested in the No.3 cell of WASTEF. The test solution for corrosion tests was prepared by adjusting the uranium and nitric acid concentrations. (author)

  4. PETER loop. Multifunctional test facility for thermal hydraulic investigations of PWR fuel elements

    International Nuclear Information System (INIS)

    Ganzmann, I.; Hille, D.; Staude, U.

    2009-01-01

    The reliable fuel element behavior during the complete fuel cycle is one of the fundamental prerequisites of a safe and efficient nuclear power plant operation. The fuel element behavior with respect to pressure drop and vibration impact cannot be simulated by means of fluid-structure interaction codes. Therefore it is necessary to perform tests using fuel element mock-ups (1:1). AREVA NP has constructed the test facility PETER (PWR fuel element tests in Erlangen) loop. The modular construction allows maximum flexibility for any type of fuel elements. Modern measuring instrumentation for flow, pressure and vibration characterization allows the analysis of cause and consequences of thermal hydraulic phenomena. PETER loop is the standard test facility for the qualification of dynamic fuel element behavior in flowing fluid and is used for failure mode analysis.

  5. A Study on the Dynamic Analysis of the Nuclear Fuel Test Rig Using 1-Way Fluid-Structure Coupled Analysis

    International Nuclear Information System (INIS)

    Yang, Tae-Ho; Hong, Jin-Tae; Ahn, Sung-Ho; Joung, Chang-Young; Heo, Sung-Ho; Jang, Seo-Yun

    2015-01-01

    1-way fluid-structure coupled analysis is used to estimate the dynamic characteristic of the fuel test rig. the motion at the bottom of the test rig is confirmed. The maximum deformation of the test rig is 0.11 mm. The structural integrity of the test rig is performed by using the comparison with the Von-mises stress of the analysis and yield stress of the material. It is evaluated that the motion at the bottom of the test rig is able to cause other structural problem. Using the 2-way fluid-structural coupled analysis, the structural integrity of the test rig will be performed in further paper. The cooling water with specific flow rate was flowed in the nuclear fuel test rig. The structural integrity of the test rig was affected by the vibration. The fluid-induced vibration test had to be performed to obtain the amplitude of the vibration on the structure. Various test systems was developed. Flow-induced vibration and pressure drop experimental tester was developed in Korea Atomic Energy Research Institute. The vibration test with high fluid flow rate was difficult by the tester. To generate the nuclear fuel test environment, coolant flow simulation system was developed. The scaled nuclear fuel test was able to be performed by the simulation system. The mock-up model of the test rig was used in the simulation system. The mock-up model in the simulation system was manufactured with scaled down full model. In this paper, the fluid induced vibration characteristic of the full model in the nuclear fuel test is studied. The hydraulic pressure on the velocity of the fluid was calculated. The static structure analysis was performed by using the pressure. The structural integrity was assessed using the results of the analysis

  6. Simulation-based Testing of Control Software

    Energy Technology Data Exchange (ETDEWEB)

    Ozmen, Ozgur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Nutaro, James J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sanyal, Jibonananda [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Olama, Mohammed M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-02-10

    It is impossible to adequately test complex software by examining its operation in a physical prototype of the system monitored. Adequate test coverage can require millions of test cases, and the cost of equipment prototypes combined with the real-time constraints of testing with them makes it infeasible to sample more than a small number of these tests. Model based testing seeks to avoid this problem by allowing for large numbers of relatively inexpensive virtual prototypes that operate in simulation time at a speed limited only by the available computing resources. In this report, we describe how a computer system emulator can be used as part of a model based testing environment; specifically, we show that a complete software stack including operating system and application software - can be deployed within a simulated environment, and that these simulations can proceed as fast as possible. To illustrate this approach to model based testing, we describe how it is being used to test several building control systems that act to coordinate air conditioning loads for the purpose of reducing peak demand. These tests involve the use of ADEVS (A Discrete Event System Simulator) and QEMU (Quick Emulator) to host the operational software within the simulation, and a building model developed with the MODELICA programming language using Buildings Library and packaged as an FMU (Functional Mock-up Unit) that serves as the virtual test environment.

  7. Development of corrosion testing equipment under heat transfer and irradiation conditions to evaluate corrosion resistance of materials used in acid recovery evaporator. Contract research

    International Nuclear Information System (INIS)

    Motooka, Takafumi; Numata, Masami; Kiuchi, Kiyoshi

    2002-01-01

    We have been evaluated the safety for corrosion of various metals applied to acid recovery evaporators by the mock-up tests using small scaled equipment and the reference tests in laboratories with small specimens. These tests have been conducted under-radioactive environment. The environment in practical reprocessing plants has many radioactive species. Therefore, the effect of irradiation on corrosion should be evaluated in detail. In this study, we have developed the corrosion testing equipment, which is employed to simulate environments in the acid recovery evaporators. This report describes the specification of corrosion testing equipment and the results of primary, reference and hot tests. Using the equipment, the corrosion test under heat transfer and irradiation conditions have been carried out for 930 hours in safety. It is expectable that useful corrosion test data in radioactive environment are accumulated with this equipment in future, and help the adequate choice of corrosion test condition in laboratories. (author)

  8. Inverse hydrochemical models of aqueous extracts tests

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, L.; Samper, J.; Montenegro, L.

    2008-10-10

    Aqueous extract test is a laboratory technique commonly used to measure the amount of soluble salts of a soil sample after adding a known mass of distilled water. Measured aqueous extract data have to be re-interpreted in order to infer porewater chemical composition of the sample because porewater chemistry changes significantly due to dilution and chemical reactions which take place during extraction. Here we present an inverse hydrochemical model to estimate porewater chemical composition from measured water content, aqueous extract, and mineralogical data. The model accounts for acid-base, redox, aqueous complexation, mineral dissolution/precipitation, gas dissolution/ex-solution, cation exchange and surface complexation reactions, of which are assumed to take place at local equilibrium. It has been solved with INVERSE-CORE{sup 2D} and been tested with bentonite samples taken from FEBEX (Full-scale Engineered Barrier EXperiment) in situ test. The inverse model reproduces most of the measured aqueous data except bicarbonate and provides an effective, flexible and comprehensive method to estimate porewater chemical composition of clays. Main uncertainties are related to kinetic calcite dissolution and variations in CO2(g) pressure.

  9. 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

  10. Environmental Risk Evaluation System (ERES) for Offshore Wind - Mock-Up of ERES, Fiscal Year 2010 Progress Report

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Richard M.; Copping, Andrea E.; Van Cleve, Frances B.

    2010-11-01

    The Environmental Risk Evaluation System (ERES) has been created to set priorities among the environmental risks from offshore wind development. This report follows the conceptual design for ERES and shows what the system would look like, using a web interface created as part of a Knowledge Management System (KMS) for offshore wind. The KMS, called Zephyrus, and ERES for offshore wind, will be populated and made operational in a later phase of the project.

  11. Neutron Diffraction Residual Strain Tensor Measurements Within The Phase IA Weld Mock-up Plate P-5

    Energy Technology Data Exchange (ETDEWEB)

    Hubbard, Camden R [ORNL

    2011-09-01

    Oak Ridge National Laboratory (ORNL) has worked with NRC and EPRI to apply neutron and X-ray diffraction methods to characterize the residual stresses in a number of dissimilar metal weld mockups and samples. The design of the Phase IA specimens aimed to enable stress measurements by several methods and computational modeling of the weld residual stresses. The partial groove in the 304L stainless steel plate was filled with weld beads of Alloy 82. A summary of the weld conditions for each plate is provided in Table 1. The plates were constrained along the long edges during and after welding by bolts with spring-loaded washers attached to the 1-inch thick Al backing plate. The purpose was to avoid stress relief due to bending of the welded stainless steel plate. The neutron diffraction method was one of the methods selected by EPRI for non-destructive through thickness strain and stress measurement. Four different plates (P-3 to P-6) were studied by neutron diffraction strain mapping, representing four different welding conditions. Through thickness neutron diffraction strain mappings at NRSF2 for the four plates and associated strain-free d-zero specimens involved measurement along seven lines across the weld and at six to seven depths. The mountings of each plate for neutron diffraction measurements were such that the diffraction vector was parallel to each of the three primary orthogonal directions of the plate: two in-plane directions, longitudinal and transverse, and the direction normal to the plate (shown in left figure within Table 1). From the three orthogonal strains for each location, the residual stresses along the three plate directions were calculated. The principal axes of the strain and stress tensors, however, need not necessarily align with the plate coordinate system. To explore this, plate P-5 was selected for examination of the possibility that the principal axes of strain are not along the sample coordinate system axes. If adequate data could be collected the goal would be to determine the strain tensor's orientation and magnitude of strain along each principle axis direction.

  12. Modelling of steady state erosion of CFC actively water-cooled mock-up for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Ogorodnikova, O.V. [Departement de Recherches sur la Fusion Controlee, Association Euratom-CEA, CEA-Cadarache, F-13108 Saint Paul Lez Durance cedex (France)], E-mail: igra32@rambler.ru

    2008-04-15

    Calculations of the physical and chemical erosion of CFC (carbon fibre composite) monoblocks as outer vertical target of the ITER divertor during normal operation regimes have been done. Off-normal events and ELM's are not considered here. For a set of components under thermal and particles loads at glancing incident angle, variations in the material properties and/or assembly of defects could result in different erosion of actively-cooled components and, thus, in temperature instabilities. Operation regimes where the temperature instability takes place are investigated. It is shown that the temperature and erosion instabilities, probably, are not a critical point for the present design of ITER vertical target if a realistic variation of material properties is assumed, namely, the difference in the thermal conductivities of the neighbouring monoblocks is 20% and the maximum allowable size of a defect between CFC armour and cooling tube is +/-90{sup o} in circumferential direction from the apex.

  13. Relation between medium fluid temperature and centroid subchannel temperatures of a nuclear fuel bundle mock-up

    International Nuclear Information System (INIS)

    Carvalho Tofani, P. de.

    1986-01-01

    The subchannel method used in nuclear fuel bundle thermal-hydraulic analysis lies in the statement that subchannel fluid temperatures are taken at mixed mean values. However, the development of mixing correlations and code assessment procedures are, sometimes in the literature, based upon the assumption of identity between lumped and local (subchannel centroid) temperature values. The present paper is concerned with the presentation of an approach for correlating lumped to centroid subchannel temperatures, based upon previously formulated models by the author, applied, applied to a nine heated tube bundle experimental data set. (Author) [pt

  14. Relation between medium fluid temperature and centroid subchannel temperatures of a nuclear fuel bundle mock-up

    International Nuclear Information System (INIS)

    Carvalho Tofani, P. de.

    1986-01-01

    The subchannel method used in nuclear fuel bundle thermal-hydraulic analysis lies in the statement that subchannel fluid temperatures are taken at mixed mean values. However, the development of mixing correlations and code assessment procedures are, sometimes in the literature, based upon the assumption of identity between lumped and local (subchannel centroid) temperature values. The present paper is concerned with the presentation of an approach for correlating lumped to centroid subchannel temperatures, based upon previously formulated models by the author, applied to a nine heated tube bundle experimental data set. (Author) [pt

  15. Modelling of steady state erosion of CFC actively water-cooled mock-up for the ITER divertor

    Science.gov (United States)

    Ogorodnikova, O. V.

    2008-04-01

    Calculations of the physical and chemical erosion of CFC (carbon fibre composite) monoblocks as outer vertical target of the ITER divertor during normal operation regimes have been done. Off-normal events and ELM's are not considered here. For a set of components under thermal and particles loads at glancing incident angle, variations in the material properties and/or assembly of defects could result in different erosion of actively-cooled components and, thus, in temperature instabilities. Operation regimes where the temperature instability takes place are investigated. It is shown that the temperature and erosion instabilities, probably, are not a critical point for the present design of ITER vertical target if a realistic variation of material properties is assumed, namely, the difference in the thermal conductivities of the neighbouring monoblocks is 20% and the maximum allowable size of a defect between CFC armour and cooling tube is +/-90° in circumferential direction from the apex.

  16. Use of mock-up training to reduce personnel exposure at the North Anna Unit 1 Steam Generator Replacement Project

    Energy Technology Data Exchange (ETDEWEB)

    Henry, H.G. [Virginia Power, Mineral, VA (United States); Reilly, B.P. [Bechtel Power Corp., Gaithersburg, MD (United States)

    1995-03-01

    The North Anna Power Station is located on the southern shore of Lake Anna in Louisa County, approximately forty miles northwest of Richmond, Virginia. The two 910 Mw nuclear units located on this site are owned by Virginia Electric and Power Company (Virginia Power) and Old Dominion Electric Cooperative and operated by Virginia Power. Fuel was loaded into Unit 1 in December 1977, and it began commercial operation in June 1978. Fuel was loaded into Unit 2 in April 1980 and began commercial operation in December 1980. Each nuclear unit includes a three-coolant-loop pressurized light water reactor nuclear steam supply system that was furnished by Westinghouse Electric Corporation. Included within each system were three Westinghouse Model 51 steam generators with alloy 600, mill-annealed tubing material. Over the years of operation of Unit 1, various corrosion-related phenomena had occurred that affected the steam generators tubing and degraded their ability to fulfill their heat transfer function. Advanced inspection and repair techniques helped extend the useful life of the steam generators, but projections based on the results of the inspections indicated that the existing steam generators tubing and degraded their ability to fullfill their heat transfer function. Advanced inspection and repair techniques helped extend the useful life of the steam generators, but projections based on the results of the inspections indicated that the existing steam generators would not last their design life and must be repaired. To this end Virginia Power determined that a steam generator replacement (SGR) program was necessary to remove the old steam generator tube bundles and lower shell sections, including the channel heads (collectively called the lower assemblies), and replace them with new lower assemblies incorporating design features that will prevent the degradation problems that the old steam generators had experienced.

  17. Optimization of the representativeness and transposition approach, for the neutronic design of experimental programs in critical mock-up

    International Nuclear Information System (INIS)

    Dos-Santos, N.

    2013-01-01

    The work performed during this thesis focused on uncertainty propagation (nuclear data, technological uncertainties, calculation biases,...) on integral parameters, and the development of a novel approach enabling to reduce this uncertainty a priori directly from the design phase of a new experimental program. This approach is based on a multi-parameter multi-criteria extension of representativeness and transposition theories. The first part of this PhD work covers an optimization study of sensitivity and uncertainty calculation schemes to different modeling scales (cell, assembly and whole core) for LWRs and FBRs. A degraded scheme, based on standard and generalized perturbation theories, has been validated for the calculation of uncertainty propagation to various integral quantities of interest. It demonstrated the good a posteriori representativeness of the EPICURE experiment for the validation of mixed UOX-MOX loadings, as the importance of some nuclear data in the power tilt phenomenon in large LWR cores. The second part of this work was devoted to methods and tools development for the optimized design of experimental programs in ZPRs. Those methods are based on multi-parameters representativeness using simultaneously various quantities of interest. Finally, an original study has been conducted on the rigorous estimation of correlations between experimental programs in the transposition process. The coupling of experimental correlations and multi-parametric representativeness approach enables to efficiently design new programs, able to answer additional qualification requirements on calculation tools. (author) [fr

  18. THYC qualification on Vatican-1 low pressure tests

    International Nuclear Information System (INIS)

    Duval, C.; Guichard, J.

    1991-06-01

    PWR cores or fuel assemblies are components of a nuclear power plant involving single and two-phase flows in rod bundles. The knowledge of the detailed two-phase and three-dimensional flow patterns is necessary to evaluate the singularity (grids) and bypass effects on the Departure from Nucleate Boiling (DNB) in reactor cores during incidental transients. For that purpose, since 1989, the VATICAN experiment has been performed at EDF as a part of the qualification program of the three-dimensional computer code THYC, developed by EDF. The qualification strategy of the THYC software for PWR cores is the following: assuming the theoretical or experimental knowledge of regular and singular pressure drops and grid turbulence sources in single-phase, pressure drop multipliers and relative velocity in two-phase flow, the VATICAN experiment allows to evaluate the diffusion phenomena in two-phase flow. It provides thermalhydraulic measurements on a mock-up of a part of 900 MWe PWR fuel assembly in single and two-phase flows, with power and quality gradients. The first configuration of the mock-up, with simple spacer grids, is studied (VATICAN-1). The specific effects of mixing spacer grids will be compared to these data through a second configuration. The last void fraction measurements, using a γ-ray technique, performed on VATICAN-1 low pressure tests allowed to qualify a set of closure relations, particularly a model of little two-phase diffusion, adapted to two-phase flows at low pressure (5.0MPa). The qualification of subcooled boiling and diffusion models will continue on next VATICAN and other experimental campaigns [fr

  19. An Airborne Parachute Compartment Test Bed for the Orion Parachute Test Program

    Science.gov (United States)

    Moore, James W.; Romero, Leah M.

    2013-01-01

    The test program developing parachutes for the Orion/MPCV includes drop tests with parachutes deployed from an Orion-like parachute compartment at a wide range of dynamic pressures. Aircraft and altitude constraints precluded the use of an Orion boilerplate capsule for several test points. Therefore, a dart-shaped test vehicle with a hi-fidelity mock-up of the Orion parachute compartment has been developed. The available aircraft options imposed constraints on the test vehicle development and concept of operations. Delivery of this test vehicle to the desired velocity, altitude, and orientation required for the test is a di cult problem involving multiple engineering disciplines. This paper describes the development of the test technique. The engineering challenges include extraction from an aircraft, reposition of the extraction parachute, and mid-air separation of two vehicles, neither of which has an active attitude control system. The desired separation behavior is achieved by precisely controlling the release point using on-board monitoring of the motion. The design of the test vehicle is also described. The trajectory simulations and other analyses used to develop this technique and predict the behavior of the test vehicle are reviewed in detail. The application of the technique on several successful drop tests is summarized.

  20. Dosimetry work and calculations in connection with the irradiation of large devices in the high flux materials testing reactor BR2

    International Nuclear Information System (INIS)

    De Raedt, C.; Leenders, L.; Tourwe, H.; Farrar, H. IV.

    1982-01-01

    For about fifteen years the high flux reactor BR2 has been involved in the testing of fast reactor fuel pins. In order to simulate the fast reactor neutron environment most devices are irradiated under cadmium screen, cutting off the thermal flux component. Extensive neutronic calculations are performed to help the optimization of the fuel bundle design. The actual experiments are preceded by irradiations of their mock-ups in BR02, the zero power model of BR2. The mock-up irradiations, supported by supplementary calculations, are performed for the determination of the main neutronic characteristics of the irradiation proper in BR2 and for the determination of the corresponding operation data. At the end of the BR2 irradiation, the experimental results, such as burn-ups, neutron fluences, helium production in the fuel pin claddings, etc. are correlated by neutronic calculations in order to examine the consistency of the post-irradiation results and to validate the routine calculation procedure and cross-section data employed. A comparison is made in this paper between neutronic calculation results and some post-irradiation data for MOL 7D, a cadmium screened sodium cooled loop containing a nineteen fuel pin bundle

  1. 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.

  2. 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

  3. Development of ultrasonic testing technique with the large transducer to inspect the containment vessel plates of nuclear power plant embedded in concrete

    International Nuclear Information System (INIS)

    Ishida, Hitoshi; Kurozumi, Yasuo; Kaneshima, Yoshiari

    2004-01-01

    The containment vessel plates embedded in concrete on Pressurized Water Reactors are inaccessible to inspect directly. Therefore, it is advisable to prepare inspection technology to detect existence and a location of corrosion on the embedded plates indirectly. In order to establish ultrasonic testing technique to be able to inspect the containment vessel plates embedded in concrete widely at the accessible point, experiments to detect artificial hollows simulating corrosion on a surface of a carbon steel plate mock-up covered with concrete simulating the embedded containment vessel plates were carried out with newly made ultrasonic transducers. We made newly low frequency (0.3 MHz and 0.5 MHz) surface shear horizontal (SH) wave transducers combined with three large active elements, which were equivalent to a 120mm width element. As a result of the experiments, the surface SH transducers could detect clearly the echo from the hollows with a depth of 9.5 mm and 19 mm at a distance of 1500mm from the transducers on the surface of the mock-up covered with concrete. Therefore, we evaluate that it is possible to detect the defects such as corrosion on the plates embedded in concrete with the newly made low frequency surface SH transducers with large elements. (author)

  4. Feasibility evaluation of x-ray imaging for measurement of fuel rod bowing in CFTL test bundles

    International Nuclear Information System (INIS)

    Baker, S.P.

    1980-06-01

    The Core Flow Test Loop (CFTL) is a high temperature, high pressure, out-of-reactor helium-circulating system. It is designed for detailed study of the thermomechanical performance, at prototypic steady-state and transient operating conditions, of electrically heated rods that simulate segments of core assemblies in the Gas-Cooled Fast Breeder reactor demonstration plant. Results are presented of a feasibility evaluation of x-ray imaging for making measurements of the displacement (bowing) of fuel rods in CFTL test bundles containing electrically heated rods. A mock-up of a representative CFTL test section consisting of a test bundle and associated piping was fabricated to assist in this evaluation

  5. Development of ultrasonic testing technique to inspect containment liners embedded in concrete on nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, H.; Kurozumi, Y. [Inst. of Nuclear Safety System, Incorporated, Mihama, Fukui (Japan); Kaneshima, Y. [The Kansai Electric Power Company, Inc., Mihama, Fukui (Japan)

    2004-07-01

    The purpose of this study is development of ultrasonic testing technique to inspect containment liners embedded in concrete on nuclear power plants. Integrity of containment liners on nuclear power plants can be secured by suitable present operation and maintenance. Furthermore, non-destructive testing technique to inspect embedded liners will ensure the integrity of the containment further. In order to develop the non-destructive testing technique, ultrasonic transducers were made newly and ultrasonic testing data acquisition and evaluation were carried out by using a mock-up. We adopted the surface shear horizontal (SH) wave, low frequency (0.3-0.5MHz), to be able to detect an echo from a defect against attenuation of ultrasonic waves due to long propagation in the liners and dispersion into concrete. We made transducers with three large active elements (40mm x 40mm) in a line which were equivalent to a 120mm width active element. Artificial hollows, {phi}200mm - 19mm depth (1/2thickness) and {phi}200mm - 9.5mm depth (1/4thickness), were made on a surface of a mock-up: carbon steel plate, 38mm thickness, 2,000mm length, 1000mm width. The surfaces of the plate were covered with concrete in order to simulate liners embedded in concrete. As a result of the examinations, the surface SH transducers could detect clearly the echo from the hollows at a distance of 1500mm. We evaluate that the newly made surface SH transducers with three elements have ability of detection of defects such as corrosion on the liners embedded in concrete. (author)

  6. The SPHINX reactor for engineering tests

    International Nuclear Information System (INIS)

    Adamov, E.O.; Artamkin, K.N.; Bovin, A.P.; Bulkin, Y.M.; Kartashev, E.F.; Korneev, A.A.; Stenbok, I.A.; Terekhov, A.S.; Khmel'Shehikov, V.V.; Cherkashov, Y.M.

    1990-01-01

    A research reactor known as SPHINX is under development in the USSR. The reactor will be used mainly to carry out tests on mock-up power reactor fuel assemblies under close-to-normal parameters in experimental loop channels installed in the core and reflector of the reactor, as well as to test samples of structural materials in ampoule and loop channels. The SPHINX reactor is a channel-type reactor with light-water coolant and moderator. Maximum achievable neutron flux density in the experimental channels (cell composition 50% Fe, 50% H 2 O) is 1.1 X 10 15 neutrons/cm 2 · s for fast neutrons (E > 0.1 MeV) and 1.7 X 10 15 for thermal neutrons at a reactor power of 200 MW. The design concepts used represent a further development of the technical features which have met with approval in the MR and MIR channel-type engineering test reactors currently in use in the USSR. The 'in-pond channel' construction makes the facility flexible and eases the carrying out of experimental work while keeping discharges of radioactivity into the environment to a low level. The reactor and all associated buildings and constructions conform to modern radiation safety and environmental protection requirements

  7. Analysis and modeling of coupled thermo-hydro-mechanical phenomena in 3D fractured media; Analyse et modelisation des phenomenes couples thermo-hydromecaniques en milieux fractures 3D

    Energy Technology Data Exchange (ETDEWEB)

    Canamon Valera, I

    2006-11-15

    This doctoral research was conducted as part of a joint France-Spain co-tutelage PhD thesis in the framework of a bilateral agreement between two universities, the Institut National Polytechnique de Toulouse (INPT) and the Universidad Politecnica de Madrid (UPM). It concerns a problem of common interest at the national and international levels, namely, the disposal of radioactive waste in deep geological repositories. The present work is devoted, more precisely, to near-field hydrogeological aspects involving mass and heat transport phenomena. The first part of the work is devoted to a specific data interpretation problem (pressures, relative humidities, temperatures) in a multi-barrier experimental system at the scale of a few meters - the 'Mock-Up Test' of the FEBEX project, conducted in Spain. Over 500 time series are characterized in terms of spatial, temporal, and/or frequency/scale-based statistical analysis techniques. The time evolution and coupling of physical phenomena during the experiment are analyzed, and conclusions are drawn concerning the behavior and reliability of the sensors. The second part of the thesis develops in more detail the 3-Dimensional (3D) modeling of coupled Thermo-Hydro-Mechanical phenomena in a fractured porous rock, this time at the scale of a hundred meters, based on the data of the 'In-Situ Test' of the FEBEX project conducted at the Grimsel Test Site in the Swiss Alps. As a first step, a reconstruction of the 3D fracture network is obtained by Monte Carlo simulation, taking into account through optimization the geomorphological data collected around the FEBEX gallery. The heterogeneous distribution of traces observed on the cylindrical wall of the tunnel is fairly well reproduced in the simulated network. In a second step, we develop a method to estimate the equivalent permeability of a many-fractured block by extending the superposition method of Ababou et al. [1994] to the case where the permeability of

  8. Analysis and modeling of coupled thermo-hydro-mechanical phenomena in 3D fractured media

    International Nuclear Information System (INIS)

    Canamon Valera, I.

    2006-11-01

    This doctoral research was conducted as part of a joint France-Spain co-tutelage PhD thesis in the framework of a bilateral agreement between two universities, the Institut National Polytechnique de Toulouse (INPT) and the Universidad Politecnica de Madrid (UPM). It concerns a problem of common interest at the national and international levels, namely, the disposal of radioactive waste in deep geological repositories. The present work is devoted, more precisely, to near-field hydrogeological aspects involving mass and heat transport phenomena. The first part of the work is devoted to a specific data interpretation problem (pressures, relative humidities, temperatures) in a multi-barrier experimental system at the scale of a few meters - the 'Mock-Up Test' of the FEBEX project, conducted in Spain. Over 500 time series are characterized in terms of spatial, temporal, and/or frequency/scale-based statistical analysis techniques. The time evolution and coupling of physical phenomena during the experiment are analyzed, and conclusions are drawn concerning the behavior and reliability of the sensors. The second part of the thesis develops in more detail the 3-Dimensional (3D) modeling of coupled Thermo-Hydro-Mechanical phenomena in a fractured porous rock, this time at the scale of a hundred meters, based on the data of the 'In-Situ Test' of the FEBEX project conducted at the Grimsel Test Site in the Swiss Alps. As a first step, a reconstruction of the 3D fracture network is obtained by Monte Carlo simulation, taking into account through optimization the geomorphological data collected around the FEBEX gallery. The heterogeneous distribution of traces observed on the cylindrical wall of the tunnel is fairly well reproduced in the simulated network. In a second step, we develop a method to estimate the equivalent permeability of a many-fractured block by extending the superposition method of Ababou et al. [1994] to the case where the permeability of the rock matrix is not

  9. Demonstration tests for manufacturing the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Shimizu, Katsusuke; Onozuka, Masanori; Usui, Yukinori; Urata, Kazuhiro; Tsujita, Yoshihiro; Nakahira, Masataka; Takeda, Nobukazu; Kakudate, Satoshi; Ohmori, Junji; Shibanuma, Kiyoshi

    2007-01-01

    Demonstration tests for manufacturing and assembly of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel have been conducted to confirm manufacturing and assembly process of the vacuum vessel (VV). The full-scale partial mock-up fabrication was planned and is in progress. The results will be available in the near future. Field-joint assembly procedure has been demonstrated using a test stand. Due to limited accessibility to the outer shell at the field joint, some operations, including alignment of the splice plates, field-joint welding, and examination, were found to be very difficult. In addition, a demonstration test on the selected back-seal structures was performed. It was found that the tested structures have insufficient sealing capabilities and need further improvement. The applicability of ultrasonic testing methods has been investigated. Although side drilled holes of 2.4 mm in diameter were detected, detection of the slit-type defects and defect characterization were found to be difficult. Feasibility test of liquid penetrant testing has revealed that the selected liquid penetrant testing (LPT) solutions have sufficient low outgas rates and are applicable to the VV

  10. Demonstration tests for manufacturing the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Katsusuke [Mitsubishi Heavy Industries, Ltd., Kobe Shipyard and Machinery Works, Wadasaki-cho 1-1-1, Hyogo-ku, Kobe 652-8585 (Japan)], E-mail: katsusuke_shimizu@mhi.co.jp; Onozuka, Masanori [Mitsubishi Heavy Industries, Ltd., Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan); Usui, Yukinori; Urata, Kazuhiro; Tsujita, Yoshihiro [Mitsubishi Heavy Industries, Ltd., Kobe Shipyard and Machinery Works, Wadasaki-cho 1-1-1, Hyogo-ku, Kobe 652-8585 (Japan); Nakahira, Masataka; Takeda, Nobukazu; Kakudate, Satoshi; Ohmori, Junji; Shibanuma, Kiyoshi [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan)

    2007-10-15

    Demonstration tests for manufacturing and assembly of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel have been conducted to confirm manufacturing and assembly process of the vacuum vessel (VV). The full-scale partial mock-up fabrication was planned and is in progress. The results will be available in the near future. Field-joint assembly procedure has been demonstrated using a test stand. Due to limited accessibility to the outer shell at the field joint, some operations, including alignment of the splice plates, field-joint welding, and examination, were found to be very difficult. In addition, a demonstration test on the selected back-seal structures was performed. It was found that the tested structures have insufficient sealing capabilities and need further improvement. The applicability of ultrasonic testing methods has been investigated. Although side drilled holes of 2.4 mm in diameter were detected, detection of the slit-type defects and defect characterization were found to be difficult. Feasibility test of liquid penetrant testing has revealed that the selected liquid penetrant testing (LPT) solutions have sufficient low outgas rates and are applicable to the VV.

  11. Tests on the integration of the ITER divertor dummy armour prototype on a simplified model of cassette body

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Canneta, A.; Cattadori, G.; Gaspari, G.P.; Merola, M.; Polazzi, G.; Vieider, G.; Zito, D.

    2001-01-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body, designed with some mechanical and hydraulic simplifications with respect to the reference body, and the actively cooled Dummy Armour Prototype (DAP). This DAP consists of the Vertical Target, the Wing and the Dump Target, manufactured by the European industry, which are integrated with the Gas Box Liner supplied by the Russian Federation Home Team. In order to simplify the manufacturing, the DAP was layered with an equivalent CuCrZr thickness simulating the real armour (CFC or W tiles). In parallel with the manufacturing activity, the ITER European HT decided to assign to ENEA the Task EU-DV1 for the 'Component Integration and Thermal-Hydraulic Testing of the ITER Divertor Targets and Wing Dummy Prototypes and Cassette Body'

  12. Tests on the integration of the ITER divertor dummy armour prototype on a simplified model of cassette body

    Energy Technology Data Exchange (ETDEWEB)

    Dell' Orco, G. E-mail: dellorco@brasimone.enea.it; Canneta, A.; Cattadori, G.; Gaspari, G.P.; Merola, M.; Polazzi, G.; Vieider, G.; Zito, D

    2001-10-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body, designed with some mechanical and hydraulic simplifications with respect to the reference body, and the actively cooled Dummy Armour Prototype (DAP). This DAP consists of the Vertical Target, the Wing and the Dump Target, manufactured by the European industry, which are integrated with the Gas Box Liner supplied by the Russian Federation Home Team. In order to simplify the manufacturing, the DAP was layered with an equivalent CuCrZr thickness simulating the real armour (CFC or W tiles). In parallel with the manufacturing activity, the ITER European HT decided to assign to ENEA the Task EU-DV1 for the 'Component Integration and Thermal-Hydraulic Testing of the ITER Divertor Targets and Wing Dummy Prototypes and Cassette Body'.

  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. Overview of the Acoustic Testing of the European Service Module Structural Test Article (E-STA)

    Science.gov (United States)

    Hughes, William; Fogt, Vince; Le Plenier, Cyprien; Duval, Francois; Durand, Jean-Francois; Staab, Lucas D.; Hozman, Aron; Mcnelis, Anne; Bittinger, Samantha; Thirkettle, Anthony; hide

    2017-01-01

    The European Space Agency (ESA) and their prime contractor Airbus Defense Space (ADS) are developing the European Service Module (ESM) for integration and utilization with other modules of NASAs Orion Multi-Purpose Crew Vehicle. As part of this development, ESA, ADS, NASA and the Lockheed Martin Company performed a series of reverberant acoustic tests in April-May 2016 on the ESM Structural Test Article (E-STA), the mechanical mock-up of the ESM designated for mechanical tests. Testing the E-STA under acoustic qualification loads verifies whether it can successfully withstand the medium and high frequency mechanical environment occurring during the vehicles lift-off and atmospheric phases of flight. The testing occurred at the Reverberant Acoustic Test Facility (RATF) at the NASA Glenn Research Centers Plum Brook Station site in Sandusky, OH, USA. This highly successful acoustic test campaign excited the E-STA to acoustic test levels as high as 149.4 dB Overall Sound Pressure Level. This acoustic testing met all the ESA and ADSs test objectives, including establishingverifying the random vibration qualification test levels for numerous hardware components of the ESM, and qualifying the ESMs Solar Array Wing electrical power system. This paper will address the test objectives, the test articles configuration, the test instrumentation and excitation levels, the RATF site and capabilities, the series of acoustic tests performed, and the technical issues faced and overcome to result in a successful acoustic test campaign for the ESM. A discussion of several test results is also included.

  15. Comparative strength analysis and thermal fatigue testing of Be/CuCrZr and Be/GlidCop joints produced by fast brazing

    International Nuclear Information System (INIS)

    Gervash, A.; Mazul, I.; Yablokov, N.; Barabash, V.; Ganenko, A.

    2000-01-01

    Proposing beryllium as plasma facing armour this paper presents the recent results obtained in Russia in the frame of such activities. Last year testing of actively cooled mock-ups produced by fast brazing of Be onto Cu-alloy heat sink allows to consider mentioned Russian method as promising for both PH-copper like CuCrZr and DS-copper like GlidCop. Summarizing recent experimental results with their previous data authors attempt to comparatively investigate a behaviour of Be/CuCrZr and Be/GlidCop joints in ITER relevant conditions. Mechanical properties, brazing zone microstructure and thermal response were taken for comparison. The shear strength for both types of joints was found as 150-200 MPa and did not depend on testing temperature. The brazing zone morphology and microhardness are presented, the thermal fatigue behaviour of investigated joints is described. All main results as well as the nearest future plans are discussed. (orig.)

  16. Material testing facilities and programs for plasma-facing component testing

    Science.gov (United States)

    Linsmeier, Ch.; Unterberg, B.; Coenen, J. W.; Doerner, R. P.; Greuner, H.; Kreter, A.; Linke, J.; Maier, H.

    2017-09-01

    Component development for operation in a large-scale fusion device requires thorough testing and qualification for the intended operational conditions. In particular environments are necessary which are comparable to the real operation conditions, allowing at the same time for in situ/in vacuo diagnostics and flexible operation, even beyond design limits during the testing. Various electron and neutral particle devices provide the capabilities for high heat load tests, suited for material samples and components from lab-scale dimensions up to full-size parts, containing toxic materials like beryllium, and being activated by neutron irradiation. To simulate the conditions specific to a fusion plasma both at the first wall and in the divertor of fusion devices, linear plasma devices allow for a test of erosion and hydrogen isotope recycling behavior under well-defined and controlled conditions. Finally, the complex conditions in a fusion device (including the effects caused by magnetic fields) are exploited for component and material tests by exposing test mock-ups or material samples to a fusion plasma by manipulator systems. They allow for easy exchange of test pieces in a tokamak or stellarator device, without opening the vessel. Such a chain of test devices and qualification procedures is required for the development of plasma-facing components which then can be successfully operated in future fusion power devices. The various available as well as newly planned devices and test stands, together with their specific capabilities, are presented in this manuscript. Results from experimental programs on test facilities illustrate their significance for the qualification of plasma-facing materials and components. An extended set of references provides access to the current status of material and component testing capabilities in the international fusion programs.

  17. Preliminary Interpretation of a Radionuclide and Colloid Tracer Test in a Granodiorite Shear Zone at the Grimsel Test Site, Switzerland

    Energy Technology Data Exchange (ETDEWEB)

    Reimus, Paul W. [Los Alamos National Laboratory

    2012-08-30

    In February and March 2012, a tracer test involving the injection of a radionuclide-colloid cocktail was conducted in the MI shear zone at the Grimsel Test Site, Switzerland, as part of the Colloids Formation and Migration (CFM) project. The colloids were derived from FEBEX bentonite, which is mined in Spain and is being considered as a potential waste package backfill in a Spanish nuclear waste repository. The tracer test, designated test 12-02 (second test in 2012), involved the injection of the tracer cocktail into borehole CFM 06.002i2 and extraction from the Pinkel surface packer at the main access tunnel wall approximately 6.1 m from the injection interval. The test configuration is depicted in Figure 1. This configuration has been used in several conservative tracer tests and two colloid-homologue tracer tests since 2007, and it is will be employed in an upcoming test involving the emplacement of a radionuclide-doped bentonite plug into CFM 06.002i2 to evaluate the swelling and erosion of the bentonite and the transport of bentonite colloids and radionuclides from the source to the extraction point at the tunnel wall. Interpretive analyses of several of the previous tracer tests, from 09-01 through 12-02 were provided in two previous Used Fuel Disposition Program milestone reports (Arnold et al., 2011; Kersting et al., 2012). However, only the data for the conservative tracer Amino-G Acid was previously analyzed from test 12-02 because the other tracer data from this test were not available at the time. This report documents the first attempt to quantitatively analyze the radionuclide and colloid breakthrough curves from CFM test 12-02. This report was originally intended to also include an experimental assessment of colloid-facilitated transport of uranium by bentonite colloids in the Grimsel system, but this assessment was not conducted because it was reported by German collaborators at the Karlsruhe Institute of Technology (KIT) that neither uranium nor

  18. Nuclear Analyses of Indian LLCB Test Blanket System in ITER

    Science.gov (United States)

    Swami, H. L.; Shaw, A. K.; Danani, C.; Chaudhuri, Paritosh

    2017-04-01

    Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System design & development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radioactive waste management, equipment maintenance & replacement strategies and nuclear safety. The nuclear behaviour of LLCB test blanket module in ITER is predicated in terms of nuclear responses such as tritium production, nuclear heating, neutron fluxes and radiation damages. Radiation shielding capability of LLCB TBS inside and outside bio-shield was also assessed to fulfill ITER shielding requirements. In order to supports the rad-waste and safety assessment, nuclear activation analyses were carried out and radioactivity data were generated for LLCB TBS components. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.l). The paper describes a comprehensive nuclear performance of LLCB TBS in ITER.

  19. SCC testing of steam generator tubes repaired by welded sleeves

    International Nuclear Information System (INIS)

    Pierson, E.; Stubbe, J.

    1993-01-01

    One way to repair steam generator tubing is to introduce a sleeve inside the tube so that it spans the corroded area and to seal it at both ends. This technique has been studied at Laborelec with a particular attention paid to the occurrence of new SCC cracks at the upper joint. Tube segments coming from the same lot of mill annealed alloy 600 were sent to six manufacturers to be sleeved by their own procedure (including TIG, laser or kinetic welding, followed or not by a stress relief heat treatment), and then tested at Laborelecin 10% NaOH at 350 degrees C. The tests were performed with and without differential pressure i.e. in capsules (Δ = 9 and 19 MPa) and in autoclave (Δp = 0). Nearly all the not stress relieved mock-ups developed through cracks in several hundred hours in auto-clave. The cracks were circumferential and situated near the weld. At 9 and 19 MPa, the time to failure decreased and longitudinal cracks appeared near the weld and at the transition zone of expanded areas. Cracks were never observed in the alloy 690 sleeve, except in the weld bead. Reference capsules (roll expaned tubes) made of the same lot of alloy 600 were tested in the same environment

  20. Demonstration tests for HTGR fuel elements and core components with test sections in HENDEL

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, Yoshiaki; Hino, Ryutaro; Inagaki, Yoshiyuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1995-03-01

    In the fuel stack test section (T{sub 1}) of the Helium Engineering Demonstration Loop (HENDEL), thermal and hydraulic performances of helium gas flows through a fuel rod channel and a fuel stack have been investigated for the High-Temperature Engineering Test Reactor (HTTR) core thermal design. The test data showed that the turbulent characteristics appearing in the Reynolds number above 2000: no typical behavior in the transition zone, and friction factors and heat transfer coefficients in the fuel channel were found to be higher than those in a smooth annular channel. Heat transfer behavior of gas flow in a fuel element channel with blockage and cross-flow through a gap between upper and lower fuel elements stacked was revealed using the mock-up models. On the other hand, demonstration tests have been performed to verify thermal and hydraulic characteristics and structural integrity related to the core bottom structure using a full-scale test facility named as the in-core structure test section (T{sub 2}). The sealing performance test revealed that the leakage of low-temperature helium gas through gaps between the permanent reflector blocks to the core was very low level compared with the HTTR design value and no change of the leakage flow rate were observed after a long term operation. The heat transfer tests including thermal transient at shutdown of gas circulators verified good insulating performance of core insulation structures in the core bottom structure and the hot gas duct; the temperature of the metal portion of these structure was below the design value. Examination of the thermal mixing characteristics indicated that the mixing of the hot helium gas started at a hot plenum and finished completely at downstream of the outlet hot gas duct. The present results obtained from these demonstration tests have been practically applied to the detailed design works and licensing procedures of the HTTR. (J.P.N.) 92 refs.

  1. Manufacturing and testing of a ITER First Wall Semi-Prototype for EUDA pre-qualification

    International Nuclear Information System (INIS)

    Banetta, S.; Bellin, B.; Lorenzetto, P.; Zacchia, F.; Boireau, B.; Bobin, I.; Boiffard, P.; Cottin, A.; Nogue, P.; Mitteau, R.; Eaton, R.; Raffray, R.; Bürger, A.; Du, J.; Linke, J.; Pintsuk, G.; Weber, T.

    2015-01-01

    Highlights: • Three ITER First Wall Small Scale Mock-ups were manufactured passing factory acceptance tests. • One of the Small Scale Mock-ups passed the thermal fatigue tests (15,000 cycles at 2 MW/m"2). • The ITER First Wall Semi-Prototype was manufactured and is being High Heat Flux tested. • Preliminary results upto 2 MW/m"2 show an overall compliance with the acceptance criteria. • Next step for EU Domestic Agency qualification is the fabrication and testing of a Full-Scale Prototype. - Abstract: This paper describes the main activities carried out in the frame of EU-DA prequalification for the supply of Normal Heat Flux (NHF) First Wall (FW) panels to ITER. A key part of these activities is the manufacturing development, the fabrication and the factory acceptance tests of a reduced scale FW prototype (Semi-Prototype (SP)) of the NHF design. The SP has a dimension of 221 mm × 665 mm, corresponding to about 1/6 of a full-scale panel, with six full-scale “fingers” and bearing a total of 84 beryllium tiles. It has been manufactured by the AREVA Company in France. The manufacturing process has made extensive use of Hot Isostatic Pressing, which was developed over more than a decade during the ITER Engineering Design Activity phase. The main manufacturing steps for the Semi-Prototype are recalled, with a summary of the lessons learned and the implications with regard to the design and manufacturing of the full-scale prototype and of the series fabrication of the EU-DA share of the ITER first wall (215 NHF panels). The fabricated SP is then tested under High Heat Flux (HHF) in the dedicated test facility of JUDITH-II in Forschungszentrum Jülich, Germany. The objective of the HHF testing is the demonstration of achieving the requested performance under thermal fatigue. The test protocol and facility qualification are presented and the behaviour of the fingers under the 7500 cycles at 2 MW/m"2 is described in detail.

  2. Manufacturing and testing of a ITER First Wall Semi-Prototype for EUDA pre-qualification

    Energy Technology Data Exchange (ETDEWEB)

    Banetta, S., E-mail: stefano.banetta@f4e.europa.eu [Fusion For Energy, Torres Diagonal Litoral, B3, Carrer Josep Pla 2, 08019 Barcelona (Spain); Bellin, B.; Lorenzetto, P.; Zacchia, F. [Fusion For Energy, Torres Diagonal Litoral, B3, Carrer Josep Pla 2, 08019 Barcelona (Spain); Boireau, B.; Bobin, I.; Boiffard, P.; Cottin, A.; Nogue, P. [AREVA NP PTCMI-F, Centre Technique, Fusion, 71200 Le Creusot (France); Mitteau, R.; Eaton, R.; Raffray, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Bürger, A.; Du, J.; Linke, J.; Pintsuk, G.; Weber, T. [Forschungszentrum Jülich, Institute of Energy and Climate Research, Jülich (Germany)

    2015-10-15

    Highlights: • Three ITER First Wall Small Scale Mock-ups were manufactured passing factory acceptance tests. • One of the Small Scale Mock-ups passed the thermal fatigue tests (15,000 cycles at 2 MW/m{sup 2}). • The ITER First Wall Semi-Prototype was manufactured and is being High Heat Flux tested. • Preliminary results upto 2 MW/m{sup 2} show an overall compliance with the acceptance criteria. • Next step for EU Domestic Agency qualification is the fabrication and testing of a Full-Scale Prototype. - Abstract: This paper describes the main activities carried out in the frame of EU-DA prequalification for the supply of Normal Heat Flux (NHF) First Wall (FW) panels to ITER. A key part of these activities is the manufacturing development, the fabrication and the factory acceptance tests of a reduced scale FW prototype (Semi-Prototype (SP)) of the NHF design. The SP has a dimension of 221 mm × 665 mm, corresponding to about 1/6 of a full-scale panel, with six full-scale “fingers” and bearing a total of 84 beryllium tiles. It has been manufactured by the AREVA Company in France. The manufacturing process has made extensive use of Hot Isostatic Pressing, which was developed over more than a decade during the ITER Engineering Design Activity phase. The main manufacturing steps for the Semi-Prototype are recalled, with a summary of the lessons learned and the implications with regard to the design and manufacturing of the full-scale prototype and of the series fabrication of the EU-DA share of the ITER first wall (215 NHF panels). The fabricated SP is then tested under High Heat Flux (HHF) in the dedicated test facility of JUDITH-II in Forschungszentrum Jülich, Germany. The objective of the HHF testing is the demonstration of achieving the requested performance under thermal fatigue. The test protocol and facility qualification are presented and the behaviour of the fingers under the 7500 cycles at 2 MW/m{sup 2} is described in detail.

  3. Results of ITER toroidal field coil cover plate welding test

    International Nuclear Information System (INIS)

    Koizumi, Norikiyo; Matsui, Kunihiro; Shimizu, Tatsuya; Nakajima, Hideo; Iijima, Ami; Makino, Yoshinobu

    2012-01-01

    In ITER Toroidal Field (TF) coils, cover plates (CP) are welded to the teeth of the radial plate (RP) to fix conductors in the grooves of the RP. Though the total length of the welds is approximately 1.5 km and the height and width of the RP are 14 and 9 m, respectively, welding deformation of smaller than 1 mm for local out-of-plane distortion and smaller than several millimeters for in-plane deformation is required. Therefore, laser welding is used for CP welding to reduce welding deformation as much as possible. However, the gap in welding joints is expected to be a maximum of 0.5 mm. Thus, a laser welding technique to enable welding of joints with a gap of 0.5 mm in width has been developed. Applying this technology, a CP welding trial using an RP mock-up was successfully performed. The achieved local flatness, that is, the flatness of the cross-section of the RP mock-up, is 0.6 mm. The analysis using inherent strains, which are derived from the welding test using flat plates, also indicates that better local flatness can be achieved if the initial distortion is zero. In addition, the welding deformation of a full-scale RP is evaluated via analysis using the inherent strain. The analytical results show that in-plane deformation is approximately 5 mm and large out-of-plane deformation, consisting of approximately 5 mm-long wave distortion and a twist of approximately 1.5 mm in the RP cross-section, is generated. It is expected that the required profile can be achieved by determining the original geometry of an RP by simulating deformation during welding. It is also expected that the required local flatness of a DP can be achieved, since out-of-plane deformation can be reduced by increasing the number of RPs turned over during CP welding. A more detailed study is required. (author)

  4. Materials development for ITER shielding and test blanket in China

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J.M., E-mail: Chenjm@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wu, J.H.; Liu, X.; Wang, P.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wang, Z.H.; Li, Z.N. [Ningxia Orient Non-ferrous Metals Group Co. Ltd., P.O. Box 105, Shizuishan (China); Wang, X.S.; Zhang, P.C. [China Academy of Engineering Physics, P.O. Box 919-71, Mianyang 621900 (China); Zhang, N.M.; Fu, H.Y.; Liu, D.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China)

    2011-10-01

    China is a member of the ITER program and is developing her own materials for its shielding and test blanket modules. The materials include vacuum-hot-pressing (VHP) Be, CuCrZr alloy, 316L(N) and China low activation ferritic/martensitic (CLF-1) steels. Joining technologies including Be/Cu hot isostatic pressing (HIP) and electron beam (EB) weldability of 316L(N) were investigated. Chinese VHP-Be showed good properties, with BeO content and ductility that satisfy the ITER requirements. Be/Cu mock-ups were fabricated for Be qualification tests at simulated ITER vertical displacement event (VDE) and heat flux cycling conditions. Fine microstructure and good mechanical strength of the CuCrZr alloy were achieved by a pre-forging treatment, while the weldability of 316L(N) by EB was demonstrated for welding depths varying from 5 to 80 mm. Fine microstructure, high strength, and good ductility were achieved in CLF-1 steel by an optimized normalizing, tempering and aging procedure.

  5. Recent results on high thermal energy load testing of beryllium for ITER first wall application

    Science.gov (United States)

    Kupriyanov, I. B.; Roedig, M.; Nikolaev, G. N.; Kurbatova, L. A.; Linke, J.; Gervash, A. A.; Giniyatulin, R. N.; Podkovyrov, V. L.; Muzichenko, A. D.; Khimchenko, L.

    2011-12-01

    In this paper, progress in the high heat flux (HHF) qualification testing of TGP-56FW beryllium grade for ITER first wall applications is presented. Two actively cooled Be/CuCrZr brazing mock-ups were tested under complex thermal loading conditions in the electron beam facility JUDITH-1 (step 1: vertical displacement event test at 40 MJ m-2, 0.3 s, 1 shot; step 2: disruption tests at 3 MJ m-2, 1 shot, Δt=5 ms; step 3: repetitive fatigue test at 80 MW m-2, 1000 shots, Δt=25 ms). After testing, metallographic investigations on the microstructure and crack morphology were carried out. The results of these studies of Be tiles are reported and discussed. The overall results of TGP-56FW grade qualification testing have demonstrated the reliable performance capability of TGP-56FW for application as the armor of the ITER first wall. In addition, the results of first experiments with TGP-56FW and S-65C beryllium grades in the QSPA-Be plasma gun facility are also reported. In these experiments, beryllium tiles (80×80×10 mm3) were tested in a hydrogen plasma stream (5 cm in diameter) with pulse duration 0.5 ms and heat loads of 0.5-2 MJ m-2. Experiments were performed at room temperature. The evolution of the surface microstructure and mass loss of beryllium exposed to up to 100 shots is presented.

  6. Recent results on high thermal energy load testing of beryllium for ITER first wall application

    International Nuclear Information System (INIS)

    Kupriyanov, I B; Nikolaev, G N; Kurbatova, L A; Roedig, M; Linke, J; Gervash, A A; Giniyatulin, R N; Podkovyrov, V L; Muzichenko, A D; Khimchenko, L

    2011-01-01

    In this paper, progress in the high heat flux (HHF) qualification testing of TGP-56FW beryllium grade for ITER first wall applications is presented. Two actively cooled Be/CuCrZr brazing mock-ups were tested under complex thermal loading conditions in the electron beam facility JUDITH-1 (step 1: vertical displacement event test at 40 MJ m - 2, 0.3 s, 1 shot; step 2: disruption tests at 3 MJ m - 2, 1 shot, Δt=5 ms; step 3: repetitive fatigue test at 80 MW m - 2, 1000 shots, Δt=25 ms). After testing, metallographic investigations on the microstructure and crack morphology were carried out. The results of these studies of Be tiles are reported and discussed. The overall results of TGP-56FW grade qualification testing have demonstrated the reliable performance capability of TGP-56FW for application as the armor of the ITER first wall. In addition, the results of first experiments with TGP-56FW and S-65C beryllium grades in the QSPA-Be plasma gun facility are also reported. In these experiments, beryllium tiles (80×80×10 mm 3 ) were tested in a hydrogen plasma stream (5 cm in diameter) with pulse duration 0.5 ms and heat loads of 0.5-2 MJ m - 2. Experiments were performed at room temperature. The evolution of the surface microstructure and mass loss of beryllium exposed to up to 100 shots is presented.

  7. Current status of the active test at RRP and development programs for the advanced melter

    International Nuclear Information System (INIS)

    Kanehira, Norio

    2016-01-01

    The vitrification facility in Rokkasho Reprocessing Plant started the active tests to solidify HAW into the glass in 2007 which was the examination of the final stage before the operation, but the active test had to be discontinued due to the trouble of glass melter operation with down of pouring by deposit of noble metals on the melter bottom. After the equipment and operating conditions were improved in response to the result of the mock-up tests, a series of active tests were restarted active tests in May, 2012. These tests were finished with enough confirmation of stability in the state such as glass temperature and controlling the noble metals. JNFL has been developed the advanced melter, Joule heated ceramic melter, and the design of the advanced melter is largely different from the existing one. For the confirmation of the advanced melter performances, the full-scale inactive tests had been performed and successfully finished. This paper describes outline of development for advanced melter in Rokkasho Reprocessing Plant. (author)

  8. Results of recent reactor-material tests on dispersal of oxide fuel from a disrupted core

    International Nuclear Information System (INIS)

    Spencer, B.W.; Wilson, R.J.; Vetter, D.L.; Erickson, E.G.; Dewey, G.

    1985-01-01

    The results of experimental investigations and related analyses are reported addressing the dispersal of molten oxide fuel from a disrupted core via various available pathways for the CRBR system. These investigations included the GAPFLOW tests in which pressure-driven and gravity drainage tests were performed using dispersal pathways mocking up the intersubassembly gaps, the CAMEL C6 and C7 tests in which molten fuel entered sodium-filled control assembly ducts under prototypic thermal-hydraulic conditions, and the Lower Internals Drainage (LID) tests in which molten fuel drained downward through simulated below-core structure (orifice plate stacks) as the bottom of control assembly ducts. The results of SHOTGUN tests addressing basic freezing of molten UO 2 and UO 2 /metal mixtures flowing through circular tubes are also reported. Test results have invariably shown the existance of stable UO 2 crusts on the inside surfaces of the flow paths. Appreciable removal of fuel was indicated prior to freezing-induced immobilization. Application of heat transfer models based upon the presence of stable, insulating fuel crusts tends to overpredict the removal process

  9. Mechanical testing of a FW panel attachment system for ITER

    International Nuclear Information System (INIS)

    Oliva, Vladislav; Vaclavik, Jaroslav; Materna, Ales; Lorenzetto, Patrick; Furmanek, Andreas

    2009-01-01

    An objective of experiments and finite element simulations was to check the stiffness, the strength and the fatigue resistance of the attachment of the First Wall panels onto a shield block of blanket modules according to the ITER 2001 design. The panel has a poloidal key at the rear side (in so-called option A with the rear access bolting) and it is attached by means of special studs located on a key-way in the shield block. Special device for a test of stud tensile pre-load relaxation during a thermal cycling was developed. True-to-scale panels, the shield block mock-up and simplified studs were fabricated and the assembly was loaded alternatively by radial moment, poloidal force or poloidal moment simulating the loading during off-normal plasma operations. Thermal cycling led to an acceptable stud pre-load relaxation. Mechanical cycling caused neither the pre-load relaxation nor the loss of the contact in the key-way nor a damage of the attachment system. The combination of poloidal moment and radial force during vertical displacement events (VDEs) seems to be a most dangerous case because it could lead to the loss of the key-key-way contact.

  10. Assessment of Pipe Wall Loss Using Guided Wave Testing

    International Nuclear Information System (INIS)

    Joo, Kyung Mun; Jin, Seuk Hong; Moon, Yong Sig

    2010-01-01

    Flow accelerated corrosion(FAC) of carbon steel pipes in nuclear power plants has been known as one of the major degradation mechanisms. It could have bad influence on the plant reliability and safety. Also detection of FAC is a significant cost to the nuclear power plant because of the need to remove and replace insulation. Recently, the interest of the guided wave testing(GWT) has grown because it allows long range inspection without removing insulation of the pipe except at the probe position. If GWT can be applied to detection of FAC damages, it will can significantly reduce the cost for the inspection of the pipes. The objective of this study was to determine the capability of GWT to identify location of FAC damages. In this paper, three kinds of techniques were used to measure the amplitude ratio between the first and the second welds at the elbow area of mock-ups that contain real FAC damages. As a result, optimal inspection technique and minimum detectability to detect FAC damages drew a conclusion

  11. The progress and results of a demonstration test of a cavern-type disposal facility

    International Nuclear Information System (INIS)

    Akiyama, Yoshihiro; Terada, Kenji; Oda, Nobuaki; Yada, Tsutomu; Nakajima, Takahiro

    2011-01-01

    The cavern-type disposal facilities for low-level waste (LLW) with relatively high radioactivity levels mainly generated from power reactor decommissioning and for part of transuranic (TRU) waste mainly from spent fuel reprocessing are designed to be constructed in a cavern 50 to 100 meters below ground, and to employ an engineered barrier system (EBS) of a combination of bentonite and cement materials in Japan. In order to advance the feasibility study for these disposal, a government-commissioned research project named Demonstration Test of Cavern-Type Disposal Facility started in fiscal 2005, and since fiscal 2007 a full-scale mock-up test facility has been constructed under actual subsurface environment. The main objective of the test is to establish construction methodology and procedures which ensure the required quality of the EBS on-site. By fiscal 2009 some parts of the facility have been constructed, and the test has demonstrated both practicability of the construction and achievement of the quality. They are respectively taken as low-permeability of less than 5x10 13 m/s and low-diffusivity of less than 1x10 -12 m 2 /s at the time of completion of construction. This paper covers the project outline and the test results obtained by the construction of some parts of a bentonite and cement materials. (author)

  12. Current status of the Demonstration Test of Underground Cavern-Type Disposal Facilities

    International Nuclear Information System (INIS)

    Akiyama, Yoshihiro; Terada, Kenji; Oda, Nobuaki; Yada, Tsutomu; Nakajima, Takahiro

    2011-01-01

    In Japan, the underground cavern-type disposal facilities for low-level waste (LLW) with relatively high radioactivity, mainly generated from power reactor decommissioning, and for certain transuranic (TRU) waste, mainly from spent fuel reprocessing, are designed to be constructed in a cavern 50-100 m underground and to employ an engineered barrier system (EBS) made of bentonite and cement materials. To advance a disposal feasibility study, the Japanese government commissioned the Demonstration Test of Underground Cavern-Type Disposal Facilities in fiscal year (FY) 2005. Construction of a full-scale mock-up test facility in an actual subsurface environment started in FY 2007. The main test objective is to establish the construction methodology and procedures that ensure the required quality of the EBS on-site. A portion of the facility was constructed by 2010, and the test has demonstrated both the practicability of the construction and the achievement of quality standards: low permeability of less than 5x10 -13 m/s and low-diffusion of less than 1x10 -12 m 2 /s at the completion of construction. This paper covers the test results from the construction of certain parts using bentonite and cement materials. (author)

  13. A performance test of a capsule for a material irradiation in the OR holes of HANARO

    International Nuclear Information System (INIS)

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

    2008-01-01

    A test for a pressure drop and a vibration was performed to develop a material capsule for an irradiation at the OR hole in HANARO. It was analyzed before the test that a diameter of a material capsule for the OR holes should be more than 49mm by an evaluation of a flow rate and pressure drop in theory. According to this estimation, 3 kinds of mock-up capsules with a diameter of 52, 54, 56 mm were made and applied to a pressure drop test. As a result of the pressure drop test, the requirement for a pressure and a flow rate in HANARO was confirmed to be satisfied for the 3 kinds of diameters. The capsules with diameters of 54, 56mm were applied to a vibration test by taking into consideration a receptive capacity of the specimens. The capsule with a diameter of 56mm satisfied the requirement for an allowable limit of the vibration acceleration applied in HANARO. The heat transfer coefficient and the temperature on the surface of a capsule were estimated. As the temperature on the surface of the capsule was calculated to be 43.7 .deg. C, the ONB condition in HANARO was satisfied

  14. Development and testing of the cooling coil cleaning end effector

    International Nuclear Information System (INIS)

    Johnson, K.I.; Mullen, O.D.; Powell, M.R.; Daly, D.S.; Engel, D.W.

    1997-01-01

    The Retrieval Process Development and Enhancement (KPD ampersand E) program has developed and tested an end effector to support the waste retrieval mission at the Idaho National Engineering and Environmental Laboratory (INEEL). The end effector was developed specifically to remove a sticky waste material from the cooling coils in the High Level Liquid Waste (HLLW) tank, and to vacuum up a sediment layer that has settled beneath the cooling coils. An extensive testing program was conducted in the hydraulic test bed (HTB) at the Pacific Northwest National Laboratory (PNNL) to evaluate the performance of the end effector under simulated in-tank conditions. A mock up of the cooling coils was installed in the test bed tank, and simulated waste materials were included to represent the sticky waste on the tubes and the particulate waste settled beneath them. The testing program focused on assessing long-duration mining strategies for cleaning the cooling coils and removing the particulate waste forms. The report describes the results of the end effector testing program at PNNL. Section 2 describes the physical characteristics of the HLLW tanks, including the layout of the cooling coils, and it also describes what is known of the waste forms in the tanks. Section 3 describes the cleaning and retrieval strategy that was used in developing the end effector design. Section 4 describes the cooling coil mockup in the hydraulic test bed. Section 5 discusses the rationale used in selecting the simulants for the tarry waste and particulate waste forms. Section 6 describes the tests that were performed to evaluate cleaning of the cooling coils and retrieval of the particulate simulant. Section 7 summarizes the cleaning and retrieval tests, assesses the relative importance of cleaning the cooling coils and retrieving the particulate waste, and suggests modifications that would simplify the end effector design

  15. Rapidly building with lightweight modules and a dry assembled foundation : used in a mock-up for home units made out of reclaimed materials

    NARCIS (Netherlands)

    Moonen, S.P.G.; Temmerman, de N; Brebbia, C.A.

    2014-01-01

    The current practice of house building is wasteful, not efficient and badly in need of a reset. Society is ready for a change towards eco-homes; yet industry stays traditional. A transition to a durable, industrial and flexible approach is required. To accomplish such, a system is developed

  16. Design of a neutron interrogation cell based on an electron accelerator and performance assessment on 220 liter nuclear waste mock-up drums

    International Nuclear Information System (INIS)

    Sari, A.; Carrel, F.; Laine, F.; Lyoussi, A.

    2013-01-01

    Radiological characterization of nuclear waste drums is an important task for the nuclear industry. The amount of actinides, such as 235 U or 239 Pu, contained in a package can be determined using non-destructive active methods based on the fission process. One of these techniques, known as neutron interrogation, uses a neutron beam to induce fission reactions on the actinides. Optimization of the neutron flux is an important step towards improving this technique. Electron accelerators enable to achieve higher neutron flux intensities than the ones delivered by deuterium-tritium generators traditionally used on neutron interrogation industrial facilities. In this paper, we design a neutron interrogation cell based on an electron accelerator by MCNPX simulation. We carry out photoneutron interrogation measurements on uranium samples placed at the center of 220 liter nuclear waste drums containing different types of matrices. We quantify impact of the matrix on the prompt neutron signal, on the ratio between the prompt and delayed neutron signals, and on the interrogative neutron half-life time. We also show that characteristics of the conversion target of the electron accelerator enable to improve significantly measurement performances. (authors)

  17. Measurement and analysis of breeding index in the first mock-up core (XXII-1(65V)) for water-cooled breeder reactor at FCA. Contract research

    International Nuclear Information System (INIS)

    Fukushima, Masahiro; Okajima, Shigeaki; Andoh, Masaki; Yamane, Tsuyoshi; Kataoka, Masaharu

    2005-03-01

    Measurements and analyses of breeding index were performed in the FCA-XXII-1(65V) core simulating reduced-moderation lightwater reactor (RMWR) with void fraction 65% of moderator at Fast Critical Assembly (FCA). The measurement of the reaction rate ratio of 238 U capture to 235 U fission (C8/F5) was made by a foil activation technique using depleted and enriched uranium foils, and the reaction rate ratios of 239 Pu fission to 235 U fission (F9/F5) and 238 U fission to 235 U fission (F8/F5) were measured using absolutely calibrated fission chambers. Cell averaged reaction correction factors were derived by the Monte Carlo code (MVP) calculations with modeling the forms and positions of fission chambers and foils. Consequently, cell averaged reaction rate ratios were determined to be C8/F5 of 0.0916±1.4%, F9/F5 of 0.759±1.2% and F8/F5 of 0.0201±0.9%. Therefore, breeding index of C8/F9 was determined to be 0.121 ± 1.8%. The analyses were made by using the JFS-3-J3.2R group constant set which is generated from the JENDL-3.2 nuclear data library. The effective cross sections were calculated by the standard cell calculation code for fast reactor, SLAROM. Three-dimensional diffusion calculations by the CITATION code with 70-group structure have been performed to estimate the reaction rate ratios in the core center. Here, effective cross sections of fuel cells in the core center were obtained using the PEACO-X code with an ultra-fine group structure to consider self-shielding effect in the resonance energy range. Calculation to experiment ratios (C/E) of F9/F5 and F8/F5 were 1.02 and 1.03, respectively. These calculated values slightly overestimate the experimental one. The calculated value of C8/F5 overestimates the experimental one by about 6%. Consequently, the C/E value of C8/F9 was 1.03. The calculated value slightly overestimates the experimental one. The calculation code system developed for the thermal reactor, SRAC code system, was also used to analyze the experiment. There was no large difference in analysis results between the fast and thermal calculation codes. (author)

  18. "Revealing hidden paint layers in oil paintings by means of scanning macro-XRF: a mock-up study based on Rembrandt's ""An old man in military costume"""

    OpenAIRE

    Alfeld, Matthias; De Nolf, Wout; Cagno, Simone; Appel, Karen; Siddons, D. Peter; Kuczewski, Anthony; Janssens, Koen; Dik, Joris; Trentelman, Karen; Walton, Marc; Sartorius, Andrea

    2013-01-01

    Over the past several decades the oeuvre of Rembrandt has been the subject of extensive art historical and scientific investigations. One of the most striking features to emerge is his frequent re-use of canvases and panels. The painting An Old Man in Military Costume (78.PB.246), in the collection of the J. Paul Getty Museum, is an example of such a re-used panel. Conventional imaging techniques revealed the presence of a second portrait under the surface portrait, but the details of this hi...

  19. Jesters, tricksters, taggers and haints: Hipping the church to the Afro-hop, pop-‘n-lock mock-up currently rocking apocalyptic Detroit

    Directory of Open Access Journals (Sweden)

    James W. Perkinson

    2017-11-01

    Full Text Available The following essay investigates the animating force of jester-humour and trickster-critique as necessary components of prophetic consciousness and social movement. Climate change devastation coupled with racialised socio-economic predation today faces social movement with a stark demand. The root-work necessary enjoins challenge of human presumption about the meaning of life at the most basic level. The locus from which such a depth-exploration will be elaborated here is postindustrial Detroit, on the part of a poet-activist-educator who will insist that ‘jesterism’ as ‘prophetic animation’ cannot merely be ‘talked about’, but begs performance and embodiment – even in the process of writing and theorising. Indigenous wisdom and folk spirituality will supply historical perspective in asserting laughter as both antidote to trauma and tactic of critique – whether looking at traditional African practices of tricksterism reincarnate in everyday street life in Detroit, medieval Christian celebrations of the Feast of Fools subverting official Church orthodoxies in feudal Europe, or the postmodern insurgence of hip-hop beats and tags in challenging corporate gentrification and church capitulation at the emblematic heart of de-industrialisation.

  20. Using inertial measurement units originally developed for biomechanics for modal testing of civil engineering structures

    Science.gov (United States)

    Hester, David; Brownjohn, James; Bocian, Mateusz; Xu, Yan; Quattrone, Antonino

    2018-05-01

    This paper explores the use of wireless Inertial Measurement Units (IMU) originally developed for bio-mechanical research applications for modal testing of civil engineering infrastructure. Due to their biomechanics origin, these devices combine a triaxial accelerometer with gyroscopes and magnetometers for orientation, as well as on board data logging capability and wireless communication for optional data streaming and to coordinate synchronisation with other IMUs in a network. The motivation for application to civil structures is that their capabilities and simple operating procedures make them suitable for modal testing of many types of civil infrastructure of limited dimension including footbridges and floors while also enabling recovering of dynamic forces generated and applied to structures by moving humans. To explore their capabilities in civil applications, the IMUs are evaluated through modal tests on three different structures with increasing challenge of spatial and environmental complexity. These are, a full-scale floor mock-up in a laboratory, a short span road bridge and a seven story office tower. For each case, the results from the IMUs are compared with those from a conventional wired system to identify the limitations. The main conclusion is that the relatively high noise floor and limited communication range will not be a serious limitation in the great majority of typical civil modal test applications where convenient operation is a significant advantage over conventional wired systems.

  1. Controlled Cold Helium Spill Test in the LHC Tunnel at CERN

    Science.gov (United States)

    Koettig, T.; Casas-Cubillos, J.; Chorowski, M.; Dufay-Chanat, L.; Grabowski, M.; Jedrusyna, A.; Lindell, G.; Nonis, M.; Vauthier, N.; van Weelderen, R.; Winkler, T.; Bremer, J.

    The helium cooled magnets of the LHC particle accelerator are installed in a confined space, formed by a 27 km circumference 3.8 m diameter underground tunnel. The vacuum enclosures of the superconducting LHC magnets are protected by a lift plate against excessive overpressure created by eventual leaks from the magnet helium bath, or from the helium supply headers. A three-meter long no stay zone has been defined centered to these plates, based on earlier scale model studies, to protect the personnel against the consequences of an eventual opening of such a lift plate. More recently several simulation studies have been carried out modelling the propagation of the resulting helium/air mixture along the tunnel in case of such a cold helium release at a rate in the range of 1 kg/s. To validate the different scale models and simulation studies, real life mock-up tests have been performed in the LHC, releasing about 1000 liter of liquid helium under standard operational tunnel conditions. Data recorded during these tests include oxygen level, temperature and flow speed as well as video recordings, taken up- and downstream of the spill point (-100 m to +200 m) with respect to the ventilation direction in the LHC tunnel. The experimental set-up and measurement results are presented. Generic effects found during the tests will be discussed to allow the transposal to possible cold helium release cases in similar facilities.

  2. Control system of test and research facilities for nuclear energy industry

    International Nuclear Information System (INIS)

    1983-01-01

    IHI manufactures several kinds of test and research facilities used for research and development of new type power reactor and solidification system of high level radioactive liquid waste and safety research of light water reactor. These facilities are usually new type plants themselves, so that their control systems have to be designed individually for each plant with the basic conception. They have many operation modes because of their purposes of research and development, so the operation has to be automatized and requires the complicated sequence control system. In addition to these requirements, the detail design is hardly fixed on schedule and often modified during the initial start up period. Therefore, the computer control system was applied to these facilities with CRT display for man-machine communication earlier than to commercial power plants, because in the computer system the control logic is not hard wired but soft programmed and can be easily modified. In this paper, two typical computer control systems, one for PWR reflood test facility and another for mock-up test facility for solidification of liquid waste, are introduced. (author)

  3. Radiant heat testing of the H1224A shipping/storage container

    Energy Technology Data Exchange (ETDEWEB)

    Harding, D.C.; Bobbe, J.G.; Stenberg, D.R.; Arviso, M.

    1994-05-01

    H1224A weapons containers have been used for years by the Departments of Energy and Defense to transport and store W78 warhead midsections. Although designed to protect the midsections only from low-energy impacts, a recent transportation risk assessment effort has identified a need to evaluate the container`s ability to protect weapons in more severe accident environments. Four radiant heat tests were performed: two each on an H1224A container (with a Mk12a Mod 6c mass mock-up midsection inside) and two on a low-cost simulated H1224A container (with a hollow Mk12 aeroshell midsections inside). For each unit tested, temperatures were recorded at numerous points throughout the container and midsection during a 4-hour 121{degrees}C (250{degrees}F) and 30-minute 1010{degrees}C (1850{degrees}F) radiant environment. Measured peak temperatures experienced by the inner walls of the midsections as a result of exposure to the high-temperature radiant environment ranged from 650{degrees} C to 980{degrees} C (1200{degrees} F to 1800{degrees}F) for the H1224A container and 770 {degrees} to 990 {degrees}C (1420{degrees} F to 1810{degrees}F) for the simulated container. The majority of both containers were completely destroyed during the high-temperature test. Temperature profiles will be used to benchmark analytical models and predict warhead midsection temperatures over a wide range of the thermal accident conditions.

  4. Progress in the integration of Test Blanket Systems in ITER equatorial port cells and in the interfaces definition

    Energy Technology Data Exchange (ETDEWEB)

    Pascal, R., E-mail: romain.pascal@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Beloglazov, S.; Bonagiri, S. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Commin, L. [CEA, IRFM, Cadarache (France); Cortes, P.; Giancarli, L.M.; Gliss, C.; Iseli, M.; Lanza, R.; Levesy, B.; Martins, J.-P. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Neviere, J.-C. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Patisson, L.; Plutino, D.; Shu, W. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Swami, H.L. [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer The design integration of two test blanket systems in ITER port cell is addressed. Black-Right-Pointing-Pointer Definition of interfaces of TBSs with building and other ITER systems is done. Black-Right-Pointing-Pointer Designs of pipe forest, bioshield plug and ancillary equipment unit are described. Black-Right-Pointing-Pointer The maintenance of the two test blanket systems in ITER port cell is considered. Black-Right-Pointing-Pointer The management of the heat and tritium releases in the TBM port cell is described. - Abstract: In the framework of the TBM Program, three ITER vacuum vessel equatorial ports (no. 16, no. 18 and no. 02) have been allocated for the testing of up to six mock-ups of six different DEMO tritium breeding blankets. Each one is called a Test Blanket System (TBS). A TBS consists mainly of the Test Blanket Module (TBM), the in-vessel component facing the plasma, and several ancillary systems, in particular the cooling system and the tritium extraction system. Each port accommodates two TBMs and therefore the two TBSs have to share the corresponding port cell. This paper deals with the design integration aspects of the two TBSs in each port cell performed at ITER Organization (IO) with the corresponding definition of interfaces with other ITER systems. The performed activities have raised several issues that are discussed in the paper and for which design solutions are proposed.

  5. Task Order 22 – Engineering and Technical Support, Deep Borehole Field Test. AREVA Summary Review Report

    Energy Technology Data Exchange (ETDEWEB)

    Denton, Mark A. [AREVA Federal Services, Charlotte, NC (United States)

    2016-01-19

    radiological release due to off-normal events are relatively low; (4) costs are relatively low; and (5) maintenance activities are relatively simple. The primary drawback associated with the wireline emplacement mode for DBD is the number of emplacement trips-in to the borehole, which results in a relatively higher probability for a drop event. Fortunately, the WPs can be engineered with impact limiters that will minimize the likelihood of a breach of the WP due to a drop. The WP designs presented in the M2 report appear to be focused on compatibility with the drill-string emplacement mode (e.g., the threaded connections). With the recommendation that the wireline emplacement mode be utilized for the DBFT, some changes may be warranted to these WPs. For example, the development of a WP release connection that is more reliable than the currently credited connection, which is considered to have a high failure probability, and the integration of an impact limiter into its design. The M2 report states the engineering demonstration of the DBFT will occur in the FTB over a 4-year period. AREVA recommends development and testing of the WP emplacement handling equipment occur separately (but concurrently, if not earlier) from the FTB at a mock-up facility. The separation of this activity would prevent schedule interference between the science and engineering thrusts of the project. Performing tests in a mock-up facility would allow additional control and observation compared to the FTB. The mock-up facility could also be utilized as a training facility for future operations. Terminal velocity and impact limiter testing would require the FTB for testing, since these areas would be difficult to reproduce in a limited depth mock-up. Although only at the end of the conceptual stage of design development, DBD appears to be a viable solution for some waste forms produced by the nuclear industry. However, regulatory requirements have yet to be established for pre- and post

  6. Simulation of loss of feedwater transient of MASLWR test facility by MARS-KS code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Juyeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is being current developed domestically also adopts helical coil steam generator, KINS has joined this ICSP to evaluate performance of domestic regulatory audit thermal-hydraulic code (MARS-KS code) in various respects including wall-to-fluid heat transfer model modification implemented in the code by independent international experiment database. In the ICSP, two types of transient experiments have been focused and they are loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels (SP-3). In the present study, KINS simulation results by the MARS-KS code (KS-002 version) for the SP-2 experiment are presented in detail and conclusions on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the loss of feedwater transient of the MASLWR test facility. Steady state run shows helical coil specific heat transfer models implemented in the code is reasonable. However, through the transient run, it is also found that three-dimensional effect within the HPC and axial conduction effect through the HTP are not well reproduced by the code.

  7. Final report on the Controlled Cold Helium Spill Test in the LHC tunnel at CERN

    CERN Document Server

    Dufay-Chanat, L; Casas-Cubillos, J; Chorowski, M; Grabowski, M; Jedrusyna, A; Lindell, G; Nonis, M; Koettig, T; Vauthier, N; van Weelderen, R; Winkler, T

    2015-01-01

    The 27 km circumference LHC underground tunnel is a space in which the helium cooled LHC magnets are installed. The vacuum enclosures of the superconducting magnets are protected by over-pressure safety relief devices that open whenever cold helium escapes either from the magnet cold enclosure or from the helium supply headers, into this vacuum enclosure. A 3-m long no stay zone around these devices is defined based on scale model studies, protecting the personnel against cold burns or asphyxia caused by such a helium release event. Recently, several simulation studies have been carried out modelling the propagation of the helium/air mixture, resulting from the opening of such a safety device, along the tunnel. The released helium flows vary in the range between 1 kg/s and 0.1 kg/s. To validate these different simulation studies, real life mock-up tests have been performed inside the LHC tunnel, releasing helium flow rates of 1 kg/s, 0.3 kg/s and 0.1 kg/s. For each test, up to 1000 liters of liquid helium wer...

  8. Full scale BWR containment LOCA response test at the INKA test facility

    International Nuclear Information System (INIS)

    Wagner, Thomas; Leyer, Stephan

    2015-01-01

    KERENA is an innovative boiling water reactor concept with passive safety systems (Generation III+) of AREVA. The reactor is an evolutionary design of operating BWRs (Generation II). In order to verify the functionality and performance of the KERENA safety concept required for the transient and accident management, the test facility “Integral Teststand Karlstein” (INKA) was built at Karlstein (Germany). It is a mock-up of the KERENA boiling water reactor containment, with integrated pressure suppression system. The complete chain of passive safety components is available. The passive components and the levels are represented in full scale. The volume scaling of the containment compartments is approximately 1:24. The reactor pressure vessel (RPV) is simulated via the steam accumulator of the Karlstein Large Valve Test Facility. This vessel provides an energy storage capacity of approximately 1/6 of the KERENA RPV and is supplied by a Benson boiler with a thermal power of 22 MW. With respect to the available power supply, the containment- and system-sizing of the facility is by far the largest one of its kind worldwide. From 2009 to 2012, several single component tests were conducted (Emergency Condenser, Containment Cooling Condenser, Core Flooding System etc.). On March 21st, 2013, the worldwide first large-scale only passively managed integral accident test of a boiling water reactor was simulated at INKA. The integral test measured the combined response of the KERENA passive safety systems to the postulated initiating event was the “Main Steam Line Break” (MSLB) inside the Containment with decay heat simulation. The results of the performed integral test (MSLB) showed that the passive safety systems alone are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. Therefore the test verified the function of those components and the interplay between them as response to an anticipated accident scenario

  9. 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…

  10. Control room design and human factors using a virtual reality based tool for design, test and training

    International Nuclear Information System (INIS)

    Lirvall, Peter

    1998-02-01

    This report describes a user-centred approach to control room design adopted by IFE for the nuclear industry. The novelty of this approach is the development of a Control Room Philosophy, and the use of Virtual Reality (VR) technology as a tool in the design process, integrated with a specially developed Design Documentation System (DDS) and a process display prototyping tool PICASSO-3. The control room philosophy identifies all functional aspects of a control centre, to define the baseline principles and guidelines for the design. The use of VR technology enables end-users of the control room design (e.g. control room operators) to specify their preferred design of the new control room, and to replace the need for a physical mock-up to test and evaluate the proposed design. The DDS, integrated with the VR design tool, guides the control room operators, through a structured approach, to document the proposed design in a complete design specification. The VR tool, specially developed by IFE, is called the VR based Design, Test and Training tool (VR DTandT). It is not only intended to visualise the design, but also to test and evaluate the design. When the design is implemented, the same model is re-used as a VR based training simulator for operators. A special feature in the VR DTandT tool is that the verification and validation (VandV) tests, concerning human factors, are according to the regulative standards for nuclear control rooms

  11. Design of helium-gas supplying facility of out-of-pile demonstration test for HTTR heat utilization system

    Energy Technology Data Exchange (ETDEWEB)

    Hino, Ryutaro; Fujisaki, Katsuo; Kobayashi, Toshiaki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    1996-09-01

    One of the objectives of the High-Temperature Engineering Test Reactor (HTTR) is to demonstrate effectiveness of high-temperature heat utilization. Prior to connect a heat utilization system to the HTTR, a series of out-of-pile demonstration test is indispensable to improve components` performance, to demonstrate operation, control and safety technologies and to verify analysis codes for design and safety evaluation. After critical review and discussion on the out-of-pile demonstration test, a test facility have been designed. In this report, a helium-gas supplying facility simulated the HTTR system was described in detail, which supplies High-temperature helium-gas of 900degC to a steam reforming facility mocking-up the HTTR heat utilization system. Components of the Helium Engineering Demonstration Loop (HENDEL) were selected to reuse in the helium-gas supplying facility in order to decrease construction cost. Structures and specifications of new components such as a high-temperature heater and a preheater were decided after evaluation of thermal and hydraulic performance and strength. (author)

  12. Integrated Suit Test 1 - A Study to Evaluate Effects of Suit Weight, Pressure, and Kinematics on Human Performance during Lunar Ambulation

    Science.gov (United States)

    Gernhardt, Michael L.; Norcross, Jason; Vos, Jessica R.

    2008-01-01

    In an effort to design the next generation Lunar suit, NASA has initiated a series of tests aimed at understanding the human physiological and biomechanical affects of space suits under a variety of conditions. The first of these tests was the EVA Walkback Test (ICES 2007-01-3133). NASA-JSC assembled a multi-disciplinary team to conduct the second test of the series, titled Integrated Suit Test 1 (IST-1), from March 6 through July 24, 2007. Similar to the Walkback Test, this study was performed with the Mark III (MKIII) EVA Technology Demonstrator suit, a treadmill, and the Partial Gravity Simulator in the Space Vehicle Mock-Up Facility at Johnson Space Center. The data collected for IST-1 included metabolic rates, ground reaction forces, biomechanics, and subjective workload and controllability feedback on both suited and unsuited (shirt-sleeve) astronaut subjects. For IST-1 the center of gravity was controlled to a nearly perfect position while the weight, pressure and biomechanics (waist locked vs. unlocked) were varied individually to evaluate the effects of each on the ability to perform level (0 degree incline) ambulation in simulated Lunar gravity. The detailed test methodology and preliminary key findings of IST-1 are summarized in this report.

  13. Electron beam facility for divertor target experiments

    International Nuclear Information System (INIS)

    Anisimov, A.; Gagen-Torn, V.; Giniyatulin, R.N.

    1994-01-01

    To test different concepts of divertor targets and bumpers an electron beam facility was assembled in Efremov Institute. It consists of a vacuum chamber (3m 3 ), vacuum pump, electron beam gun, manipulator to place and remove the samples, water loop and liquid metal loop. The following diagnostics of mock-ups is stipulated: (1) temperature distribution on the mock-up working surface (scanning pyrometer and infra-red imager); (2) temperature distribution over mocked-up thickness in 3 typical cross-sections (thermo-couples); (3) cracking dynamics during thermal cycling (acoustic-emission method), (4) defects in the mock-up before and after tests (ultra-sonic diagnostics, electron and optical microscopes). Carbon-based and beryllium mock-ups are made for experimental feasibility study of water and liquid-metal-cooled divertor/bumper concepts

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-02-15

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

  15. Manufacturing and High Heat Flux Testing of Brazed Flat-Type W/CuCrZr Plasma Facing Components

    Science.gov (United States)

    Lian, Youyun; Liu, Xiang; Feng, Fan; Chen, Lei; Cheng, Zhengkui; Wang, Jin; Chen, Jiming

    2016-02-01

    Water-cooled flat-type W/CuCrZr plasma facing components with an interlayer of oxygen-free copper (OFC) have been developed by using vacuum brazing route. The OFC layer for the accommodation of thermal stresses was cast onto the surface of W at a temperature range of 1150 °C-1200 °C in a vacuum furnace. The W/OFC cast tiles were vacuum brazed to a CuCrZr heat sink at 940 °C using the silver-free filler material CuMnSiCr. The microstructure, bonding strength, and high heat flux properties of the brazed W/CuCrZr joint samples were investigated. The W/Cu joint exhibits an average tensile strength of 134 MPa, which is about the same strength as pure annealed copper. High heat flux tests were performed in the electron beam facility EMS-60. Experimental results indicated that the brazed W/CuCrZr mock-up experienced screening tests of up to 15 MW/m2 and cyclic tests of 9 MW/m2 for 1000 cycles without visible damage. supported by National Natural Science Foundation of China (No. 11205049) and the National Magnetic Confinement Fusion Science Program of China (No. 2011GB110004)

  16. Measurement of the momentum transferred between contacting bodies during the LISA test-mass release phase—uncertainty estimation

    International Nuclear Information System (INIS)

    De Cecco, M; Bortoluzzi, D; Da Lio, M; Baglivo, L; Benedetti, M

    2009-01-01

    The requirements for the Laser Interferometer Space Antenna (LISA) test-mass (TM) release phase are analysed in view of the building up of a testing facility aimed at on-Earth qualification of the release mechanism. Accordingly, the release of the TM to free-fall must provide a linear momentum transferred to the TM not exceeding 10 −5 kg m s −1 . In order to test this requirement, a double pendulum system has been developed. The mock-ups of the TM and the release-dedicated plunger are brought into contact and then the latter is quickly retracted. During and after release, the TM motion is measured by a laser interferometer. The transferred momentum is estimated from the free oscillations following the plunger retraction by means of a Wiener–Kolmogorov optimal filter. This work is aimed at modelling the measurement chain, taking into account procedure, instruments, mechanisms and data elaboration in order to estimate the uncertainty associated with the transferred momentum measurement by means of Monte Carlo simulation

  17. The development and testing of the thermal break divertor monoblock target design delivering 20 MW m-2 heat load capability

    Science.gov (United States)

    Fursdon, M.; Barrett, T.; Domptail, F.; Evans, Ll M.; Luzginova, N.; Greuner, N. H.; You, J.-H.; Li, M.; Richou, M.; Gallay, F.; Visca, E.

    2017-12-01

    The design and development of a novel plasma facing component (for fusion power plants) is described. The component uses the existing ‘monoblock’ construction which consists of a tungsten ‘block’ joined via a copper interlayer to a through CuCrZr cooling pipe. In the new concept the interlayer stiffness and conductivity properties are tuned so that stress in the principal structural element of the component (the cooling pipe) is reduced. Following initial trials with off-the-shelf materials, the concept was realized by machined features in an otherwise solid copper interlayer. The shape and distribution of the features were tuned by finite element analyses subject to ITER structural design criterion in-vessel components (SDC-IC) design rules. Proof of concept mock-ups were manufactured using a two stage brazing process verified by tomography and micrographic inspection. Full assemblies were inspected using ultrasound and thermographic (SATIR) test methods at ENEA and CEA respectively. High heat flux tests using IPP’s GLADIS facility showed that 200 cycles at 20 MW m-2 and five cycles at 25 MW m-2 could be sustained without apparent component damage. Further testing and component development is planned.

  18. The development and testing of the thermal break divertor monoblock target design delivering 20 MW m−2 heat load capability

    International Nuclear Information System (INIS)

    Fursdon, M; Barrett, T; Domptail, F; Evans, Ll M; Luzginova, N; Greuner, N H; You, J-H; Li, M; Richou, M; Gallay, F; Visca, E

    2017-01-01

    The design and development of a novel plasma facing component (for fusion power plants) is described. The component uses the existing ‘monoblock’ construction which consists of a tungsten ‘block’ joined via a copper interlayer to a through CuCrZr cooling pipe. In the new concept the interlayer stiffness and conductivity properties are tuned so that stress in the principal structural element of the component (the cooling pipe) is reduced. Following initial trials with off-the-shelf materials, the concept was realized by machined features in an otherwise solid copper interlayer. The shape and distribution of the features were tuned by finite element analyses subject to ITER structural design criterion in-vessel components (SDC-IC) design rules. Proof of concept mock-ups were manufactured using a two stage brazing process verified by tomography and micrographic inspection. Full assemblies were inspected using ultrasound and thermographic (SATIR) test methods at ENEA and CEA respectively. High heat flux tests using IPP’s GLADIS facility showed that 200 cycles at 20 MW m −2 and five cycles at 25 MW m −2 could be sustained without apparent component damage. Further testing and component development is planned. (paper)

  19. Manufacturing and testing of relevant scale mockup based on monoblock concept

    International Nuclear Information System (INIS)

    Di Pietro, E.; Orsini, A.; Sacchetti, M.; Libera, S.; Cardella, A.; Vieider, G.

    1993-01-01

    The results obtained from small-scale mockups manufactured on the monoblock design concept have proven that the solution appears promising for a conventional divertor operating with heat fluxes in the range 10 to 15 MW/m 2 with a thermal fatigue cycle exceeding 1000 cycles at full power. The divertor mock-up consists of six half meter-long armored tubes obtained by brazing CFC to TZM molybdenum alloy. Two types of CFC were used to investigate the advantages of 3-d CFCs with respect to more conventional and cheaper 2-d CFC. The brazing process utilizes three variants of a process developed in laboratory trials and based on selected combinations of active braze filler/CFC surface conditioning procedures. The supporting structure is based on the sliding support concept intended to assure a compromise between the requested thermal stability of the component and the buildup of secondary stresses deriving from mechanical constraints. The FE thermal and thermal mechanical analysis of the divertor mockup structure is reported and the critical areas of sliding support are highlighted for comparison with experimental results. The main results of NDE and experimental high heat flux tests are reported and discussed

  20. Malfunction tests and vibration analysis of P.W.R. internal structures

    International Nuclear Information System (INIS)

    Puyal, C.; Carre, J.C.; Epstein, A.

    1987-01-01

    To diagnose changes liable to occur in the vibration behavior of internals, it is important to understand the influence of changes in the mechanical properties of elements on the output signals obtained from neutron chambers placed out of core and accelerometers fixed to the reactor vessel. To do this, the effects of changes liable to occur in the hold-down springs and the flexures were simulated on the SAFRAN loop, using a representative hydroelastic mock-up. The results obtained experimentally on SAFRAN for different characteristics of the hold-down spring, which lies between the upper part of the core barrel and the vessel head, have been published. In this paper, we propose to present the results of the investigation of the fracture of one or more flexures which connect the cylindrical thermal shield to the core barrel. This work is in two parts: a) Computation based on a hydroelastic model using the substructuration computer program TRISTANA of the CASTEM system. b) Tests simulating flexure fracture: 1 - in air, for an understanding of the mechanisms involved; 2 - on the SAFRAN loop with a representative flow in order to estimate the strains liable to exist on the vibration signatures recorded on displacement transducers and accelerometers. Good agreement was observed between the computation results with the theoretical model employed and those obtained experimentally [fr

  1. Manufacturing and testing in reactor relevant conditions of brazed plasma facing components of the ITER divertor

    International Nuclear Information System (INIS)

    Bisio, M.; Branca, V.; Marco, M. Di; Federici, A.; Grattarola, M.; Gualco, G.; Guarnone, P.; Luconi, U.; Merola, M.; Ozzano, C.; Pasquale, G.; Poggi, P.; Rizzo, S.; Varone, F.

    2005-01-01

    A fabrication route based on brazing technology has been developed for the realization of the high heat flux components for the ITER vertical target and Dome-Liner. The divertor vertical target is armoured with carbon fiber reinforced carbon and tungsten in the lower straight part and in the upper curved part, respectively. The armour material is joined to heat sinks made of precipitation hardened copper-chromium-zirconium alloy. The plasma facing units of the dome component are based on a tungsten flat tile design with hypervapotron cooling. An innovative brazing technique based on the addition of carbon fibers to the active brazing alloy, developed by Ansaldo Ricerche for applications in the field of the energy production, has been used for the carbon fiber composite to copper joint to reduce residual stresses. The tungsten-copper joint has been realized by direct casting. A proper brazing thermal cycle has been studied to guarantee the required mechanical properties of the precipitation hardened alloy after brazing. The fabrication route of plasma facing components for the ITER vertical target and dome based on the brazing technology has been proved by means of thermal fatigue tests performed on mock-ups in reactor relevant conditions

  2. Approach to the assessment of the performance of nondestructive test methods in the manufacture of nuclear power station equipment

    International Nuclear Information System (INIS)

    Michaut, J.P.

    1996-01-01

    The safety of a nuclear power station lies largely in the possibility of ensuring, at the time of in service inspections on major equipment, that the extent of faults which may appear or develop is not greater than that of faults detrimental to behavior in service. This assurance is based on performance demonstration of the nondestructive test methods used for inspecting the equipment in service. This is the subject of numerous studies in various countries. To ensure that manufacturing faults likely to downgrade the safety of the equipment are not discovered in service, it seems desirable to make sure that the performance of the nondestructive test (NDT) methods which are going to be used in manufacture will be at least as high as those used in service and that they are therefore capable of guaranteeing detection of faults clearly less important than really harmful faults. The performance of NDT methods and their consistency with those which can be used in service is evaluated before the start of manufacture on a mock-up representative of the equipment itself. Information is given on research in progress on the bimetal welding of a pressurizer spray nozzle

  3. The State of the Art Report on the Development and Manufacturing Technology of Test Blanket Module

    International Nuclear Information System (INIS)

    Lee, J. S.; Jeong, Y. H.; Park, S. Y.; Lee, M. H.; Choi, B. K.; Baek, J. H.; Park, J. Y.; Kim, J. H.; Kim, H. G.; Kim, K. H.

    2006-07-01

    The main objective of the present R and D on breeder blanket is the development of test blanket modules (TBMs) to be installed and tested in International Thermonuclear Experimental Reactor (ITER). In the program of the blanket development, a blanket module test in the ITER is scheduled from the beginning of the ITER operation, and the performance test of TBM in ITER is the most important milestone for the development of the DEMO blanket. The fabrication of TBMs has been required to test the basic performance of the DEMO blanket, i.e., tritium production/recovery, high-grade heat generation and radiation shielding. Therefore, the integration of the TBM systems into ITER has been investigated with the aim to check the safety, reliability and compatibility under nuclear fusion state. For this reason, in the Test Blanket Working Group (TBWG) as an activity of the International Energy Association (IEA), a variety of ITER TBMs have been proposed and investigated by each party: helium-cooled ceramic (WSG-1), helium-cooled LiPb (WSG-2), water-cooled ceramic (WSG-3), self-cooled lithium (WSG-4) and self-cooled molten salt (WSG-5) blanket systems. Because we are still deficient in investigation of TBM development, the need of development became pressing. In this report, for the development of TBM sub-module and mock-up, it is necessary to analyze and examine the state of the art on the development of manufacturing technology of TBM in other countries. And we will be applied as basic data to establish a manufacturing technology

  4. Dynamic test of the ITER blanket key and ceramic insulated pad

    International Nuclear Information System (INIS)

    Khomyakov, S.; Sysoev, G.; Strebkov, Yu.; Kucherov, A.; Ioki, K.

    2010-01-01

    The dynamic testing of the blanket module's key integrated into ITER vacuum vessel portion has been performed in 2008 to investigate its capability to react the electro-magnetic (EM) loads. The preliminary analysis showed the large dynamic amplification factor (DAF) of the reactions because of technological gaps between the blanket module and key. Shock load may yield the bronze pads, which protect the blanket electrical insulation from damage. However the dynamic analysis of such particularly non-linear system needs an experimental ground and confirmation. Toward this end, as well as demonstration of the key reliability, the special test facility has been made, and the full-scale mock-up of the inboard intermodular key was tested. So as not to scale non-linear dynamic parameters, 1-ton mass was built on the single flexible support. The key was welded in a 60-mm thick steel plate modeled with a fragment of the VV. The different gaps were set in between the bronze pad of the key and the mass shock worker. This system (supplemented with some additional constraints) has natural oscillations like as the 4-ton module built on four flexible supports. Thus the most critical radial torque might be modeled with a straight force. The objectives of the test were as follows: dynamic response, DAF and damping factor determination; measurement of the strain oscillations in the key's base and in the weld seam; comparison of the measured data with computation results. The paper will present the analytical grounds of the testing conditions, test facility description, analytical adaptation of the facility, experimental results, its comparison with analysis and discussion, and guidelines for the next experimental phase.

  5. Development of ultrasonic testing technique with a large transducer to inspect the containment vessel plates embedded in concrete for corrosion on nuclear power plant (2)

    International Nuclear Information System (INIS)

    Ishida, Hitoshi

    2005-01-01

    The containment vessel plates embedded in concrete on Pressurized Water Reactors are inaccessible to inspect directly. Therefore, it is advisable to prepare inspection technology to detect existence and a location of corrosion on the embedded plates indirectly. The purpose of this study is establishment of ultrasonic testing technique to be able to inspect the containment vessel plates embedded in concrete widely from the accessible point. Experiments to detect artificial hollows simulating corrosion and stud bolts which hold the mold of concrete on a surface of a carbon steel plate mock-up covered with concrete were carried out with newly made low frequency (0.3MHz and 0.5MHz) 90 degrees refraction angle shear horizontal (SH) wave transducers combined with three active elements, which were equivalent to a 120 mm width element. As the results: (1) The echoes from the artificial hollows with a depth of 19 mm and 9.5mm at a distance of 1.5 m and the stud bolts with a diameter of 8mm at a distance of 0.7 - 1.7m could be discriminated clearly. (2) The multiple echoes bouncing three times between the front side and the back side of the plate, which was equivalent to a distance of about 12m, could be discriminated. (3) A divergence angle and a -6dB divergence angle of the large element (combined three elements) transducer were about 7 degrees and about 3 degrees. (4) The echoes from the hollows with a depth of 9.5m could be detected at a distance of 3.6 m with a reflection at the side wall of the mock-up. (5) It was estimated that the maximum distance of detection of the echo from the stud bolt with a diameter of 8mm was about 2.9 ∼ 3.6 m. Therefore we evaluate that the large element transducer can propagate the SH wave to about a half of a distance to the bottom of the embedded containment vessel and it is possible to detect the defects such as corrosion to a distance of 3.6 m. (author)

  6. Control and Non-Payload Communications (CNPC) Prototype Radio - Generation 2 Security Architecture Lab Test Report

    Science.gov (United States)

    Iannicca, Dennis C.; McKim, James H.; Stewart, David H.; Thadhani, Suresh K.; Young, Daniel P.

    2015-01-01

    NASA Glenn Research Center, in cooperation with Rockwell Collins, is working to develop a prototype Control and Non-Payload Communications (CNPC) radio platform as part of NASA Integrated Systems Research Program's (ISRP) Unmanned Aircraft Systems (UAS) Integration in the National Airspace System (NAS) project. A primary focus of the project is to work with the FAA and industry standards bodies to build and demonstrate a safe, secure, and efficient CNPC architecture that can be used by industry to evaluate the feasibility of deploying a system using these technologies in an operational capacity. GRC has been working in conjunction with these groups to assess threats, identify security requirements, and to develop a system of standards-based security controls that can be applied to the current GRC prototype CNPC architecture as a demonstration platform. The security controls were integrated into a lab test bed mock-up of the Mobile IPv6 architecture currently being used for NASA flight testing, and a series of network tests were conducted to evaluate the security overhead of the controls compared to the baseline CNPC link without any security. The aim of testing was to evaluate the performance impact of the additional security control overhead when added to the Mobile IPv6 architecture in various modes of operation. The statistics collected included packet captures at points along the path to gauge packet size as the sample data traversed the CNPC network, round trip latency, jitter, and throughput. The effort involved a series of tests of the baseline link, a link with Robust Header Compression (ROHC) and without security controls, a link with security controls and without ROHC, and finally a link with both ROHC and security controls enabled. The effort demonstrated that ROHC is both desirable and necessary to offset the additional expected overhead of applying security controls to the CNPC link.

  7. Thermo-mechanical tests on W7-X current lead flanges

    International Nuclear Information System (INIS)

    Dhard, Chandra Prakash; Rummel, Thomas; Zacharias, Daniel; Bykov, Victor; Moennich, Thomas; Buscher, Klaus-Peter

    2013-01-01

    Highlights: • There are significant mechanical loads on the cryostat and radial flanges for W7-X current leads. • These are due to evacuation of W7-X cryostat, cool-down of cold mass, electro-magnetic forces and self weight of leads. • The actual mechanical loads were reduced to simplify the experimental set-up. • The tests were carried out on mock-up flanges test assembly at ambient temperature and at 77 K. • The thermo-mechanical tests on W7-X current lead flanges validate the design and joints of these flanges to the leads. -- Abstract: Fourteen pieces of high temperature superconducting current leads (CL) arranged in seven pairs, will be installed on the outer vessel of Wendelstein 7-X (W7-X) stellarator. In order to support the CL, it is provided with two glass fiber reinforce plastic (GFRP) flanges, namely, the lower cryostat flange (CF) remaining at room temperature and upper radial flange (RF) at about 5 K. Both the flanges i.e. CF and RF experience high mechanical loads with respect to the CL, due to the evacuation of W7-X cryostat, cool-down of cold mass including the CL, electro-magnetic forces due to current and plasma operations and self weight of CL. In order to check the integrity of these flanges for such mechanical loads, thermo-mechanical tests were carried out on these flanges at room temperatures and at liquid nitrogen (LN2) temperatures. The details of test set-up, results and modeling are described in the paper

  8. Orion Crew Module / Service Module Structural Weight and Center of Gravity Simulator and Vehicle Motion Simulator Hoist Structure for Orion Service Module Umbilical Testing

    Science.gov (United States)

    Ascoli, Peter A.; Haddock, Michael H.

    2014-01-01

    An Orion Crew Module Service Module Structural Weight and Center of Gravity Simulator and a Vehicle Motion Simulator Hoist Structure for Orion Service Module Umbilical Testing were designed during a summer 2014 internship in Kennedy Space Centers Structures and Mechanisms Design Branch. The simulator is a structure that supports ballast, which will be integrated into an existing Orion mock-up to simulate the mass properties of the Exploration Mission-1 flight vehicle in both fueled and unfueled states. The simulator mimics these configurations through the use of approximately 40,000 lbf of steel and water ballast, and a steel support structure. Draining four water tanks, which house the water ballast, transitions the simulator from the fueled to unfueled mass properties. The Ground Systems Development and Operations organization will utilize the simulator to verify and validate equipment used to maneuver and transport the Orion spacecraft in its fueled and unfueled configurations. The second design comprises a cantilevered tripod hoist structure that provides the capability to position a large Orion Service Module Umbilical in proximity to the Vehicle Motion Simulator. The Ground Systems Development and Operations organization will utilize the Vehicle Motion Simulator, with the hoist structure attached, to test the Orion Service Module Umbilical for proper operation prior to installation on the Mobile Launcher. Overall, these two designs provide NASA engineers viable concepts worthy of fabricating and placing into service to prepare for the launch of Orion in 2017.

  9. Final report on the Controlled Cold Helium Spill Test in the LHC tunnel at CERN

    International Nuclear Information System (INIS)

    Dufay-Chanat, L; Bremer, J; Casas-Cubillos, J; Koettig, T; Vauthier, N; Van Weelderen, R; Winkler, T; Chorowski, M; Grabowski, M; Jedrusyna, A; Lindell, G; Nonis, M

    2015-01-01

    The 27 km circumference LHC underground tunnel is a space in which the helium cooled LHC magnets are installed. The vacuum enclosures of the superconducting magnets are protected by over-pressure safety relief devices that open whenever cold helium escapes either from the magnet cold enclosure or from the helium supply headers, into this vacuum enclosure. A 3-m long no stay zone around these devices is defined based on scale model studies, protecting the personnel against cold burns or asphyxia caused by such a helium release event. Recently, several simulation studies have been carried out modelling the propagation of the helium/air mixture, resulting from the opening of such a safety device, along the tunnel. The released helium flows vary in the range between 1 kg/s and 0.1 kg/s. To validate these different simulation studies, real life mock-up tests have been performed inside the LHC tunnel, releasing helium flow rates of 1 kg/s, 0.3 kg/s and 0.1 kg/s. For each test, up to 1000 liters of liquid helium were released under standard operational tunnel conditions. The data recorded include oxygen concentration, temperature and flow speed measurements, and video footage used to assess qualitatively the visibility. These measurements have been made in the up- and downstream directions, with respect to the air ventilation flow, of the spill point.This paper presents the experimental set-up under which these release tests were made, the effects of these releases on the atmospheric tunnel condition as a function of the release flow rate. We discuss the modification to the personnel access conditions to the LHC tunnel that are presently implemented as a result of these tests. (paper)

  10. Final report on the Controlled Cold Helium Spill Test in the LHC tunnel at CERN

    Science.gov (United States)

    Dufay-Chanat, L.; Bremer, J.; Casas-Cubillos, J.; Chorowski, M.; Grabowski, M.; Jedrusyna, A.; Lindell, G.; Nonis, M.; Koettig, T.; Vauthier, N.; van Weelderen, R.; Winkler, T.

    2015-12-01

    The 27 km circumference LHC underground tunnel is a space in which the helium cooled LHC magnets are installed. The vacuum enclosures of the superconducting magnets are protected by over-pressure safety relief devices that open whenever cold helium escapes either from the magnet cold enclosure or from the helium supply headers, into this vacuum enclosure. A 3-m long no stay zone around these devices is defined based on scale model studies, protecting the personnel against cold burns or asphyxia caused by such a helium release event. Recently, several simulation studies have been carried out modelling the propagation of the helium/air mixture, resulting from the opening of such a safety device, along the tunnel. The released helium flows vary in the range between 1 kg/s and 0.1 kg/s. To validate these different simulation studies, real life mock-up tests have been performed inside the LHC tunnel, releasing helium flow rates of 1 kg/s, 0.3 kg/s and 0.1 kg/s. For each test, up to 1000 liters of liquid helium were released under standard operational tunnel conditions. The data recorded include oxygen concentration, temperature and flow speed measurements, and video footage used to assess qualitatively the visibility. These measurements have been made in the up- and downstream directions, with respect to the air ventilation flow, of the spill point. This paper presents the experimental set-up under which these release tests were made, the effects of these releases on the atmospheric tunnel condition as a function of the release flow rate. We discuss the modification to the personnel access conditions to the LHC tunnel that are presently implemented as a result of these tests.

  11. 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.)

  12. Simulation of power maneuvering experiment of MASLWR test facility by MARS-KS code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ju Yeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    In the present study, KINS simulation result by the MARS-KS code (KS-002 version) for the SP-3 experiment is presented in detail and conclusion on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the power maneuvering experiment of the MASLWR test facility. Steady run shows the helical coil specific heat transfer model of the code is reasonable. However, identified discrepancy of the primary mass flowrate at transient run shows code performance for pressure drop needs to be improved considering sensitivity of the flowrate to the pressure drop at natural circulation. Since 2009, IAEA has conducted a research program entitled as ICSP (International Collaborative Standard Problem) on integral PWR design to evaluate current the state of the art of thermal-hydraulic code in simulating natural circulation flow within integral type reactor. In this ICSP, experimental data obtained from MASLWR (Multi-Application Small Light Water Reactor) test facility located at Oregon state university in the US have been simulated by various thermal-hydraulic codes of each participant of the ICSP and compared among others. MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is currently being developed in Korea also adopts a helical coil steam generator, Korea Institute of Nuclear Safety (KINS) has joined this ICSP to assess the applicability of a domestic regulatory audit thermal-hydraulic code (i. e. MARS-KS code) for the SMART reactor including wall-to-fluid heat transfer model modification based on independent international experiment data. In the ICSP, two types of transient experiments have been focused and they are loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels (SP-3)

  13. Simulation of power maneuvering experiment of MASLWR test facility by MARS-KS code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ju Yeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    In this ICSP, experimental data obtained from MASLWR (Mulit-Application Small Light Water Reactor) test facility located at Oregon state university in the US have been simulated by various thermal-hydraulic codes of each participant of the ICSP and compared among others. MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is currently being developed in Korea also adopts a helical coil steam generator, Korea Institute of Nuclear Safety (KINS) has joined this ICSP to assess the applicability of a domestic regulatory audit thermal-hydraulic code (i. e. MARS-KS code) for the SMART reactor including wall-to-fluid heat transfer model modification based on independent international experiment data. In the ICSP, two types of transient experiments have been focused and they are 1) loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels. In the present study, KINS simulation result by the MARS-KS code (KS-002 version) for the SP-3 experiment is presented in detail and conclusion on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the power maneuvering experiment of the MASLWR test facility. Steady run shows the helical coil specific heat transfer model of the code is reasonable. However, identified discrepancy of the primary mass flowrate at transient run shows code performance for pressure drop needs to be improved considering sensitivity of the flowrate to the pressure drop at natural circulation.

  14. EFRT M-12 Issue Resolution: Caustic-Leach Rate Constants from PEP and Laboratory-Scale Tests

    Energy Technology Data Exchange (ETDEWEB)

    Mahoney, Lenna A.; Rassat, Scot D.; Eslinger, Paul W.; Aaberg, Rosanne L.; Aker, Pamela M.; Golovich, Elizabeth C.; Hanson, Brady D.; Hausmann, Tom S.; Huckaby, James L.; Kurath, Dean E.; Minette, Michael J.; Sundaram, S. K.; Yokuda, Satoru T.

    2010-01-01

    Pacific Northwest National Laboratory (PNNL) has been tasked by Bechtel National Inc. (BNI) on the River Protection Project-Hanford Tank Waste Treatment and Immobilization Plant (RPP-WTP) project to perform research and development activities to resolve technical issues identified for the Pretreatment Facility (PTF). The Pretreatment Engineering Platform (PEP) was designed, constructed and operated as part of a plan to respond to issue M12, “Undemonstrated Leaching Processes” of the External Flowsheet Review Team (EFRT) issue response plan.( ) The PEP is a 1/4.5-scale test platform designed to simulate the WTP pretreatment caustic leaching, oxidative leaching, ultrafiltration solids concentration, and slurry washing processes. The PEP replicates the WTP leaching processes using prototypic equipment and control strategies. The PEP also includes non-prototypic ancillary equipment to support the core processing. The work described in this report addresses caustic leaching under WTP conditions, based on tests performed with a Hanford waste simulant. Because gibbsite leaching kinetics are rapid (gibbsite is expected to be dissolved by the time the final leach temperature is reached), boehmite leach kinetics are the main focus of the caustic-leach tests. The tests were completed at the laboratory-scale and in the PEP, which is a 1/4.5-scale mock-up of key PTF process equipment. Two laboratory-scale caustic-leach tests were performed for each of the PEP runs. For each PEP run, unleached slurry was taken from the PEP caustic-leach vessel for one batch and used as feed for both of the corresponding laboratory-scale tests.

  15. Saturation of compacted bentonite under repository conditions: long-term experimental evidences

    International Nuclear Information System (INIS)

    Villar, M.V.; Martin, P.L.; Gomez-Espina, R.; Garcia-Sineriz, J.L.; Barcena, I.; Lloret, A.

    2010-01-01

    Document available in extended abstract form only. A current design for engineered barriers in the context of high-level radioactive waste disposal includes bentonite compacted blocks initially unsaturated. The heat released by the waste will induce high temperatures in the bentonite barrier. It is expected that full saturation of the buffer be reached before the dissipation of the thermal gradient. However, it still remains unclear whether the high temperatures around the canister would hinder the full saturation of the inner part of the barrier or just delay it. This paper summarises the information gathered in the last 15 years on the saturation of compacted FEBEX bentonite by means of different scale laboratory tests, a big-scale mock-up test and a real-scale in situ test, that were performed in order to simulate the conditions of the clay barrier in the repository and better understand the hydration/heating processes and their consequences on bentonite performance. FEBEX is a Spanish bentonite composed mainly of montmorillonite (about 92%). In the tests it has been used compacted with its hygroscopic water content (14%) at dry densities between 1.6 and 1.7 g/cm 3 , which is the range expected in the repository. For these densities the saturated permeability of the bentonite is about 3.10 -14 m/s and its swelling pressure 8 MPa. The FEBEX in situ test is being performed under natural conditions and at full scale within a drift excavated in the underground laboratory managed by NAGRA at the Grimsel Test Site (Switzerland). The thickness of the bentonite barrier is of 65 cm, and the surface heater temperature is 100 C. After five years of heating, and according to the sensors measurements, the bentonite closer to the heater had water contents below the initial ones, although they were recovering after the intense initial drying. On the contrary, for the same period of time, the sensors located at the same distance from the gallery wall, but in an area not

  16. Seismic test facilities at the ENEA Casaccia Research Center; Prove sismiche con le tavole vibranti al centro ricerche Enea Casaccia

    Energy Technology Data Exchange (ETDEWEB)

    De Canio, G. [ENEA, Divisione Servizi Tecnologici, Centro Ricerche Casaccia, Rome (Italy)

    2000-07-01

    The main experimental facilities for seismic tests at the ENEA C.R. Casaccia laboratories consist of two high performance shake table for three axial seismic tests of structures up to 10 ton mass and 3g acceleration applied at the Center of Gravity at 1m from the base table. The activities are principally devoted to the dynamic characterization and vibration tests for mechanical and aero spatial structures, and the experimental analysis of innovative systems for the seismic isolation and retrofitting of civil, industrial, and historical buildings; together with the seismic tests of sub-structures and scaled mock-ups, in order to evaluate the isolation/dissipation performance of the anti-seismic devices, and the failure modes of the structural parts of the building. [Italian] Le principali attrezzature per le prove sismiche presso i laboratori del C.R. Casaccia consistono di due tavole vibranti triassali per prove su strutture fino a 10t di peso con una accelerazione di 3g applicata al centro di gravita' posto ad 1 m di altezza dal piano della tavola. Le principali attivita' riguardano: (a) test di caratterizzazione dinamica e prove di vibrazioni per strutture meccaniche ed aerospaziali; (b) l'analisi sperimentale di sistemi innovativi per l'isolamento sismico ed il consolidamento di strutture civili, industriali e storico monumentali, e le prove sismiche di elementi strutturali e di modelli in scala per la valutazione della capacita' di dissipazione dei dispositivi antisismici e le modalita' di formazione delle fratture.

  17. High RF power test of a CFC antenna module for lower hybrid current drive

    International Nuclear Information System (INIS)

    Maebara, S.; Seki, M.; Ikeda, Y.; Kiyono, K.; Suganuma, K.; Imai, T.; Goniche, M.; Bibet, Ph.; Brossaud, J.; Cano, V.; Kazarian-Vibert, F.; Froissard, P.; Rey, G.

    1998-01-01

    A mock-up of a 3.7 GHz Lower Hybrid Current Drive (LHCD) antenna module was fabricated from Carbon Fibre Composite (CFC) for the development of heat resistive low Z front facing the plasma. This 2 divided waveguide module is made from CFC plates and rods which are Cu-plated to reduce the RF losses. The withstand-voltage, the RF properties and the outgassing rates for long pulses and high RF power were tested at the Lower Hybrid test bed facility of Cadarache. A reference module made from Dispersion Strengthened Copper (DSC) was also fabricated. After the short pulse conditioning, long pulses with a power density ranging between 50 and 150 MW/m 2 were performed with no breakdowns on the CFC module. It was also checked that the highest power density, up to 150 MW/m 2 , could be transmitted when the waveguides are filled with H2 at a pressure of 5 x 10 -2 Pa. During a long pulse, the power reflection coefficient remains low in the 0.8-1.3 % range and no significant change in the reflection coefficient is measured after the thermal cycling provided by the long pulse operation. From thermocouple measurements, RF losses of the copper coated CFC and the DSC modules were compared. No significant differences were measured. From pressure measurements, it was found that the outgassing rate of Cu-plated CFC is about 6-7 times larger than of DSC at 300 deg.C. It is concluded that a CFC module is an attractive candidate for the hardening of the tip of the LHCD antenna. (author)

  18. Production and Testing of Kiln-cast Glass Components for an Interlocking, Dry-assembled Transparent Bridge

    NARCIS (Netherlands)

    Bristogianni, T.; Oikonomopoulou, F.; Veer, F.A.; Snijder, A.H.; Nijsse, R.

    A pedestrian glass bridge, located at the TU Delft campus site, is being designed by the TU Delft Glass & Transparency Lab. Specifically, the arch-formed bridge consists of cast glass, dry-assembled, interlocking components. To validate the shape of the components, glass mock-ups in 1:2 scale

  19. MSR Fetch Rover Capability Development at the Canadian Space Agency

    Science.gov (United States)

    Picard, M.; Hipkin, V.; Gingras, D.; Allard, P.; Lamarche, T.; Rocheleau, S. G.; Gemme, S.

    2018-04-01

    Describes Fetch Rover technology testing during CSA's 2016 Mars Sample Return Analogue Deployment which demonstrated autonomous navigation to 'cache depots' of M-2020-like sample tubes, acquisition of six such tubes, and transfer to a MAV mock up.

  20. Calculation methods of Structure-Soil-Structure Interaction (3SI) for embedded buildings: Application to NUPEC tests

    International Nuclear Information System (INIS)

    Clouteau, D.; Broc, D.; Devesa, G.; Guyonvarh, V.; Massin, P.

    2012-01-01

    This work aims at improving and validating methods coupling Finite Element (FE) and Boundary Element (BE) Methods in the context of Soil-Structure Interaction (SSI) and Structure-Soil-Structure Interaction (3SI) tests performed by NUPEC on mock-up structures built on an unmade ground. Several cases have been tested: single and juxtaposed buildings, shallow and embedded foundations, with various loading conditions: forced and natural seismic loadings. The numerical simulations of forced vibration tests are in good agreement with the results of the NUPEC experiments in the case of two embedded buildings either in terms of amplitude and resonance. The numerical simulation of seismic response tests by FEM and BEM allows for a proper choice of the 'reference point' where the computed and the experimental displacements coincide. A parametric analysis of Structure-Soil-Structure Interaction carried out by the FEM has allowed to determine the influence of some parameters on SSI. Most of them like the position of the building in the excavation, the direction of the load, the quality of the contact between the sidewalls of the buildings and the soil for embedded foundations, do not show to have a strong influence on the dynamic system behaviour, which is mainly governed by the stiffness of the first soil layer. As far as 3SI is concerned, this paper shows that when the cross interaction has a small effect on the building response in the case of surface foundations, it has a strong influence in the case of embedded foundations with an important decrease of the response at the top of the buildings. (authors)

  1. International Benchmark based on Pressurised Water Reactor Sub-channel and Bundle Tests. Volume III: Departure from Nucleate Boiling

    International Nuclear Information System (INIS)

    Rubin, Adam; Avramova, Maria; Velazquez-Lozada, Alexander

    2016-03-01

    This report summarised the second phase of the Nuclear Energy Agency (NEA) and the Nuclear Regulatory Commission (NRC) Benchmark Based on NUPEC PWR Sub-channel and Bundle Tests (PSBT), which was intended to provide data for the verification of Departure from Nucleate Boiling (DNB) prediction in existing thermal-hydraulics codes and provide direction in the development of future methods. This phase was composed of three exercises; Exercise 1: fluid temperature benchmark, Exercise 2: steady-state rod bundle benchmark and Exercise 3: transient rod bundle benchmark. The experimental data provided to the participants of this benchmark is from a series of void measurement tests using full-size mock-up tests for both BWRs and PWRs. These tests were performed from 1987 to 1995 by the Nuclear Power Engineering Corporation (NUPEC) in Japan and made available by the Japan Nuclear Energy Safety Organisation (JNES) for the purposes of this benchmark, which was organised by Pennsylvania State University. Nine institutions from seven countries participated in this benchmark. Nine different computer codes were used in Exercise 1, 2 and 3. Among the computer codes were porous media, sub-channel and systems thermal-hydraulic code. The improvement between FLICA-OVAP (sub-channel) and FLICA (sub-channel) was noticeable. The main difference between the two was that FLICA-OVAP implicitly assigned flow regime based on drift flux, while FLICA assumes single phase flows. In Exercises 2 and 3, the codes were generally able to predict the Departure from Nucleate Boiling (DNB) power as well as the axial location of the onset of DNB (for the steady-state cases) and the time of DNB (for the transient cases). It was noted that the codes that used the Electric-Power-Research- Institute (EPRI) Critical-Heat-Flux (CHF) correlation had the lowest mean error in Exercise 2 for the predicted DNB power

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  3. Intercomparison of liquid metal fast reactor seismic analysis codes. V.1: Validation of seismic analysis codes using reactor core experiments. Proceedings of a research co-ordination meeting held in Vienna, 16-17 November 1993

    International Nuclear Information System (INIS)

    1995-05-01

    The Research Co-ordination Meeting held in Vienna, 16-17 November 1993, was attended by participants from France, India, Italy, Japan and the Russian Federation. The meeting was held to discuss and compare the results obtained by various organizations for the analysis of Italian tests on PEC mock-up. The background paper by A. Martelli, et al., Italy, entitled Fluid-Structure Interaction Experiments of PEC Core Mock-ups and Numerical Analysis Performed by ENEA presented details on the Italian PEC (Prova Elementi di Combustibile, i.e. Fuel Element Test Facility) test data for the benchmark. Several papers were presented on the analytical investigations of the PEC reactor core experiments. The paper by M. Morishita, Japan, entitled Seismic Response Analysis of PEC Reactor Core Mock-up, gives a brief review of the Japanese data on the Monju mock-up core experiment which had been distributed to the participating countries through the IAEA. Refs, figs and tabs

  4. The high-heat-flux test facilities in the joint stock company “D.V. Efremov Institute of Electrophysical Apparatus”

    Energy Technology Data Exchange (ETDEWEB)

    Volodin, A., E-mail: volodin@sintez.niiefa.spb.su [JSC “NIIEFA”, 196641 St. Petersburg (Russian Federation); Kuznetcov, V.; Davydov, V.; Kokoulin, A.; Komarov, A.; Mazul, I.; Mudyugin, B.; Ovchinnikov, I.; Stepanov, N.; Rulev, R.; Eremkin, A.; Rogov, A.; Prianikov, V. [JSC “NIIEFA”, 196641 St. Petersburg (Russian Federation); Fedosov, A. [ITER Organization, Building 81/124, TKM, Internal Components Division, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2015-10-15

    Highlights: • The IDTF was created for the high heat flux tests of the PFUs of the ITER divertor. • At the present on the TSEFEY-M a brazing of fingers a FW semi-prototype is performing. • The IDTF and TSEFEY-M facilities are ready for the HHF testing of the ITER components. - Abstract: The current ITER design involves beryllium and tungsten as plasma facing materials for in-vessel components. Due to a high number of operating cycles and to the expected surface heat loads, thermal fatigue is one of the most damaging mechanisms for the plasma facing components (PFCs) of the ITER machine. Therefore, it is essential to perform an assessment of the behavior of PFCs under cycling heat loads to demonstrate the fitness for purpose of the selected technologies. This article summarizes the features of high heat flux facilities designed and constructed in the Efremov Institute for the performance of high heat flux (HHF) tests under ITER procurements as well as related R&D works. The TSEFEY-M facility was commissioned in 1994. The main purpose of this facility is thermal fatigue testing of mock-ups with various plasma-facing materials (carbon fiber reinforced composite (CFC), tungsten, beryllium, etc.) and with various cooling agents (water or gas). The ITER divertor test facility (IDTF) was created in the framework of ITER project, specifically for the HHF tests of the vertical targets (inner and outer) and domes of the ITER divertor. After commissioning in 2008, the IDTF facility was qualified in 2012–2013 for HHF tests of ITER PFCs.

  5. WANTO 32: Proceedings of the 32nd Weapons Agencies Nondestructive Testing Organization meeting

    International Nuclear Information System (INIS)

    Majzlik, E.J. Jr.

    1991-02-01

    The Thirty-Second Weapons Agencies Nondestructive Testing Organization (WANTO) meeting was held at the Savannah River Site (SRS) on November 27--29, 1990. The meeting was hosted by Edward J. Majzlik, Jr., Savannah River Laboratory, Westinghouse Savannah River Company (WSRC). The Chairman of the WANTO Steering Committee is L.E. (Larry) Bryant, Los Alamos National Laboratory. This report is the sole proceedings of the meeting and includes the agenda, attendance, steering committee report, interim reports and technical presentation summaries. This report is the first to present the meeting proceedings in an unclassified form. The reader should contact individual authors directly for any additional information desired. The meeting was organized to provide coverage of a wide variety of NDE subjects relevant to the Department of Energy (DOE) Nuclear Weapons Complex (NWC). Approximately 60 technical experts and managers representing 11 DOE weapons agencies and 3 DOE offices attended. A total of 56 technical presentations and 3 special NDE workshops were included in the agenda along with two tours of the SRS Reactor Mock-up Facility. Invited presentations included five speakers who reported on the DOE sponsored conference Concurrent Engineering and the NDE Role held at Sandia National Laboratory, Albuquerque, New Mexico, October 30-November 1, 1990 and X-ray Film Retention and Disposal presented by Orville R. Pratt, DOE/AL-Martin Marietta. The WANTO Steering Committee report for the Thirty-Second Meeting immediately follows this section. The meeting Agenda and Attendance are presented in Appendices A and B, respectively. Technical presentation summaries and abstracts are presented in Appendix C. For cases in which a summary was not provided, a telephone number is offered for direct contact with the author/presenter. Interim Activity Reports are compiled in Appendix D. Special Workshop reports are presented in Appendix E. (JF)

  6. The European ITER test blanket modules: Progress in development of fabrication technologies towards standardization

    Energy Technology Data Exchange (ETDEWEB)

    Zmitko, Milan, E-mail: milan.zmitko@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla 2, Barcelona (Spain); Thomas, Noël [ATMOSTAT, F-94815 Villejuif (France); LiPuma, Antonella; Forest, Laurent [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Cogneau, Laurence [CEA-DRT, 38000 Grenoble (France); Rey, Jörg; Neuberger, Heiko [Karlsruhe Institute of Technology (KIT), Postfach 3640, Karlsruhe (Germany); Poitevin, Yves [Fusion for Energy (F4E), Josep Pla 2, Barcelona (Spain)

    2016-11-01

    Highlights: • Significant progress on the development of welding procedures for European TBM achieved. • Fabrication processes feasibility based on diffusion and fusion welding demonstrated. • An optimized welding scenario/sequence for TBM box assembly identified. • Future qualification of pF/WPS proposed through realization of a number of QMUs. - Abstract: The paper reviews progress achieved in development of fabrication technologies and procedures applied for manufacturing of the TBM sub-components, like, HCLL and HCPB cooling plates, HCLL/HCPB stiffening plates, and HCLL/HCPB first wall and side caps. The used technologies are based on fusion and diffusion welding techniques taking into account specificities of the EUROFER97 steel. Development of a standardized procedure complying with professional codes and standards (RCC-MRx), a preliminary fabrication/welding procedure specification (pF/WPS), is described based on fabrication and non-destructive and destructive characterization of feasibility mock-ups (FMU) aimed at assessing the suitability of a fabrication process for fulfilling the design and fabrication specifications. The main FMUs characterization results are reported (e.g. pressure resistance and helium leak tightness tests, mechanical properties and microstructure at the weld joints, geometrical characteristics of the sub-components and internal cooling channels) and the key pF/WPS steps and parameters are outlined. Also, fabrication procedures for the TBM box assembly are presently under development for the establishment of an optimized assembly sequence/scenario and development of standardized welding procedure specifications. In conclusions, further steps towards the pF/WPS qualification are briefly discussed.

  7. Performance test of micro-fission chambers for in-vessel neutron monitoring of ITER

    International Nuclear Information System (INIS)

    Yamauchi, Michinori; Nishitani, Takeo; Ochiai, Kentaro; Morimoto, Yuichi; Hori, Jun-ichi; Ebisawa, Katsuyuki; Kasai, Satoshi

    2002-03-01

    A micro-fission chamber with 12 mg UO 2 and a dummy chamber without uranium were fabricated and the performance was tested. They are designed to be installed inside the vacuum vessel of the compact ITER (ITER-FEAT) for neutron monitoring. The vacuum leak rate of the dummy chamber with MI cable, resistances of chambers between central conductor and outer sheath, and mechanical strength up to 50G acceleration were confirmed to meet the design criteria. The gamma-ray sensitivity was measured for the dummy chamber with the 60 Co gamma-ray irradiation facility at JAERI Takasaki. The output signals for gamma-rays in Campbelling mode were estimated to be less than 0.1% of those by neutrons at the location behind the blanket module in ITER-FEAT. The detector response for 14 MeV neutrons was investigated with the FNS facility. Excellent linearity between count rates, square of Campbelling voltage and neutron fluxes was confirmed in the temperature range from 20degC (room) to 250degC. However, a positive dependence of 14 MeV neutron count rates on temperature was observed, which might be caused by the increase in the pulse height with temperature rise. Effects of a change of surrounding materials were evaluated by the sensitivity measurements of the micro-fission chamber inserted into the shielding blanket mock-up. The sensitivity was enhanced by slow-downed neutrons, which agreed with the calculation result by MCNP-4C code. As a result, it was concluded that the developed micro-fission chamber is applicable for ITER-FEAT. (author)

  8. Abnormal reactions in a evaporator in a fuel reprocessing plant

    International Nuclear Information System (INIS)

    Kida, Takashi; Umeda, Miki; Sugikawa, Susumu

    2003-01-01

    In order to evaluate a self-accelerated reaction in an evaporator in a fuel reprocessing plant due to organic-nitric acid reactions, a development of a calculation code is under way. Mock-up tests were performed to investigate the fluid dynamic behavior of the organic solvent in the evaporator. Based on these results, the model of the calculation code was constructed. This report describes the results of mock-up tests and the model of the calculation code. (author)

  9. Optimization of armour geometry and bonding techniques for tungsten-armoured high heat flux components

    International Nuclear Information System (INIS)

    Giniyatulin, R.N.; Komarov, V.L.; Kuzmin, E.G.; Makhankov, A.N.; Mazul, I.V.; Yablokov, N.A.; Zhuk, A.N.

    2002-01-01

    Joining of tungsten with copper-based cooling structure and armour geometry optimization are the major aspects in development of the tungsten-armoured plasma facing components (PFC). Fabrication techniques and high heat flux (HHF) tests of tungsten-armoured components have to reflect different PFC designs and acceptable manufacturing cost. The authors present the recent results of tungsten-armoured mock-ups development based on manufacturing and HHF tests. Two aspects were investigated--selection of armour geometry and examination of tungsten-copper bonding techniques. Brazing and casting tungsten-copper bonding techniques were used in small mock-ups. The mock-ups with armour tiles (20x5x10, 10x10x10, 20x20x10, 27x27x10) mm 3 in dimensions were tested by cyclic heat fluxes in the range of (5-20) MW/m 2 , the number of thermal cycles varied from hundreds to several thousands for each mock-up. The results of the tests show the applicability of different geometry and different bonding technique to corresponding heat loading. A medium-scale mock-up 0.6-m in length was manufactured and tested. HHF tests of the medium-scale mock-up have demonstrated the applicability of the applied bonding techniques and armour geometry for full-scale PFC's manufacturing

  10. Results of high heat flux testing of W/CuCrZr multilayer composites with percolating microstructure for plasma-facing components

    International Nuclear Information System (INIS)

    Greuner, Henri; Zivelonghi, Alessandro; Böswirth, Bernd; You, Jeong-Ha

    2015-01-01

    Highlights: • Improvement of the performance of plasma-facing components made of W and CuCrZr. • Functionally graded composite at the interface of W and CuCrZr to mitigate the CTE. • A three-layer composite system (W volume fraction: 70/50/30%) was developed. • Design of water-cooled divertor components up to 20 MW/m"2 heat load for e.g. DEMO. • HHF tests up to 20 MW/m"2 were successfully performed. - Abstract: Reliable joining of tungsten to copper is a major issue in the design of water-cooled divertor components for future fusion reactors. One of the suggested advanced engineering solutions is to use functionally graded composite interlayers. Recently, the authors have developed a novel processing route for fabricating multi-layer graded W/CuCrZr composites. Previous characterization confirmed that the composite materials possess enhanced strength compared to the matrix alloy and shows reasonable ductility up to 300 °C indicating large potential to extend the operation temperature limit. Furthermore, a three-layer composite system (W volume fraction: 70/50/30%) was developed as a graded interlayer between the W armour and CuCrZr heat sink. In this study, we investigated the structural performance of the graded joint. Three water-cooled mock-ups of a flat tile type component were fabricated using electron beam welding and thermally loaded at the hydrogen neutral beam test facility GLADIS. Cycling tests at 10 MW/m"2 and screening tests up to 20 MW/m"2 were successfully performed and confirmed the expected thermal performance of the compound. The measured temperature values were in good agreement with the prediction of finite element analysis. Microscopic investigation confirmed the structural integrity of the newly developed functionally graded composite after these tests.

  11. Results of high heat flux testing of W/CuCrZr multilayer composites with percolating microstructure for plasma-facing components

    Energy Technology Data Exchange (ETDEWEB)

    Greuner, Henri, E-mail: henri.greuner@ipp.mpg.de; Zivelonghi, Alessandro; Böswirth, Bernd; You, Jeong-Ha

    2015-10-15

    Highlights: • Improvement of the performance of plasma-facing components made of W and CuCrZr. • Functionally graded composite at the interface of W and CuCrZr to mitigate the CTE. • A three-layer composite system (W volume fraction: 70/50/30%) was developed. • Design of water-cooled divertor components up to 20 MW/m{sup 2} heat load for e.g. DEMO. • HHF tests up to 20 MW/m{sup 2} were successfully performed. - Abstract: Reliable joining of tungsten to copper is a major issue in the design of water-cooled divertor components for future fusion reactors. One of the suggested advanced engineering solutions is to use functionally graded composite interlayers. Recently, the authors have developed a novel processing route for fabricating multi-layer graded W/CuCrZr composites. Previous characterization confirmed that the composite materials possess enhanced strength compared to the matrix alloy and shows reasonable ductility up to 300 °C indicating large potential to extend the operation temperature limit. Furthermore, a three-layer composite system (W volume fraction: 70/50/30%) was developed as a graded interlayer between the W armour and CuCrZr heat sink. In this study, we investigated the structural performance of the graded joint. Three water-cooled mock-ups of a flat tile type component were fabricated using electron beam welding and thermally loaded at the hydrogen neutral beam test facility GLADIS. Cycling tests at 10 MW/m{sup 2} and screening tests up to 20 MW/m{sup 2} were successfully performed and confirmed the expected thermal performance of the compound. The measured temperature values were in good agreement with the prediction of finite element analysis. Microscopic investigation confirmed the structural integrity of the newly developed functionally graded composite after these tests.

  12. Determination of extra trajectory parameters of projectile layout motion

    Science.gov (United States)

    Ishchenko, A.; Burkin, V.; Faraponov, V.; Korolkov, L.; Maslov, E.; Diachkovskiy, A.; Chupashev, A.; Zykova, A.

    2017-11-01

    The paper presents a brief description of the experimental track developed and implemented on the base of the RIAMM TSU for external trajectory investigations on determining the main aeroballistic parameters of various shapes projectiles, in the wide velocity range. There is comparison between the experimentally obtained dependence of the fin-stabilized projectile mock-up aerodynamic drag coefficient on the Mach number with the 1958 aerodynamic drag law and aerodynamic tests of the same mock-up

  13. Experimental Investigation Into Thermal Siphon Used as an Intermediate Circuit of an Integrated Cooling System Reactor

    International Nuclear Information System (INIS)

    Adamovich, L.A.; Gabaraev, B.A.; Solovjev, S.L.; Shpansky, S.B.

    2002-01-01

    In the paper the results of study in heat transfer capacity of the thermosyphon mock-up which is considered as an intermediate circuit of the reactor under design, are presented. The mock-up design, the test rig and the experimental results are described. It is shown that the simplest mathematical model describes the processes of power transfer by the thermosyphon under certain conditions. (authors)

  14. R

    Directory of Open Access Journals (Sweden)

    Patil Yashashri

    2015-06-01

    Full Text Available This paper is focused on various aspects of the development and testing of water cooled divertor PFCs. Divertor PFCs are mainly designed to absorb the heat and particle fluxes outflowing from the core plasma of fusion devices like ITER. The Divertor and First Wall Technology Development Division at the Institute for Plasma Research (IPR, India, is extensively working on development and testing of divertor plasma facing components (PFCs. Tungsten and graphite macro-brush type test mock-ups were produced using vacuum brazing furnace technique and tungsten monoblock type of test mock-ups were obtained by hot radial pressing (HRP technique. Heat transfer performance of the developed test mock-ups was tested using high heat flux tests with different heat load conditions as well as the surface temperature monitoring using transient infrared thermography technique. Recently we have established the High Heat Flux Test Facility (HHFTF at IPR with an electron gun EH300V (M/s Von Ardenne Anlagentechnik GmbH, Germany having maximum power 200 kW. Two tungsten monoblock type test mock-ups were probed using HHFTF. Both of the test mock-ups successfully sustained 316 thermal cycles during high heat flux (HHF tests. The test mock-ups were non-destructively tested using infrared thermography before and after the HHF tests. In this note we describe the detailed procedure used for testing macro-brush and monoblock type test mock-ups using in-house transient infrared thermography set-up. An acceptance criteria limit was defined for small scale macro-brush type of mock-ups using DTrefmax value and the surface temperature measured during the HHF tests. It is concluded that the heat transfer behavior of a plasma facing component was checked by the HHF tests followed by transient IR thermography. The acceptance criteria DTrefmax limit for a graphite macro-brush mock-up was found to be ~3°C while for a tungsten macro-brush mock-up it was ~5°C.

  15. Design and Testing of Novel Lethal Ovitrap to Reduce Populations of Aedes Mosquitoes: Community-Based Participatory Research between Industry, Academia and Communities in Peru and Thailand.

    Science.gov (United States)

    Paz-Soldan, Valerie A; Yukich, Josh; Soonthorndhada, Amara; Giron, Maziel; Apperson, Charles S; Ponnusamy, Loganathan; Schal, Coby; Morrison, Amy C; Keating, Joseph; Wesson, Dawn M

    2016-01-01

    Dengue virus (and Chikungunya and Zika viruses) is transmitted by Aedes aegypti and Aedes albopictus mosquitoes and causes considerable human morbidity and mortality. As there is currently no vaccine or chemoprophylaxis to protect people from dengue virus infection, vector control is the only viable option for disease prevention. The purpose of this paper is to illustrate the design and placement process for an attractive lethal ovitrap to reduce vector populations and to describe lessons learned in the development of the trap. This study was conducted in 2010 in Iquitos, Peru and Lopburi Province, Thailand and used an iterative community-based participatory approach to adjust design specifications of the trap, based on community members' perceptions and feedback, entomological findings in the lab, and design and research team observations. Multiple focus group discussions (FGD) were held over a 6 month period, stratified by age, sex and motherhood status, to inform the design process. Trap testing transitioned from the lab to within households. Through an iterative process of working with specifications from the research team, findings from the laboratory testing, and feedback from FGD, the design team narrowed trap design options from 22 to 6. Comments from the FGD centered on safety for children and pets interacting with traps, durability, maintenance issues, and aesthetics. Testing in the laboratory involved releasing groups of 50 gravid Ae. aegypti in walk-in rooms and assessing what percentage were caught in traps of different colors, with different trap cover sizes, and placed under lighter or darker locations. Two final trap models were mocked up and tested in homes for a week; one model was the top choice in both Iquitos and Lopburi. The community-based participatory process was essential for the development of novel traps that provided effective vector control, but also met the needs and concerns of community members.

  16. Design and Testing of Novel Lethal Ovitrap to Reduce Populations of Aedes Mosquitoes: Community-Based Participatory Research between Industry, Academia and Communities in Peru and Thailand.

    Directory of Open Access Journals (Sweden)

    Valerie A Paz-Soldan

    Full Text Available Dengue virus (and Chikungunya and Zika viruses is transmitted by Aedes aegypti and Aedes albopictus mosquitoes and causes considerable human morbidity and mortality. As there is currently no vaccine or chemoprophylaxis to protect people from dengue virus infection, vector control is the only viable option for disease prevention. The purpose of this paper is to illustrate the design and placement process for an attractive lethal ovitrap to reduce vector populations and to describe lessons learned in the development of the trap.This study was conducted in 2010 in Iquitos, Peru and Lopburi Province, Thailand and used an iterative community-based participatory approach to adjust design specifications of the trap, based on community members' perceptions and feedback, entomological findings in the lab, and design and research team observations. Multiple focus group discussions (FGD were held over a 6 month period, stratified by age, sex and motherhood status, to inform the design process. Trap testing transitioned from the lab to within households.Through an iterative process of working with specifications from the research team, findings from the laboratory testing, and feedback from FGD, the design team narrowed trap design options from 22 to 6. Comments from the FGD centered on safety for children and pets interacting with traps, durability, maintenance issues, and aesthetics. Testing in the laboratory involved releasing groups of 50 gravid Ae. aegypti in walk-in rooms and assessing what percentage were caught in traps of different colors, with different trap cover sizes, and placed under lighter or darker locations. Two final trap models were mocked up and tested in homes for a week; one model was the top choice in both Iquitos and Lopburi.The community-based participatory process was essential for the development of novel traps that provided effective vector control, but also met the needs and concerns of community members.

  17. YUCCA Mountain Project - Argonne National Laboratory, Annual Progress Report, FY 1997 for activity WP 1221 unsaturated drip condition testing of spent fuel and unsaturated dissolution tests of glass.

    Energy Technology Data Exchange (ETDEWEB)

    Bates, J. K.; Buck, E. C.; Emery, J. W.; Finch, R. J.; Finn, P. A.; Fortner, J.; Hoh, J. C.; Mertz, C.; Neimark, L. A.; Wolf, S. F.; Wronkiewicz, D. J.

    1998-09-18

    {sub 2} in contact with small volumes of water within a several month period when the radiolysis product H{sub 2}O{sub 2} is added to the groundwater solution. The test setup has been mocked up for operation with spent fuel in the hot-cell.

  18. YUCCA Mountain Project - Argonne National Laboratory, Annual Progress Report, FY 1997 for activity WP 1221 unsaturated drip condition testing of spent fuel and unsaturated dissolution tests of glass

    International Nuclear Information System (INIS)

    Bates, J.K.; Buck, E.C.; Emery, J.W.; Finch, R.J.; Finn, P.A.; Fortner, J.; Hoh, J.C.; Mertz, C.; Neimark, L.A.; Wolf, S.F.; Wronkiewicz, D.J.

    1998-01-01

    small volumes of water within a several month period when the radiolysis product H 2 O 2 is added to the groundwater solution. The test setup has been mocked up for operation with spent fuel in the hot-cell

  19. Propagation of thermal neutrons in mock-up screw-shaped steel elements with water protection; Propagation des neutrons thermiques dans des fausses cartouches d'acier en helice dans une protection d'eau. Programme tournesol 3

    Energy Technology Data Exchange (ETDEWEB)

    Devillers, C L; Lanore, J M [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1967-07-01

    This report treats the streaming of thermal neutrons in a cylindrical duct in light water. The duct contains a spiral iron shield. Transmission and reflection matrices are used to describe the probabilities for the thermal neutrons to be absorbed or to be scattered on the surfaces. The neutron paths across the void are represented by geometrical matrices. The numerical resolution is performed by the Monte-Carlo method. (authors) [French] Dans ce rapport on traite un probleme de fuites de neutrons thermiques dans un canal cylindrique plonge dans l'eau et obture par un ecran helicoidal en acier. On utilise des matrices de transmission-reflexion pour decrire les probabilites d'absorption et de diffusion des neutrons sur les parois et l'helicoide et des matrices de correspondance geometrique pour representer la propagation dans le vide. La resolution numerique se fait par une methode de Monte-Carlo. (auteur)

  20. COBACORE Community Based Comprehensive Recovery : D3.1: Mock-ups of interface of COBACORE workspace and functional behaviour of CORBCORE - WP3 Concept development and support mechanisms

    NARCIS (Netherlands)

    Neef, R.M.; Streefkerk, J.W.; Buul-Besseling, K. van; Rijken, M.

    2014-01-01

    D3.1 describes the core feature set of the COBACORE platform, the underlying assumptions and information elements, and introduces a number of interaction patterns. D3.1 is a guiding document that serves as a blueprint for the actual implementation of the COBACORE platform, and that can be used by

  1. Revealing hidden paint layers in oil paintings by means of scanning macro-XRF : A mock-up study based on Rembrandt's “An old man in military costume”

    NARCIS (Netherlands)

    Alfeld, M.; De Nolf, W.; Cagno, S.; Appel, K.; Siddons, D.P.; Kuczewski, A.; Janssens, K.; Dik, J.; Trentelman, K.; Walton, M.; Sartorius, A.

    2012-01-01

    Over the past several decades the oeuvre of Rembrandt has been the subject of extensive art historical and scientific investigations. One of the most striking features to emerge is his frequent re-use of canvases and panels. The painting An Old Man in Military Costume (78.PB.246), in the collection

  2. 'Optimal conditions for group-dynamic challenges' : The results of mock-up research on group-dynamics during the January 2014 Juuka Finland ‘Ice Dome’ building by university students initiated by the Eindhoven Technical University

    NARCIS (Netherlands)

    Sanders, F.C.; Overtoom, M.E.

    2016-01-01

    Society counts a growing number of group-dynamic challenges like civilian movements, resident initia-tive, self steering teams on the work floor and innovation team challenges. The basis driving force is governments that draw back, increasing competition in business and empowerment of people.

  3. 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

  4. Development work for the manufacture of a blanket shield prototype for ITER

    International Nuclear Information System (INIS)

    Boireau, B.; Boudot, C.; Cottin, A.; Lorenzetto, P.; Jacquinot, F.; Bucci, P.; Gillia, O.; Vidotto, F.

    2006-01-01

    In the frame of the blanket development for ITER, an R-and-D programme was implemented for the manufacture of a shield prototype by powder Hot Isostatic Pressing (HIPping). The shield consists of a Stainless steel forged block drilled and machined, at the back of which 3D bent tubes are HIPped inside a powder layer. This paper describes the development work through the manufacturing of several mock ups that leads us to be confident for the shield prototype manufacturing. The paper is divided into 2 parts, the first one related to the machining development and validation, the second one relating to the HIP development and validation. A partial full scale mock up for the machining development (machined PFSMU) was manufactured with no particular problems, all the main identified difficulties in machining like deep drilling and castellation machining where overcome and the mock up was conform to the specification. The manufacturing of a HIPped PFSMU for the HIP development was done after the manufacturing of smaller mock-ups each representing a particular detailed design point. A computer simulation work gave us some design recommendation, and the compared analysis of the numerical simulation and experimental results lead us to predict the distortions on the PFSMU HIPped mock up. The HIPped distortions that were the main uncertainty were assessed through small mock ups and bigger one. The mechanical characteristics of the joints are conform to the specification. Associated to the mock up manufacturing is the ultrasonic test development which consists in designing and manufacturing a miniaturized probe travelling inside the bent tube after the HIP cycle to examine the joint tube / powder among others. This ultrasonic development allowed the examination of the HIPped PFSMU mock up that concluded this development work. (author)

  5. Simulation of a hypothetical core disruptive accident in the mars test-facility

    International Nuclear Information System (INIS)

    Robbe, M.F.; Lepareux, M.

    2001-01-01

    In France, a large experimental programme MARA/MARS was undertaken in the 80's to estimate the mechanical consequences of an HCDA (Hypothetical Core Disruptive Accident) and to validate the SIRIUS computer code used at that time for the numerical simulations. At the end of the 80's, it was preferred to add a HCDA sodium-bubble-argon tri-component constitutive law to the general ALE fast dynamics finite element CASTEM-PLEXUS code rather than going on developing and using the specialized SIRIUS code. The experimental results of the MARA programme were used in the 90's to validate and qualify the CASTEM-PLEXUS code. A first series of computations of the tests MARA 8, MARA 10 and MARS was realised. The simulations showed a rather good agreement between the experimental and computed results for the MARA 8 and MARA 10 tests - even if there were some discrepancies - but the prediction of the MARS structure displacements and strains was overestimated. This conservatism was supposed to come from the fact that several MARS non axisymmetric structures like core elements, pumps and heat exchangers were not represented in the CASTEM-PLEXUS model. These structures, acting as porous barriers, had a protective effect on the mock-up containment by absorbing energy and slowing down the fluid impacting the containment. For these reasons, we developed in CASTEM-PLEXUS a new HCDA constitutive law taking into account the presence of the internal structures (without meshing them) by means of an equivalent porosity method. In other respects, the process used for dealing with the fluid-structure coupling in CASTEM-PLEXUS was improved. Thus a second series of simulations of the tests MARA8 and MARA10 was realised. A simulation of the test MARS was carried out too with the same simplified representation of the peripheral structures as in order to estimate the improvement provided by the new fluid-structure coupling. This paper presents a third numerical simulation of the MARS test with the

  6. Corrosion of several components of the in-situ test performed in a deep geological granite disposal site

    International Nuclear Information System (INIS)

    Madina, Virginia; Azkarate, Inaki; Insausti, Mikel

    2004-01-01

    The corrosion damage experienced by different components in a deep geological disposal in a granite formation has been analysed. This in-situ test is part of the Full-scale Engineered Barriers EXperiment project (FEBEX) carried out in Grimsel (Switzerland). Two heaters, simulating the canister and the heat generated, were installed horizontally inside the guide tubes or liners and surrounded by highly compacted bentonite blocks. Coupons of several candidate metals for manufacturing HLW containers were introduced in these bentonite blocks, as well as sensors in order to monitor different physicochemical parameters during the test. The in- situ test began in July 1996 and in June 2002 one of the heaters, a section of the liner, several corrosion coupons and four sensors were extracted. The studied heater is a carbon steel cylinder with welded lids, with a wall thickness of 100 mm and 4.54 m long. The liner consists of a perforated carbon steel tube, 970 mm in diameter and 15 mm thick. Corrosion coupons were made of carbon steel, stainless steel, titanium, copper and cupronickel alloys. Two extensometer type sensors with an outer protection tube made of austenitic stainless steel were also analysed. Visual inspection of the above mentioned components, optical and scanning electron microscope study, together with EDS and XRD analyses of corrosion products, have been performed in order to analyse the corrosion suffered by these components. This has been complemented with the chemical and microbiological characterisation of bentonite samples. Results obtained in the study indicate a slight generalised corrosion for the heater, liner and corrosion coupons. The low humidity content of the bentonite surrounding the liner and the corrosion coupons, is the responsible of this practical absence of corrosion. The sensors studied show, however, an important corrosion damage. The sulphur rich corrosion products, the presence of Sulphate Reducing Bacteria (SRB) in the bentonite

  7. Concrete/Febex Bentonite Interaction: Results On Short-Term Column Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Escribano, A.; Turrero, M.J.; Torres, E.; Martin, P.L. [CIEMAT, Environmental Department, Avda. Complutense, 22, 28040 Madrid (Spain)

    2008-07-01

    Interaction between the alkaline pore fluids from the concrete engineered barriers and the bentonite at the repository conditions may generate products that can diffuse through the porous structure of the bentonite affecting their properties. A comprehensive study based on series of short term experiments is being performed to provide experimental evidences on the physical, chemical and mineralogical changes during the concrete-compacted bentonite interaction. Samples were analyzed by XRD, SEM-EDS and FTIR. Measurements of swelling capacity, specific surface area and chemical analysis for cation exchange capacity and soluble salts analyses were also performed. (authors)

  8. Concrete/Febex Bentonite Interaction: Results On Short-Term Column Experiments

    International Nuclear Information System (INIS)

    Escribano, A.; Turrero, M.J.; Torres, E.; Martin, P.L.

    2008-01-01

    Interaction between the alkaline pore fluids from the concrete engineered barriers and the bentonite at the repository conditions may generate products that can diffuse through the porous structure of the bentonite affecting their properties. A comprehensive study based on series of short term experiments is being performed to provide experimental evidences on the physical, chemical and mineralogical changes during the concrete-compacted bentonite interaction. Samples were analyzed by XRD, SEM-EDS and FTIR. Measurements of swelling capacity, specific surface area and chemical analysis for cation exchange capacity and soluble salts analyses were also performed. (authors)

  9. 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

  10. Safety of confinement of Super Phenix: MARS test

    International Nuclear Information System (INIS)

    Falgayrettes, M.F.G.; Hamon, P.; Fiche, C.

    1984-08-01

    The protection of people and property must be assured by every situation around an industrial power plant. That is why the French Commissariat a l'Energie Atomique has defined the size of the confinement of Super Phenix to withstand the worst highly hypothetical accident. This report presents the study of the strength of the confinement carried out on a reactor mock-up. The latter is presented in film. The solution which have been adopted for the problem encountered are emphasized. Finally consequences are examined [fr

  11. Integrated programme of research into the behaviour of the clay engineered barrier: an example from Nagra's Grimsel test site

    International Nuclear Information System (INIS)

    Biggin, C.; Alexander, R.; Kickmaier, W.; McKinley, I.G.

    2003-01-01

    Many designs for the disposal of higher activity radioactive wastes include bentonite clay as part of the engineered barrier system (EBS). Generally, the EBS is characterised by the use of large quantities of rather simple, well-understood materials, leading to increased confidence in the predicted long-term behaviour of the EBS (see, for example, Alexander and McKinley, 1999). Despite