WorldWideScience

Sample records for irradiation performance test

  1. Irradiation test and performance evaluation of DUPIC fuel

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Song, K. C.; Moon, J. S.

    2002-05-01

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

  2. SP-100 Fuel Pin Performance: Results from Irradiation Testing

    Science.gov (United States)

    Makenas, Bruce J.; Paxton, Dean M.; Vaidyanathan, Swaminathan; Marietta, Martin; Hoth, Carl W.

    1994-07-01

    A total of 86 experimental fuel pins with various fuel, liner, and cladding candidate materials have been irradiated in the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF) reactor as part of the SP-100 fuel pin irradiation testing program. Postirradiation examination results from these fuel pins are key in establishing performance correlations and demonstrating the lifetime and safety of the reactor fuel system. This paper provides a brief description of the in-reactor fuel pin tests and presents the most recent irradiation data on the performance of wrought rhenium (Re) liner material and high density UN fuel at goal burnup of 6 atom percent (at. %). It also provides an overview of the significant variety of other fuel/liner/cladding combinations which were irradiated as part of this program and which may be of interest to more advanced efforts.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  4. Irradiated fuel performance evaluation technology development

    International Nuclear Information System (INIS)

    Koo, Yang Hyun; Bang, J. G.; Kim, D. H.

    2012-01-01

    Alpha version performance code for dual-cooled annular fuel under steady state operation, so called 'DUOS', has been developed applying performance models and proposed methodology. Furthermore, nonlinear finite element module which could be integrated into transient/accident fuel performance code was also developed and evaluated using commercial FE code. The first/second irradiation and PIE test of annular pellet for dual-cooled annular fuel in the world have been completed. In-pile irradiation test DB of annular pellet up to burnup of 10,000 MWd/MTU through the 1st test was established and cracking behavior of annular pellet and swelling rate at low temperature were studied. To do irradiation test of dual-cooled annular fuel under PWR's simulating steady-state conditions, irradiation test rig/rod design/manufacture of mock-up/performance test have been completed through international collaboration program with Halden reactor project. The irradiation test of large grain pellets has been continued from 2002 to 2011 and completed successfully. Burnup of 70,000 MWd/MTU which is the highest burnup among irradiation test pellets in domestic was achieved

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

    International Nuclear Information System (INIS)

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

    2008-03-01

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

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

    International Nuclear Information System (INIS)

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

    2008-03-01

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

  7. Post irradiation test report of irradiated DUPIC simulated fuel

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Jung, I. H.; Moon, J. S. and others

    2001-12-01

    The post-irradiation examination of irradiated DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) simulated fuel in HANARO was performed at IMEF (Irradiated Material Examination Facility) in KAERI during 6 months from October 1999 to March 2000. The objectives of this post-irradiation test are i) the integrity of the capsule to be used for DUPIC fuel, ii) ensuring the irradiation requirements of DUPIC fuel at HANARO, iii) performance verification in-core behavior at HANARO of DUPIC simulated fuel, iv) establishing and improvement the data base for DUPIC fuel performance verification codes, and v) establishing the irradiation procedure in HANARO for DUPIC fuel. The post-irradiation examination performed are γ-scanning, profilometry, density, hardness, observation the microstructure and fission product distribution by optical microscope and electron probe microanalyser (EPMA)

  8. HANARO fuel irradiation test(II)

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-04-01

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

  9. HANARO fuel irradiation test (II): revision

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-04-01

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

  10. HRB-22 irradiation phase test data report

    International Nuclear Information System (INIS)

    Montgomery, F.C.; Acharya, R.T.; Baldwin, C.A.; Rittenhouse, P.L.; Thoms, K.R.; Wallace, R.L.

    1995-03-01

    Irradiation capsule HRB-22 was a test capsule containing advanced Japanese fuel for the High Temperature Test Reactor (HTTR). Its function was to obtain fuel performance data at HTTR operating temperatures in an accelerated irradiation environment. The irradiation was performed in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL). The capsule was irradiated for 88.8 effective full power days in position RB-3B of the removable beryllium (RB) facility. The maximum fuel compact temperature was maintained at or below the allowable limit of 1300 degrees C for a majority of the irradiation. This report presents the data collected during the irradiation test. Included are test thermocouple and gas flow data, the calculated maximum and volume average temperatures based on the measured graphite temperatures, measured gaseous fission product activity in the purge gas, and associated release rate-to-birth rate (R/B) results. Also included are quality assurance data obtained during the test

  11. Performance and Irradiation Tests of the 0.3$\\mu$m CMOS TDC for the ATLAS MDT

    OpenAIRE

    Arai, Y; Fukuda, M; Emura, T

    1999-01-01

    ATLAS Muon TDC test-element group chip (AMTTEG) was developed and tested to confirm the performance of critical circuits of the TDC and measure radiation tolerance of the process. The chip was fabricated in a 0.3 mm CMOS Gate-Array technology. Measurements of critical elements of the chip such as the PLL, and data buffering circuits demonstrated adequate performance. The effect of gamma-ray irradiation, using a Co60 source, and neutron irradiation, using PROSPERO reactor in France, were also ...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-12-01

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

  13. A Performance Evaluation of a Notebook PC under a High Dose-Rate Gamma Ray Irradiation Test

    Directory of Open Access Journals (Sweden)

    Jai Wan Cho

    2014-01-01

    Full Text Available We describe the performance of a notebook PC under a high dose-rate gamma ray irradiation test. A notebook PC, which is small and light weight, is generally used as the control unit of a robot system and loaded onto the robot body. Using TEPCO’s CAMS (containment atmospheric monitoring system data, the gamma ray dose rate before and after a hydrogen explosion in reactor units 1–3 of the Fukushima nuclear power plant was more than 150 Gy/h. To use a notebook PC as the control unit of a robot system entering a reactor building to mitigate the severe accident situation of a nuclear power plant, the performance of the notebook PC under such intense gamma-irradiation fields should be evaluated. Under a similar dose-rate (150 Gy/h gamma ray environment, the performances of different notebook PCs were evaluated. In addition, a simple method for a performance evaluation of a notebook PC under a high dose-rate gamma ray irradiation test is proposed. Three notebook PCs were tested to verify the method proposed in this paper.

  14. Irradiation tests report of the 32nd cycle in 'JOYO'

    International Nuclear Information System (INIS)

    1998-09-01

    This report summarizes the operating and irradiation data of the experimental reactor 'JOYO' 32nd cycle, and estimates the 33rd cycle irradiation condition. Irradiation tests in the 31st cycle are as follows: (1) B-type irradiation rig (B9). (a) High burn up performance tests of MONJU' fuel pins, advanced austenitic steel cladding fuel pins, large diameter fuel pins, ferrite steel cladding fuel pins (in collaboration with the USA) and large diameter annular pellet fuel pins. (b) Mixed carbide and nitride fuel pins irradiation tests (in collaboration with JAERI). (2) C-type irradiation rig (C4F). (a) High burn up performance test of advanced austenitic steel cladding fuel pins (in collaboration with France). (3) C-type irradiation rig (C6D). (a) Large diameter fuel pins irradiation test. (4) Absorber Materials Irradiation Rig (AMIR-6). (a) Run to absorber pin's cladding breach. (5) Absorber Materials Irradiation Rig (AMIR-8). (a) High-temperature shroud and Na-bond elements tests. (6) Core Materials Irradiation Rig (CMIR-5-1). (a) Core materials irradiation tests. (7) Structure Materials Irradiation Rigs (SMIR). (a) Material irradiation tests (in collaboration with universities). (b) Surveillance back up tests for MONJU'. (8) MAterial testing RIg with temperature COntrol (MARICO-1). (a) Material irradiation tests (in collaboration with universities), (b) Creep rupture tests of the core materials for the demonstration reactor. (9) Upper core structure irradiation Plug Rig (UPR-1-5). (a) Upper core neutron spectrum effect and accelerated irradiation effect. The maximum burn-up driver assembly 'PFD503' reached 65,600 MWd/t (pin average). (author)

  15. Capsule Development and Utilization for Material Irradiation Tests

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-15

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

  16. Post irradiation examination on test fuel pins for PWR

    International Nuclear Information System (INIS)

    Fogaca Filho, N.; Ambrozio Filho, F.

    1981-01-01

    Certain aspects of irradiation technology on test fuel pins for PWR, are studied. The results of post irradiation tests, performed on test fuel pins in hot cells, are presented. The results of the tests permit an evaluation of the effects of irradiation on the fuel and cladding of the pin. (Author) [pt

  17. Capsule Development and Utilization for Material Irradiation Tests

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-15

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

  18. Updated Results of Ultrasonic Transducer Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  19. Irradiation Testing of Ultrasonic Transducers

    International Nuclear Information System (INIS)

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

    2013-06-01

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

  20. Irradiation performance updates on Korean advanced fuels for PWRs

    International Nuclear Information System (INIS)

    Jang, Y.K.; Jeon, K.L.; Kim, Y.H.; Yoo, J.S.; Kim, J.I.; Shin, J.C.; Chung, J.G.; Park, J.R.; Chung, S.K.; Kim, T.W.; Yoon, Y.B.; Park, K.M.; Yoo, M.J.; Kim, M.S.; Lee, T.H.

    2010-01-01

    The developments of advanced nuclear fuels for PWRs were started in 1999 and in 2001, respectively: PLUS7 TM for eight operating optimized power reactors of 1000 MWe class (OPR1000) and four advanced power reactors of 1400 MWe class (APR1400) under construction, and 16ACE7 TM and 17ACE7 TM for an operating 16x16 Westinghouse type plant and six operating 17x17 Westinghouse type plants. The design targets were as follows: batch average burnup up to 55 GWD/MTU, over 10% thermal margin increase, improvement of the mechanical integrity of higher seismic capability, higher debris or grid fretting wear performance, higher control rod insertion capability, increase of neutron economy, improvement of manufacturability, solving incomplete rod insertion (IRI) issue and top nozzle screw failure issue, etc. in comparison of the existing nuclear fuels. The irradiation tests using each four LTAs (Lead Test Assemblies) during 3 cycles were completed in three Korean nuclear reactors until 2009. The eight irradiation performance items which are assembly growth, rod growth, grid width growth, assembly bow, rod bow, assembly twist, rod diameter and cladding oxidation were examined in pool-side after each cycle and evaluated. The irradiation tests could be continued by expecting the good performances for next cycle from the previous cycle. After 2 cycle irradiations, the region implementation could be started in 15 nuclear power plants. Even though the verifications using the LTAs were completed, each surveillance program was launched and the irradiation performance data were being updated during region implementation. In addition to pool-side examinations (PSEs) by assembly-wise during irradiation tests, six rod-wise performance items were also examined in pool-side using each LTA after discharge. All performance items met their design criteria as a result of the evaluation. Even though the interesting ones among the irradiation performance parameters were assembly and grid growths

  1. Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

    International Nuclear Information System (INIS)

    Loelgen, R.H.; Cundy, M.R.; Schuele, W.

    1977-01-01

    A review is given of investigations on the rate of phase changes during neutron and electron irradiation in many different fcc alloys showing either precipitation or ordering. The diffusion rate was determined as a function of the irradiation flux, the irradiation temperature and the irradiation dose. It was found that the radiation enhanced diffusion in all the investigated alloys is nearly temperature independent and linearly dependent on the flux. From these results conclusions were drawn concerning the properties of point defects and diffusion mechanisms rate determining during irradiation, which appears to be of a common nature for fcc alloys having a similar structure to those investigated. It has been recognized that the same dependencies which are found for the diffusion rate were also observed for the irradiation creep rate in stainless steels, as reported in literature. On the basis of this observation a combination of measurements is suggested, of radiation enhanced diffusion and radiation enhanced creep in stainless steel alloys. The diffusion tests will be performed at the Euratom Joint Research Centre in Ispra, Italy, and the irradiation creep tests will be carried out in the High Flux Reactor /9/ of the Euratom Joint Research Centre in Petten, The Netherlands. In order to investigate irradiation creep on many samples at a time two special rigs were developed which are distinguished only by the mode of stress applied to the steel specimens. In the first type of rig about 50 samples can be tested uniaxially under tension with various combinations of irradiation temperature and stress. The second type of rig holds up to 70 samples which are tested in bending, again with various combinations of irradiation temperature and stress

  2. FFTF utilization for irradiation testing

    International Nuclear Information System (INIS)

    Corrigan, D.C.; Julyk, L.J.; Hoth, C.W.; McGuire, J.C.; Sloan, W.R.

    1980-01-01

    FFTF utilization for irradiation testing is beginning. Two Fuels Open Test Assemblies and one Vibration Open Test Assembly, both containing in-core contact instrumentation, are installed in the reactor. These assemblies will be used to confirm plant design performance predictions. Some 100 additional experiments are currently planned to follow these three. This will result in an average core loading of about 50 test assemblies throughout the early FFTF operating cycles

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-09-15

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

  4. In-cell facility for performing mechanical-property tests on irradiated cladding

    International Nuclear Information System (INIS)

    Yaggee, F.L.; Haglund, R.C.; Mattas, R.F.

    1978-11-01

    A new facility was developed for testing cladding sections of LWR fuel rods. This facility and the accompanying test procedures have improved the level of in-cell mechanical-testing capabilities, making them comparable to existing capabilities for unirradiated cladding. The new facility is currently being used to study the susceptibility of irradiated Zircaloy cladding from LWR fuel rods to iodine stress-corrosion cracking. Preliminary testing results indicate a systematic effect of temperature, stress and irradiation on the susceptibility of annealed and stress-relieved Zircaloy-2. Experimental data obtained to date are being used to develop a stress-corrosion cracking model for LWR fuel rod failure. SEM examination of the undisturbed fracture surface of specimens that failed by pinhole leakage provides useful information on crack propagation and morphology

  5. Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

    International Nuclear Information System (INIS)

    Loelgen, R.H.; Cundy, M.R.; Schuele, W.

    1977-01-01

    A review is given of investigations on the rate of phase changes during neutron and electron irradiation in many different fcc alloys showing either precipitation or ordering. The diffusion rate was determined as a function of the irradiation flux, the irradiation temperature and the irradiation dose. It was found that the radiation enhanced diffusion in all the investigated alloys is nearly temperature independent and linearly dependent on the flux. From these results conclusions were drawn concerning the properties of point defects and diffusion mechanisms rate determining during irradiation, which appears to be of a common nature for fcc alloys having a similar structure to those investigated. It has been recognized that the same dependencies which are found for the diffusion rate were also observed for the irradiation creep rate in stainless steels, as reported in literature. On the basis of this obervation a combination of measurements is suggested, of radiation enhanced diffusion and radiation enhanced creep in stainless steel alloys. Measurements of radiation enhanced diffusion are less time consuming and expensive than irradiation creep tests and information on this property can be obtained rather quickly, prior to the selection of stainless steel alloys for creep tests. In order to investigate irradiation creep on many samples at a time two special rigs were developed which are distinguished only by the mode of stress applied to the steel specimens. Finally, a few uniaxial tensile creep tests will be performed in fully instrumented rigs. (Auth.)

  6. The Advanced Test Reactor Irradiation Facilities and Capabilities

    International Nuclear Information System (INIS)

    S. Blaine Grover; Raymond V. Furstenau

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-01

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

  8. Tensile and fracture toughness test results of neutron irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R.; Moons, F.; Puzzolante, J.L. [Centre d`Etude de l`Energie Nucleaire, Mol (Belgium)

    1998-01-01

    Tensile and fracture toughness test results of four Beryllium grades are reported here. The flow and fracture properties are investigated by using small size tensile and round compact tension specimens. Irradiation was performed at the BR2 material testing reactor which allows various temperature and irradiation conditions. The fast neutron fluence (>1 MeV) ranges between 0.65 and 2.45 10{sup 21} n/cm{sup 2}. In the meantime, un-irradiated specimens were aged at the irradiation temperatures to separate if any the effect of temperature from irradiation damage. Test results are analyzed and discussed, in particular in terms of the effects of material grade, test temperature, thermal ageing and neutron irradiation. (author)

  9. Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

    International Nuclear Information System (INIS)

    Biemiller, E.C.; Fyfitch, S.; Campbell, C.A.

    1993-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 1800 deg F) and to simulate Regulatory Guide 1.99 database materials (austenitized at 1600 deg. F). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (U. of Michigan Test Reactor) which had never been used for this type of irradiation program. Materials taken from plate surface locations (vs. 1/4T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, is maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (500 deg. F and 550 deg. F) to determine the effect of irradiation temperature on embrittlement. The preliminary results of the irradiation program show an increase in T 30 shift of 69 deg. F for a decrease in irradiation temperature of 50 deg. F. The results suggest that for nickel bearing steels, the superior toughness of plate surface material is maintained after irradiation and for the copper content tested, nickel had no apparent effect on irradiation response. No apparent microstructure

  10. AGR-2 Irradiation Test Final As-Run Report

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Program

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) program. The objectives of the AGR-2 experiment are to: 1. Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. 2. Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. 3. Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tristructural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S.-produced fuel.

  11. AGR-2 Irradiation Test Final As-Run Report

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Program

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel.

  12. Endurance test of DUPIC irradiation test rig-003

    Energy Technology Data Exchange (ETDEWEB)

    Moon, J.S; Yang, M.S.; Lee, C.Y.; Ryu, J.S.; Jeon, H.G

    2001-04-01

    This report presents the pressure drop, vibration and endurance test results for DUPIC Irradiation Test Rig-003 which was design and fabricated by KAERI. From the pressure drop and vibration test results, it is verified that DUPIC Irradiation Test Rig-003 satisfied the limit conditions of HANARO. And, remarkable wear is not observed in DUPIC Irradiation Test Rig-003 during 40 endurance test days.

  13. Irradiation performance of HTGR recycle fissile fuel

    International Nuclear Information System (INIS)

    Homan, F.J.; Long, E.L. Jr.

    1976-08-01

    The irradiation performance of candidate HTGR recycle fissile fuel under accelerated testing conditions is reviewed. Failure modes for coated-particle fuels are described, and the performance of candidate recycle fissile fuels is discussed in terms of these failure modes. The bases on which UO 2 and (Th,U)O 2 were rejected as candidate recycle fissile fuels are outlined, along with the bases on which the weak-acid resin (WAR)-derived fissile fuel was selected as the reference recycle kernel. Comparisons are made relative to the irradiation behavior of WAR-derived fuels of varying stoichiometry and conclusions are drawn about the optimum stoichiometry and the range of acceptable values. Plans for future testing in support of specification development, confirmation of the results of accelerated testing by real-time experiments, and improvement in fuel performance and reliability are described

  14. Integrity Assessment of HANARO Irradiation Capsule for Long-Term Irradiation Testing

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Kee Nam; Cho, Man Soon; Yang, Sung Woo; Shin, Yoon Taek; Park, Seng Jae; Yang, Tae Ho; Jun, Byung Hyuk; Kim, Myong Seop [KAERI, Daejeon (Korea, Republic of); Hong, Sang Hyun [Chungnam University, Daejeon (Korea, Republic of)

    2016-05-15

    The capsule technology was basically developed for irradiation testing under a commercial reactor operation environment. Most irradiation testing using capsules has been performed at around 300 .deg. C within four reactor operation cycles (about 100 days equivalent to 1.5 dpa (displacement for atom)) at HANARO. Based on the accumulated experience as well as the sophisticated requirements of users, HANARO has recently been required to support national R and D projects requiring much higher neutron fluence. To scope the user requirements for higher neutron irradiation fluence, several efforts using an instrumented capsule have been applied at HANARO. In this paper, the applied stresses on the capsule are estimated because the capsule was suspected to be susceptible to fatigue failure during irradiation testing. In addition, the on-going design improvements of the irradiation capsule for higher neutron irradiation fluence at HANARO are described. The applied stresses on the rod tip were analyzed using the ANSYS program. The applied stresses on the rod tip can be classified into stresses by the designed bottom spring, by the upward flowing coolant, by the capsule vibration, and by the welding residual stress. The maximal stresses due to the first three factors were estimated as 5.4 MPa, 132.9 MPa, and 161 MPa, respectively. These stresses do not exceed the known fatigue strength of stainless steels (∼300 MPa). Residual stress by welding is another possible stress and it is known to occur at up to about 300 MPa.

  15. New JMTR irradiation test plan on fuels and materials

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Nishiyama, Yutaka; Chimi, Yasuhiro; Sasajima, Hideo; Ogiyanagi, Jin; Nakamura, Jinichi; Suzuki, Masahide; Kawamura, Hiroshi

    2009-01-01

    In order to maintain and enhance safety of light water reactors (LWRs) in long-term and up-graded operations, proper understanding of irradiation behavior of fuels and materials is essentially important. Japanese government and the Japan Atomic Energy Agency (JAEA) have decided to refurbish the Japan Materials Testing Reactor (JMTR) and to install new tests rigs, in order to play an active role for solving irradiation related issues on plant aging and high-duty uses of the current LWRs and on development of next-generation reactors. New tests on fuel integrity under simulated abnormal transients and high-duty irradiation conditions are planned in the JMTR. Power ramp tests of newdesign fuel rods will also be performed in the first stage of the program, which is expected to start in year 2011 after refurbishment of the JMTR. Combination of the JMTR tests with simulated reactivity initiated accident tests in the Nuclear Safety Research Reactor (NSRR) and loss of coolant accident tests in hot laboratories would serve as the integrated fuel safety research on the high performance fuels at extended burnups, covering from the normal to the accident conditions, including abnormal transients. For the materials irradiation, fracture toughness of reactor vessel steels and stress corrosion cracking behavior of stainless steels are being studied in addition to basic irradiation behavior of nuclear materials such as hafnium. The irradiation studies would contribute not only to solve the current problems but also to identify possible seeds of troubles and to make proactive responses. (author)

  16. DUPIC fuel irradiation test and performance evaluation; the performance analysis of pellet-cladding contact fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ho, K. I.; Kim, H. M.; Yang, K. B.; Choi, S. J. [Suwon University, Whasung (Korea)

    2002-04-01

    Thermal and mechanical models were reviewed, and selected for the analysis of nuclear fuel performance in reactor. 2 dimensional FEM software was developed. Thermal models-gap conductances, thermal conductivity of pellets, fission gas release, temperature distribution-were set and packaged into a software. Both thermal and mechanical models were interrelated to each other, and the final results, fuel performance during irradiation is obtained by iteration calculation. Also, the contact phenomena between pellet and cladding was analysed by mechanical computer software which was developed during this work. dimensional FEM program was developed which estimate the mechanical behavior and the thermal behaviors of nuclear fuel during irradiation. Since there is a importance during the mechanical deformation analysis in describing pellet-cladding contact phenomena, simplified 2 dimensional calculation method is used after the contact. The estimation of thermal fuel behavior during irradiation was compared with the results of other. 8 refs., 17 figs. (Author)

  17. Metallographic analysis of irradiated RERTR-3 fuel test specimens

    International Nuclear Information System (INIS)

    Meyer, M. K.; Hofman, G. L.; Strain, R. V.; Clark, C. R.; Stuart, J. R.

    2000-01-01

    The RERTR-3 irradiation test was designed to investigate the irradiation behavior of aluminum matrix U-MO alloy dispersion fuels under high-temperature, high-fission-rate conditions. Initial postirradiation examination of RERTR-3 fuel specimens has concentrated on binary U-MO atomized fuels. The rate of matrix aluminum depletion was found to be higher than predictions based on low temperature irradiation data. Wavelength Dispersive X-ray Spectroscopy (WDS) indicates that aluminum is present in the interior of the fuel particles. WDS data is supported by a mass and volume balance calculation performed on the basis of image analysis results. The depletion of matrix aluminum seems to have no detrimental effects on fuel performance under the conditions tested to date

  18. Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

    International Nuclear Information System (INIS)

    Biemiller, E.C.; Fyfitch, Stephen; Campbell, C.A.

    1994-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 982 o C (1800 o F)) and to simulate Regulatory Guide 1.99 database materials (austenitized at 871 o C (1600 o F)). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (University of Michigan Test Reactor) which had never been used before for this type of irradiation program. Materials taken from plate surface locations (versus 1/4 T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, are maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (260 o C and 288 o C) to determine the effect of irradiation temperature on embrittlement. (Author)

  19. Preliminary test results for post irradiation examination on the HTTR fuel

    International Nuclear Information System (INIS)

    Ueta, Shohei; Umeda, Masayuki; Sawa, Kazuhiro; Sozawa, Shizuo; Shimizu, Michio; Ishigaki, Yoshinobu; Obata, Hiroyuki

    2007-01-01

    The future post-irradiation program for the first-loading fuel of the HTTR is scheduled using the HTTR fuel handling facilities and the Hot Laboratory in the Japan Materials Testing Reactor (JMTR) to confirm its irradiation resistance and to obtain data on its irradiation characteristics in the core. This report describes the preliminary test results and the future plan for a post-irradiation examination for the HTTR fuel. In the preliminary test, fuel compacts made with the same SiC-coated fuel particle as the first loading fuel were used. In the preliminary test, dimension, weight, fuel failure fraction, and burnup were measured, and X-ray radiograph, SEM, and EPMA observations were carried out. Finally, it was confirmed that the first-loading fuel of the HTTR showed good quality under an irradiation condition. The future plan for the post-irradiation tests was described to confirm its irradiation performance and to obtain data on its irradiation characteristics in the HTTR core. (author)

  20. Irradiation effects test series test IE-1 test results report

    International Nuclear Information System (INIS)

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

    1977-03-01

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

  1. An X-Ray facility to perform irradiation tests and TID studies on electronics and detectors

    CERN Document Server

    Brundu, Davide; Cadeddu, Sandro; Wyllie, Ken; Ciambrone, Paolo

    2018-01-01

    The X-Ray irradiation system of the LHCb group, installed in Cagliari, is presented; with a particular focus on the setup configuration and dose rate calibration. The system can be used to perform Total Ionizing Dose (TID) studies for detectors, readout and front-end electronics. It was already used to test the nSYNC chip, an ASIC for the readout of the LHCb upgraded muon system.

  2. Irradiation testing of high density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-10-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U 2 Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions

  3. Summary of ALSEP Test Loop Solvent Irradiation Testing

    Energy Technology Data Exchange (ETDEWEB)

    Peterman, Dean Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Olson, Lonnie Gene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    Separating the minor actinide elements (americium and curium) from the fission product lanthanides is an important step in closing the nuclear fuel cycle. Isolating the minor actinides will allow transmuting them to short lived or stable isotopes in fast reactors, thereby reducing the long-term hazard associated with these elements. The Actinide Lanthanide Separation Process (ALSEP) is being developed by the DOE-NE Material Recovery and Waste Form Development Campaign to accomplish this separation with a single process. To develop a fundamental understanding of the solvent degradation mechanisms for the ALSEP Process, testing was performed in the INL Radiolysis/Hydrolysis Test Loop for the extraction section of the ALSEP flowsheet. This work culminated in the completion of the level two milestone (M2FT-16IN030102021) "Complete ALSEP test loop solvent irradiation test.” This report summarizes the testing performed and the impact of radiation on the ALSEP Process performance as a function of dose.

  4. JOYO-1 Irradiation Test Campaign Technical Close-out, For Information

    International Nuclear Information System (INIS)

    G. Borges

    2006-01-01

    The JOYO-1 irradiation testing was designed to screen the irradiation performance of candidate cladding, structural and reflector materials in support of space reactor development. The JOYO-1 designation refers to the first of four planned irradiation tests in the JOYO reactor. Limited irradiated material performance data for the candidate materials exists for the expected Prometheus-1 duration, fluences and temperatures. Materials of interest include fuel element cladding and core materials (refractory metal alloys and silicon carbide (Sic)), vessel and plant structural materials (refractory metal alloys and nickel-base superalloys), and control and reflector materials (BeO). Key issues to be evaluated were long term microstructure and material property stability. The JOYO-1 test campaign was initiated to irradiate a matrix of specimens at prototypical temperatures and fluences anticipated for the Prometheus-1 reactor [Reference (1)]. Enclosures 1 through 9 describe the specimen and temperature monitors/dosimetry fabrication efforts, capsule design, disposition of structural material irradiation rigs, and plans for post-irradiation examination. These enclosures provide a detailed overview of Naval Reactors Prime Contractor Team (NRPCT) progress in specific areas; however, efforts were in various states of completion at the termination of NRPCT involvement with and restructuring of Project Prometheus

  5. JOYO-1 Irradiation Test Campaign Technical Close-out, For Information

    Energy Technology Data Exchange (ETDEWEB)

    G. Borges

    2006-01-31

    The JOYO-1 irradiation testing was designed to screen the irradiation performance of candidate cladding, structural and reflector materials in support of space reactor development. The JOYO-1 designation refers to the first of four planned irradiation tests in the JOYO reactor. Limited irradiated material performance data for the candidate materials exists for the expected Prometheus-1 duration, fluences and temperatures. Materials of interest include fuel element cladding and core materials (refractory metal alloys and silicon carbide (Sic)), vessel and plant structural materials (refractory metal alloys and nickel-base superalloys), and control and reflector materials (BeO). Key issues to be evaluated were long term microstructure and material property stability. The JOYO-1 test campaign was initiated to irradiate a matrix of specimens at prototypical temperatures and fluences anticipated for the Prometheus-1 reactor [Reference (1)]. Enclosures 1 through 9 describe the specimen and temperature monitors/dosimetry fabrication efforts, capsule design, disposition of structural material irradiation rigs, and plans for post-irradiation examination. These enclosures provide a detailed overview of Naval Reactors Prime Contractor Team (NRPCT) progress in specific areas; however, efforts were in various states of completion at the termination of NRPCT involvement with and restructuring of Project Prometheus.

  6. Irradiation performance of metallic fuels

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Porter, D.L.; Batte, G.L.; Hofman, G.L.

    1989-01-01

    Argonne National Laboratory has been working for the past five years to develop and demonstrate the Integral Fast Reactor (IFR) concept. The concept involves a closed system for fast-reactor power generation and on-site fuel reprocessing, both designed specifically around the use of metallic fuel. The Experimental Breeder Reactor-II (EBR-II) has used metallic fuel for all of its 25-year life. In 1985, tests were begun to examine the irradiation performance of advanced-design metallic fuel systems based on U-Zr or U-Pu-Zr fuels. These tests have demonstrated the viable performance of these fuel systems to high burnup. The initial testing program will be described in this paper. 2 figs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-04-15

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

  8. Neutron irradiation test of copper alloy/stainless steel joint materials

    International Nuclear Information System (INIS)

    Yamada, Hirokazu; Kawamura, Hiroshi

    2006-01-01

    As a study about the joint technology of copper alloy and stainless steel for utilization as cooling piping in International Thermonuclear Experimental Reactor (ITER), Al 2 O 3 -dispersed strengthened copper or CuCrZr was jointed to stainless steel by three kinds of joint methods (casting joint, brazing joint and friction welding method) for the evaluation of the neutron irradiation effect on joints. A neutron irradiation test was performed to three types of joints and each copper alloy. The average value of fast neutron fluence in this irradiation test was about 2 x 10 24 n/m 2 (E>1 MeV), and the irradiation temperature was about 130degC. As post-irradiation examinations, tensile tests, hardness tests and observation of fracture surface after the tensile tests were performed. All type joints changed to be brittle by the neutron irradiation effect like each copper alloy material, and no particular neutron irradiation effect due to the effect of joint process was observed. On the casting and friction welding, hardness of copper alloy near the joint boundary changed to be lower than that of each copper alloy by the effect of joint procedure. However, tensile strength of joints was almost the same as that of each copper alloy before/after neutron irradiation. On the other hand, tensile strength of joints by brazing changed to be much lower than CuAl-25 base material by the effect of joint process before/after neutron irradiation. Results in this study showed that the friction welding method and the casting would be able to apply to the joint method of piping in ITER. This report is based on the final report of the ITER Engineering Design Activities (EDA). (author)

  9. Irradiation and testing of compact ignition tokamak toroidal field coil insulation materials

    International Nuclear Information System (INIS)

    Kanemoto, G.K.; Sherick, M.J.; Sparks, D.C.

    1990-05-01

    This report documents the results of an irradiation and testing program performed on behalf of Martin Marietta Energy Systems, Inc. in support of the Compact Ignition Tokamak Research and Development program. The purpose of the irradiation and testing program was to determine the effects of neutron and gamma irradiation on the mechanical and electrical properties of candidate toroidal field coil insulation materials. Insulation samples were irradiated in the Advanced Test Reactor (ATR) in a large I-hole. The insulation samples were irradiated within a lead shield to reduce exposure to gamma radiation to better approximate the desired ration of neutron to gamma exposure. Two different exposure levels were specified for the insulation samples. To accomplish this, the samples were encapsulated in two separate aluminum capsules; the capsules positioned at the ATR core mid-plane and at the top of the fueled region to take advantage of the axial cosine distribution of the neutron and gamma flux; and by varying the length of irradiation time of the two capsules. Disassembly of the irradiated capsules and testing of the insulation samples were performed at the Test Reactor Area (TRA) Hot Cell Facilities. Testing of the samples included shear compression static, shear compression fatigue, flexure static, and electrical resistance measurements

  10. Irradiation testing of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-01-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 'microplates'. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U10Mo-0.05Sn, U2Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of approximately 40 and 80 at.% U 235 . Of particular interest are the extent of reaction of the fuel and matrix phases and the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions. (author)

  11. Development of small scale mechanical testing techniques on ion beam irradiated 304 SS

    International Nuclear Information System (INIS)

    Reichardt, A.; Abad, M.D.; Hosemann, P.; Lupinacci, A.; Kacher, J.; Minor, A.; Jiao, Z; Chou, P.

    2015-01-01

    Austenitic stainless steels are widely used for structural components in light water reactors, however uncertainty in their susceptibility to irradiation assisted stress corrosion cracking (IASCC) has made long term performance predictions difficult. In addition, the testing of reactor irradiated materials has proven challenging due to the long irradiation times required, limited sample availability, and unwanted activation. To address these problems, we apply recently developed techniques in nano-indentation and micro-compression testing to small volume samples of 10 dpa proton-beam irradiated 304 stainless steel. Cross sectional nano-indentation was performed on both proton beam irradiated and non-irradiated samples at temperatures ranging from 22 to 300 C. degrees to determine the effects of irradiation and operating temperature on hardening. Micro-compression tests using 2 μm x 2 μm x 5 μm focused-ion beam milled pillars were then performed in situ in an electron microscope to allow for a more accurate look at stress-strain behavior along with real-time observations of localized mechanical deformation. Large sudden slip events and significant increase in yield strength are observed in irradiated micro-compression samples at room temperature. Elevated temperature nano-indentation results reveal the possibility of thermally-activated changes in deformation mechanism for irradiated specimens. Since the deformation mechanism information provided by micro-compression testing can provide valuable information about IASCC susceptibility, future work will involve ex situ micro-compression tests at reactor operating temperature

  12. Technical review on irradiation tests and post-irradiation examinations in JMTR

    International Nuclear Information System (INIS)

    2017-07-01

    The Japan Materials Testing Reactor (JMTR) has been contributing to various R and D activities in the nuclear research such as the fundamental research of nuclear materials/ fuels, safety research and development of power reactors, radio isotope (RI) production since its beginning of the operation in 1968. Irradiation technologies and post irradiation examination (PIE) technologies are the important factors for irradiation test research. Moreover, these technologies induce the breakthrough in area of nuclear research. JMTR has been providing unique capabilities for the irradiation test research for about 40 years since 1968. In future, any needs for irradiation test research used irradiation test reactors will continue, such as R and D of generation 4 power reactors, fundamental research of materials/fuels, RI production. Now, decontamination and new research reactor construction are common issue in the world according to aging. This situation is the same in Japan. This report outlines irradiation and PIE technologies developed at JMTR in 40 years to contribute to the technology transfer and human resource development. We hope that this report will be used for the new research rector design as well as the irradiation test research and also used for the human resource development of nuclear engineers in future. (author)

  13. An investigation of neutron irradiation test on superplastic zirconia-ceramic materials

    International Nuclear Information System (INIS)

    Shibata, Taiju; Ishihara, Masahiro; Baba, Shinichi; Hayashi, Kimio

    2000-05-01

    A neutron irradiation test on superplastic ceramic materials at high temperature has been proposed as an innovative basic research on high-temperature engineering using the High Temperature Engineering Test Reactor (HTTR). For the effective execution of the test, we reviewed the superplastic deformation mechanism of ceramic materials and discussed neutron irradiation effects on the superplastic deformation process of stabilized Tetragonal Zirconia Polycrystal (TZP), which is a representative superplastic ceramic material. As a result, we pointed out that the decrease in the activation energy for superplastic deformation is expected by the radiation-enhanced diffusion. We selected a fast neutron fluence of 5x10 20 n/cm 2 and an irradiation temperature of about 600degC as test conditions for the first irradiation test on TZP and decided to perform a preliminary irradiation test by the Japan Materials Testing Reactor (JMTR). Moreover, we estimated the radioactivity of irradiated TZP and indicated that it is in the order of 10 10 Bq/g (about 0.3 Ci/g) immediately after irradiation to a thermal neutron fluence of 3x10 20 n/cm 2 and that it decays to about 1/100 in a year. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-05-15

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

  15. Development of endplug welding technology for irradiation testing capsule

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-10-01

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

  16. LVDT Development for High Temperature Irradiation Test and Application

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chul Yong; Ban, Chae Min; Choo, Kee Nam; Jun, Byung Hyuk [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The LVDT (Linear Variable Differential Transformer) is used to measure the elongation and pressure of a nuclear fuel rod, or the creep and fatigue of the material during a reactor irradiation test. This device must be a radiation-resistant LVDT for use in a research reactor. Norway Halden has LVDTs for an irradiation test by the own development and commercialized. But Halden's LVDTs have limited the temperature of the use until to 350 .deg. C. So, KAERI has been developing a new LVDT for high temperature irradiation test. This paper describes the design of a LVDT, the fabrication process of a LVDT, and the result of the performance test. The designed LVDT uses thermocouple cable for coil wire material and one MI cable as signal cable. This LVDT for a high temperature irradiation test can be used until a maximum of 900 .deg. C. Welding is a very important factor for the fabrication of an LVDT. We are using a 150W fiber laser welding system that consists of a welding head, monitoring vision system and rotary index.

  17. Irradiation and performance evaluation of DUPIC fuel

    International Nuclear Information System (INIS)

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

    2000-05-01

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

  18. Irradiation and performance evaluation of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  19. The 3rd irradiation test plan of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Song, K. C.; Park, J. H. and others

    2001-05-01

    The objective of the 3rd irradiation test of DUPIC fuel at the HANARO is to estimate the in-core behaviour of a DUPIC pellet that is irradiated up to more than average burnup of CANDU fuel. The irradiation of DUPIC fuel is planned to start at May 21, 2001, and will be continued at least for 8 months. The burnup of DUPIC fuel through this irradiation test is thought to be more than 7,000 MWd/tHE. The DUPIC irradiation rig instrumented with three SPN detectors will be used to accumulate the experience for the instrumented irradiation and to estimate the burnup of irradiated DUPIC fuel more accurately. Under normal operating condition, the maximum linear power of DUPIC fuel was estimated as 55.06 kW/m, and the centerline temperature of a pellet was calculated as 2510 deg C. In order to assess the integrity of DUPIC fuel under the accident condition postulated at the HANARO, safety analyses on the locked rotor and reactivity insertion accidents were carried out. The maximum centerline temperature of DUPIC fuel was estimated 2590 deg C and 2094 deg C for each accident, respectively. From the results of the safety analysis, the integrity of DUPIC fuel during the HANARO irradiation test will be secured. The irradiated DUPIC fuel will be transported to the IMEF. The post-irradiation examinations are planned to be performed at the PIEF and IMEF.

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

    International Nuclear Information System (INIS)

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

    1977-10-01

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

  1. Experimental data report for Test TS-2 reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo

    1993-02-01

    This report presents experimental data for Test TS-2 which was the second test in a series of Reactivity Initiated Accident (RIA) condition test using pre-irradiated BWR fuel rods, performed at the Nuclear Safety Research Reactor (NSRR) in February, 1990. Test fuel rod used in the Test TS-2 was a short sized BWR (7x7) type rod which was fabricated from a commercial rod irradiated at Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79% and a burnup of 21.3Gwd/tU (bundle average). A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 72±5cal/g·fuel (66±5cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and, results of pre and post pulse irradiation examinations are described in this report. (author)

  2. AGR-1 Irradiation Test Final As-Run Report, Rev. 3

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-01-01

    This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained 12 compacts of a single type, or variant, of the AGR coated fuel. No fuel particles failed during the AGR-1 irradiation. Final burnup values on a per compact basis ranged from 11.5 to 19.6 %FIMA, while fast fluence values ranged from 2.21 to 4.39 x 1025 n/m2 (E >0.18 MeV). We’ll say something here about temperatures once thermal recalc is done. Thermocouples performed well, failing at a lower rate than expected. At the end of the irradiation, nine of the originally-planned 19 TCs were considered functional. Fission product release-to-birth (R/B) ratios were quite low. In most capsules, R/B values at the end of the irradiation were at or below

  3. Evaluation of burnup characteristics and energy deposition during NSRR pulse irradiation tests on irradiated BWR fuels

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio

    2000-11-01

    Pulse irradiation tests of irradiated fuel are performed in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior under Reactivity Initiated Accident Conditions (RIA). The severity of the RIA is represented by energy deposition or peak fuel enthalpy during the power excursion. In case of the irradiated fuel tests, the energy deposition varies depending both on the amounts and distribution of residual fissile and neutron absorbing fission products generated during the base irradiation. Thus, proper fuel burnup characterization, especially for low enriched commercial fuels, is important, because plutonium (Pu) takes a large part of fissile and its generation depends on the neutron spectrum during the base irradiation. Fuel burnup calculations were conducted with ORIGEN2, RODBURN and SWAT codes for the BWR fuels tested in the NSRR. The calculation results were compared with the measured isotope concentrations and used for the NSRR neutron calculations to evaluate energy depositions of the test fuel. The comparison of the code calculations and the measurements revealed that the neutron spectrum change due to difference in void fraction altered Pu generation and energy deposition in the NSRR tests considerably. With the properly evaluated neutron spectrum, the combined burnup and NSRR neutron calculation gave reasonably good evaluation of the energy deposition. The calculations provided radial distributions of the fission product accumulation during the base irradiation and power distribution during the NSRR pulse irradiation, which were important for the evaluation of both burnup characteristics and fission gas release behavior. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-03-01

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

  5. Genotoxicity test of irradiated spice mixture by dominant lethal test

    Energy Technology Data Exchange (ETDEWEB)

    Barna, J

    1986-03-01

    Dominant lethal test (DLT) was performed in Sprague Dawley male rats prefed with 25% irradiated spice mixture which was composed of 55% non-pungent ground paprika, 14% black pepper, 9% allspice, 9% coriander, 7% marjoram, 4% cumin, 2% nutmeg. Microbial count of the spice mixture was reduced with 15 kGy from a sup(60)Co source. Control groups received spice-free or untreated spice diet or were administered to cyclophosphamide i.p., respectively. DTL parameters altered significantly in the latter group but neither untreated nor irradiated spice mixture proved to be germ cell mutagens. 24 refs.; 8 figs.

  6. Postirradiation examination results for the Irradiation Effects Scoping Test 2

    International Nuclear Information System (INIS)

    Mehner, A.S.

    1977-01-01

    The postirradiation examination results are reported for two rods from the second scoping test (IE-ST-2) of the Nuclear Regulatory Commission Irradiation Effects Program. The rods were irradiated in the in-pile test loop of the Power Burst Facility at the Idaho National Engineering Laboratory. Rod IE-005 was fabricated from fresh fuel and cladding previously irradiated in the Saxton Reactor. Rod IE-006, fabricated from fresh fuel and unirradiated cladding, was equipped with six developmental cladding surface thermocouples. The rods were preconditioned, power ramped, and then subjected to film boiling operation. The performance of the rods and the developmental thermocouples are evaluated from the post irradiation examination results. The effects of prior irradiation damage in cladding are discussed in relation to fuel rod behavior during a power ramp and subsequent film boiling operation

  7. Post-Irradiation Examination Test of the Parts of X-Gen Nuclear Fuel Assembly

    International Nuclear Information System (INIS)

    Ahn, S. B.; Ryu, W. S.; Choo, Y. S.

    2008-08-01

    The mechanical properties of the parts of a nuclear fuel assembly are degraded during the operation of the reactor, through the mechanism of irradiation damage. The properties changes of the parts of the fuel assembly should be quantitatively estimated to ensure the safety of the fuel assembly and rod during the operation. The test techniques developed in this report are used to produce the irradiation data of the grid 1x1 cell spring, the grid 1x1 cell, the spring on one face of the 1x1 cell, the inner/outer strip of the grid and the welded part. The specimens were irradiated in the CT test hole of HANARO of a 30 MW thermal output at 300 deg. C during about 100 days From the spring test of mid grid 1x1 cell and grid plate, the irradiation effects can be examined. The irradiation effects on the irradiation growth also were occurred. The buckling load of mid grid 1x1 cell does not change with a neutron irradiation. From the tensile tests, the strengths increased but the elongations decreased due to an irradiation. The tensile test and microstructure examination of the spot and fillet welded parts are performed for the evaluation of an irradiation effects. Through these tests of components, the essential data on the fuel assembly design could be obtained. These results will be used to update the irradiation behavior databases, to improve the performance of fuel assembly, and to predict the service life of the fuel assembly in a reactor

  8. Irradiation and examination results of the AC-3 mixed-carbide test

    International Nuclear Information System (INIS)

    Mason, R.E.; Hoth, C.W.; Stratton, R.W.; Botta, F.

    1992-01-01

    The AC-3 test was a cooperative Swiss/US irradiation test of mixed-carbide, (U,Pr)C, fuel pins in the Fast Flux Test Facility. The test included 25 Swiss-fabricated sphere-pac-type fuel pins and 66 U.S. fabricated pellet-type fuel pins. The test was designed to operate at prototypical fast reactor conditions to provide a direct comparison of the irradiation performance of the two fuel types. The test design and fuel fabrication processes used for the AC-3 test are presented

  9. Performance of irradiated CVD diamond micro-strip sensors

    International Nuclear Information System (INIS)

    Adam, W.; Berdermann, E.; Bergonzo, P.; Bertuccio, G.; Bogani, F.; Borchi, E.; Brambilla, A.; Bruzzi, M.; Colledani, C.; Conway, J.; D'Angelo, P.; Dabrowski, W.; Delpierre, P.; Deneuville, A.; Dulinski, W.; Eijk, B. van; Fallou, A.; Fizzotti, F.; Foulon, F.; Friedl, M.; Gan, K.K.; Gheeraert, E.; Hallewell, G.; Han, S.; Hartjes, F.; Hrubec, J.; Husson, D.; Kagan, H.; Kania, D.; Kaplon, J.; Kass, R.; Koeth, T.; Krammer, M.; Logiudice, A.; Lu, R.; Mac Lynne, L.; Manfredotti, C.; Meier, D.; Mishina, M.; Moroni, L.; Noomen, J.; Oh, A.; Pan, L.S.; Pernicka, M.; Peitz, A.; Perera, L.; Pirollo, S.; Procario, M.; Riester, J.L.; Roe, S.; Rousseau, L.; Rudge, A.; Russ, J.; Sala, S.; Sampietro, M.; Schnetzer, S.; Sciortino, S.; Stelzer, H.; Stone, R.; Suter, B.; Tapper, R.J.; Tesarek, R.; Trischuk, W.; Tromson, D.; Vittone, E.; Walsh, A.M.; Wedenig, R.; Weilhammer, P.; Wetstein, M.; White, C.; Zeuner, W.; Zoeller, M.; Plano, R.; Somalwar, S.V.; Thomson, G.B.

    2002-01-01

    CVD diamond detectors are of interest for charged particle detection and tracking due to their high radiation tolerance. In this article, we present, for the first time, beam test results from recently manufactured CVD diamond strip detectors and their behavior under low doses of electrons from a β-source and the performance before and after intense (>10 15 /cm 2 ) proton- and pion-irradiations. We find that low dose irradiation increase the signal-to-noise ratio (pumping of the signal) and slightly deteriorate the spatial resolution. Intense irradiation with protons 2.2x10 15 p/cm 2 lowers the signal-to-noise ratio slightly. Intense irradiation with pions 2.9x10 15 π/cm 2 lowers the signal-to-noise ratio more. The spatial resolution of the diamond sensors improves after irradiations

  10. Performance of irradiated CVD diamond micro-strip sensors

    Science.gov (United States)

    Adam, W.; Berdermann, E.; Bergonzo, P.; Bertuccio, G.; Bogani, F.; Borchi, E.; Brambilla, A.; Bruzzi, M.; Colledani, C.; Conway, J.; D'Angelo, P.; Dabrowski, W.; Delpierre, P.; Deneuville, A.; Dulinski, W.; van Eijk, B.; Fallou, A.; Fizzotti, F.; Foulon, F.; Friedl, M.; Gan, K. K.; Gheeraert, E.; Hallewell, G.; Han, S.; Hartjes, F.; Hrubec, J.; Husson, D.; Kagan, H.; Kania, D.; Kaplon, J.; Kass, R.; Koeth, T.; Krammer, M.; Logiudice, A.; Lu, R.; mac Lynne, L.; Manfredotti, C.; Meier, D.; Mishina, M.; Moroni, L.; Noomen, J.; Oh, A.; Pan, L. S.; Pernicka, M.; Peitz, A.; Perera, L.; Pirollo, S.; Procario, M.; Riester, J. L.; Roe, S.; Rousseau, L.; Rudge, A.; Russ, J.; Sala, S.; Sampietro, M.; Schnetzer, S.; Sciortino, S.; Stelzer, H.; Stone, R.; Suter, B.; Tapper, R. J.; Tesarek, R.; Trischuk, W.; Tromson, D.; Vittone, E.; Walsh, A. M.; Wedenig, R.; Weilhammer, P.; Wetstein, M.; White, C.; Zeuner, W.; Zoeller, M.; Plano, R.; Somalwar, S. V.; Thomson, G. B.

    2002-01-01

    CVD diamond detectors are of interest for charged particle detection and tracking due to their high radiation tolerance. In this article, we present, for the first time, beam test results from recently manufactured CVD diamond strip detectors and their behavior under low doses of electrons from a β-source and the performance before and after intense (>10 15/cm 2) proton- and pion-irradiations. We find that low dose irradiation increase the signal-to-noise ratio (pumping of the signal) and slightly deteriorate the spatial resolution. Intense irradiation with protons 2.2×10 15 p/ cm2 lowers the signal-to-noise ratio slightly. Intense irradiation with pions 2.9×10 15 π/ cm2 lowers the signal-to-noise ratio more. The spatial resolution of the diamond sensors improves after irradiations.

  11. Design and fabrication of irradiation testing capsule for research reactor materials

    International Nuclear Information System (INIS)

    Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu

    2012-01-01

    Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed

  12. Design and fabrication of irradiation testing capsule for research reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed.

  13. Performance Analysis Review of Thorium TRISO Coated Particles during Manufacture, Irradiation and Accident Condition Heating Tests

    International Nuclear Information System (INIS)

    2015-03-01

    Thorium, in combination with high enriched uranium, was used in all early high temperature reactors (HTRs). Initially, the fuel was contained in a kernel of coated particles. However, particle quality was low in the 1960s and early 1970s. Modern, high quality, tristructural isotropic (TRISO) fuel particles with thorium oxide and uranium dioxide (UO 2 ) had been manufactured since 1978 and were successfully demonstrated in irradiation and accident tests. In 1980, HTR fuels changed to low enriched uranium UO 2 TRISO fuels. The wide ranging development and demonstration programme was successful, and it established a worldwide standard that is still valid today. During the process, results of the thorium work with high quality TRISO fuel particles had not been fully evaluated or documented. This publication collects and presents the information and demonstrates the performance of thorium TRISO fuels.This publication is an outcome of the technical contract awarded under the IAEA Coordinated Research Project on Near Term and Promising Long Term Options for Deployment of Thorium Based Nuclear Energy, initiated in 2012. It is based on the compilation and analysis of available results on thorium TRISO coated particle performance in manufacturing and during irradiation and accident condition heating tests

  14. Design and Testing for a New Thermosyphon Irradiation Vehicle

    Energy Technology Data Exchange (ETDEWEB)

    Felde, David K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McDuffee, Joel Lee [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) requires most materials and all fuel experiments to be placed in a pressure containment vessel to ensure that internal contaminants such as fission products cannot be released into the primary coolant. It also requires that all experiments be capable of withstanding various accident conditions (e.g., loss of coolant) without generating vapor bubbles on the surface of the experiment in the primary coolant. These requirements are intended to artificially increase experiment temperatures by introducing a barrier between the experimental materials and the HFIR coolant, and by reducing heat loads to the HFIR primary coolant, thus ensuring that no boiling can occur. A proposed design for materials irradiation would remove these limitations by providing the required primary containment with an internal cooling flow. This would allow for experiments to be irradiated without concern for coolant contamination (e.g., from cladding failure of advanced fuel pins) or for specimen heat load. This report describes a new materials irradiation experiment design that uses a thermosyphon cooling system to allow experimental materials direct access to a liquid coolant. The new design also increases the range of conditions that can be tested in HFIR. This design will provide a unique capability to validate the performance of current and advanced fuels and materials. Because of limited supporting data for this kind of irradiation vehicle, a test program was initiated to obtain operating data that can be used to (1) qualify the vehicle for operation in HFIR and (2) validate computer models used to perform design- and safety-basis calculations. This report also describes the test facility and experimental data, and it provides a comparison of the experimental data to computer simulations. A total of 51 tests have been completed: four tests with pure steam, 12 tests with argon, and 35 tests with helium. A total

  15. Status of fuel irradiation tests in HANARO

    International Nuclear Information System (INIS)

    Kim, Hark Rho; Lee, Choong Sung; Lee, Kye Hong; Jun, Byung Jin; Lee, Ji Bok

    1999-01-01

    Since 1996 after finishing the long-term operational test, HANARO (High-Flux Advanced Neutron Application Reactor) has been extensively used for material irradiation tests, beam application research, radioisotope production and neutron activation analysis. This paper presents the fuel irradiation test activities which are now conducted or have been finished in HANARO. KAERI developed LEU fuel using an atomization method for the research reactors. Using this LEU, we have set up and conducted three irradiation programs: (1) medium power irradiation test using a short-length mini-assembly made of 3.15 gU/cc U 3 Si, (2) high power irradiation tests using full-length test assemblies made of 3.15 gU/cc U 3 Si, and (3) irradiation test using a short-length mini-plate made of 4.8 gU/cc U 3 Si 2 . DUPIC (Direct Use of spent PWR fuels in CANDU Reactors) simulation fuel pellets, of which compositions are very similar to DUPIC pellets to keep the similarity in the thermo-mechanical property, were developed. Three mini-elements including 5 pellets each were installed in a capsule. This capsule has been irradiated for 2 months and unloaded from the HANARO core at the end of September 1999. Another very important test is the HANARO fuel qualification program at high power, which is required to resolve the licensing issue. This test is imposed on the HANARO operation license due to insufficient test data under high power environment. To resolve this licensing issue, we have been carrying out the required irradiation tests and PIE (Post-irradiation Examination) tests. Through this program, it is believed that the resolution of the licensing issue is achieved. In addition to these programs, several fuel test plans are under way. Through these vigorous activities of fuel irradiation test programs, HANARO is sure to significantly contribute to the national nuclear R and D programs. (author)

  16. Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

    International Nuclear Information System (INIS)

    Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

    1998-01-01

    This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle

  17. Mechanical performance of irradiated beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Dalle-Donne, M.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik

    1998-01-01

    For the Helium Cooled Pebble Bed (HCPB) Blanket, which is one of the two reference concepts studied within the European Fusion Technology Programme, the neutron multiplier consists of a mixed bed of about 2 and 0.1-0.2 mm diameter beryllium pebbles. Beryllium has no structural function in the blanket, however microstructural and mechanical properties are important, as they might influence the material behavior under neutron irradiation. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating it. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from these irradiation experiments, emphasizing the effects of irradiation of essential material properties and trying to elucidate the processes controlling the property changes. The microstructure, the porosity distribution, the impurity content, the behavior under compression loads and the compatibility of the beryllium pebbles with lithium orthosilicate (Li{sub 4}SiO{sub 4}) during the in-pile irradiation are presented and critically discussed. Qualitative information on ductility and creep obtained by hardness-type measurements are also supplied. (author)

  18. Experimental data report for Test TS-1 Reactivity Initiated Accident Test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Horiki, Ohichiro; Yamahara, Takeshi; Ichihashi, Yoshinori; Kikuchi, Teruo

    1992-01-01

    This report presents experimental data for Test TS-1 which was the first in a series of tests, simulating Reactivity Initiated Accident (RIA) conditions using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in October, 1989. Test fuel rod used in the Test TS-1 was a short-sized BWR (7 x 7) type rod which was fabricated from a commercial rod provided from Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79 % and burnup of 21.3 GWd/t (bundle average). Pulse irradiation was performed at a condition of stagnant water cooling, atmospheric pressure and ambient temperature using a newly developed double container-type capsule. Energy deposition of the rod in this test was evaluated to be about 61 cal/g·fuel (55 cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, fuel burnup measurements, transient behavior of the test rod during pulse irradiation and results of post pulse irradiation examinations are contained in this report. (author)

  19. Fusion Materials Irradiation Test Facility: experimental capabilities and test matrix

    International Nuclear Information System (INIS)

    Opperman, E.K.

    1982-01-01

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

  20. Storage tests with irradiated and non-irradiated onions

    International Nuclear Information System (INIS)

    Gruenewald, T.; Rumpf, G.; Troemel, I.; Bundesforschungsanstalt fuer Ernaehrung, Karlsruhe

    1978-07-01

    The results of several test series on the storage of irradiated and non-irradiated German grown onion are reported. Investigated was the influence of the irradiation conditions such as time and dose and of the storage conditions on sprouting, spoilage, browning of the vegetation centres, composition of the onions, strength and sensorial properties of seven different onion varieties. If the onions were irradiated during the dormancy period following harvest, a dose of 50 Gy (krad) was sufficient to prevent sprouting. Regarding the irradiated onions, it was not possible by variation of the storage conditions within the limits set by practical requirements to extend the dormancy period or to prevent browning of the vegetation centres, however. (orig.) 891 MG 892 RSW [de

  1. Performance of irradiated CVD diamond micro-strip sensors

    Energy Technology Data Exchange (ETDEWEB)

    Adam, W.; Berdermann, E.; Bergonzo, P.; Bertuccio, G.; Bogani, F.; Borchi, E.; Brambilla, A.; Bruzzi, M.; Colledani, C.; Conway, J.; D' Angelo, P.; Dabrowski, W.; Delpierre, P.; Deneuville, A.; Dulinski, W.; Eijk, B. van; Fallou, A.; Fizzotti, F.; Foulon, F.; Friedl, M.; Gan, K.K.; Gheeraert, E.; Hallewell, G.; Han, S.; Hartjes, F.; Hrubec, J.; Husson, D.; Kagan, H.; Kania, D.; Kaplon, J.; Kass, R.; Koeth, T.; Krammer, M.; Logiudice, A.; Lu, R.; Mac Lynne, L.; Manfredotti, C.; Meier, D. E-mail: dirk.meier@cern.ch; Mishina, M.; Moroni, L.; Noomen, J.; Oh, A.; Pan, L.S.; Pernicka, M.; Peitz, A.; Perera, L.; Pirollo, S.; Procario, M.; Riester, J.L.; Roe, S.; Rousseau, L.; Rudge, A.; Russ, J.; Sala, S.; Sampietro, M.; Schnetzer, S.; Sciortino, S.; Stelzer, H.; Stone, R.; Suter, B.; Tapper, R.J.; Tesarek, R.; Trischuk, W.; Tromson, D.; Vittone, E.; Walsh, A.M.; Wedenig, R.; Weilhammer, P.; Wetstein, M.; White, C.; Zeuner, W.; Zoeller, M.; Plano, R.; Somalwar, S.V.; Thomson, G.B

    2002-01-11

    CVD diamond detectors are of interest for charged particle detection and tracking due to their high radiation tolerance. In this article, we present, for the first time, beam test results from recently manufactured CVD diamond strip detectors and their behavior under low doses of electrons from a {beta}-source and the performance before and after intense (>10{sup 15}/cm{sup 2}) proton- and pion-irradiations. We find that low dose irradiation increase the signal-to-noise ratio (pumping of the signal) and slightly deteriorate the spatial resolution. Intense irradiation with protons 2.2x10{sup 15} p/cm{sup 2} lowers the signal-to-noise ratio slightly. Intense irradiation with pions 2.9x10{sup 15} {pi}/cm{sup 2} lowers the signal-to-noise ratio more. The spatial resolution of the diamond sensors improves after irradiations.

  2. Performance of irradiated CVD diamond micro-strip sensors

    CERN Document Server

    Adam, W; Bergonzo, P; Bertuccio, G; Bogani, F; Borchi, E; Brambilla, A; Bruzzi, Mara; Colledani, C; Conway, J; D'Angelo, P; Dabrowski, W; Delpierre, P A; Deneuville, A; Dulinski, W; van Eijk, B; Fallou, A; Fizzotti, F; Foulon, F; Friedl, M; Gan, K K; Gheeraert, E; Hallewell, G D; Han, S; Hartjes, F G; Hrubec, Josef; Husson, D; Kagan, H; Kania, D R; Kaplon, J; Kass, R; Koeth, T W; Krammer, Manfred; Lo Giudice, A; Lü, R; MacLynne, L; Manfredotti, C; Meier, D; Mishina, M; Moroni, L; Noomen, J; Oh, A; Pan, L S; Pernicka, Manfred; Peitz, A; Perera, L P; Pirollo, S; Procario, M; Riester, J L; Roe, S; Rousseau, L; Rudge, A; Russ, J; Sala, S; Sampietro, M; Schnetzer, S R; Sciortino, S; Stelzer, H; Stone, R; Suter, B; Tapper, R J; Tesarek, R J; Trischuk, W; Tromson, D; Vittone, E; Walsh, A M; Wedenig, R; Weilhammer, Peter; Wetstein, M; White, C; Zeuner, W; Zöller, M

    2002-01-01

    CVD diamond detectors are of interest for charged particle detection and tracking due to their high radiation tolerance. In this article we present, for the first time, beam test results from recently manufactured CVD diamond strip detectors and their behavior under low doses of electrons from a $\\beta$-source and the performance before and after intense ($>10^{15}/{\\rm cm^2}$) proton- and pion-irradiations. We find that low dose irradiations increase the signal-to-noise ratio (pumping of the signal) and slightly deteriorate the spatial resolution. Intense irradiations with protons ($2.2\\times 10^{15}~p/{\\rm cm^2}$) lowers the signal-to-noise ratio slightly. Intense irradiation with pions ($2.9\\times 10^{15}~\\pi/{\\rm cm^2}$) lowers the signal-to-noise ratio more. The spatial resolution of the diamond sensors improves after irradiations.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-10-01

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

  4. Performance of a proton irradiation chamber

    International Nuclear Information System (INIS)

    Agosteo, S.; Borsato, E.; Dal Corso, F.; Fazzi, A.; Gonella, F.; Introini, M.V.; Lippi, I.; Lorenzoli, M.; Modenese, L.; Montecassiano, F.; Pegoraro, M.; Pola, A.; Varoli, V.; Zotto, P.

    2012-01-01

    A Proton Irradiation Chamber aiming to perform radiation tests of electronic components was developed. The precision on the measurement of the ion currents was pushed beyond the resolution of the picoammeter by means of a series of collimators on the beam showing a linear correlation among the currents measured on them and the smaller, not measurable, current on the target. As an example of the obtained results the tests done on a Si microdosimeter and a power p-MOS are reported.

  5. Performance of a proton irradiation chamber

    Energy Technology Data Exchange (ETDEWEB)

    Agosteo, S. [INFN, Sezione di Milano, via Celoria 16, 20133 Milano (Italy); Politecnico di Milano, Dipartimento di Energia, Sezione di Ingegneria Nucleare-CeSNEF, via Ponzio 34/3 20133 Milano (Italy); Borsato, E. [INFN, Sezione di Padova, via Marzolo 8, 35131 Padova (Italy); Universita di Padova, Dipartimento di Fisica, via Marzolo 8, 35131 Padova (Italy); Dal Corso, F. [INFN, Sezione di Padova, via Marzolo 8, 35131 Padova (Italy); Fazzi, A. [INFN, Sezione di Milano, via Celoria 16, 20133 Milano (Italy); Politecnico di Milano, Dipartimento di Energia, Sezione di Ingegneria Nucleare-CeSNEF, via Ponzio 34/3 20133 Milano (Italy); Gonella, F. [INFN, Sezione di Padova, via Marzolo 8, 35131 Padova (Italy); Introini, M.V. [INFN, Sezione di Milano, via Celoria 16, 20133 Milano (Italy); Politecnico di Milano, Dipartimento di Energia, Sezione di Ingegneria Nucleare-CeSNEF, via Ponzio 34/3 20133 Milano (Italy); Lippi, I. [INFN, Sezione di Padova, via Marzolo 8, 35131 Padova (Italy); Lorenzoli, M. [INFN, Sezione di Milano, via Celoria 16, 20133 Milano (Italy); Politecnico di Milano, Dipartimento di Energia, Sezione di Ingegneria Nucleare-CeSNEF, via Ponzio 34/3 20133 Milano (Italy); Modenese, L.; Montecassiano, F.; Pegoraro, M. [INFN, Sezione di Padova, via Marzolo 8, 35131 Padova (Italy); Pola, A.; Varoli, V. [INFN, Sezione di Milano, via Celoria 16, 20133 Milano (Italy); Politecnico di Milano, Dipartimento di Energia, Sezione di Ingegneria Nucleare-CeSNEF, via Ponzio 34/3 20133 Milano (Italy); Zotto, P., E-mail: pierluigi.zotto@pd.infn.it [INFN, Sezione di Padova, via Marzolo 8, 35131 Padova (Italy); Universita di Padova, Dipartimento di Fisica, via Marzolo 8, 35131 Padova (Italy)

    2012-02-01

    A Proton Irradiation Chamber aiming to perform radiation tests of electronic components was developed. The precision on the measurement of the ion currents was pushed beyond the resolution of the picoammeter by means of a series of collimators on the beam showing a linear correlation among the currents measured on them and the smaller, not measurable, current on the target. As an example of the obtained results the tests done on a Si microdosimeter and a power p-MOS are reported.

  6. AGR 3/4 Irradiation Test Final As Run Report

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-06-01

    Several fuel and material irradiation experiments have been planned for the Idaho National Laboratory Advanced Reactor Technologies Technology Development Office Advanced Gas Reactor Fuel Development and Qualification Program (referred to as the INL ART TDO/AGR fuel program hereafter), which supports the development and qualification of tristructural-isotropic (TRISO) coated particle fuel for use in HTGRs. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development and validation of fuel performance and fission product transport models and codes, and provide irradiated fuel and materials for post irradiation examination and safety testing (INL 05/2015). AGR-3/4 combined the third and fourth in this series of planned experiments to test TRISO coated low enriched uranium (LEU) oxycarbide fuel. This combined experiment was intended to support the refinement of fission product transport models and to assess the effects of sweep gas impurities on fuel performance and fission product transport by irradiating designed-to-fail fuel particles and by measuring subsequent fission metal transport in fuel-compact matrix material and fuel-element graphite. The AGR 3/4 fuel test was successful in irradiating the fuel compacts to the burnup and fast fluence target ranges, considering the experiment was terminated short of its initial 400 EFPD target (Collin 2015). Out of the 48 AGR-3/4 compacts, 42 achieved the specified burnup of at least 6% fissions per initial heavy-metal atom (FIMA). Three capsules had a maximum fuel compact average burnup < 10% FIMA, one more than originally specified, and the maximum fuel compact average burnup was <19% FIMA for the remaining capsules, as specified. Fast neutron fluence fell in the expected range of 1.0 to 5.5×1025 n/m2 (E >0.18 MeV) for all compacts. In addition, the AGR-3/4 experiment was globally successful in keeping the

  7. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  8. Neutron Flux Characterization of Irradiation Holes for Irradiation Test at HANARO

    Directory of Open Access Journals (Sweden)

    Yang Seong Woo

    2016-01-01

    Full Text Available The High flux Advanced Neutron Application ReactOr (HANARO is a unique research reactor in the Republic of Korea, and has been used for irradiation testing since 1998. To conduct irradiation tests for nuclear materials, the irradiation holes of CT and OR5 have been used due to a high fast-neutron flux. Because the neutron flux must be accurately calculated to evaluate the neutron fluence of irradiated material, it was conducted using MCNP. The neutron flux was measured using fluence monitor wires to verify the calculated result. Some evaluations have been conducted, however, more than 20% errors have frequently occurred at the OR irradiation hole, while a good agreement between the calculated and measured data was shown at the CT irradiation hole.

  9. Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for Kijang research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Man; Tahk, Young Wook; Jeong, Yong Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); and others

    2017-08-15

    The construction project of the Kijang research reactor (KJRR), which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm{sup 3}, was selected to achieve higher fuel efficiency and performance than are possible when using U{sub 3}Si{sub 2}/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1)], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm{sup 3}), were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination) results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

  10. Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for kijang research reactor

    Directory of Open Access Journals (Sweden)

    Jong Man Park

    2017-08-01

    Full Text Available The construction project of the Kijang research reactor (KJRR, which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm3, was selected to achieve higher fuel efficiency and performance than are possible when using U3Si2/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm3, were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

  11. Experimental data report for test TS-3 Reactivity Initiated Accident test in the NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo; Sobajima, Makoto.

    1993-09-01

    This report presents experimental data for Test TS-3 which was the third test in a series of Reactivity Initiated Accident (RIA) tests using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in September, 1990. Test fuel rod used in the Test TS-3 was a short-sized BWR (7 x 7) type rod which was re-fabricated from a commercial rod irradiated in the Tsuruga Unit 1 power reactor of Japan Atomic Power Co. The fuel had an initial enrichment of 2.79 % and a burnup of 26 Gwd/tU. A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 94 ± 4 cal/g · fuel (88 ± 4 cal/g · fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and results of pre-pulse and post-pulse irradiation examinations are described in this report. (author)

  12. HFR irradiation testing of light water reactor (LWR) fuel

    International Nuclear Information System (INIS)

    Markgraf, J.F.W.

    1985-01-01

    For the materials testing reactor HFR some characteristic information with emphasis on LWR fuel rod testing capabilities and hot cell investigation is presented. Additionally a summary of LWR fuel irradiation programmes performed and forthcoming programmes are described. Project management information and a list of publications pertaining to LWR fuel rod test programmes is given

  13. Irradiation test plan of the simulated DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-11-01

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

  14. Vibration test report on the instrumented capsule for fuel irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Jeong Soo; Yoon, D. B.; Wu, J. S.; Oh, J. M.; Park, S. J.; Cho, M. S.; Kim, B. G.; Kang, Y. W

    2003-01-01

    The fluid-induced vibration level of instrumented capsule, which was manufactured for fuel irradiation test at the reactor core of HANARO, was investigated. For this purpose, the instrumented capsule was loaded at the OR site of the HANARO design verification test facility that could simulate identical flow condition as the HANARO core. Then, vibration signals of the instrumented capsule subjected to various flow conditions were measured by using vibration sensors. In time domain analysis, maximum amplitudes and RMS values of the measured acceleration and displacement signals were obtained. By using frequency domain analysis, frequency components of the fluid-induced vibration were analyzed. In addition, natural frequencies of the instrumented capsule were obtained by performing modal test. The frequency analysis results showed that the natural frequency components near 7.5Hz and 17.5Hz were dominant in the fluid-induced vibration signal. The maximum amplitude of the accelerations was measured as 12.04m/s{sup 2} that is within the allowable vibrational limit(18.99m/s{sup 2})of the reactor structure. Also, the maximum displacement amplitude was calculated as 0.191mm. Since these vibration levels are remarkably low, excessive vibration is not expected when the irradiation test of the instrumented capsule is performed at the HANARO core.

  15. Irradiation performance of uranium-molybdenum alloy dispersion fuels

    International Nuclear Information System (INIS)

    Almeida, Cirila Tacconi de

    2005-01-01

    The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm 3 were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm 3 showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

  16. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Grover, S. Blaine

    2009-01-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  17. Status on the construction of the fuel irradiation test facility

    International Nuclear Information System (INIS)

    Park, Kook Nam; Sim, Bong Shick; Lee, Chung Young; Yoo, Seong Yeon

    2005-01-01

    As a facility to examine general performance of nuclear fuel under irradiation condition in HANARO, Fuel Test Loop(FTL) has been developed which can accommodate 3 fuel pins at the core irradiation hole(IR1 hole) taking consideration user's test requirement. 3-Pin FTL consists of In-Pile Test Section (IPS) and Out-of- Pile System (OPS). Test condition in IPS such as pressure, temperature and the water quality, can be controlled by OPS. 3-Pin FTL Conceptual design was set up in 2001 and had completed detail design including a design requirement and basic Piping and Instrument Diagram (P and ID) in 2004. The safety analysis report was prepared and submitted in early 2005 to the regulatory body(KINS) for review and approval of FTL. In 2005, the development team is going to purchase and manufacture hardware and make a contract for construction work. In 2006, the development team is going to install an FTL system performance test shall be done as a part of commissioning. After a 3-Pin FTL development which is expected to be finished by the 2007, FTL will be used for the irradiation test of the new PWR-type fuel and the usage of HANARO will be enhanced

  18. Non-destructive tests of capsules for JMTR irradiation examination

    International Nuclear Information System (INIS)

    Tanaka, Hidetaka; Nagao, Yoshiharu; Sato, Masashi; Osawa, Kenji

    2007-03-01

    Irradiation examination are increasing in advanced irradiation research for accurate prediction control and evaluation of irradiation parameter such as neutron fluence, etc. by using JMTR. Irradiation capsule internals are therefore structurally complicated recently. This report described the procedure of non destructive tests such as radiographic test, penetrant test, ultrasonic test, etc. for inspection of irradiation capsules in JMTR, and the result of Test-case of confirmation procedure for internal parts of irradiation capsules. (author)

  19. Fatigue performance of copper and copper alloys before and after irradiation with fission neutrons

    International Nuclear Information System (INIS)

    Singh, B.N.; Toft, P.; Stubbins, J.F.

    1997-05-01

    The fatigue performance of pure copper of the oxygen free, high conductivity (OFHC) grade and two copper alloys (CuCrZr and CuAl-25) was investigated. Mechanical testing and microstructural analysis were carried out to establish the fatigue life of these materials in the unirradiated and irradiated states. The present report provides the first information on the ability of these copper alloys to perform under cyclic loading conditions when they have undergone significant irradiation exposure. Fatigue specimens of OFHC-Cu, CuCrZr and CuAl-25 were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of ∼2.5 x 10 17 n/m 2 s (E > 1 MeV) to fluence levels of 1.5 - 2.5 x 10 24 n/m 2 s (E > 1 MeV) at ∼47 and 100 deg. C. Specimens irradiated at 47 deg. C were fatigue tested at 22 deg. C, whereas those irradiated at 100 deg. C were tested at the irradiation temperature. The major conclusion of the present work is that although irradiation causes significant hardening of copper and copper alloys, it does not appear to be a problem for the fatigue life of these materials. In fact, the present experimental results clearly demonstrate that the fatigue performance of the irradiated CuAl-25 alloy is considerably better in the irradiated than that in the unirradiated state tested both at 22 and 100 deg. C. This improvement, however, is not so significant in the case of the irradiated OFHC-copper and CuCrZr alloy tested at 22 deg. C. These conclusions are supported by the microstructural observations and cyclic hardening experiments. (au) 4 tabs., 26 ills., 10 refs

  20. Development of an End-plug Welding Technology for an Instrumented Fuel Irradiation Test

    International Nuclear Information System (INIS)

    Kim, Soo Sung; Lee, Chul Yong; Shin, Yoon Taek; Choo, Kee Nam

    2010-01-01

    The irradiation test of end-plug specimens was planned for the evaluation of nuclear fuels performance. To establish the fabrication process, and for satisfying the requirements of the irradiation test, an orbital-GTA weld machine for the specimens of the dual rods was developed, and the preliminary welding experiments for optimizing the process conditions of the specimens of the dual rods were performed. Dual rods with a 9.5mm diameter and a 0.6mm wall thickness of the cladding tubes and end-plugs have been used and the optimum conditions of the pin-hole welding have also been selected. This paper describes the experimental results of the GTA welds of the specimens of the dual rods and the metallography examinations of the GTA welded specimens for various welding conditions for the instrumented fuel irradiation test. These investigations satisfied the requirements of the instrumented irradiation test and the GTA welds for the specimens of the dual rods at the HANARO research reactor

  1. Improvement of irradiation facilities performance in JMTR

    International Nuclear Information System (INIS)

    Kanno, Masaru; Sakurai, Susumu; Honma, Kenzo; Sagawa, Hisashi; Nakazaki, Chousaburo

    1999-01-01

    Various kinds of irradiation facilities are installed in the JMTR for the purpose of irradiation tests on fuels and materials and of producing radioisotopes. The irradiation facilities have been improved so far at every opportunity of new irradiation requirements and of renewing them which reached the design lifetime. Of these irradiation facilities, improvements of the power ramping test facility (BOCA/OSF-1 facility) and the hydraulic rabbit No.2 (HR-2 facility) are described here. (author)

  2. AGR-5/6/7 Irradiation Test Predictions using PARFUME

    Energy Technology Data Exchange (ETDEWEB)

    Skerjanc, William F. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-09-14

    PARFUME, (PARticle FUel ModEl) a fuel performance modeling code used for high temperature gas-cooled reactors (HTGRs), was used to model the Advanced Gas Reactor (AGR)-5/6/7 irradiation test using predicted physics and thermal hydraulics data. The AGR-5/6/7 test consists of the combined fifth, sixth, and seventh planned irradiations of the AGR Fuel Development and Qualification Program. The AGR-5/6/7 test train is a multi-capsule, instrumented experiment that is designed for irradiation in the 133.4-mm diameter north east flux trap (NEFT) position of Advanced Test Reactor (ATR). Each capsule contains compacts filled with uranium oxycarbide (UCO) unaltered fuel particles. This report documents the calculations performed to predict the failure probability of tristructural isotropic (TRISO)-coated fuel particles during the AGR-5/6/7 experiment. In addition, this report documents the calculated source term from the driver fuel. The calculations include modeling of the AGR-5/6/7 irradiation that is scheduled to occur from October 2017 to April 2021 over a total of 13 ATR cycles, including nine normal cycles and four Power Axial Locator Mechanism (PALM) cycle for a total between 500 – 550 effective full power days (EFPD). The irradiation conditions and material properties of the AGR-5/6/7 test predicted zero fuel particle failures in Capsules 1, 2, and 4. Fuel particle failures were predicted in Capsule 3 due to internal particle pressure. These failures were predicted in the highest temperature compacts. Capsule 5 fuel particle failures were due to inner pyrolytic carbon (IPyC) cracking causing localized stresses concentrations in the SiC layer. This capsule predicted the highest particle failures due to the lower irradiation temperature. In addition, shrinkage of the buffer and IPyC layer during irradiation resulted in formation of a buffer-IPyC gap. The two capsules at the two ends of the test train, Capsules 1 and 5 experienced the smallest buffer-IPyC gap

  3. Design of water feeding system for IASCC irradiation tests at JMTR

    International Nuclear Information System (INIS)

    Kanno, Masaru; Nabeya, Hideaki; Mori, Yuichiro

    2001-12-01

    In relation to the aging of light water reactors (LWRs), the irradiation assisted stress corrosion cracking (IASCC) has been regarded as a significant and urgent issue for the reliability of in-core components and materials of LWRs, and the irradiation research is now under schedule. It is essential for IASCC studies to irradiated materials under well-controlled conditions simulating LWR in-core environment. Therefore, a new water feeding system to supply high temperature water into irradiation capsules in the Japan Materials Testing Reactor (JMTR) has been designed and will be installed in near future. This report describes the specification and performance of the water feeding system that is designed to supply high temperature water to simulate BWR conditions in irradiation capsules. This design work was performed in the fiscal year 1999. (author)

  4. Present status of irradiation tests on tritium breeding blanket for fusion reactor

    International Nuclear Information System (INIS)

    Futamura, Yoshiaki; Sagawa, Hisashi; Shimakawa, Satoshi; Tsuchiya, Kunihiko; Kuroda, Toshimasa; Kawamura, Hiroshi.

    1994-01-01

    To develop a tritium breeding blanket for a fusion reactor, irradiation tests in fission reactors are indispensable for obtaining data on irradiation effects on materials, and neutronics/thermal characteristics and tritium production/recovery performance of the blanket. Various irradiation tests have been conducted in the world, especially to investigate tritium release characteristics from tritium breeding and neutron multiplier materials, and materials integrity under irradiation. In Japan, VOM experiments at JRR-2 for ceramic breeders and experiments at JMTR for ceramic breeders and beryllium as a neutron multiplier have been performed. Several universities have also investigated ceramic breeders. In the EC, the EXOTIC experiments at HFR in the Netherlands and the SIBELIUS, the LILA, the LISA and the MOZART experiments for ceramic breeders have carried out. In Canada, NRU has been used for the CRITIC experiments. The TRIO experiments at ORR(ORNL), experiments at RTNS-II, FUBR and ATR have been conducted in the USA. The last two are experiments with high neutron fluence aiming at investigating materials integrity under irradiation. The BEATRIX-I and -II experiments have proceeded under international collaboration of Japan, Canada, the EC and the USA. This report shows the present status of these irradiation tests following a review of the blanket design in the ITER CDA(Conceptual Design Activity). (author)

  5. PIE Report on the KOMO-3 Irradiation Test Fuels

    International Nuclear Information System (INIS)

    Park, Jong Man; Ryu, H. J.; Yang, J. H.

    2009-04-01

    In the KOMO-3, in-reactor irradiation test had been performed for 12 kinds of dispersed U-Mo fuel rods, a multi wire fuel rod and a tube fuel rod. In this report we described the PIE results on the KOMO-3 irradiation test fuels. The interaction layer thickness between fuel particle and matrix could be reduced by using a large size U-Mo fuel particle or introducing Al-Si matrix or adding the third element in the U-Mo particle. Monolithic fuel rod of multi-wire or tube fuel was also effective in reducing the interaction layer thickness

  6. Fuel performance analysis for the HAMP-1 mini plate test

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Byoung Jin; Tahka, Y. W.; Yim, J. S.; Lee, B. H. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    U-7wt%Mo/Al- 5wt%Si dispersion fuel with 8gU/cm{sup 3} is chosen to achieve more efficiency and higher performance than the conventional U{sub 3}Si{sub 2} fuel. As part of the fuel qualification program for the KiJang research reactor (KJRR), three irradiation tests with mini-plates are on the way at the High-flux Advanced Neutron Application Reactor (HANARO). The first test among three HANARO Mini-Plate Irradiation tests (HAMP-1, 2, 3) has completed. PLATE code has been initially developed to analyze the thermal performance of high density U-Mo/Al dispersion fuel plates during irradiation [1]. We upgraded the PLATE code with the latest irradiation results which were implemented by corrosion, thermal conductivity and swelling model. Fuel performance analysis for HAMP-1 was conducted with updated PLATE. This paper presents results of performance evaluation of the HAMP-1. Maximum fuel temperature was obtained 136 .deg., which is far below the preset limit of 200 .deg. for the irradiation test. The meat swelling and corrosion thickness was also confirmed that the developed fuel would behave as anticipated.

  7. Mechanical compression tests of beryllium pebbles after neutron irradiation up to 3000 appm helium production

    Energy Technology Data Exchange (ETDEWEB)

    Chakin, V., E-mail: vladimir.chakin@kit.edu [Karlsruhe Institute of Technology, Institite for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rolli, R.; Moeslang, A. [Karlsruhe Institute of Technology, Institite for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

    2015-04-15

    Highlights: • Compression tests of highly neutron irradiated beryllium pebbles have been performed. • Irradiation hardening of beryllium pebbles decreases the steady-state strain-rates. • The steady-state strain-rates of irradiated beryllium pebbles exceed their swelling rates. - Abstract: Results: of mechanical compression tests of irradiated and non-irradiated beryllium pebbles with diameters of 1 and 2 mm are presented. The neutron irradiation was performed in the HFR in Petten, The Netherlands at 686–968 K up to 1890–2950 appm helium production. The irradiation at 686 and 753 K cause irradiation hardening due to the gas bubble formation in beryllium. The irradiation-induced hardening leads to decrease of steady-state strain-rates of irradiated beryllium pebbles compared to non-irradiated ones. In contrary, after irradiation at higher temperatures of 861 and 968 K, the steady-state strain-rates of the pebbles increase because annealing of irradiation defects and softening of the material take place. It was shown that the steady-state strain-rates of irradiated beryllium pebbles always exceed their swelling rates.

  8. NRI experimental facility for the testing of irradiation assisted stress corrosion cracking

    International Nuclear Information System (INIS)

    Ruscak, M.; Chvatal, P.; Zamboch, M.

    1998-01-01

    IASCC influencing reactor internals of both BWR and PWR reactors is a complex phenomenon covering influences of material structure, neutron fluence, neutron flux, chemistry of environment, gamma radiation and mechanical stress. To evaluate such degradation, tests should be performed under conditions similar to those in real structure. Nuclear Research Institute has built several experimental facilities in order to be able to test IASCC degradation of materials. Basically, reactor water loops, both PWR and BWR, could be used to model environmental conditions including gamma and neutron irradiation. Pre-irradiation can be done in irradiation channels under well controlled temperature conditions. During the experiment, in-pile conditions can be compared with those out of pile. It enables to clarify pure influence of irradiation. For testing of irradiated specimens, hot cell facility has been developed for slow strain rate tests. The paper will show all above mentioned facilities as well as some of the results observed with them. (author)

  9. Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

    Energy Technology Data Exchange (ETDEWEB)

    Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.; Holdaway, K. K.; Housley, G. K.; Rabin, B. H.

    2016-10-01

    The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, and other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-10-01

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

  11. Education and training by utilizing irradiation test reactor simulator

    International Nuclear Information System (INIS)

    Eguchi, Shohei; Koike, Sumio; Takemoto, Noriyuki; Tanimoto, Masataka; Kusunoki, Tsuyoshi

    2016-01-01

    The Japan Atomic Energy Agency, at its Japan Materials Testing Reactor (JMTR), completed an irradiation test reactor simulator in May 2012. This simulator simulates the operation, irradiation test, abnormal transient change during operation, and accident progress events, etc., and is able to perform operation training on reactor and irradiation equipment corresponding to the above simulations. This simulator is composed of a reactor control panel, process control panel, irradiation equipment control panel, instructor control panel, large display panel, and compute server. The completed simulator has been utilized in the education and training of JMTR operators for the purpose of the safe and stable operation of JMTR and the achievement of high operation rate after resuming operation. For the education and training, an education and training curriculum has been prepared for use in not only operation procedures at the time of normal operation, but also learning of fast and accurate response in case of accident events. In addition, this simulator is also being used in operation training for the purpose of contributing to the cultivation of human resources for atomic power in and out of Japan. (A.O.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-01-15

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

  13. The Defect Inspection on the Irradiated Fuel Rod by Eddy Current Test

    International Nuclear Information System (INIS)

    Koo, D. S.; Park, Y. K.; Kim, E. K.

    1996-01-01

    The eddy current test(ECT) probe of differential encircling coil type was designed and fabricated, and the optimum condition of ECT was derived for the examination of the irradiated fuel rod. The correlation between ECT test frequency and phase and amplitude was derived by performing the test of the standard rig that includes inner notches, outer notches and through-holes. The defect of through-hole was predicted by ECT at the G33-N2 fuel rod irradiated in the Kori-1 nuclear power reactor. The metallographic examination on the G33-N2 fuel rod was Performed at the defect location predicted by ECT. The result of metallographic examination for the G33-N2 fuel rod was in good agreement with that of ECT. This proves that the evaluation for integrity of irradiated fuel rod by ECT is reliable

  14. RECH-1 test fuel irradiation status report

    International Nuclear Information System (INIS)

    Marin, J.; Lisboa, J.; Olivares, L.; Chavez, J.

    2005-01-01

    Since May 2003, one RECH-1 fuel element has been submitted to irradiation at the HFR-Petten, Holland. By November 2004 the irradiation has achieved its pursued goal of 55% burn up. This irradiation qualification service will finish in the year 2005 with PIE tests, as established in a contractual agreement between the IAEA, NRG, and CCHEN. This report presents the objectives and the current results of this fuel qualification under irradiation. Besides, a brief description of CHI/4/021, IAEA's Technical Cooperation Project that has supported this irradiation test, is also presented here. (author)

  15. Disk-bend ductility tests for irradiated materials

    International Nuclear Information System (INIS)

    Klueh, R.L.; Braski, D.N.

    1984-01-01

    We modified the HEDL disk-bend test machine and are using it to qualitatively screen alloys that are susceptible to embrittlement caused by irradiation. Tests designed to understand the disk-bend test in relation to a uniaxial test are discussed. Selected results of tests of neutron-irradiated material are also presented

  16. A method to evaluate fission gas release during irradiation testing of spherical fuel - HTR2008-58184

    International Nuclear Information System (INIS)

    Van Der Merwet, H.; Venter, J.

    2008-01-01

    The evaluation of fission gas release from spherical fuel during irradiation testing is critical to understand expected fuel performance under real reactor conditions. Online measurements of Krypton and Xenon fission products explain coated particle performance and contributions from graphitic matrix materials used in fuel manufacture and irradiation rig materials. Methods that are being developed to accurately evaluate fission gas release are described here together with examples of evaluations performed on irradiation tests HFR-K5, -K6 and EU1bis. (authors)

  17. Design and fuel fabrication processes for the AC-3 mixed-carbide irradiation test

    International Nuclear Information System (INIS)

    Latimer, T.W.; Chidester, K.M.; Stratton, R.W.; Ledergerber, G.; Ingold, F.

    1992-01-01

    The AC-3 test was a cooperative U.S./Swiss irradiation test of 91 wire-wrapped helium-bonded U-20% Pu carbide fuel pins irradiated to 8.3 at % peak burnup in the Fast Flux Test Facility. The test consisted of 25 pins that contained spherepac fuel fabricated by the Paul Scherrer Institute (PSI) and 66 pins that contained pelletized fuel fabricated by the Los Alamos National Laboratory. Design of AC-3 by LANL and PSI was begun in 1981, the fuel pins were fabricated from 1983 to 1985, and the test was irradiated from 1986 to 1988. The principal objective of the AC-3 test was to compare the irradiation performance of mixed-carbide fuel pins that contained either pelletized or sphere-pac fuel at prototypic fluence and burnup levels for a fast breeder reactor

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-15

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

  19. Status of irradiation testing and PIE of MOX (Pu-containing) fuel

    International Nuclear Information System (INIS)

    Dimayuga, F.C.; Zhou, Y.N.; Ryz, M.A.

    1995-01-01

    This paper describes AECL's mixed oxide (MOX) fuel-irradiation and post-irradiation examination (PIE) program. Post-irradiation examination results of two major irradiation experiments involving several (U, Pu)O 2 fuel bundles are highlighted. One experiment involved bundles irradiated to burnups ranging fro 400 to 1200 MWh/kgHe in the Nuclear Power Demonstration (NPD) reactor. The other experiment consisted of several (U, Pu)O 2 bundles irradiated to burnups of up to 500 Mwh/kgHe in the National Research Universal (NRU) reactor. Results of these experiments demonstrate the excellent performance of CANDU MOX fuel. This paper also outlines the status of current MOX fuel irradiation tests, including the irradiation of various (U, Pu)O 2 bundles. The strategic importance of MOX fuel to CANDU fuel-cycle flexibility is discussed. (author)

  20. Development of a Device for a Material Irradiation Test in the OR Test Hole

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-05-15

    To develop a technology and a device for the irradiation test for utilization of the OR/IP holes according to the various requirements of users, the properties of the OR/IP holes were investigated and an irradiation device for the OR hole was designed and fabricated. The OR-4, 5 and the IP-9, 10, 11 holes were selected as those suitable to irradiation tests among the test holes located in the out core area. The conceptual design was performed to design a device to irradiate materials using the OR and IP holes. The capsule for the OR holes is fixed by pressing the protection tube using a clamping device, on the other hand the IP capsule is inserted in the hole without a special clamping device. In the basic design of the irradiation device for the OR hole, the capsules having the outside diameter of 50, 52, 54, 56mm were reviewed theoretically to investigate if they meet the hydraulic and vibration conditions required in the HANARO. The results of the pressure drop test showed that the 3 kinds of capsules having diameter of 52, 54, 56mm satisfied the requirement for the pressure difference and flow rate in HANARO. The capsule of {phi}56mm out of the above three satisfied the vibration condition and was finally selected giving consideration of a capacity of specimens. The capsule having a diameter of {phi}56mm was fabricated and the flow rate was measured. Using the velocity data measured at the out-core facility, the heat transfer coefficient, and the temperature on the surface of the capsule was evaluated to confirm it less than the ONB temperature. As a result, the capsule of {phi}56mm was selected for the irradiation test at the OR holes.

  1. Investigation of neutron fluence using fluence monitors for irradiation test at WWR-K

    International Nuclear Information System (INIS)

    Romanova, N.K.; Takemoto, N.

    2013-01-01

    Irradiation test of a Si ingot is planned using WWR-K in Institute of Nuclear Physics Republic of Kazakhstan (INP RK) to develop an irradiation technology for Si semiconductor production by Neutron Transmutation Doping (NTD) method in the framework of an international cooperation between INP RK and Japan Atomic Energy Agency (JAEA), Japan. It is possible to irradiate the Si ingot of 6 inch in diameter at the K-23 irradiation channel in the WWR-K. The preliminary irradiation test using 4 Al ingots was performed to evaluate the actual neutronic irradiation field at the K-23 channel in the WWR-K. Each Al ingot has the same dimension as the Si ingot, and 15 fluence monitors are equipped in it. Iron wire and aluminum-cobalt wire are inserted into them, and it is possible to evaluate both fast and thermal neutron fluxes by measurement of these radiation activities after irradiation. This report described the results of the preliminary irradiation test and the neutronic calculations by Monte Carlo method in order to evaluate the neutronic irradiation field in the irradiation position for the silicon ingot at the channel in the WWR-K. (authors)

  2. Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

    International Nuclear Information System (INIS)

    Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio

    2002-10-01

    An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)

  3. Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2002-10-01

    An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)

  4. AGR-2 irradiation test final as-run report, Rev. 1

    International Nuclear Information System (INIS)

    2014-01-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO 2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities; (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing; and, (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO 2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO 2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The

  5. AGR-2 Irradiation Test Final As-Run Report, Rev 2

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a

  6. Design and manufacturing of non-instrumented capsule for advanced PWR fuel pellet irradiation test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Lee, C. B.; Song, K. W. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    This project is preparing to irradiation test of the developed large grain UO{sub 2} fuel pellet in HANARO for pursuit fuel safety and high burn-up in 'Advanced LWR Fuel Technology Development Project' as a part Nuclear Mid and Long-term R and D Program. On the basis test rod is performed the nuclei property and preliminary fuel performance analysis, test rod and non-instrumented capsule are designed and manufactured for irradiation test in HANARO. This non-instrumented irradiation capsule of Advanced PWR Fuel pellet was referred the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO(DUPIC Rig-001) and 18-element HANARO fuel, was designed to ensure the integrity and the endurance of non-instrumented capsule during the long term(2.5 years) irradiation. To irradiate the UO{sub 2} pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. This non-instrumented irradiation capsule will be based to develope the non-instrumented capsule for the more long term irradiation in HANARO. 22 refs., 13 figs., 5 tabs. (Author)

  7. Inert medium (helium) irradiation testing of pressure tube samples

    International Nuclear Information System (INIS)

    Ancuta, M.; Radu, V.; Stefan, V.; Preda, M.

    2001-01-01

    Irradiation tests currently performed in C-5 capsule aim at obtaining data and information concerning behavior to irradiation of pressure tubes of CANDU type fuel channel, to evidence the factors limiting operation life span. A calculation code for analysis and prediction of pressure tube behavior should be based upon periodical inspection results, post irradiation examination of the removed from reactor pressure tubes as well as on the experimental results obtained with materials subjected to irradiation conditions identical with the operational ones. Mechanical behavior analysis should focus both complex thermal-mechanical type stresses and mechanical properties alteration under irradiation. The experimental results should be applied: - to evaluate the irradiation effects upon mechanical properties of Zr-2.5% Nb exposed to fluences up to 10 21 n·cm -2 ; - to gather data concerning the real stress / real deformation characteristic from which characteristic quantities can be deduced as, for instance, elasticity modulus, plasticity modulus, exponent of stress term in the Tsu-Berteles relation, to be used within the CANTUP simulation code describing pressure tube behavior, currently developed at INR Pitesti; - to develop prediction methods of pressure tube behavior and merging with in-service inspection procedure in order to forecast the life span and the proper timing for replacement before major failures occur. The samples irradiated in C-5 capsule were extracted from the ends of Zr-2.5% Nb pressure tubes resulting from Cernavoda NPP Unit 1. The samples for tensile tests were extracted on longitudinal and transversal directions of the pressure tube. The tests were carried out under following conditions: - test environment temperature, 260 - 280 deg.C; - testing medium, helium at 1 - 6 b pressure; - neutron flux (E n > 1 MeV), 1 - 2 · 10 13 ncm -2 s -1 ; - neutron fluence (E n > 1 MeV), 4 · 10 20 ncm -2 . The following characteristics were obtained from tensile

  8. The Design and Manufacturing Report of Plug Type Non-Instrumented Rig for Irradiation Test in HANARO OR Hole

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Bang, Je Geon; Lim, Ik Sung; Kim, Sun Ki; Yang, Yong Sik; Song, Kun Woo

    2008-09-15

    This project is developed the plug type non-instrumented irradiation test rig of the advanced nuclear fuel in HANARO for pursuit advanced performance in High Performance Fuel Technology Development as a part Nuclear Mid and Long-term R and D Program. This irradiation rig was confirmed the integrity and HANARO core compatibility by the optimum design and the thermal hydraulic out-pile test in FIVPET. The characteristic of plug type non-instrument rig is to possible irradiation test of variable in-pile condition and reduced the wastes for reusable as function. This plug type non-instrumented rig was satisfied the quality assurance requirements and written out the end of manufacturing report. This plug type non-instrumented rig is adopt to the irradiation test for nuclear fuel irradiation test in HANARO OR hole.

  9. Gamma irradiation testing of prototype ITER in-vessel magnetic pick-up coils

    International Nuclear Information System (INIS)

    Vermeeren, Ludo; Leysen, Willem

    2013-01-01

    Highlights: ► We tested five prototype ITER in-vessel coils up to a gamma dose of 72 MGy. ► Before and after irradiation thermal tests were also performed from 30 °C till 130 °C. ► The continuity resistances and the insulation resistances were continuously monitored. ► The observed behavior of all coils was satisfactory in all conditions. ► For the further design the mechanical robustness should be taken into account. -- Abstract: To fulfill the requirements for ITER in-vessel magnetic diagnostics, several coil prototypes have been developed, aiming at minimizing the disturbing effects of temperature gradients and radiation induced phenomena. As a first step in the radiation resistance testing of these prototypes, an in-situ high dose rate gamma radiation test on a selection of prototypes was performed. The aim of this test was to get a first experimental feedback regarding the behavior of the pick-up coil prototypes under radiation. Five prototypes (a coil wound with glass-insulated copper wire, two LTCC coils and two HTCC coils) were irradiated at a dose rate of 46 kGy/h up to a total dose of 72 MGy and at a temperature of 50 °C. During the irradiation, the continuity resistances and the insulation resistances were continuously measured. Before and after irradiation reference data were recorded as a function of temperature (from 30 °C to 130 °C). This paper includes the results of the temperature and irradiation tests and a discussion of the behavior of the prototype coils in terms of electrical and mechanical properties

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

    International Nuclear Information System (INIS)

    1997-03-01

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

  11. Genotoxicity test of irradiated foods

    International Nuclear Information System (INIS)

    Tanaka, Noriho

    2004-01-01

    Safety tests of radiation irradiated foods started as early as from 1967 in Japan and genotoxicity tests in the Hatano Res. Inst., from 1977. The latter is unique in the world and is reviewed in this paper. Tests included those for the initial injury of DNA, mutagenicity, chromosomal aberration and transformation with use of bacteria, cultured mammalian cells and animals (for chromosomal aberration, micronucleus formation and dominant lethality). Foods tested hitherto were onion, rice, wheat and flour, Vienna sausage, fish sausage (kamaboko), mandarian orange, potato, black pepper and red capsicum, of which extract or powder was subjected to the test. Irradiation doses and its purposes were 0.15-6 kGy γ-ray ( 60 Co) or electron beam by the accelerator (only for the orange), and suppression of germination, pesticide action or sterilization, respectively. Genotoxicity of all foods under tested conditions is shown negative. (N.I.)

  12. IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

    Directory of Open Access Journals (Sweden)

    M.K. MEYER

    2014-04-01

    Full Text Available High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  13. Irradiation performance of U-Mo monolithic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, M. K.; Gan, J.; Jue, J. F.; Keiser, D. D.; Perez, E.; Robinson, A.; Wachs, D. M.; Woolstenhulme, N. [Idaho National Laboratory, Idaho (Korea, Republic of); Kim, Y.S.; Hofman, G. L. [Argonne National Laboratory, Lemont (United States)

    2014-04-15

    High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  14. Safety analysis report of the irradiation test of Type-B bundle

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Choong Sung; Lim, I. C.; Lee, B. C.; Ryu, J. S.; Kim, H. R

    2000-06-01

    The HANARO fuel, U{sub 3}Si-A1, has been developed by AECL and tested in NRU reactor. In the course of the fuel qualification tests, only one case was performed under the higher power condition than maximum linear power which was expected in the design stage. The Korea regulatory body, KINS imposed that HANARO shall be operated at the power level less than 24MW which is 80% of the design full power until HANARO shows the repetitive performance of the fuel at the power condition abov e 112.8KW/m. To resolve this imposition, KAERI designed two types of special test bundles: two non-instrumented(Type-A) and one instrumented(Type-B) test bundles. Two Type-A bundles were irradiated in HANARO: one of them has finished PIE and the other is under PIE. Type-B bundle was loaded in the core during 1.32 day at 1996, but outstanding FIV(flow induced vibration) was observed at the pool top because of long guide tube attached to the top of the bundle. The successful installation of the chimney fastener to fix the guide tube resulted in conducting the irradiation test of Type-B bundle again. The test will start at mid- July, 2000. In order to safely do the Type-B irradiation test, the safety analysis for the nuclear, mechanical and thermal-hydraulic aspects was performed. The reactivity worth and the maximum 1 near power predicted by VENTURE are 6.3mk/k and 121.6kW/m, respectively. Thermal margins for normal and transient conditions using MATRA-h, are assessed to satisfy the safety criteria.

  15. Insulation interlaminar shear strength testing with compression and irradiation

    International Nuclear Information System (INIS)

    McManamy, T.J.; Brasier, J.E.; Snook, P.

    1989-01-01

    The Compact Ignition Tokamak (CIT) project identified the need for research and development for the insulation to be used in the toroidal field coils. The requirements included tolerance to a combination of high compression and shear and a high radiation dose. Samples of laminate-type sheet material were obtained from commercial vendors. The materials included various combinations of epoxy, polyimide, E-glass, S-glass, and T-glass. The T-glass was in the form of a three-dimensional weave. The first tests were with 50 x 25 x 1 mm samples. These materials were loaded in compression and then to failure in shear. At 345-MPa compression, the interlaminar shear strength was generally in the range of 110 to 140 MPa for the different materials. A smaller sample configuration was developed for irradiation testing. The data before irradiation were similar to those for the larger samples but approximately 10% lower. Limited fatigue testing was also performed by cycling the shear load. No reduction in shear strength was found after 50,000 cycles at 90% of the failure stress. Because of space limitations, only three materials were chosen for irradiation: two polyimide systems and one epoxy system. All used boron-free glass. The small shear/compression samples and some flexure specimens were irradiated to 4 x 10 9 and 2 x 10 10 rad in the Advanced Technology Reactor at Idaho National Engineering Laboratory. A lead shield was used to ensure that the majority of the dose was from neutrons. The shear strength with compression before and after irradiation at the lower dose was determined. Flexure strength and the results from irradiation at the higher dose level will be available in the near future. 7 refs., 7 figs., 2 tabs

  16. Needs of in-situ materials testing under neutron irradiation

    International Nuclear Information System (INIS)

    Noda, K.; Hishinuma, A.; Kiuchi, K.

    1989-01-01

    Under neutron irradiation, the component atoms of materials are displaced as primary knock-on atoms, and the energy of the primary knock-on atoms is consumed by electron excitation and nuclear collision. Elementary irradiation defects accumulate to form damage structure including voids and bubbles. In situ test under neutron irradiation is necessary for investigating into the effect of irradiation on creep behavior, the electric properties of ceramics, transport phenomena and so on. The in situ test is also important to investigate into the phenomena related to the chemical reaction with environment during irradiation. Accelerator type high energy neutron sources are preferable to fission reactors. In this paper, the needs and the research items of in situ test under neutron irradiation using a D-Li stripping type high energy neutron source on metallic and ceramic materials are described. Creep behavior is one of the most important mechanical properties, and depends strongly on irradiation environment, also it is closely related to microstructure. Irradiation affects the electric conductibity of ceramics and also their creep behavior. In this way, in situ test is necessary. (K.I.)

  17. MAFF sponsored research: detection tests for irradiated food

    International Nuclear Information System (INIS)

    Blackburn, C.M.; Holley, P.A.; Pryke, D.C.

    1993-01-01

    In their 1986 report on the safety and wholesomeness of irradiated food the UK Advisory Committee on Irradiated and Novel Foods (ACINF) recognised that a generally applicable test to determine if a food had been irradiated was not available. The committee considered that, although not a pre-requisite, the existence of a detection test would be a useful supplement to a control system and do much to reassure consumers; with this in mind ACINF recommended that detection methods should be kept under review. As a consequence, in 1987 the Ministry initiated a comprehensive R and D detection test programme. Over fifty papers have been published to date as a result of this programme. MAFF (Ministry Of Agriculture Fisheries and Food) has also been involved in other research associated with irradiation and food safety, some of which is described in this paper. This paper aims to give an overview of recent work funded under the food irradiation programme. Twelve projects have been supported over the last two years, ten of which involved the development of detection tests for irradiated food. A summary of these projects is presented: - Thermoluminescence; - Electron Spin Resonance; - 2-alkylcyclobutanones; -Determination Of Hydrogen; - Differential Scanning Calorimetry; - Limulus Amoebocyte Lysate; - DNA; - Pesticide Breakdown; - Neutron Irradiation; -Future Plans. (orig./vhe)

  18. Transcriptome profiling of mice testes following low dose irradiation

    DEFF Research Database (Denmark)

    Belling, Kirstine C.; Tanaka, Masami; Dalgaard, Marlene Danner

    2013-01-01

    ABSTRACT: BACKGROUND: Radiotherapy is used routinely to treat testicular cancer. Testicular cells vary in radio-sensitivity and the aim of this study was to investigate cellular and molecular changes caused by low dose irradiation of mice testis and to identify transcripts from different cell types...... in the adult testis. METHODS: Transcriptome profiling was performed on total RNA from testes sampled at various time points (n = 17) after 1 Gy of irradiation. Transcripts displaying large overall expression changes during the time series, but small expression changes between neighbouring time points were...... selected for further analysis. These transcripts were separated into clusters and their cellular origin was determined. Immunohistochemistry and in silico quantification was further used to study cellular changes post-irradiation (pi). RESULTS: We identified a subset of transcripts (n = 988) where changes...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-08-15

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

  20. Design principles and field performance of a solar spectral irradiance meter

    Energy Technology Data Exchange (ETDEWEB)

    Tatsiankou, V.; Hinzer, K.; Haysom, J.; Schriemer, H.; Emery, K.; Beal, R.

    2016-08-01

    A solar spectral irradiance meter (SSIM), designed for measuring the direct normal irradiance (DNI) in six wavelength bands, has been combined with models to determine key atmospheric transmittances and the resulting spectral irradiance distribution of DNI under all sky conditions. The design principles of the SSIM, implementation of a parameterized transmittance model, and field performance comparisons of modeled solar spectra with reference radiometer measurements are presented. Two SSIMs were tested and calibrated at the National Renewable Energy Laboratory (NREL) against four spectroradiometers and an absolute cavity radiometer. The SSIMs' DNI was on average within 1% of the DNI values reported by one of NREL's primary absolute cavity radiometers. An additional SSIM was installed at the SUNLAB Outdoor Test Facility in September 2014, with ongoing collection of environmental and spectral data. The SSIM's performance in Ottawa was compared against a commercial pyrheliometer and a spectroradiometer over an eight month study. The difference in integrated daily spectral irradiance between the SSIM and the ASD spectroradiometer was found to be less than 1%. The cumulative energy density collected by the SSIM over this duration agreed with that measured by an Eppley model NIP pyrheliometer to within 0.5%. No degradation was observed.

  1. HFR irradiation testing of fusion materials

    International Nuclear Information System (INIS)

    Conrad, R.; von der Hardt, P.; Loelgen, R.; Scheurer, H.; Zeisser, P.

    1984-01-01

    The present and future role of the High Flux Reactor Petten for fusion materials testing has been assessed. For practical purposes the Tokamak-based fusion reactor is chosen as a point of departure to identify material problems and materials data needs. The identification is largely based on the INTOR and NET design studies, the reported programme strategies of Japan, the U.S.A. and the European Communities for technical development of thermonuclear fusion reactors and on interviews with several experts. Existing and planned irradiation facilities, their capabilities and limitations concerning materials testing have been surveyed and discussed. It is concluded that fission reactors can supply important contributions for fusion materials testing. From the point of view of future availability of fission testing reactors and their performance it appears that the HFR is a useful tool for materials testing for a large variety of materials. Prospects and recommendations for future developments are given

  2. Irradiation Testing of TRISO-Coated Particle Fuel in Korea

    International Nuclear Information System (INIS)

    Kim, Bong Goo; Yeo, Sunghwan; Jeong, Kyung-Chai; Eom, Sung-Ho; Kim, Yeon-Ku; Kim, Woong Ki; Lee, Young Woo; Cho, Moon Sung; Kim, Yong Wan

    2014-01-01

    In Korea, coated particle fuel is being developed to support development of a VHTR. At the end of March 2014, the first irradiation test in HANARO at KAERI to demonstrate and qualify TRISO-coated particle fuel for use in a VHTR was terminated. This experiment was conducted in an inert gas atmosphere without on-line temperature monitoring and control, or on-line fission product monitoring of the sweep gas. The irradiation device contained two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The duration of irradiation testing at HANARO was about 135 full power days from last August 2013. The maximum average power per particle was about 165 mW/particle. The calculated peak burnup of the TRISO-coated fuel was a little less than 4 atom percent. Post-irradiation examination is being carried out at KAERI’s Irradiated Material Examination Facility beginning in September of 2014. This paper describes characteristics of coated particle fuel, the design of the test rod and irradiation device for this coated particle fuel, and discusses the technical results of irradiation testing at HANARO. (author)

  3. Testing of Performance of Optical Fibers Under Irradiation in Intense Radiation Fields, When Subjected to High Temperature

    International Nuclear Information System (INIS)

    Blue, Thomas; Windl, Wolfgang; Dickerson, Bryan

    2013-01-01

    The primary objective of this project is to measure and model the performance of optical fibers in intense radiation fields when subjected to very high temperatures. This research will pave the way for fiber optic and optically based sensors under conditions expected in future high-temperature gas-cooled reactors. Sensor life and signal-to-noise ratios are susceptible to attenuation of the light signal due to scattering and absorbance in the fibers. This project will provide an experimental and theoretical study of the darkening of optical fibers in high-radiation and high-temperature environments. Although optical fibers have been studied for moderate radiation fluence and flux levels, the results of irradiation at very high temperatures have not been published for extended in-core exposures. Several previous multi-scale modeling efforts have studied irradiation effects on the mechanical properties of materials. However, model-based prediction of irradiation-induced changes in silica'@@s optical transport properties has only recently started to receive attention due to possible applications as optical transmission components in fusion reactors. Nearly all damage-modeling studies have been performed in the molecular-dynamics domain, limited to very short times and small systems. Extended-time modeling, however, is crucial to predicting the long-term effects of irradiation at high temperatures, since the experimental testing may not encompass the displacement rate that the fibers will encounter if they are deployed in the VHTR. The project team will pursue such extended-time modeling, including the effects of the ambient and recrystallization. The process will be based on kinetic MC modeling using the concept of amorphous material consisting of building blocks of defect-pairs or clusters, which has been successfully applied to kinetic modeling in amorphized and recrystallized silicon. Using this procedure, the team will model compensation for rate effects, and

  4. Irradiation testing of coated particle fuel at Hanaro

    International Nuclear Information System (INIS)

    Goo Kim, Bong; Sung Cho, Moo; Kim, Yong Wan

    2014-01-01

    TRISO-coated particle fuel is developing to support development of VHTR in Korea. From August 2013, the first irradiation testing of coated particle fuel was begun to demonstrate and qualify TRISO fuel for use in VHTR in the HANARO at KAERI. This experiment is currently undergoing under the atmosphere of a mixed inert gas without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one contains nine fuel compacts and the other five compacts and eight graphite specimens. Each compact has 263 coated particles. After a peak burn-up of about 4 at% and a peak fast neutron fluence of about 1.7 x 10 21 n/cm 2 , PIE will be carried out at KAERI's Irradiated Material Examination Facility. This paper is described characteristics of coated particle fuel, the design of test rod and irradiation device for coated particle fuel, discusses the technical results for irradiation testing at HANARO. (authors)

  5. Germination test for identification of gamma-irradiated bean seeds

    International Nuclear Information System (INIS)

    Wesolowska, B.; Ignatowicz, S.

    1993-01-01

    The feasibility of germination test for the practical detection of irradiated beans has not been investigated. The objective of this study was to determine if the relationship between the root growth rate and radiation dose could be used to produce a rapid analytical method for identification of irradiated beans. Such detection method could be potentially used for both (a) identification of irradiated food, and (b) for quarantine inspection (to certify that the agricultural product has been irradiated, and the pests present in it do not pose a quarantine risk). Results presented in this paper indicate that the germination test is not always capable of discriminating satisfactorily between irradiated and unirradiated samples of bean seeds, because the sensitivity of the test is often higher than the low doses which are suggested for disinfestation purposes. However, using the germination test, an unexperienced person can easily discriminate untreated bean seeds from those irradiated with 0.3-1.5 kGy doses of gamma radiation. (orig./vhe)

  6. Evaluation of fuel rods behavior - under irradiation test

    International Nuclear Information System (INIS)

    Lameiras, F.S.; Terra, J.L.; Pinto, L.C.M.; Dias, M.S.; Pinheiro, R.B.

    1981-04-01

    By the accompanying of the irradiation of instrumented test fuel rods simulating the operational conditions in reactors, plus the results of post - irradiation exams, tests, evaluation and calibration of analitic modelling of such fuel rods is done. (E.G.) [pt

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  8. Nitride fuels irradiation performance data base

    International Nuclear Information System (INIS)

    Brozak, D.E.; Thomas, J.K.; Peddicord, K.L.

    1987-01-01

    An irradiation performance data base for nitride fuels has been developed from an extensive literature search and review that emphasized uranium nitride, but also included performance data for mixed nitrides [(U,Pu)N] and carbonitrides [(U,Pu)C,N] to increase the quantity and depth of pin data available. This work represents a very extensive effort to systematically collect and organize irradiation data for nitride-based fuels. The data base has many potential applications. First, it can facilitate parametric studies of nitride-based fuels to be performed using a wide range of pin designs and operating conditions. This should aid in the identification of important parameters and design requirements for multimegawatt and SP-100 fuel systems. Secondly, the data base can be used to evaluate fuel performance models. For detailed studies, it can serve as a guide to selecting a small group of pin specimens for extensive characterization. Finally, the data base will serve as an easily accessible and expandable source of irradiation performance information for nitride fuels

  9. Stress corrosion testing of irradiated cladding tubes

    International Nuclear Information System (INIS)

    Lunde, L.; Olshausen, K.D.

    1980-01-01

    Samples from two fuel rods with different cladding have been stress corrosion tested by closed-end argon-iodine pressurization at 320 0 C. The fuel rods with stress relieved and recrystallized Zircaloy-2 had received burnups of 10.000 and 20.000 MWd/ton UO 2 , respectively. It was found that the SCC failure stress was unchanged or slightly higher for the irradiated than for the unirradiated control tubes. The tubes failed consistently in the end with the lowest irradiation dose. The diameter increase of the irradiated cladding during the test was 1.1% for the stress-relieved samples and 0.24% for the recrystallized samples. SEM examination revealed no major differences between irradiated and unirradiated cladding. A ''semi-ductile'' fracture zone in recrystallized material is described in some detail. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-05-01

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

  11. Atomic force microscopy and mechanical testing of bovine pericardium irradiated to radiotherapy doses

    International Nuclear Information System (INIS)

    Daar, Eman; Kaabar, W.; Woods, E.; Lei, C.; Nisbet, A.; Bradley, D.A.

    2014-01-01

    Within the context of radiotherapy our work investigates the feasibility of identifying changes in structural and biomechanical properties of pericardium resulting from exposure to penetrating photon irradiation. Collagen fibres extracted from bovine pericardium were chosen as a model of pericardium extracellular matrix as these form the main fibrous component of the medium. Tests of mechanical properties, controlled by the various structural elements of the tissues, were performed on frontal pericardium, including uni-axial tests and atomic force microscopy (AFM). While the irradiated collagen fibres showed no significant change in D-band spacing up to doses of 80 Gy, the fibre width was found to increase by 34±9% at 80 Gy when compared with that for un-irradiated samples. - Highlights: • Methods for identifying changes in tissue biophysical properties following photon irradiation. • Tests made using collagen fibres extracted from bovine pericardium. • Sensitivity of uni-axial tests and atomic force microscopy (AFM) investigated. • Radiotherapy doses investigated up to 80 Gy, delivered by 6 MV photons

  12. Irradiation performance of full-length metallic IFR fuels

    International Nuclear Information System (INIS)

    Tsai, H.; Neimark, L.A.

    1992-07-01

    An assembly irradiation of 169 full-length U-Pu-Zr metallic fuel pins was successfully completed in FFTF to a goal burnup of 10 at.%. All test fuel pins maintained their cladding integrity during the irradiation. Postirradiation examination showed minimal fuel/cladding mechanical interaction and excellent stability of the fuel column. Fission-gas release was normal and consistent with the existing data base from irradiation testing of shorter metallic fuel pins in EBR-II

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-08-15

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

  14. Technology Implementation Plan: Irradiation Testing and Qualification for Nuclear Thermal Propulsion Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rader, Jordan D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    This document is a notional technology implementation plan (TIP) for the development, testing, and qualification of a prototypic fuel element to support design and construction of a nuclear thermal propulsion (NTP) engine, specifically its pre-flight ground test. This TIP outlines a generic methodology for the progression from non-nuclear out-of-pile (OOP) testing through nuclear in-pile (IP) testing, at operational temperatures, flows, and specific powers, of an NTP fuel element in an existing test reactor. Subsequent post-irradiation examination (PIE) will occur in existing radiological facilities. Further, the methodology is intended to be nonspecific with respect to fuel types and irradiation or examination facilities. The goals of OOP and IP testing are to provide confidence in the operational performance of fuel system concepts and provide data to program leadership for system optimization and fuel down-selection. The test methodology, parameters, collected data, and analytical results from OOP, IP, and PIE will be documented for reference by the NTP operator and the appropriate regulatory and oversight authorities. Final full-scale integrated testing would be performed separately by the reactor operator as part of the preflight ground test.

  15. Long duration performance of high temperature irradiation resistant thermocouples

    International Nuclear Information System (INIS)

    Rempe, J.; Knudson, D.; Condie, K.; Cole, J.; Wilkins, S.C.

    2007-01-01

    Many advanced nuclear reactor designs require new fuel, cladding, and structural materials. Data are needed to characterize the performance of these new materials in high temperature, radiation conditions. However, traditional methods for measuring temperature in-pile degrade at temperatures above 1100 C degrees. To address this instrumentation need, the Idaho National Laboratory (INL) developed and evaluated the performance of a high temperature irradiation-resistant thermocouple that contains alloys of molybdenum and niobium. To verify the performance of INL's recommended thermocouple design, a series of high temperature (from 1200 to 1800 C) long duration (up to six months) tests has been initiated. This paper summarizes results from the tests that have been completed. Data are presented from 4000 hour tests conducted at 1200 and 1400 C that demonstrate the stability of this thermocouple (less than 2% drift). In addition, post test metallographic examinations are discussed which confirm the compatibility of thermocouple materials throughout these long duration, high temperature tests. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-08-15

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

  17. Design and manufacturing of instrumented capsule (02F-06K/02F-11K) for nuclear fuel irradiation test in HANARO

    International Nuclear Information System (INIS)

    Kim, Bong Goo; Kang, Y. H.; Cho, M. S.; Sohn, J. M.; Choo, K. N.; Kim, D. S.; Oh, J. M.; Shin, Y.T.; Park, S.J.; Kim, Y. J.; Seo, C.G.; Ryu, J.S.; Cho, Y. G.

    2003-02-01

    To measure the characteristics of nuclear fuel during irradiation test, it is necessary to develop the instrumented capsule for the nuclear fuel irradiation test. Then considering the requirements for the nuclear fuel irradiation test and the compatibility with OR test hole in HANARO as well as the requirements for HANARO operation and related equipments, the instrumented capsule for the nuclear fuel irradiation test was designed and successfully manufactured. The structural integrity of the capsule design was verified by performing nuclear physics, structural and thermal analyses. And, not only out-of-pile tests such as pressure drop test, vibration test, endurance test, were performed in HANARO design verification test facility, but the mechanical and hydraulic safety of the capsule and the compatibility of the capsule with HANARO was verified

  18. Development of irradiation rig in HTTR and dosimetry method. I-I type irradiation equipment

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Taiju; Kikuchi, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Miyamoto, Satoshi; Ogura, Kazutomo [Japan Atomic Power Co., Tokyo (Japan)

    2002-12-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated, helium gas-cooled test reactor with a maximum power of 30 MW. The HTTR aims not only to establish and upgrade the technological basis for the HTGRs but also to perform the innovative basic research on high temperature engineering with high temperature irradiation fields. It is planned that the HTTR is used to perform various engineering tests such as the safety demonstration test, high temperature test operation and irradiation test with large irradiation fields at high temperatures. This paper describes the design of the I-I type irradiation equipment developed as the first irradiation rig for the HTTR and does the planned dosimetry method at the first irradiation test. It was developed to perform in-pile creep test on a stainless steel with large standard size specimens in the HTTR. It can give great loads on the specimens stably and can control the irradiation temperature precisely. The in-core creep properties on the specimens are measured by newly developed differential transformers and the irradiation condition in the core is monitored by thermocouples and self-powered neutron detectors (SPNDs), continuously. The irradiated neutron fluence is assessed by neutron fluence monitors of small metallic wires after the irradiation. The obtained data at the first irradiation test can strongly be contributed to upgrade the technological basis for the HTGRs, since it is the first direct measurement of the in-core irradiation environments of the HTTR. (author)

  19. Conceptual design for simulator of irradiation test reactors

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Ohto, Tsutomu; Magome, Hirokatsu; Izumo, Hironobu; Hori, Naohiko

    2012-03-01

    A simulator of irradiation test reactors has been developed since JFY 2010 for understanding reactor behavior and for upskilling in order to utilize a nuclear human resource development (HRD) and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR, one of the irradiation test reactors, and it simulates operation, irradiation tests and various kinds of accidents caused by the reactor and irradiation facility. The development of the simulator is sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. The training using the simulator will be started for the nuclear HRD from JFY 2012. This report summarizes the result of the conceptual design of the simulator in JFY 2010. (author)

  20. Irradiation capsule for testing magnetic fusion reactor first-wall materials at 60 and 2000C

    International Nuclear Information System (INIS)

    Conlin, J.A.

    1985-08-01

    A new type of irradiation capsule has been designed, and a prototype has been tested in the Oak Ridge Research Reactor (ORR) for low-temperature irradiation of Magnetic Fusion Reactor first-wall materials. The capsule meets the requirements of the joint US/Japanese collaborative fusion reactor materials irradiation program for the irradiation of first-wall fusion reactor materials at 60 and 200 0 C. The design description and results of the prototype capsule performance are presented

  1. Irradiation test of borosilicate glass burnable poison

    International Nuclear Information System (INIS)

    Feng Mingquan; Liao Zumin; Yang Mingjin; Lu Changlong; Huang Deyang; Zeng Wangchun; Zhao Xihou

    1991-08-01

    The irradiation test and post-irradiation examinations for borosilicate glass burnable poison are introduced. Examinations include visual examination, measurement of dimensions and density, and determination of He gas releasing and 10 B burnup. The corrosion and phenomenon of irradiation densification are also discussed. Two type glass samples have been irradiated with different levels of neutron flux. It proved that the GG-17 borosilicate glass can be used as burnable poison to replace the 10 B stainless steel in the Qinshan Nuclear Power Plant, and it is safe, economical and reasonable

  2. Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-15

    A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

  3. Storage tests on irradiated deep-frozen chickens

    International Nuclear Information System (INIS)

    Gruenewald, T.

    1975-01-01

    Salmonellae infections in deep-frozen roasting chicken can be dealt with by ionising radiation as this process involves hardly any heating of the product. Deep-frozen chickens irradiated with doses up to 800 krad were stored at -30 0 C for two years and were regularly submitted to sensory tests. There was no significant difference in quality between the irradiated samples and the non-irradiated controls. (orig.) [de

  4. Computational analysis of modern HTGR fuel performance and fission product release during the HFR-EU1 irradiation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl, E-mail: k.verfondern@fz-juelich.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Xhonneux, André, E-mail: xhonneux@lrst.rwth-aachen.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Nabielek, Heinz, E-mail: heinznabielek@me.com [Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Allelein, Hans-Josef, E-mail: h.j.allelein@fz-juelich.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, Chair for Reactor Safety and Reactor Technology, 52072 Aachen (Germany)

    2014-07-01

    Highlights: • HFR-EU1 irradiation test demonstrates high quality of HTGR spherical fuel elements. • Irradiation performance is in good agreement with German fuel performance modeling. • International benchmark exercise expected first particle to fail at ∼13–17% FIMA. • EOL silver release is predicted to be in the percentage range. • EOL cesium and strontium are expected to remain at a low level. - Abstract: Various countries engaged in the development and fabrication of modern HTGR fuel have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under HTGR operating and accident conditions. Verification and validation studies are conducted by code-to-code benchmarking and code-to-experiment comparisons as part of international exercises. The methodology developed in Germany since the 1980s represents valuable and efficient tools to describe fission product release from spherical fuel elements and TRISO fuel performance, respectively, under given conditions. Continued application to new results of irradiation and accident simulation testing demonstrates the appropriateness of the models in terms of a conservative estimation of the source term as part of interactions with HTGR licensing authorities. Within the European irradiation testing program for HTGR fuel and as part of the former EU RAPHAEL project, the HFR-EU1 irradiation experiment explores the potential for high performance of the presently existing German and newly produced Chinese fuel spheres under defined conditions up to high burnups. The fuel irradiation was completed in 2010. Test samples are prepared for further postirradiation examinations (PIE) including heatup simulation testing in the KÜFA-II furnace at the JRC-ITU, Karlsruhe, to be conducted within the on-going ARCHER Project of the European Commission. The paper will describe the application of the German computer models to the HFR-EU1 irradiation test and

  5. Spherical nanoindentation of proton irradiated 304 stainless steel: A comparison of small scale mechanical test techniques for measuring irradiation hardening

    Science.gov (United States)

    Weaver, Jordan S.; Pathak, Siddhartha; Reichardt, Ashley; Vo, Hi T.; Maloy, Stuart A.; Hosemann, Peter; Mara, Nathan A.

    2017-09-01

    Experimentally quantifying the mechanical effects of radiation damage in reactor materials is necessary for the development and qualification of new materials for improved performance and safety. This can be achieved in a high-throughput fashion through a combination of ion beam irradiation and small scale mechanical testing in contrast to the high cost and laborious nature of bulk testing of reactor irradiated samples. The current work focuses on using spherical nanoindentation stress-strain curves on unirradiated and proton irradiated (10 dpa at 360 °C) 304 stainless steel to quantify the mechanical effects of radiation damage. Spherical nanoindentation stress-strain measurements show a radiation-induced increase in indentation yield strength from 1.36 GPa to 2.72 GPa and a radiation-induced increase in indentation work hardening rate of 10 GPa-30 GPa. These measurements are critically compared against Berkovich nanohardness, micropillar compression, and micro-tension measurements on the same material and similar grain orientations. The ratio of irradiated to unirradiated yield strength increases by a similar factor of 2 when measured via spherical nanoindentation or Berkovich nanohardness testing. A comparison of spherical indentation stress-strain curves to uniaxial (micropillar and micro-tension) stress-strain curves was achieved using a simple scaling relationship which shows good agreement for the unirradiated condition and poor agreement in post-yield behavior for the irradiated condition. The disagreement between spherical nanoindentation and uniaxial stress-strain curves is likely due to the plastic instability that occurs during uniaxial tests but is absent during spherical nanoindentation tests.

  6. Irradiation effects test Series Scoping Test 1: test results report

    International Nuclear Information System (INIS)

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

    1977-09-01

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

  7. 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

  8. Accelerated irradiation test of Gundremmingen reactor vessel trepan material

    Energy Technology Data Exchange (ETDEWEB)

    Hawthorne, J.R. [Materials Engineering Associates, Inc., Lanham, MD (United States)

    1992-08-01

    Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279{degrees}C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed.

  9. Accelerated irradiation test of gundremmingen reactor vessel trepan material

    International Nuclear Information System (INIS)

    Hawthorne, J.R.

    1992-08-01

    Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279 degrees C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed

  10. Development of Micro-welding Technology of Cladding Tube with Temperature Sensor for Nuclear Fuel Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Soo Sung; Lee, C. Y.; Kim, W. K.; Lee, J. W.; Lee, D. Y

    2006-01-15

    Laser welding technology is widely used to fabricate some products of nuclear fuel in the nuclear industry. Especially, micro-laser welding is one of the key technology to be developed to fabricate precise products of fuel irradiation test. We have to secure laser welding technology to perform various instrumentations for fuel irradiation test. The instrumented fuel irradiation test at a research reactor is needed to evaluate the performance of the developed nuclear fuel. The fuel elements can be designed to measure the center line temperature of fuel pellets during the irradiation test by using temperature sensor. The thermal sensor was composed of thermocouple and sensor sheath. Micro-laser welding technology was adopted to seal between seal tube and sensor sheath with thickness of 0.15mm. The soundness of weld area has to be confirmed to prevent fission gas of the fuel from leaking out of the element during the fuel irradiation test. In this study, fundamental data for micro-laser welding technology was proposed to seal temperature sensor sheath of the instrumented fuel element. And, micro-laser welding for dissimilar metals between sensor sheath and seal tube was characterized by investigating welding conditions. Moreover, the micro-laser welding technology is closely related to advanced industry. It is expected that the laser material processing technology will be adopted to various applications in the industry.

  11. Testing of Performance of Optical Fibers Under Irradiation in Intense Radiation Fields, When Subjected to Very High Temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Blue, Thomas [The Ohio State Univ., Columbus, OH (United States); Windl, Wolfgang [The Ohio State Univ., Columbus, OH (United States); Dickerson, Bryan [Luna Innovations, Inc. (United States)

    2013-01-03

    The primary objective of this project is to measure and model the performance of optical fibers in intense radiation fields when subjected to very high temperatures. This research will pave the way for fiber optic and optically based sensors under conditions expected in future high-temperature gas-cooled reactors. Sensor life and signal-to-noise ratios are susceptible to attenuation of the light signal due to scattering and absorbance in the fibers. This project will provide an experimental and theoretical study of the darkening of optical fibers in high-radiation and high-temperature environments. Although optical fibers have been studied for moderate radiation fluence and flux levels, the results of irradiation at very high temperatures have not been published for extended in-core exposures. Several previous multi-scale modeling efforts have studied irradiation effects on the mechanical properties of materials. However, model-based prediction of irradiation-induced changes in silica's optical transport properties has only recently started to receive attention due to possible applications as optical transmission components in fusion reactors. Nearly all damage-modeling studies have been performed in the molecular-dynamics domain, limited to very short times and small systems. Extended-time modeling, however, is crucial to predicting the long-term effects of irradiation at high temperatures, since the experimental testing may not encompass the displacement rate that the fibers will encounter if they are deployed in the VHTR. The project team will pursue such extended-time modeling, including the effects of the ambient and recrystallization. The process will be based on kinetic MC modeling using the concept of amorphous material consisting of building blocks of defect-pairs or clusters, which has been successfully applied to kinetic modeling in amorphized and recrystallized silicon. Using this procedure, the team will model compensation for rate effects, and

  12. Performance evaluation of gamma irradiated SiR-EPDM blends

    Energy Technology Data Exchange (ETDEWEB)

    Deepalaxmi, R., E-mail: deepalaxmivaithi@gmail.com; Rajini, V.

    2014-07-01

    Highlights: • The effects of gamma irradiation on SiR-EPDM blend are examined. • Cross-linking reaction is dominant in blends C, D and E, due to higher EPDM content. • The tensile strength and hardness of blend E is improved by gamma irradiation. • The blend C and EPDM rich blends (D, E) are found to have superior performance. • Among C, D and E, suitable blend can be selected for a particular NPP application. - Abstract: Cable insulation materials (CIM) should perform their safety functions throughout their installed life in nuclear power plants (NPP). The CIM will be exposed to gamma irradiation at the installed locations. In order to forecast long-term performance of CIM, the short time accelerated testing was carried out. Due to its good mechanical strength, ethylene propylene diene monomer (EPDM) is widely used as CIM. Silicone rubber (SiR) is used in high temperature environments, due to its good di-electric properties/hydrophobicity. The blending of these two polymers may result in the improvement in their specific properties. This paper analyses the effects of gamma irradiation on the five different compositions (90-10; 70-30; 50-50; 30-70; 10-90) of SiR-EPDM blends. The blends were exposed to four different doses (25 Mrad, 100 Mrad, 200 Mrad and 250 Mrad) of gamma irradiation. The electrical and mechanical parameters like volume resistivity (VRY), surface resistivity (SRY), tensile strength (TS), elongation at break (EB), hardness (H) of the virgin and gamma irradiated blends were determined as per ASTM/IEC standards. The nature of degradation was investigated using Fourier transform infrared spectroscopy (FTIR). The simultaneous occurrence of cross-linking and chain scission is found to be the mechanism for ageing in SiR-EPDM blends. The electrical parameters such as volume resistivity and surface resistivity of all the blends are found to improve for all doses of gamma irradiation. To validate the influence of cross-linking reaction of the Si

  13. Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra

    International Nuclear Information System (INIS)

    Duran, I.; Bolshakova, I.; Holyaka, R.; Viererbl, L.; Lahodova, Z.; Sentkerestiova, J.; Bem, P.

    2010-01-01

    We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10 16 cm -2 was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

  14. Performance evaluation of large U-Mo particle dispersed fuel irradiated in HANARO

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Oh, Seok Jin; Jang, Se Jung; Yu, Byung Ok; Lee, Choong Seong; Seo, Chul Gyo; Chae, Hee Taek; Kim, Chang Kyu

    2008-01-01

    U-Mo/Al dispersion fuel is being developed as advanced fuel for research reactors. Irradiation behavior of U-Mo/Al dispersion fuel has been studied to evaluate its fuel performance. One of the performance limiting factors is a chemical interaction between the U-Mo particle and the Al matrix because the thermal conductivity of fuel meat is decreased with the interaction layer growth. In order to overcome the interaction problem, large-sized U-Mo particles were fabricated by controlling the centrifugal atomization conditions. The fuel performance behavior of U-Mo/Al dispersion fuel was estimated by using empirical models formulated based on the microstructural analyses of the post-irradiation examination (PIE) on U-Mo/Al dispersion fuel irradiated in HANARO reactor. Temperature histories of U-Mo/Al dispersion fuel during irradiation tests were estimated by considering the effect of an interaction layer growth on the thermal conductivity of the fuel meat. When the fuel performances of the dispersion fuel rods containing U-Mo particles with various sizes were compared, fuel temperature was decreased as the average U-Mo particle size was increases. It was found that the dispersion of a larger U-Mo particle was effective for mitigating the thermal degradation which is associated with an interaction layer growth. (author)

  15. Utilization of half-embryo test to identify irradiated beans

    International Nuclear Information System (INIS)

    Villavicencio, Anna Lucia C.H.; Mancini-Filho, Jorge

    1996-01-01

    Germination tests were carried out in irradiated and non-irradiated bean seeds which allow to observe characteristically variations on the shoots and roots. The methodology used in this work, is based upon biological changes which occur in two Brazilian beans, Phaseolus vulgaris L., var. carioca and Vigna unguiculata (L.) Walp, var. macacar, irradiated in a 60 Co source, with doses of 0,0.5, 1.0, 2.5, 5.0 and 10.0 kGy. The shoots and roots were observed during 3 days of culturing period under specified conditions. The differences observed in these two varieties were analysed immediately after irradiation and after 6 months of storage period at room temperature. Irradiated half-embryos showed markedly reduced root grow and almost totally retarded shoot elongation. Differences between irradiated and nonirradiated half-embryo could be observed after irradiation when different beans and storage time were varied. The shoots of half-embryos irradiated with more than 2.5 kGy did not undergo any elongation, whereas, the shoots of non-irradiated or those beans irradiated under 1.0 kGy elongated significantly within the 3 day test period. (author)

  16. Crack-arrest tests on two irradiated high-copper welds

    International Nuclear Information System (INIS)

    Iskander, S.K.; Corwin, W.R.; Nanstad, R.K.

    1994-03-01

    The objective of the Heavy-Section Steel Irradiation Program Sixth Irradiation Series is to determine the effect of neutron irradiation on the shift and shape of the lower-bound curve to crack-arrest toughness data. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plate. Crack-arrest specimens fabricated from these welds were irradiated at a nominal temperature of 288 degrees C to an average fluence of 1.9 x 10 19 neutrons/cm 2 (>1 MeV). This is the second report giving the results of the tests on irradiated duplex-type crack-arrest specimens. A previous report gave results of tests on irradiated weld-embrittled-type specimens. Charpy V-notch (CVN) specimens irradiated in the same capsules as the crack-arrest specimens were also tested, and a 41-J transition temperature shift was determined from these specimens. open-quotes Mean close-quote curves of the same form as the American Society of Mechanical Engineers (ASME) K la curve were fit to the data with only the open-quotes reference temperatureclose quotes as a parameter. The shift between the mean curves agrees well with the 41-J transition temperature shift obtained from the CVN specimen tests. Moreover, the four data points resulting from tests on the duplex crack-arrest specimens of the present study did not make a significant change to mean curve fits to either the previously obtained data or all the data combined

  17. International fusion materials irradiation facility and neutronic calculations for its test modules

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.

    1997-01-01

    The International Fusion Material Irradiation Facility (IFMIF) is a projected high intensity neutron source for material testing. Neutron transport calculations for the IFMIF project are performed for variety of here explained reasons. The results of MCNP neutronic calculations for IFMIF test modules with NaK and He cooled high flux test cells are presented in this paper. (author). 3 refs., 2 figs., 3 tabs

  18. Development of high performance lubricant through the compatibilization of polyalphaolefin, polyurea and irradiated polytetrafluoroethylene

    International Nuclear Information System (INIS)

    Ratao, Natalia T.; Zen, Heloisa A.; Lugao, Ademar B.

    2013-01-01

    High performance lubricants are essential for shipping and oil exploration industries for working under severe conditions of temperature, pressure and contamination. Polytetrafluoroethylene (PTFE) is widely used as a lubricating additive in greases for presenting the lowest known coefficient of friction, is commercially found irradiated in air, by this way can be grinded and oxygenated terminal groups can be formed. PTFE processed by ionizing radiation with 4μ average particle size was supplied by Uniflon and compared with the ultraviolet-irradiated PTFE with the same average particle size, from Dupont. The characterization of functional proprieties by drop point, oil separation and four ball test was conducted with a polyurea base grease additive with both types of PTFE. The chemical characterization was done with Differential Scanning Calorimetry (DSC). The drop point and oil separation tests showed that the Uniflon irradiated PTFE improved the thermal resistance and stability of polyurea grease due to the greater presence of the end groups increase the compatibility of the additive with the fluid. Four ball tests showed that the grease with Uniflon irradiated PTFE also has a better antiwear and extreme pressure performance because of the greater polarity of the molecules improving the dispersion and anchoring of the PTFE in metal surface voids, reducing wear. The PTFE processed by ionizing radiation showed greater compatibility with polyurea grease and polyalphaolefin fluid and the grease functional characteristics met the requirements for a high performance lubricant. (author)

  19. Irradiation capability of Japanese materials test reactor for water chemistry experiments

    International Nuclear Information System (INIS)

    Hanawa, Satoshi; Hata, Kuniki; Chimi, Yasuhiro; Nishiyama, Yutaka; Nakamura, Takehiko

    2012-09-01

    Appropriate understanding of water chemistry in the core of LWRs is essential as chemical species generated due to water radiolysis by neutron and gamma-ray irradiation govern corrosive environment of structural materials in the core and its periphery, causing material degradation such as stress corrosion cracking. Theoretical model calculation such as water radiolysis calculation gives comprehensive understanding of water chemistry at irradiation field where we cannot directly monitor. For enhancement of the technology, accuracy verification of theoretical models under wide range of irradiation conditions, i.e. dose rate, temperature etc., with well quantified in-pile measurement data is essential. Japan Atomic Energy Agency (JAEA) has decided to launch water chemistry experiments for obtaining data that applicable to model verification as well as model benchmarking, by using an in-pile loop which will be installed in the Japan Materials Testing Reactor (JMTR). In order to clarify the irradiation capability of the JMTR for water chemistry experiments, preliminary investigations by water radiolysis / ECP model calculations were performed. One of the important irradiation conditions for the experiments, i.e. dose rate by neutron and gamma-ray, can be controlled by selecting irradiation position in the core. In this preliminary study, several representative irradiation positions that cover from highest to low absorption dose rate were chosen and absorption dose rate at the irradiation positions were evaluated by MCNP calculations. As a result of the calculations, it became clear that the JMTR could provide the irradiation conditions close to the BWR. The calculated absorption dose rate at each irradiation position was provided to water radiolysis calculations. The radiolysis calculations were performed under various conditions by changing absorption dose rate, water chemistry of feeding water etc. parametrically. Qualitatively, the concentration of H 2 O 2 , O 2 and

  20. POST-IRRADIATION ANALYSES OF U-MO DISPERSION FUEL RODS OF KOMO TESTS AT HANARO

    Directory of Open Access Journals (Sweden)

    H.J. RYU

    2013-12-01

    Full Text Available Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4–5 g-U/cm3 were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr, additional protective coatings (silicide or nitride, and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

  1. Irradiation Tests on Bipolar Front-End Preamplifier Using the Harris UHF1 Technology

    CERN Document Server

    Goyot, Michel

    1997-01-01

    A low noise, low power, radiation hard, full custom integrated circuit has been studied for coupling with photodetectors of the CMS electromagnetic calorimeter. A new version of the prototype preamplifier has been designed with improved performances ( noise, dynamic range, consumption). Irradiation tests on this monolithic circuit have been performed at CERN with a Xray SEIFERT RP149 generator up to 10Mrad ( SiO2). Tests results are presented.

  2. Design considerations of the irradiation test vehicle for the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    An irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) is being jointly developed by the Lockheed Martin Idaho Technologies Company (LMIT) and the U.S. Fusion Program. The vehicle is intended for neutron irradiation testing of candidate structural materials, including vanadium-based alloys, silicon carbide composites, and low activation steels. It could possibly be used for U.S./Japanese collaboration in the Jupiter Program. The first test train is scheduled to be completed by September 1998. In this report, we present the functional requirements for the vehicle and a preliminary design that satisfies these requirements.

  3. Design considerations of the irradiation test vehicle for the advanced test reactor

    International Nuclear Information System (INIS)

    Tsai, H.; Gomes, I.C.; Smith, D.L.

    1997-01-01

    An irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) is being jointly developed by the Lockheed Martin Idaho Technologies Company (LMIT) and the U.S. Fusion Program. The vehicle is intended for neutron irradiation testing of candidate structural materials, including vanadium-based alloys, silicon carbide composites, and low activation steels. It could possibly be used for U.S./Japanese collaboration in the Jupiter Program. The first test train is scheduled to be completed by September 1998. In this report, we present the functional requirements for the vehicle and a preliminary design that satisfies these requirements

  4. Investigation of TIG welding characteristics with a dual cooled rod for the fuel irradiation test

    International Nuclear Information System (INIS)

    Kim, Soo Sung; Kim, Hyung Kyu

    2008-01-01

    To establish the fabrication process, and for satisfying the requirements of the irradiation test, an TIG(Tungsten Inert Gas) welding machine for the dual cooled rods specimens was developed, and the preliminary welding experiments were performed to optimize the welding process conditions. Cladding tubes of 15.9 and 9 mm for the outer and inner diameters, respectively with a 0.57 mm thickness and end caps were used for the specimens. This paper describes the experimental results of the TIG welds and the micrograph examinations of the TIG welded specimens corresponding to various welding conditions for the dual cooled fuel irradiation test. The investigations revealed that the present TIG process satisfied the requirements for the fuel irradiation test in the HANARO research reactor

  5. Coating effect of chitosan irradiation on performance of strawberries (Fragaria x ananassa Duchesne)

    International Nuclear Information System (INIS)

    Gatot Trimulyadi Rekso; Adjat Sudradjat

    2016-01-01

    Strawberry is a horticultural product which has high economic value, however, strawberries are easily damaged. Therefore, required proper post harvest handling for keep up the appearances of strawberries, one of them is the edible coating technique using chitosan. The purpose of this research was to maintaining the appearance of strawberry. Chitosan is irradiated at a dose of 0 kGy, 5 kGy, and 10 kGy, and then characterized. Observations were carried on for five days at room temperature including: the testing of the performance of the fruit that was observed visually and analysis of fruit weight loss. The results showed that the performance of strawberries is coating with irradiated chitosan of 10 kGy is still good. The percentage of weight loss strawberries coated with irradiated chitosan smaller than the strawberries without coating with chitosan (control). Strawberries coated with irradiated chitosan has a shelf life and weight loss better than control. (author)

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

    International Nuclear Information System (INIS)

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

    2004-11-01

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

  7. Capsule Development and Utilization for Material Irradiation Tests

    International Nuclear Information System (INIS)

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

    2003-05-01

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

  8. A germination test: an easy approach to know the irradiation

    International Nuclear Information System (INIS)

    Khawar, A.; Bhatti, I.A.; Bhatti, H.N.

    2010-01-01

    Food irradiation is an evolving preserving technique that provides a shield against the spoilage and might have a potential to ensure the food safety and security world wide. In the present study, feasibility to apply germination test to distinguish an un-irradiated and irradiated samples of wheat, maize, chickpea and black eye beans was checked. Samples were irradiated to the absorbed doses ranging from 0-10 kGy using Co-60 gamma irradiator and were germinated in plant growth chamber. Root and shoot lengths were measured at 7th day after gamma radiation treatment. In all the irradiated samples root and shoot lengths were decreased with the increase in radiation absorbed doses. The seeds irradiated to the absorbed doses more than 2 kGy were not germinated. Germination test proved as an easy and simple method to detect irradiation in wheat, maize, chickpea and black eye beans irradiated even at low absorbed doses. (author)

  9. Early works on the nuclear microprobe for microelectronics irradiation tests at the CEICI (Sevilla, Spain)

    International Nuclear Information System (INIS)

    Palomo, F.R.; Morilla, Y.; Mogollon, J.M.; Garcia-Lopez, J.; Labrador, J.A.; Aguirre, M.A.

    2011-01-01

    Particle radiation effects are a fundamental problem in the use of numerous electronic devices for space applications, which is aggravated with the technology shrinking towards smaller and smaller scales. The suitability of low-energy accelerators for irradiation testing is being considered nowadays. Moreover, the possibility to use a nuclear microprobe, with a lateral resolution of a few microns, allows us to evaluate the behavior under ion irradiation of specific elements in an electronic device. The CEICI is the new CEnter for Integrated Circuits Irradiation tests, created into the facilities at the Centro Nacional de Aceleradores (CNA) in Sevilla-Spain. We have verified that our 3 MV Tandem accelerator, typically used for ion beam characterization of materials, is also a valuable tool to perform irradiation experiments in the low LET (Linear Energy Transfer) region.

  10. Preliminary Beam Irradiation Test for RI Production Targets at KOMAC

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Sang Pil; Kwon, Hyeok Jung; Kim, Han Sung; Cho, Yong Sub; Seol, Kyung Tae; Song, Young Gi; Kim, Dae Il; Jung, Myung Hwan; Kim, Kye Ryung; Min, Yi Sub [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The new beamline and target irradiation facility has been constructed for the production of therapeutic radio-isotope. Sr-82 and Cu-67 were selected as the target isotope in this facility, they are promising isotope for the PET imaging and cancer therapy. For the facility commissioning, the irradiation test for the prototype-target was conducted to confirm the feasibility of radio-isotope production, the proto-type targets are made of RbCl pellet and the natural Zn metal for Sr-82 and Cu-67 production respectively, In this paper, an introduction to the RI production targetry system and the results of the preliminary beam irradiation test are discussed. the low-flux beam irradiation tests for proto-type RI target have been conducted. As a result of the beam irradiation tests, we could obtain the evidence of Sr-82 and Cu-67 production, have confirmed the feasibility of Sr-82 and Cu-67 production at KOMAC RI production facility.

  11. Preliminary Beam Irradiation Test for RI Production Targets at KOMAC

    International Nuclear Information System (INIS)

    Yoon, Sang Pil; Kwon, Hyeok Jung; Kim, Han Sung; Cho, Yong Sub; Seol, Kyung Tae; Song, Young Gi; Kim, Dae Il; Jung, Myung Hwan; Kim, Kye Ryung; Min, Yi Sub

    2016-01-01

    The new beamline and target irradiation facility has been constructed for the production of therapeutic radio-isotope. Sr-82 and Cu-67 were selected as the target isotope in this facility, they are promising isotope for the PET imaging and cancer therapy. For the facility commissioning, the irradiation test for the prototype-target was conducted to confirm the feasibility of radio-isotope production, the proto-type targets are made of RbCl pellet and the natural Zn metal for Sr-82 and Cu-67 production respectively, In this paper, an introduction to the RI production targetry system and the results of the preliminary beam irradiation test are discussed. the low-flux beam irradiation tests for proto-type RI target have been conducted. As a result of the beam irradiation tests, we could obtain the evidence of Sr-82 and Cu-67 production, have confirmed the feasibility of Sr-82 and Cu-67 production at KOMAC RI production facility

  12. Thrombogenicity tests on ar-irradiated polycarbonate foils

    Energy Technology Data Exchange (ETDEWEB)

    Trindade, Gustavo F.; Rizzutto, Marcia A.; Silva, Tiago F.; Moro, Marcos V.; Added, Nemitala; Tabacniks, Manfredo H., E-mail: g.ferraz@usp.br [Universidade de Sao Paulo (USP), Sao Paulo, SP (Brazil). Inst. de Fisica; Delgado, Adriana O. [Universidade Federal de Sao Carlos (UFSCAR), Sorocaba, SP (Brazil); Cunha, Tatiana F. [Biosintesis P and D do Brasil, Sao Paulo, SP (Brazil); Higa, Olga Z. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Biotecnologia

    2013-07-01

    Understanding polymer surface properties is extremely important for the most wide range of their applications, from basic coating to the most complex composites and biomaterials. Low energy ion beam irradiation of polymer can improve such surface properties. By modifying its surface biocompatibility, polymers are excellent candidates for biomaterials, due to its malleability and low weight, when compared to metals. In this work, we irradiated 30-μm Bisphenol-A Polycarbonate foils with 23-keV Argon ion beam at six different doses. Aluminium foils were simultaneously irradiated in order to measure the doses by Rutherford Backscattering Spectroscopy. The surface modifications after the argon ion beam irradiation were analyzed by water contact angle measurements and atomic force microscopy. Platelet adhesion tests were used in order to investigate thrombogenicity, showing a growing tendency with the irradiated Argon dose. (author)

  13. Thrombogenicity tests on ar-irradiated polycarbonate foils

    International Nuclear Information System (INIS)

    Trindade, Gustavo F.; Rizzutto, Marcia A.; Silva, Tiago F.; Moro, Marcos V.; Added, Nemitala; Tabacniks, Manfredo H.; Cunha, Tatiana F.; Higa, Olga Z.

    2013-01-01

    Understanding polymer surface properties is extremely important for the most wide range of their applications, from basic coating to the most complex composites and biomaterials. Low energy ion beam irradiation of polymer can improve such surface properties. By modifying its surface biocompatibility, polymers are excellent candidates for biomaterials, due to its malleability and low weight, when compared to metals. In this work, we irradiated 30-μm Bisphenol-A Polycarbonate foils with 23-keV Argon ion beam at six different doses. Aluminium foils were simultaneously irradiated in order to measure the doses by Rutherford Backscattering Spectroscopy. The surface modifications after the argon ion beam irradiation were analyzed by water contact angle measurements and atomic force microscopy. Platelet adhesion tests were used in order to investigate thrombogenicity, showing a growing tendency with the irradiated Argon dose. (author)

  14. Irradiation tests on bipolar front-end preamplifier using the Harris UHF1 technology

    Energy Technology Data Exchange (ETDEWEB)

    Goyot, M

    1997-09-01

    A low noise, low power, radiation hard, full custom integrated circuit has been studied for coupling with photodetectors of the CMS electromagnetic calorimeter. A new version of the prototype preamplifier has been designed with improved performances (noise, dynamic range, consumption). Irradiation tests on this monolithic circuit have been performed at CERN with a X ray SEIFERT RP149 generator up to 10 Mrad equivalent silicon oxide. Tests results are presented. (author) 5 refs.

  15. Tensile tests and metallography of brazed AISI 316L specimens after irradiation

    International Nuclear Information System (INIS)

    Groot, P.; Franconi, E.

    1994-01-01

    Stainless steel type 316L tensile specimens were vacuum brazed with three kinds of alloys: BNi-5, BNi-6, and BNi-7. The specimens were irradiated up to 0.7 dpa at 353 K in the High Flux Reactor at JRC Petten, the Netherlands. Tensile tests were performed at a constant displacement rate of 10 -3 s -1 at room temperature in the ECN hot cell facility. BNi-5 brazed specimens showed ductile behaviour. Necking and fractures were localized in the plate material. BNi-6 and BNi-7 brazed specimens failed brittle in the brazed zone. This was preceded by uniform deformation of the plate material. Tensile test results of irradiated specimens showed higher stresses due to radiation hardening and a reduction of the elongation of the plate material compared to the reference. SEM examination of the irradiated BNi-6 and BNi-7 fracture surfaces showed nonmetallic phases. These phases were not found in the reference specimens. ((orig.))

  16. Fusion materials irradiation test facility: description and status

    International Nuclear Information System (INIS)

    Trego, A.L.; Parker, E.F.; Hagan, J.W.

    1982-01-01

    The Fusion Materials Irradiation Test (FMIT) Facility will generate a high-flux, high-energy neutron source that will provide a fusion-like radiation environment for fusion reactor materials development. The neutrons will be produced in a nuclear stripping reaction by impinging a 35 MeV beam of deuterons from an Alvarez-type linear accelerator on a flowing lithium target. The target will be located in a test cell which will provide an irradiation volume of over 750l within which 10 cm 3 will have an average neutron flux of greater than 1.4 x 10 15 n/cm 2 -s and 500 cm 3 an average flux of greater than 2.2 by 10 14 n/cm 2- s with an expected availability factor greater than 65%. The projected fluence within the 10 cm 3 high flux region of FMIT will effect damage upon the materials test specimens to 30 dpa (displacements per atom) for each 90 day irradiation period. This irradiation flux volume will be at least 500 times larger than that of any other facility with comparable neutron energy and will fully meet the fusion materials damage research objective of 100 dpa within three years for the first round of tests

  17. Impact of neutron irradiation on mechanical performance of FeCrAl alloy laser-beam weldments

    Science.gov (United States)

    Gussev, M. N.; Cakmak, E.; Field, K. G.

    2018-06-01

    Oxidation-resistant iron-chromium-aluminum (FeCrAl) alloys demonstrate better performance in Loss-of-Coolant Accidents, compared with austenitic- and zirconium-based alloys. However, further deployment of FeCrAl-based materials requires detailed characterization of their performance under irradiation; moreover, since welding is one of the key operations in fabrication of light water reactor fuel cladding, FeCrAl alloy weldment performance and properties also should be determined prior to and after irradiation. Here, advanced C35M alloy (Fe-13%Cr-5%Al) and variants with aluminum (+2%) or titanium carbide (+1%) additions were characterized after neutron irradiation in Oak Ridge National Laboratory's High Flux Isotope Reactor at 1.8-1.9 dpa in a temperature range of 195-559 °C. Specimen sets included as-received (AR) materials and specimens after controlled laser-beam welding. Tensile tests with digital image correlation (DIC), scanning electron microscopy-electron back scatter diffraction analysis, fractography, and x-ray tomography analysis were performed. DIC allowed for investigating local yield stress in the weldments, deformation hardening behavior, and plastic anisotropy. Both AR and welded material revealed a high degree of radiation-induced hardening for low-temperature irradiation; however, irradiation at high-temperatures (i.e., 559 °C) had little overall effect on the mechanical performance.

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-10-01

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

  1. Irradiation performance of HTGR fuel in HFIR experiment HRB-13

    International Nuclear Information System (INIS)

    Tiegs, T.N.

    1982-03-01

    Irradiation capsule HRB-13 tested High-Temperature Gas-Cooled Reactor (HTGR) fuel under accelerated conditions in the High Flux Isotope Reactor (HFIR) at ORNL. The ORNL part of the capsule was designed to provide definitive results on how variously misshapen kernels affect the irradiation performance of weak-acid-resin (WAR)-derived fissile fuel particles. Two batches of WAR fissile fuel particles were Triso-coated and shape-separated into four different fractions according to their deviation from spericity, which ranged from 9.6 to 29.7%. The fissile particles were irradiated for 7721 h. Heavy-metal burnups ranged from 80 to 82.5% FIMA (fraction of initial heavy-metal atoms). Fast neutron fluences (>0.18 MeV) ranged from 4.9 x 10 25 neutrons/m 2 to 8.5 x 10 25 neutrons/m 2 . Postirradiation examination showed that the two batches of fissile particles contained chlorine, presumably introduced during deposition of the SiC coating

  2. Visual observations of a degraded bundle of irradiated fuel: the Phebus FPT1 test

    International Nuclear Information System (INIS)

    Barrachin, M.; Bottomley, P.D.

    1999-01-01

    The international Phebus-FP (Fission Product) project is managed by the Institut de Protection et Surete Nucleaire in collaboration with Electricite de France (EDF), the European Commission (EC), the USNRC (USA), COG (Canada), NUPEC and JAERI (Japan), KAERI (South Korea), PSI and HSK (Switzerland). It is designed to measure the source-term and to study the degradation of irradiated UO 2 fuel in conditions typical of a severe loss of coolant accident in a pressurised water reactor (PWR). In the first test (FPT0), performed in December '93, a bundle of 20 fresh fuel rods and a central Ag-In-Cd control rod underwent a short 15-day irradiation to generate fission products before testing in the Phebus reactor in Cadarache. The second test (FPT1) was performed in July '96, in the same conditions and geometry, but using irradiated fuel (-23 GWd/tU). In the FPT1 test, the bundle was heated to an estimated 3000 K over a period of 30 minutes in order to induce a substantial liquefaction of the bundle. After the test, the bundle was embedded in epoxy and cut at different levels to investigate the mechanisms of the core degradation. This paper reports the visual observations of the degraded FPT1 bundle, very preliminary interpretations about the scenario of degradation and a comparison between the behaviour of the fuel in the FPT0 and FPT1 tests. (author)

  3. Post-irradiation examination and R and D programs using irradiated fuels at KAERI

    International Nuclear Information System (INIS)

    Chun, Yong Bum; Min, Duck Kee; Kim, Eun Ka and others

    2000-12-01

    This report describes the Post-Irradiation Examination(PIE) and R and D programs using irradiated fuels at KAERI. The objectives of post-irradiation examination (PIE) for the PWR irradiated fuels, CANDU fuels, HANARO fuels and test fuel materials are to verify the irradiation performance and their integrity as well as to construct a fuel performance data base. The comprehensive utilization program of the KAERI's post-irradiation examination related nuclear facilities such as Post-Irradiation Examination Facility (PIEF), Irradiated Materials Examination Facility (IMEF) and HANARO is described

  4. Post-irradiation examination and R and D programs using irradiated fuels at KAERI

    International Nuclear Information System (INIS)

    Chun, Yong Bum; So, Dong Sup; Lee, Byung Doo; Lee, Song Ho; Min, Duck Kee

    2001-09-01

    This report describes the Post-Irradiation Examination(PIE) and R and D programs using irradiated fuels at KAERI. The objectives of post-irradiation examination (PIE) for the PWR irradiated fuels, CANDU fuels, HANARO fuels and test fuel materials are to verify the irradiation performance and their integrity as well as to construct a fuel performance data base. The comprehensive utilization program of the KAERI's post-irradiation examination related nuclear facilities such as Post-Irradiation Examination Facility (PIEF), Irradiated Materials Examination Facility (IMEF) and HANARO is described

  5. Insulation irradiation test programme for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    McManamy, T.J.; Kanemoto, G.; Snook, P.

    1991-01-01

    In a programme to evaluate the effects of radiation exposure on the electrical insulation for the toroidal field coils of the Compact Ignition Tokamak, three types of boron-free insulation were irradiated at room temperature in the Advanced Technology Reactor (ATR) and tested at the Idaho National Engineering Laboratory. The materials were Spaulrad-S, Shikishima PG5-1 and Shikishima PG3-1. The first two use a bismaleimide resin and the third an aromatic amine hardened epoxy. Spaulrad-S is a two-dimensional (2-D) weave of S-glass, while the others are 3-D weaves of T-glass. Flexure and shear/compression samples were irradiated to ≅ 5 x 10 9 and 3 x 10 10 rad with 35-40% of the total dose from neutrons. The shear/compression samples were tested in pairs by applying an average compression of 345 MPa and then a shear load. After static tests were completed, fatigue testing was performed by cycling the shear load for up to 30000 cycles with a constant compression. The static shear strength of the samples that did not fail was then determined. Generally, shear strengths of the order of 120 MPa were measured. The behaviour of the flexure and shear/compression samples was significantly different; large reductions in the flexure strength were observed, while the shear strength stayed the same or increased slightly. The 3-D weave material demonstrated higher strength and significantly less radiation damage than the 2-D material in flexure but performed almost identically when tested with combined shear and compression. The epoxy system was much more sensitive to fatigue damage than the bismaleimide materials. No swelling was measured; however, the epoxy samples did twist slightly. (author)

  6. Post-irradiation analysis of low enriched U-Mo/Al dispersions fuel miniplate tests, RERTR 4 and 5

    International Nuclear Information System (INIS)

    Hofman, G.L.; Finlay, M.R.; Kim, Y.S.

    2005-01-01

    Interpretation of the post irradiation data of U-Mo/Al dispersion fuel mini plates irradiated in the Advanced Test Reactor to a maximum U-235 burn up of 80% are presented. The analyses addresses fuel swelling and porosity formation as these fuel performance issues relate to fuel fabrication and irradiation parameters. Specifically, mechanisms involved in the formation of porosity observed in the U-Mo/Al interaction phase are discussed and, means of mitigating or eliminating this irradiation phenomenon are offered. (author)

  7. Database on Performance of Neutron Irradiated FeCrAl Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Briggs, Samuel A. [Univ. of Wisconsin, Madison, WI (United States); Littrell, Ken [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Parish, Chad M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    The present report summarizes and discusses the database on radiation tolerance for Generation I, Generation II, and commercial FeCrAl alloys. This database has been built upon mechanical testing and microstructural characterization on selected alloys irradiated within the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) up to doses of 13.8 dpa at temperatures ranging from 200°C to 550°C. The structure and performance of these irradiated alloys were characterized using advanced microstructural characterization techniques and mechanical testing. The primary objective of developing this database is to enhance the rapid development of a mechanistic understanding on the radiation tolerance of FeCrAl alloys, thereby enabling informed decisions on the optimization of composition and microstructure of FeCrAl alloys for application as an accident tolerant fuel (ATF) cladding. This report is structured to provide a brief summary of critical results related to the database on radiation tolerance of FeCrAl alloys.

  8. Application of half-embryo test to identify irradiated fresh fruits

    International Nuclear Information System (INIS)

    Abdelbary, N.A.; EL agamawy, M.R.

    2004-01-01

    Some countries already permit the irradiation of foods to extend its storage life and to control pests, therefore, a faster and significantly more uniform identification method are needed. Half-embryo test is based on the inhibition of shooting due to gamma irradiation since biological systems are sensitive to low doses of gamma irradiation. The intact fruits, apples, lemons, oranges and watermelons were obtained from the local market and irradiated directly with doses of 0.5, 0.75, 1.5 and 3 KGy. Shooting was defined as the elongation of the shoot to the extent of at least 1 mm length in apples and watermelon, while 0.5 mm length in citrus fruits. Root and shoot growth was stimulated most strongly by the addition of benzyladenine (2.5 mg/l) as a growth hormone. Shooting started after 1-3 days and reached to 90 % after 4 days. A long lasting half-embryo test (4-5 days) was capable to discriminate between irradiated and non-irradiated fruits. Growth of half-embryo and the changes were almost the same in all non-irradiated fruits under study. Growth of half-embryo irradiated with a dose of 0.5 KGy or more almost has totally retarded elongation of both root and shoot. Practically, it was observed that small-developed shoots showed slight elongation and afterward they were decayed. If shooting percentage after 1-3 days is less than 20% in apples, 40% in oranges and 30% in lemons and watermelons, the fruits are classified as i rradiated u nder 0.5 KGy as a detection limit dose of the irradiation. Irradiation caused obvious changes in root and shoot growth of half-embryos studied. Roots of non-irradiated half-embryos grew well in all fruits under study and those irradiated with 0.5 KGy or more were obviously reduced. In the same way, shoots of non-irradiated half-embryo grew well and shooting percentage reached to 50 % after 1-2 days and those fruits irradiated with 0.5 KGy or more were reduced. It is recommended to employ the half-embryo test as a practical technique

  9. Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

    Science.gov (United States)

    Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

    2018-04-01

    A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature- and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS). The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code.

  10. IFMIF [International Fusion Materials Irradiation Facility], an accelerator-based neutron source for fusion components irradiation testing: Materials testing capabilities

    International Nuclear Information System (INIS)

    Mann, F.M.

    1988-08-01

    The International Fusion Materials Irradiation Facility (IFMIF) is proposed as an advanced accelerator-based neutron source for high-flux irradiation testing of large-sized fusion reactor components. The facility would require only small extensions to existing accelerator and target technology originally developed for the Fusion Materials Irradiation Test (FMIT) facility. At the extended facility, neutrons would be produced by a 0.1-A beam of 35-MeV deuterons incident upon a liquid lithium target. The volume available for high-flux (>10/sup 15/ n/cm/sup 2/-s) testing in IFMITF would be over a liter, a factor of about three larger than in the FMIT facility. This is because the effective beam current of 35-MeV deuterons on target can be increased by a factor of ten to 1A or more. Such an increase can be accomplished by funneling beams of deuterium ions from the radio-frequency quadruple into a linear accelerator and by taking advantage of recent developments in accelerator technology. Multiple beams and large total current allow great variety in available testing. For example, multiple simultaneous experiments, and great flexibility in tailoring spatial distributions of flux and spectra can be achieved. 5 refs., 2 figs., 1 tab

  11. Performance evaluation of the European Committee Standardization Method EN1785 for the detection of irradiated processed food

    International Nuclear Information System (INIS)

    Tsutsumi, Tomoaki; Adachi, Rika; Takatsuki, Satoshi; Matsuda, Rieko; Teshima, Reiko; Nei, Daisuke; Kameya, Hiromi; Todoriki, Setsuko; Kikuchi, Masahiro; Kobayashi, Yasuhiko

    2014-01-01

    2-dodecylcyclobutanone (DCB) and 2-tetradecylcyclobutanone (TCB) are specific radiolytic products in irradiated lipid-containing food and can be used to detect irradiated foods. We previously reported a performance evaluation test in a single laboratory for qualitative detection methods using DCB and TCB as markers for irradiated foods. EN1785, a European Committee Standardization Method, has been widely used for qualitative detection of DCB and TCB in irradiated foods. Here, we evaluated the applicability of EN1785 to camembert cheese, liquid whole egg, sausage and eel grilled without seasoning by using our evaluation test. In this test, extracted lipids from the food were used as negative samples. The lipids spiked with DCB and TCB were used as positive samples. For each food type, 4 negative and 16 positive samples were analyzed by EN1785. All of the negative samples were judged negative and all of the positive samples were judged positive. Thus, EN1785 should be able to detect irradiation in the tested food. Additionally, to confirm the ability of the validated method to detect irradiated food, the same type of food examined above, both unirradiated and irradiated at doses of 0.5-4 kGy, were analyzed by the method. All of the unirradiated samples were judged negative and all of the irradiated samples were judged positive. (author)

  12. Design Improvements of a Fuel Capsule for Re-irradiation Tests

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  13. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  14. Neutron irradiation test of depleted CMOS pixel detector prototypes

    International Nuclear Information System (INIS)

    Mandić, I.; Cindro, V.; Gorišek, A.; Hiti, B.; Kramberger, G.; Mikuž, M.; Zavrtanik, M.; Hemperek, T.; Daas, M.; Hügging, F.; Krüger, H.; Pohl, D.-L.; Wermes, N.; Gonella, L.

    2017-01-01

    Charge collection properties of depleted CMOS pixel detector prototypes produced on p-type substrate of 2 kΩ cm initial resistivity (by LFoundry 150 nm process) were studied using Edge-TCT method before and after neutron irradiation. The test structures were produced for investigation of CMOS technology in tracking detectors for experiments at HL-LHC upgrade. Measurements were made with passive detector structures in which current pulses induced on charge collecting electrodes could be directly observed. Thickness of depleted layer was estimated and studied as function of neutron irradiation fluence. An increase of depletion thickness was observed after first two irradiation steps to 1 · 10 13 n/cm 2 and 5 · 10 13 n/cm 2 and attributed to initial acceptor removal. At higher fluences the depletion thickness at given voltage decreases with increasing fluence because of radiation induced defects contributing to the effective space charge concentration. The behaviour is consistent with that of high resistivity silicon used for standard particle detectors. The measured thickness of the depleted layer after irradiation with 1 · 10 15 n/cm 2 is more than 50 μm at 100 V bias. This is sufficient to guarantee satisfactory signal/noise performance on outer layers of pixel trackers in HL-LHC experiments.

  15. Design verification test of instrumented capsule (02F-11K) for nuclear fuel irradiation in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Goo; Sohn, J. M.; Oh, J. M. [and others

    2004-01-01

    An instrumented capsule is being developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in HANARO. The instrumented capsule for measuring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. The instrumented capsule includes three test fuel rods installed thermocouple to measure fuel centerline temperature and three SPNDs (Self-Powered Neutron Detector) to monitor the neutron flux. Its stability was verified by out-of-pile performance test, and its safety evaluation was also shown that the safety requirements were satisfied. And then, to verify the design of the instrumented capsule in the test hole, it was successfully irradiated in the test hole of HANARO from March 14, 2003 to June 1, 2003 (53.8 full power days at 24 MWth). During irradiation, the centerline temperature of PWR UO{sub 2} fuel pellets fabricated by KEPCO Nuclear Fuel Company and the neutron flux were continuously measured and monitored. The test fuel rods were irradiated at less than 350 W/cm to 5.13 GWD/MTU with fuel centerline peak temperature below 1,375 .deg. C. The structural stability of the capsule was satisfied by the naked eye in service pool of HANARO. The capsule and test fuel rods were dismantled and test fuel rods were examined at the hot cell of IMEF (Irradiated Material Examination Facility)

  16. Irradiation performance of U-Pu-Zr metal fuels for liquid-metal-cooled reactors

    International Nuclear Information System (INIS)

    Tsai, H.; Cohen, A.B.; Billone, M.C.; Neimark, L.A.

    1994-10-01

    This report discusses a fuel system utilizing metallic U-Pu-Zr alloys which has been developed for advanced liquid metal-cooled reactors (LMRs). Result's from extensive irradiation testing conducted in EBR-II show a design having the following key features can achieve both high reliability and high burnup capability: a cast nominally U-20wt %Pu-10wt %Zr slug with the diameter sized to yield a fuel smear density of ∼75% theoretical density, low-swelling tempered martensitic stainless steel cladding, sodium bond filling the initial fuel/cladding gap, and an as-built plenum/fuel volume ratio of ∼1.5. The robust performance capability of this design stems primarily from the negligible loading on the cladding from either fuel/cladding mechanical interaction or fission-gas pressure during the irradiation. The effects of these individual design parameters, e.g., fuel smear density, zirconium content in fuel, plenum volume, and cladding types, on fuel element performance were investigated in a systematic irradiation experiment in EBR-II. The results show that, at the discharge burnup of ∼11 at. %, variations on zirconium content or plenum volume in the ranges tested have no substantial effects on performance. Fuel smear density, on the other hand, has pronounced but countervailing effects: increased density results in greater cladding strain, but lesser cladding wastage from fuel/cladding chemical interaction

  17. Continuous parameter determination of irradiated nuclear fuels in the test-reactor

    International Nuclear Information System (INIS)

    Bevilacqua, A.; Junod, E.; Mas, P.; Perdreau, R.

    1977-01-01

    During the irradiation tests of nuclear fuels, the flux level may often be variable by shifting the loops in a high neutron-gradient. So integral fluence measurements are no longer sufficient. The self-powered neutron detectors allow to finely scan instantaneous fluxes. More than 100 such SPN detectors are used on the experiments irradiated in the SILOE reactor. The treatment of the large amount of information is following. A first minicomputer scans all the measurement lines at a variable frequence (10 min to 1 hr) and writes rough voltage values on a magnetic disk. A second computer does a sorting of these values for each set of SPND corresponding to an experiment. At the present time, the main treatment is performed by batch processing by some FORTRAN codes to get time-evolving quantities, such as effective flux, fission power, burn-up, fission product activities, etc. The future development of the system will allow some of these calculations to be performed directly on the second computer in such a manner to control the movements of the loops automatically in view of a given irradiation program

  18. Comparison of DNA comet assay and germination test (half-embryo-test) in gamma-irradiated cherry seeds

    International Nuclear Information System (INIS)

    Todoroki, Setsuko; Hayashi, Toru

    2002-01-01

    Cherry fruits were irradiated with gamma-rays at doses up to 200Gy (effective dose for disinfestation of codling moth), and DNA strand break in seed embryos was investigated by using alkaline comet assay. Immediately after irradiation (≥100Gy), DNA from embryos produced comets with a long and wide tail due to fragmentation. In control cells, DNA relaxed and produced comet with very short tail (with few strand break). After 72h storage, DNA from fruits irradiated at 200 Gy showed comets with little tail and tail moment of comets was same as un-irradiated control. These results indicate that the strand breaks of DNA caused by irradiation in fresh seed embryo are repaired during storage. On the contrary, the ability of germination lost by irradiation did not restored, a dose of 100Gy and more retarded shoot elongation. In cherries irradiated at 100Gy, the shooting percentage was less than 50% at 4th day after incubation. Germination test (Half embryo test) can be discriminate between irradiated and un-irradiated cherries. (author)

  19. Recent irradiation tests for future nuclear system at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Man Soon; Choo, Kee Nam; Yang, Seong Woo; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-05-15

    The capsule at HANARO is a device that evaluates the irradiation effects of nuclear materials and fuels, which can reproduce the environment of nuclear power plants and accelerate to reach to the end of life condition. As the integrity assessment and the extension of lifetime of nuclear power plants are recently considered as important issues in Korea, the requirements for irradiation test are gradually being increased. The capacity and capability irradiation tests at HANARO are becoming important because Korea strives to develop SFR (Sodium-cooled Fast Reactor) and VHTR (Very High Temperature Reactor) among the future nuclear system and to export the research reactors and to develop the fusion reactor technology.

  20. Performance evaluation of UO2-Zr fuel in power ramp tests

    International Nuclear Information System (INIS)

    Knudsen, P.; Bagger, C.

    1977-01-01

    In power reactors using UO 2 -Zr fuel, rapid power increases may lead to failures in fuel pins that have been irradiated at steady or decreasing heat loads. This paper presents results which extend the experience with power ramp performance of high burn-up fuel pins. A test fuel element containing both pellet and vipac UO 2 -Zr fuel pins was irradiated in the HBWR at Halden for effectively 2 1/2 years to an average burn-up of 21,000 MWD/te UO 2 at gradually decreasing power levels. The subsequent non-destructive characterization revealed formation of transverse cracks in the vipac fuel columns. After the HBWR irradiation, five of the fuel pins were power ramp tested individually in the DR 3 Reactor at Riso. The ramp rates in this test series were in the range 3-60 W/cm min. The maximum local heat loads seen in the ramp tests were 20-120% above the highest levels experienced at the same axial positions during the HBWR irradiation. Three pellets and one vipac fuel pin failed, whereas another vipac pin gave no indication of clad penetration. Profilometry after the ramp testing indicated the formation of small ridges for both types of fuel pins. For vipac fuel, the ridges were less regularly distributed along the pin length than for pellet fuel. Neutron radiography revealed the formation of additional transverse and longitudinal fuel cracks during the power ramps for both types of fuel pins. The observed failures seemed to be marginal since little or no indication as to the locations of the clad penetrations could be derived from the non-destructive post-irradiation examinations. The cases have been analyzed by means of the Danish fuel performance codes. The calculations, which are in general agreement with the observations, are discussed. The results of the investigations indicate qualitative similarities in over power performance of the two fuel types

  1. On-line irradiation testing of a Giant Magneto-Resistive (GMR) sensor

    Energy Technology Data Exchange (ETDEWEB)

    Olfert, J.; Luloff, B.; MacDonald, D.; Lumsden, R., E-mail: jeff.olfert@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    Magneto-resistive sensors are rapidly gaining favour for magnetic field sensing applications owing to their high sensitivity, small size, and low cost. Their metallic, nonsemiconductor construction makes them excellent candidates for use in the harsh environments present in nuclear and space applications. In this work, a commercially available magneto-resistive sensor was irradiated up to a total gamma dose of 2 MGy (200 Mrad), and online testing was performed to monitor the sensor throughout the irradiation to detect any degradation. No significant evidence of degradation of the sensor characteristics was observed. A very small (< 1%) change in the bridge balance of the sensor as a function of accumulated dose was detected. (author)

  2. Irradiation tests of radiation resistance optical fibers for fusion diagnostic application

    Science.gov (United States)

    Kakuta, Tsunemi; Shikama, Tatsuo; Nishitani, Takeo; Yamamoto, Shin; Nagata, Shinji; Tsuchiya, Bun; Toh, Kentaro; Hori, Junichi

    2002-11-01

    To promote development of radiation-resistant core optical fibers, the ITER-EDA (International Thermonuclear Experimental Reactor-Engineering Design Activity) recommended carrying out international round-robin irradiation tests of optical fibers to establish a reliable database for their applications in the ITER plasma diagnostics. Ten developed optical fibers were irradiation-tested in a Co-60 gamma cell, a Japan Materials Testing Reactor (JMTR). Also, some of them were irradiation tested in a fast neutron irradiation facility of FNS (Fast Neutron Source), especially to study temperature dependence of neutron-associated irradiation effects. Included were several Japanese fluorine doped fibers and one Japanese standard fiber (purified and undoped silica core), as well as seven Russian fibers. Some of Russian fibers were drawn by Japanese manufactures from Russian made pre-form rods to study effects of manufacturing processes to radiation resistant properties. The present paper will describe behaviors of growth of radiation-induced optical transmission loss in the wavelength range of 350-1750nm. Results indicate that role of displacement damages by fast neutrons are very important in introducing permanent optical transmission loss. Spectra of optical transmission loss in visible range will depend on irradiation temperatures and material parameters of optical fibers.

  3. Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

    1986-09-01

    Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed

  4. Performance degradation of ferrofluidic feedthroughs in a mixed irradiation field

    Science.gov (United States)

    Simos, Nikolaos; Fernandes, S.; Mittig, Wolfgang; Pellemoine, Frederique; Avilov, M.; Kostin, M.; Mausner, L.; Ronningen, R.; Schein, M.; Bollen, G.

    2017-01-01

    Ferrofluidic feedthrough (FF) rotary seals containing either NdFeB or SmCo-type permanent magnets have been considered for use in the target and beam dump systems of the Facility for Rare Isotope Beams (FRIB). To evaluate their performance under irradiation three FF seals were irradiated in a mixed field consisting of fast neutrons, protons and γ-rays to an average absorbed dose of 0.2, 2.0, and 20.0 MGy at the Brookhaven Linac Isotope Producer facility (BLIP). The radiation types and energy profiles mimic those expected at the FRIB facility. Degradation of the operational performance of these devices due to irradiation is expected to be the result of the de-magnetization of the permanent magnets contained within the seal and the changes in the ferrofluid properties. Post-irradiation performance was evaluated by determining the ferrofluidic seal vacuum tightness and torque under static and dynamic conditions. The study revealed that the ferrofluidic feedthrough seal irradiated to a dose of 0.2 MGy maintained its vacuum tightness under both static and rotational condition while the one irradiated to a dose of 2.0 MGy exhibited signs of ferrofluid damage but no overall performance loss. At 20 MGy dose the effects of irradiation on the ferrofluid properties (viscosity and particle agglomeration) were shown to be severe. Furthermore, limited de-magnetization of the annular shaped Nd2Fe14B and Sm2Co17 magnets located within the irradiated FFs was observed for doses of 0.2 MGy and 20 MGy respectively.

  5. Performance of irradiated silicon microstrip detectors

    International Nuclear Information System (INIS)

    Catacchini, E.; Civinini, C.; D'Alessandro, R.; Focardi, E.; Lenzi, M.; Meschini, M.; Parrini, G.; Pieri, M.

    1999-01-01

    Silicon microstrip devices to be installed in Large Hadron Collider (LHC) tracking detectors will have to operate in a high radiation environment. We report on performance studies of silicon microstrip detectors irradiated with neutrons or protons, up to fluences comparable to the first ten years of running at LHC. Obtained results show that irradiated detectors can still be operated with satisfactory signal-to-noise ratio,and in the case of inhomogeneously type inverted detector a very good position resolution is achieved regardless of the zone crossed by the particle

  6. Irradiated cocoa tested in the wing spot assay in Drosophila melanogaster

    International Nuclear Information System (INIS)

    Zimmering, S.; Olvera, O.; Cruces, M.P.; Pimentel, E.; Arceo, C.; Rosa, M.E. de la; Guzman, J.

    1992-01-01

    The result of treatment of Drosophila melanogaster with irradiated cocoa as scored in the somatic wing spot test is described. The test has been used previously in the evaluation of irradiated food and has registrated a significantly greater number of positives among chemicals tested than germ line counterparts. Irradiated cocoa has thus far been reported negative in other mutagenicity assays including those employing salmonella and Drosophila germ cells and mammalian cells. The wing spot test as described in Graf et al. was employed. Females of the genotype mwh were mated with flr 3 /TM3; Ser males. (author). 9 refs.; 1 tab

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

    International Nuclear Information System (INIS)

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

    1977-08-01

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

  8. East Area Irradiation Test Facility: Preliminary FLUKA calculations

    CERN Document Server

    Lebbos, E; Calviani, M; Gatignon, L; Glaser, M; Moll, M; CERN. Geneva. ATS Department

    2011-01-01

    In the framework of the Radiation to Electronics (R2E) mitigation project, the testing of electronic equipment in a radiation field similar to the one occurring in the LHC tunnel and shielded areas to study its sensitivity to single even upsets (SEU) is one of the main topics. Adequate irradiation test facilities are therefore required, and one installation is under consideration in the framework of the PS East area renovation activity. FLUKA Monte Carlo calculations were performed in order to estimate the radiation field which could be obtained in a mixed field facility using the slowly extracted 24 GeV/c proton beam from the PS. The prompt ambient dose equivalent as well as the equivalent residual dose rate after operation was also studied and results of simulations are presented in this report.

  9. Irradiation test plan of instrumented capsule(05F-01K) for nuclear fuel irradiation in Hanaro (Revision 1)

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Jae Min; Kim, B. G.; Choi, M. H. (and others)

    2006-09-15

    An instrumented capsule was developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel pellet elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in HANARO. The instrumented capsule for measuring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. And then, to verify the design of the instrumented capsule in the test hole, it was successfully irradiated in the test hole of HANARO from March 14, 2003 to June 1, 2003 (53.84 full power days at 24 MW). In the year of 2004, 3 test fuel rods and the 03F-05K instrumented fuel capsule were designed and fabricated to measure fuel centerline temperature, internal pressure of fuel rod, and fuel axial deformation during irradiation test. Now, this capsule was successfully irradiated in the test hole OR5 of HANARO reactor from April 27, 2004 to October 1, 2004 (59.5 full power days at 24-30 MW). The capsule and fuel rods have been be dismantled and fuel rods have been examined at the hot cell of IMEF. The instrumented fuel capsule (05F-01K) was designed and manufactured for a design verification test of the dual instrumented fuel rods. The irradiation test of the 05F-01K instrumented fuel capsule will be carried out at the OR5 vertical experimental hole of HANARO.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-03-15

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

  11. Study on irradiation of freshening ginseng and toxicity test

    International Nuclear Information System (INIS)

    Wang Ziwen; Xu Dechun; Yang Wanqi

    1991-01-01

    The ginsengs irradiated by 1 or 2 kGy of γ-rays have been stored for 6 months under room temperature. Its freshening rates was 86.67% and 88.33% respectively. The saponin content was maintained. The irradiated ginsengs had the vigour of sap fully and beautiful colour. Therefore they can be stored much longer for sell. The toxicity test showed that there was no toxicity for irradiated ginsengs

  12. Irradiation Performance of HTGR Fuel in WWR-K Research Reactor

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-01-01

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

  14. Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

    2018-04-01

    A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature-and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS).The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code. (c) 2018 Elsevier B.V. All rights reserved.

  15. Status of development and irradiation performance of advanced proliferation resistant MTR fuel at NUKEM

    International Nuclear Information System (INIS)

    Hrovat, M.; Hassel, H.-W.; Wehner, E.

    1985-01-01

    This paper describes the current status of development and irradiation performance of fuel elements for Material Test and Research (MTR) Reactors with Medium Enriched Uranium (MEU, ≤ 45 % 235-U) and Low Enriched Uranium (LEU, ≤ 20 % 235-U). (author)

  16. Testbeam and laboratory test results of irradiated 3D CMS pixel detectors

    Energy Technology Data Exchange (ETDEWEB)

    Bubna, Mayur [Purdue University, Department of Physics, West Lafayette, IN 47907-1396 (United States); Purdue University, School of Electrical and Computer Engineering, West Lafayette, IN 47907-1396 (United States); Alagoz, Enver, E-mail: enver.alagoz@cern.ch [Purdue University, Department of Physics, West Lafayette, IN 47907-1396 (United States); Cervantes, Mayra; Krzywda, Alex; Arndt, Kirk [Purdue University, Department of Physics, West Lafayette, IN 47907-1396 (United States); Obertino, Margherita; Solano, Ada [Istituto Nazionale di Fisica Nucleare, Sezione di Torino, 10125 Torino (Italy); Dalla Betta, Gian-Franco [INFN Padova (Gruppo Collegato di Trento) (Italy); Dipartimento di Ingegneria e Scienzadella Informazione, Universitá di Trento, I-38123 Povo di Trento (Italy); Menace, Dario; Moroni, Luigi [Istituto Nazionale di Fisica Nucleare, Sezione di Milano Bicocca (Italy); Universitá degli Studi di Milano Bicocca, 20126 Milano (Italy); Uplegger, Lorenzo; Rivera, Ryan [Fermi National Accelerator Laboratory, Batavia, IL 60510-0500 (United States); Osipenkov, Ilya [Texas A and M University, Department of Physics, College Station, TX 77843-4242 (United States); Andresen, Jeff [Fermi National Accelerator Laboratory, Batavia, IL 60510-0500 (United States); Bolla, Gino; Bortoletto, Daniela [Purdue University, Department of Physics, West Lafayette, IN 47907-1396 (United States); Boscardin, Maurizio [Centro per i Materiali e i Microsistemi Fondazione Bruno Kessler (FBK), Trento, I-38123 Povo di Trento (Italy); Marie Brom, Jean [Strasbourg IPHC, Institut Pluriedisciplinaire Hubert Curien, F-67037 Strasbourg Cedex (France); Brosius, Richard [State University of New York at Buffalo (SUNY), Department of Physics, Buffalo, NY 14260-1500 (United States); Chramowicz, John [Fermi National Accelerator Laboratory, Batavia, IL 60510-0500 (United States); and others

    2013-12-21

    The CMS silicon pixel detector is the tracking device closest to the LHC p–p collisions, which precisely reconstructs the charged particle trajectories. The planar technology used in the current innermost layer of the pixel detector will reach the design limit for radiation hardness at the end of Phase I upgrade and will need to be replaced before the Phase II upgrade in 2020. Due to its unprecedented performance in harsh radiation environments, 3D silicon technology is under consideration as a possible replacement of planar technology for the High Luminosity-LHC or HL-LHC. 3D silicon detectors are fabricated by the Deep Reactive-Ion-Etching (DRIE) technique which allows p- and n-type electrodes to be processed through the silicon substrate as opposed to being implanted through the silicon surface. The 3D CMS pixel devices presented in this paper were processed at FBK. They were bump bonded to the current CMS pixel readout chip, tested in the laboratory, and testbeams carried out at FNAL with the proton beam of 120 GeV/c. In this paper we present the laboratory and beam test results for the irradiated 3D CMS pixel devices. -- Highlights: •Pre-irradiation and post-irradiation electrical properties of 3D sensors and 3D diodes from various FBK production batches were measured and analyzed. •I–T measurements of gamma irradiated diodes were analyzed to understand leakage current generation mechanism in 3D diodes. •Laboratory measurements: signal to noise ratio and charge collection efficiency of 3D sensors before and after irradiation. •Testbeam measurements: pre- and post-irradiation pixel cell efficiency and position resolution of 3D sensors.

  17. Irradiation tests of optoelectronic components for LHC inner-detectors

    International Nuclear Information System (INIS)

    Dawson, I.; Oglesby, S.J.; Dowell, J.D.; Homer, R.J.; Kenyon, I.R.; Shaylor, H.R.; Wilson, J.A.

    1997-01-01

    Two kinds of optical-link technologies have been investigated for the readout of data at LHC experiments; one based on LEDs and the other on Multi-Quantum-Well modulators. Presented in this paper are the results of irradiating LEDs and MQW modulators with 1 MeV-equivalent neutrons and 24 GeV protons. The devices were biased and the performances of the optical links were monitored throughout the tests. The fluences achieved were ∝5 x 10 14 n cm -2 and ∝6 x 10 13 p cm -2 . (orig.)

  18. Pre-irradiation tests on U-Si alloys

    International Nuclear Information System (INIS)

    Howe, L.M.; Bell, L.G.

    1958-05-01

    Pre-irradiation tests of hardness, density, electrical resistivity, and corrosion resistance as well as metallographic and X-ray examinations were undertaken on U-Si core material, which had been co-extruded in Zr--2, in order that the effect of irradiation on alloys in the epsilon range could be assessed. In addition, a study of the epsilonization of arc-melted material was undertaken in order to rain familiarity with the epsilonization process and to obtain information on the corrosion behaviour of epsilonized material. Sheathed U-Si samples in the epsilonized and de-epsilonized conditions have been irradiated in the X-2 loop, with a water temperature of 275 o C. The samples have been examined after 250 MWD/Tonne and show no dimensional change. (author)

  19. Irradiation test of FPGA for BES III

    International Nuclear Information System (INIS)

    Chen Yixin; Liang Hao; Xue Jundong; Liu Baoying; Liu Qiang; Yu Xiaoqi; Zhou Yongzhao; Hou Long

    2005-01-01

    The irradiation effect of FPGA, applied in Front-end Electronics for experiments of High-Energy Physics, is a serious problem. The performance of FPGA, used in the front-end card of Muon Counters of BES III project, needs to be evaluated under irradiation. SEUs on Altera ACEX 1K FPGA, observed in the experiment under the irradiation of γ ray, 14 and 2.5 MeV neutrons, was investigated. The authors calculated involved cross-section and provided reasonable analysis and evaluation for the result of the experiment. The conclusion about feasibility of applying ACEX 1K FPGA in the front-end card of the readout system of Muon Counters for BES III was given. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-03-15

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  2. National food irradiation programme of Japan

    International Nuclear Information System (INIS)

    Fujimaki, M.

    1982-01-01

    The present state of studies on feasibility and wholesomeness of irradiated food is presented. Irradiation projects were realized of potatoes, onions, wheat, Vienna sausages, fish-paste products, and mandarine oranges. Mutagenecity tests with Salmonella or E. coli, chromosome aberration tests, dominant lethal tests, fibroblasts and micronucleus tests, and toxicity tests performed in amimals fed with irradiated food showed no positive results

  3. Development of post-irradiation test facility for domestic production of 99Mo

    International Nuclear Information System (INIS)

    Taguchi, Taketoshi; Yonekawa, Minoru; Kato, Yoshiaki; Kurosawa, Makoto; Nishikata, Kaori; Ishida, Takuya; Kawamata, Kazuo

    2013-01-01

    JMTR focus on the activation method. By carrying out the preliminary tests using irradiation facilities existing, and verification tests using the irradiation facility that has developed in the cutting-edge research and development strategic strengthening business, as irradiation tests towards the production of 99 Mo, we have been conducting research and development that can contribute to supply about 25% for 99 Mo demand in Japan and the stable supply of radiopharmaceutical. This report describes a summary of the status of the preliminary tests for the production of 99 Mo: Maintenance of test equipment in the facility in JMTR hot laboratory in preparation for research and development for the production of 99 Mo in JMTR and using MoO 3 pellet irradiated at Kyoto University Research Reactor Institute (KUR). (author)

  4. Developing Ultra-small Scale Mechanical Testing Methods and Microstructural Investigation Procedures for Irradiated Materials

    Energy Technology Data Exchange (ETDEWEB)

    Hosemann, Peter; Kaoumi, Djamel

    2018-04-02

    -beam irradiations have been utilized for decades to foster the understanding of materials’ behavior under radiation, and significant efforts at comparing ion-beam irradiations to neutron irradiations are ongoing [1]. While extensive microstructural and chemical characterizations of neutron-irradiated and ion-irradiated materials are essential to the understanding of the underlying physics of materials’ degradation in nuclear environments, the ultimate test is the mechanical performance of a material under the anticipated condition, since it is the final criterion for a material to be accepted for use in a specific nuclear component. Again, standard, large-scale, bulk evaluations are key for the licensing of materials in a specific component, but additional, more basic scientific testing can accelerate the process by targeting specific areas of interest. Small-scale mechanical testing has been applied on nuclear materials for decades [2]. Traditionally the driving forces to use non-standard-size samples are the limited space in reactors, the availability of new alloys, and a reduction in radioactive-materials volume. Shear punch testing [3,5], sub-sized micro tensile testing [4], sub-sized compact tension and charpy testing [6,7], micro bulge testing [8], and micro hardness testing [3] have been used. Small-scale mechanical testing also allows the targeting of specific regions of interest, be they single grains to evaluate a specific deformation mechanism [9], grain boundaries, heat-affected zones in welds, or any other specific critical area of interest. With further reducing of the sample size, it also holds the promise to obtain quantitative data from ion-beam irradiations and to compare such data to the microstructural changes observed. Over the last few decades, a number of small-scale mechanical characterization techniques have been developed and utilized for irradiated materials. In addition to the above-mentioned sample test techniques at the mm and sub mm length scale

  5. Testing capabilities of Los Alamos National Laboratory for irradiated materials

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  6. Estimation of γ irradiation induced genetic damage by Ames test

    International Nuclear Information System (INIS)

    Hosoda, Eiko

    1999-01-01

    Mutation by 60 Co γ irradiation was studied in five different histidine-requiring auxotrophs of Salmonella typhimurium. The strains TA98 (sensitive to frameshift) and TA100 (sensitive to base-pair substitution) were irradiated (10-84 Gy and 45-317 Gy, respectively) and revertants were counted. TA98 exhibited radiation-induced revertants, 2.8 fold of spontaneous revertants, although no significant increase was detected in TA100. Then, three other frameshift-sensitive strains TA1537, TA1538 and TA94 were irradiated in a dose of 61-167 Gy. Only in TA94, revertants increased 3.5 fold. Since spontaneous revertants are known to be independent of cell density, a decrease of bacterial number by γ irradiation was confirmed not to affect the induced revertants by dilution test. Thus the standard Ames Salmonella assay identified γ irradiation was confirmed not to affect the induced revertants by dilution test. Thus the standard Ames Salmonella assay identified γ irradiation as a mutagenetic agent. The mutagenicity of dinitropyrene, a mutagen widely existing in food, and dismutagenicity of boiling water insoluble fraction of Hizikia fusiforme, edible marine alga, were tested on γ induced revertant formation in TA98 and TA94. Dinitropyrene synergistically increased γ induced revertants and Hizikia insoluble fraction reduced the synergistic effect of dinitropyrene dependently on the concentration. (author)

  7. A Study on the High Temperature Irradiation Test Possibility for the HANARO Outer Core Region

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-01-15

    1. Information on the neutron flux levels and the gamma heat of the concerned test holes, which have been produced from a series of nuclear analysis and tests performed at KAERI since 1993, were collected and analyzed to develop the nuclear data for the concerned test holes of HANARO and to develop the new design concepts of a capsule for the high temperature irradiation devices. 2. From the literature survey and analysis about the system design characteristics of the new concepts of irradiation devices in the ATR and MIT reactor, U.S. and the JHR reactor, France, which are helpful in understanding the key issues for the on-going R and D programmes related to a SFR and a VHTR, the most important parameters for the design of high temperature irradiation devices are identified as the neutron spectrum, the heat generation density, the fuel and cladding temperature, and the coolant chemistry. 3. From the thermal analysis of a capsule by using a finite element program ANSYS, high temperature test possibility at the OR and IP holes of HANARO was investigated based on the data collected from a literature survey. The OR holes are recommended for the tests of the SFR and VHTR nuclear materials. The IP holes could be applicable for an intermediate temperature irradiation of the SWR and LMR materials. 4. A thermal analysis for the development of a capsule with a new configuration was also performed. The size of the center hole, which is located at the thermal media of a capsule, did not cause specimen temperature changes. The temperature differences are found to be less than 2%. The introduction of an additional gap in the thermal media was able to contribute to an increase in the specimen temperature by up to 27-90 %.

  8. Identification of gamma irradiated apples by the half-embryo test

    International Nuclear Information System (INIS)

    Miranda, Gabriel C.; Bujan, Alfonso; Leiva, Carlos H.; Yusef, Maria V.

    2003-01-01

    The half-embryo test was applied to irradiated apples (var. Red delicious).The irradiation of apples caused obvious changes in the growth of the half-embryo. A dose of 100 Gy or more, inhibits the epicotyl development and with 50 Gy dose is possible to observe a great contrast with the non-irradiated apples. If the epicotyl development is less than 4 cm., the apples are identified as irradiated. The assessment can be made after 7 days. (author)

  9. Proton and Neutron Irradiation Tests of Readout Electronics of the ATLAS Hadronic Endcap Calorimeter

    CERN Document Server

    Menke, Sven; The ATLAS collaboration

    2012-01-01

    The readout electronics of the ATLAS Hadronic Endcap Calorimeter will have to withstand the about ten times larger radiation environment of the future high-luminosity LHC (HL-LHC) compared to their design values. The GaAs ASIC which comprises the heart of the readout electronics has been exposed to neutron and proton radiation with fluences up to ten times the total expected fluences for ten years of running of the HL-LHC. Neutron tests where performed at the NPI in Rez, Czech Republic, where a 36 MeV proton beam is directed on a thick heavy water target to produce neutrons. The proton irradiation was done with 200 MeV protons at the PROSCAN area of the Proton Irradiation Facility at the PSI in Villigen, Switzerland. In-situ measurements of S-parameters in both tests allow the evaluation of frequency dependent performance parameters - like gain and input impedance - as a function of the fluence. The linearity of the ASIC response has been measured directly in the neutron tests with a triangular input pulse of...

  10. Proton and Neutron Irradiation Tests of Readout Electronics of the ATLAS Hadronic Endcap Calorimeter

    CERN Document Server

    INSPIRE-00106910

    2012-01-01

    The readout electronics of the ATLAS Hadronic Endcap Calorimeter will have to withstand the about ten times larger radiation environment of the future high-luminosity LHC (HL-LHC) compared to their design values. The GaAs ASIC which comprises the heart of the readout electronics has been exposed to neutron and proton radiation with fluences up to ten times the total expected fluences for ten years of running of the HL-LHC. Neutron tests were performed at the NPI in Rez, Czech Republic, where a 36 MeV proton beam is directed on a thick heavy water target to produce neutrons. The proton irradiation was done with 200 MeV protons at the PROSCAN area of the Proton Irradiation Facility at the PSI in Villigen, Switzerland. In-situ measurements of S-parameters in both tests allow the evaluation of frequency dependent performance parameters - like gain and input impedance - as a function of the fluence. The linearity of the ASIC response has been measured directly in the neutron tests with a triangular input pulse of ...

  11. Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

    Energy Technology Data Exchange (ETDEWEB)

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01

    Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

  12. Results from the CDE phase activity on neutron dosimetry for the international fusion materials irradiation facility test cell

    CERN Document Server

    Esposito, B; Maruccia, G; Petrizzi, L; Bignon, G; Blandin, C; Chauffriat, S; Lebrun, A; Recroix, H; Trapp, J P; Kaschuck, Y

    2000-01-01

    The international fusion materials irradiation facility (IFMIF) project deals with the study of an accelerator-based, deuterium-lithium source, producing high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials for fusion energy reactors. IFMIF would also provide calibration and validation of data from fission reactor and other accelerator based irradiation tests. This paper describes the activity on neutron/gamma dosimetry (necessary for the characterization of the specimens' irradiation) performed in the frame of the IFMIF conceptual design evaluation (CDE) neutronics tasks. During the previous phase (conceptual design activity (CDA)) the multifoil activation method was proposed for the measurement of the neutron fluence and spectrum and a set of suitable foils was defined. The cross section variances and covariances of this set of foils have now been used for tests on the sensitivity of the IFMIF neutron spectrum determination to cross section uncertainties...

  13. GfW-handbook for irradiation test guidelines for radiation hardness of electronic components

    International Nuclear Information System (INIS)

    Braeunig, D.; Wulf, F.; Gaebler, W.; Boden, A.

    1982-12-01

    The purpose of the report is to propose irradiation test methods so that a standardized application of the methods can lead to a better comparison of test results. The interaction of different radiation species with matter - ionization and displacement - is described. Application of appropriate radiation sources, dosimetry problems, and shielding for simulating space radiation effects by laboratory testing is discussed. The description and characteristics of the irradiation sources are presented. Flowcharts of the planning and running of irradiation tests are given. Guidelines for running the tests are established, test methods and test circuits are proposed. The test system offers the capability of measuring devices also of high complexity up to microprocessors. The test results are collected regularly and are published in GfW-Handbook TN53/08, 'Data Compilation of Irradiation Tested Electronic Components'. (orig./HP) [de

  14. Quality engineering in FFTF irradiation tests

    International Nuclear Information System (INIS)

    Caplinger, W.H.

    1980-01-01

    The design and fabrication of an irradiation test for the Fast Flux Test Facility are planned, controlled and documented in accordance with the Department of Energy standards. Tests built by Westinghouse Hanford Company are further controlled and guided by a series of increasingly specific documents, including guidelines for program control, procedures for engineering operations, standard practices and detailed operating procedures. In response to this guidance, a series of five documents is prepared covering each step of the experiment from conception through fabrication and assembly. This paper describes the quality assurance accompanying these five steps

  15. Evaluation of toughness degradation by small punch (SP) tests for neutron irradiated structural steels

    International Nuclear Information System (INIS)

    Misawa, Toshihei; Hamaguchi, Yoshikazu; Kimura, Akihiko; Eto, Motokuni; Suzuki, Masahide; Nakajima, Nobuya.

    1992-01-01

    The small punch (SP) test as one of the useful small specimen testing technique (SSTT) has been developed to evaluate the fracture toughness, ductile-brittle transition temperature (DBTT) and tensile properties for neutron irradiated structural materials. The SP tests using the miniaturized specimens of φ3 mm TEM disk and 10 mm 2 coupon were performed for six kinds of ferritic steels of F-82, F-82H, HT-9, JFMS, 2.25-1Mo and SQV2A. It was shown that the temperature dependence of SP fracture energies with scatter in miniaturized testing can give reliable information on the DBTT by use of the statistical analysis based on the Weibull distribution. A good correlation between the DBTT of the SP tests and that of the standard CVN test has been obtained for the various nuclear ferritic steels. The SP test was performed for cryogenic austenitic steels as a way of evaluating elastic-plastic fracture toughness, J IC , on the basis of a universal empirical relationship between J IC and SP equivalent fracture strain, ε-bar qf . The SP testing using the neutron irradiated specimens of 2.25Cr-1Mo, F-82, F-82H and HT-9 steels was successfully applied and presented the neutron radiation induced changes on the DBTT, fracture toughness and tensile properties. (author)

  16. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    Energy Technology Data Exchange (ETDEWEB)

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  17. HIGH-TEMPERATURE SAFETY TESTING OF IRRADIATED AGR-1 TRISO FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Stempien, John D.; Demkowicz, Paul A.; Reber, Edward L.; Chrisensen, Cad L.

    2016-11-01

    High-Temperature Safety Testing of Irradiated AGR-1 TRISO Fuel John D. Stempien, Paul A. Demkowicz, Edward L. Reber, and Cad L. Christensen Idaho National Laboratory, P.O. Box 1625 Idaho Falls, ID 83415, USA Corresponding Author: john.stempien@inl.gov, +1-208-526-8410 Two new safety tests of irradiated tristructural isotropic (TRISO) coated particle fuel have been completed in the Fuel Accident Condition Simulator (FACS) furnace at the Idaho National Laboratory (INL). In the first test, three fuel compacts from the first Advanced Gas Reactor irradiation experiment (AGR-1) were simultaneously heated in the FACS furnace. Prior to safety testing, each compact was irradiated in the Advanced Test Reactor to a burnup of approximately 15 % fissions per initial metal atom (FIMA), a fast fluence of 3×1025 n/m2 (E > 0.18 MeV), and a time-average volume-average (TAVA) irradiation temperature of about 1020 °C. In order to simulate a core-conduction cool-down event, a temperature-versus-time profile having a peak temperature of 1700 °C was programmed into the FACS furnace controllers. Gaseous fission products (i.e., Kr-85) were carried to the Fission Gas Monitoring System (FGMS) by a helium sweep gas and captured in cold traps featuring online gamma counting. By the end of the test, a total of 3.9% of an average particle’s inventory of Kr-85 was detected in the FGMS traps. Such a low Kr-85 activity indicates that no TRISO failures (failure of all three TRISO layers) occurred during the test. If released from the compacts, condensable fission products (e.g., Ag-110m, Cs-134, Cs-137, Eu-154, Eu-155, and Sr-90) were collected on condensation plates fitted to the end of the cold finger in the FACS furnace. These condensation plates were then analyzed for fission products. In the second test, five loose UCO fuel kernels, obtained from deconsolidated particles from an irradiated AGR-1 compact, were heated in the FACS furnace to a peak temperature of 1600 °C. This test had two

  18. Flexural strength of proof-tested and neutron-irradiated silicon carbide

    Science.gov (United States)

    Price, R. J.; Hopkins, G. R.

    1982-08-01

    Proof testing before service is a valuable method for ensuring the reliability of ceramic structures. Silicon carbide has been proposed as a very low activation first-wall and blanket structural material for fusion devices, where it would experience a high flux of fast neutrons. Strips of three types of silicon carbide were loaded in four-point bending to a stress sufficient to break about a third of the specimens. Groups of 16 survivors were irradiated to 2 × 10 26n/ m2 ( E>0.05 MeV) at 740°C and bend tested to failure. The strength distribution of chemically vapor-deposited silicon carbide (Texas Instruments) was virtually unchanged by irradiation. The mean strength of sintered silicon carbide (Carborundum Alpha) was reduced 34% by irradiation, while the Weibull modulus and the truncated strength distribution characteristic of proof-tested material were retained. Irradiation reduced the mean strength of reaction-bonded silicon carbide (Norton NC-430) by 58%, and the spread in strength values was increased. We conclude that for the chemically vapor-deposited and the sintered silicon carbide the benefits of proof testing to eliminate low strength material are retained after high neutron exposures.

  19. Eurados trial performance test for photon dosimetry

    DEFF Research Database (Denmark)

    Stadtmann, H.; Bordy, J.M.; Ambrosi, P.

    2001-01-01

    Within the framework of the EURADOS Action entitled Harmonisation and Dosimetric Quality Assurance in Individual Monitoring for External Radiation, trial performance tests for whole-body and extremity personal dosemeters were carried out. Photon, beta and neutron dosemeters were considered....... This paper summarises the results of the whole-body photon dosemeter test. Twenty-six dosimetry services from all EU Member States and Switzerland participated. Twelve different radiation fields were used to simulate various workplace irradiation fields. Dose values from 0.4 mSv to 80 mSv were chosen. From...

  20. Irradiation and beam tests qualification for ATLAS IBL Pixel Modules

    International Nuclear Information System (INIS)

    Rubinskiy, Igor

    2013-01-01

    The upgrade for the ATLAS detector will have different steps towards HL-LHC. The first upgrade for the Pixel Detector will consist in the construction of a new pixel layer which will be installed during the first shutdown of the LHC machine (foreseen for 2013–2014). The new detector, called Insertable B-Layer (IBL), will be inserted between the existing Pixel Detector and a new (smaller radius) beam-pipe at a radius of 33 mm. The IBL will require the development of several new technologies to cope with the increase in the radiation damage and the pixel occupancy and also to improve the physics performance, which will be achieved by reduction of the pixel size and of the material budget. Two different promising silicon sensor technologies (Planar n-in-n and 3D) are currently under investigation for the Pixel Detector. An overview of the sensor technologies' qualification with particular emphasis on irradiation and beam tests is presented. -- Highlights: ► The ATLAS inner tracker will be extended with a so called Insertable B-Layer (IBL). ► The IBL modules are required to withstand irradiation up to 5×10 15 n eq /cm 2 . ► Two types of silicon pixel detector technologies (Planar and 3D) were tested in beam. ► The irradiated sensor efficiency exceeds 97% both with and without magnetic field. ► The leakage current, power dissipation, module active area ratio requirements are met.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  2. FMIT - the fusion materials irradiation test facility

    International Nuclear Information System (INIS)

    Liska, D.J.

    1980-01-01

    A joint effort by the Hanford Engineering Development Laboratory (HEDL) and Los Alamos Scientific Laboratory (LASL) has produced a preliminary design for a Fusion Materials Irradiation Test Facility (FMIT) that uses a high-power linear accelerator to fire a deuteron beam into a high-speed jet of molten lithium. The result is a continuous energy spectrum of neutrons with a 14-MeV average energy which can irradiate material samples to projected end-of-life levels in about 3 years, with a total accumulated fluence of 10 21 to 10 22 n/cm 2

  3. Mechanical Tests Plan after Neutron Irradiation for SMART SG Tube Materials in a Hot Cell

    International Nuclear Information System (INIS)

    Ahn, Sang Bok; Baik, Seung Jai; Kim, Do Sik; Yoo, Byung Ok; Jung, Yang Hong; Song, Woong Sub; Choo, Kee Nam; Park, Jin Seok; Lee, Yong Sun; Ryu, Woo Seog

    2010-01-01

    An advanced integral PWR, SMART (System- Integrated Modular Advanced ReacTor) is being developed in KAERI. It has compact size and a relatively small power rating compared to a conventional reactor. The main components such as the steam generators, main circulation pumps are located in the reactor vessel. Therefore they are damaged from neutron irradiations generated from nuclear fuel fissions during operation. The SMART SG tubes which are 17 mm in a diameter and 2.5 mm in a thickness will be made of Alloy 690. To ensure the operation safety the post irradiation examinations is necessary to evaluate the deterioration levels of various original properties. Specially the amount of mechanical properties change should be reflected and revised to design data. For that tensile, fracture, hardness test are planned and under preparations. In this paper the detailed plans are reviewed. Three kinds of materials having different heat treatment procedures are prepared to fabricate specimens. The capsules installed the specimens are going to be irradiated in HANARO. Finally the tests for them will be performed in IMEF, Irradiated Materials Examination Facility at KAERI

  4. Effects of ion beam irradiation on semiconductor devices

    Energy Technology Data Exchange (ETDEWEB)

    Nashiyama, Isamu; Hirao, Toshio; Itoh, Hisayoshi; Ohshima, Takeshi [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment

    1997-03-01

    Energetic heavy-ion irradiation apparatus has been developed for single-event effects (SEE) testing. We have applied three irradiation methods such as a scattered-ion irradiation method, a recoiled-atom irradiation method, and a direct-beam irradiation method to perform SEE testing efficiently. (author)

  5. Irradiation testing of miniature fuel plates for the RERTR program

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R L; Martin, M M [Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States)

    1983-08-01

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. The objective of these tests is to screen various candidate fuel materials as to their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% {sup 235}U in place of highly enriched fuel for these reactors would reduce the potential for {sup 235}U diversion. Fuel materials currently being evaluated in this first phase of these screening tests include aluminum-base dispersion-type fuel plates with fuel cores of 1) high uranium content U{sup 3}){sup 8}-Al being developed by ORNL, 2) high uranium content UAI{sub x}-Al being developed by EG and G Idaho, Inc., and 3) very high uranium content U{sub 3}Si-Al- being developed by ANL. The miniplates are 115-mm long by 50-mm wide with overall plate thicknesses of 1.27 or 1.52 mm. The fuel core dimensions vary according to overall plate thicknesses with a minimal clad thickness requirement of 0.20 mm. Sixty such miniplates (thirty of each thickness) can be irradiated in one test facility. The irradiation test facility, designated as HFED-1 is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The peak neutron flux measured for this experiment is 1.96 x 10{sup 18} neutrons m{sub -2} s{sub -1}. The various types of miniplates will achieve burnups of up to approximately 2.2x10{sup 27} fissions/m{sup 3} of fuel, which will require approximately eight full power months of irradiation. During reactor shutdown periods, the experiment is removed from the reactor, moved to a special poolside station, disassembled, and inspected

  6. Meso-scale modeling of irradiated concrete in test reactor

    International Nuclear Information System (INIS)

    Giorla, A.; Vaitová, M.; Le Pape, Y.; Štemberk, P.

    2015-01-01

    Highlights: • A meso-scale finite element model for irradiated concrete is developed. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • Confrontation with expansion and damage obtained from experiments is successful. • Effects of paste shrinkage, creep and ductility are discussed. - Abstract: A numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale is detailed in this paper. Irradiation experiments in test reactor (Elleuch et al., 1972), i.e., in accelerated conditions, are simulated. Concrete is considered as a two-phase material made of elastic inclusions (aggregate) subjected to thermal and irradiation-induced swelling and embedded in a cementitious matrix subjected to shrinkage and thermal expansion. The role of the hardened cement paste in the post-peak regime (brittle-ductile transition with decreasing loading rate), and creep effects are investigated. Radiation-induced volumetric expansion (RIVE) of the aggregate cause the development and propagation of damage around the aggregate which further develops in bridging cracks across the hardened cement paste between the individual aggregate particles. The development of damage is aggravated when shrinkage occurs simultaneously with RIVE during the irradiation experiment. The post-irradiation expansion derived from the simulation is well correlated with the experimental data and, the obtained damage levels are fully consistent with previous estimations based on a micromechanical interpretation of the experimental post-irradiation elastic properties (Le Pape et al., 2015). The proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.

  7. Meso-scale modeling of irradiated concrete in test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giorla, A. [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Vaitová, M. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic); Le Pape, Y., E-mail: lepapeym@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Štemberk, P. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic)

    2015-12-15

    Highlights: • A meso-scale finite element model for irradiated concrete is developed. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • Confrontation with expansion and damage obtained from experiments is successful. • Effects of paste shrinkage, creep and ductility are discussed. - Abstract: A numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale is detailed in this paper. Irradiation experiments in test reactor (Elleuch et al., 1972), i.e., in accelerated conditions, are simulated. Concrete is considered as a two-phase material made of elastic inclusions (aggregate) subjected to thermal and irradiation-induced swelling and embedded in a cementitious matrix subjected to shrinkage and thermal expansion. The role of the hardened cement paste in the post-peak regime (brittle-ductile transition with decreasing loading rate), and creep effects are investigated. Radiation-induced volumetric expansion (RIVE) of the aggregate cause the development and propagation of damage around the aggregate which further develops in bridging cracks across the hardened cement paste between the individual aggregate particles. The development of damage is aggravated when shrinkage occurs simultaneously with RIVE during the irradiation experiment. The post-irradiation expansion derived from the simulation is well correlated with the experimental data and, the obtained damage levels are fully consistent with previous estimations based on a micromechanical interpretation of the experimental post-irradiation elastic properties (Le Pape et al., 2015). The proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.

  8. Irradiation test HT-31: high-temperature irradiation behavior of LASL-made extruded fuel rods and LASL-made coated particles

    International Nuclear Information System (INIS)

    Wagner, P.; Reiswig, R.D.; Hollabaugh, C.M.; White, R.W.; Davidson, K.V.; Schell, D.H.

    1977-04-01

    Three LASL-made extruded graphite and coated particle fuel rods have been irradiated in the Oak Ridge National Laboratory High Fluence Isotope Reactor test HT-31. Test conditions were about 9 x 10 21 nvt(E > .18 MeV) at 1250 0 C. The graphite matrix showed little or no effect of the irradiation. LASL-made ZrC containing coated particles with ZrC coats and ZrC-doped pyrolytic carbon coats showed no observable effects of the irradiation

  9. Identification of gamma irradiated pulse seed (Lens sp.) based on germination test

    International Nuclear Information System (INIS)

    Chaudhuri, Sadhan K.

    2001-01-01

    The germination test of pulse seed provided a reliable method for the identification of lentil seeds that had been subjected to irradiation. Root and shoot lengths were found more sensitive to the gamma irradiation than the germination percentages. The critical dose that prevented the root elongation varied from 0.1 kGy to 0.5 kGy. Germination percentage was reduced drastically above 0.2 kGy. Above 1.0 kGy dose, the lentil seeds did not germinate. The sensitivity of lentil seeds to gamma irradiation was inversely proportional to moisture content of the seeds. In addition, storage period up to 12 months had little effect on irradiation the induced reduction of root and shoot lengths. Thus, this test can determine the difference between irradiated and non-irradiated lentil seeds even 12 months after gamma irradiation. (author)

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

    International Nuclear Information System (INIS)

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

    1977-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-01

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

  12. Effects of non-steady irradiation conditions on fusion materials performance

    International Nuclear Information System (INIS)

    Matsui, H.; Fukumoto, K.; Nagumo, T.; Nita, N.

    2001-01-01

    During startup of fusion reactors, materials are exposed to neutron irradiation under non-steady temperature condition. Since the temperature of irradiation has decisive effects on the microstructural evolution, the non-steady temperature will have important consequences in the performance of fusion reactor materials. In the present study, a series of vanadium based alloys have been irradiated with neutrons in a temperature cycling condition. It has been found from this study that cavity number density is much greater in temperature cycled specimens than in steady temperature irradiation. Keeping the upper temperature constant, cavity number density is greater for smaller difference between the upper and the lower temperature. It follows that relatively small temperature excursions may have rather significant effects on the fusion material performance in service. (author)

  13. Irradiation test plan of oxidation-resistant graphite in WWR-K Research Reactor

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Sakaba, Nariaki; Osaki, Hirotaka; Kato, Hideki; Fujitsuka, Kunihiro; Muto, Takenori; Gizatulin, Shamil; Shaimerdenov, Asset; Dyussambayev, Daulet; Chakrov, Petr

    2014-01-01

    Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR) which is a graphite-moderated and helium gas-cooled reactor. In the case of air ingress accident in HTGR, SiO_2 protective layer is formed on the surface of SiC layer in TRISO CFP and oxidation of SiC does not proceed and fission products are retained inside the fuel particle. A new safety concept for the HTGR, called Naturally Safe HTGR, has been recently proposed. To enhance the safety of Naturally Safe HTGR ultimately, it is expected that oxidation-resistant graphite is used for graphite components to prevent the TRISO CFPs and fuel compacts from failure. SiC coating is one of candidate methods for oxidation-resistant graphite. JAEA and four graphite companies launched R&Ds to develop the oxidation-resistant graphite and the International Science and Technology Center (ISTC) partner project with JAEA and INP was launched to investigate the irradiation effects on the oxidation-resistant graphite. To determine grades of the oxidation-resistant graphite which will be adopted as irradiation test, a preliminary oxidation test was carried out. This paper described the results of the preliminary oxidation test, the plan of out-of-pile test, irradiation test and post-irradiation test (PIE) of the oxidation-resistant graphite. The results of the preliminary oxidation test showed that the integrity of the oxidation resistant graphite was confirmed and that all of grades used in the preliminary test can be adopted as the irradiation test. Target irradiation temperature was determined to be 1473 (K) and neutron fluence was determined to be from 0.54 × 10"2"5through 1.4 × 10"2"5 (/m"2, E>0.18MeV). Weight change, oxidation rate, activation energy, surface condition, etc. will be evaluated in out-of-pile test and weight change, irradiation effect on oxidation rate and activation energy, surface condition, etc. will be evaluated in PIE. (author)

  14. Irradiation tests on PHWR type fuel elements in TRIGA research reactor of INR Pitesti

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore [Institute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.; Sorescu, Ion [Institute for Nuclear Research, Pitesti (Romania). TRIGA Reactor Loop Facility; Parvan, Marcel [Institute for Nuclear Research, Pitesti (Romania). Hot Cells Lab.

    2012-12-15

    Nine PHWR type fuel elements with reduced length were irradiated in loop A of the TRIGA Research Reactor of INR Pitesti. The primary objective of the test was to determine the performance of nuclear fuel fabricated at INR Pitesti at high linear powers in pressurized water conditions. Six fuel elements were irradiated with a ramp power history, achieving a maximum power of 45 kW/m during pre-ramp and of 64 kW/m in the ramp. The maximum discharge burnup was of 216 MWh/kgU. Another three fuel elements with reduced length were irradiated with declining power history. At the beginning of irradiation the fuel elements achieved a maximum linear power of 66 kW/m. The maximum fuel power was about 1.3 times the maximum expected in PHWR. The maximum discharge burnup was 205 MWh/kgU. The elements were destructively examined in the hot cells of INR Pitesti. Temperature-sensitive parameters such as UO{sub 2} grain growth, fission-gas release and sheath deformations were examined. The tests proved the feasibility of irradiating PHWR type fuel elements at linear powers up to 66 kW/m under pressurized water conditions and demonstrated the possibility of more flexible operation of this fuel in power reactors. This paper presents the results of the investigation. (orig.)

  15. Respiratory function tests after mantle irradiation in patients with Hodgkin's disease

    Energy Technology Data Exchange (ETDEWEB)

    Cionini, L; Pacini, P; De Paola, E; Corrado, A; De Luca Cardillo, C; Mungai, V; Biti, G P; Ponticelli, P

    1984-01-01

    Pulmonary function tests were performed in 43 patients with Hodgkin's disease before mantle irradiation and at 3, 6, 9, 12 and 15 or more months thereafter. Treatment was given with a telecobalt unit to a total dose of 36 to 42 Gy, the higher dose being reserved for cases with considerable mediastinal involvement. The functional parameters explored included static and dynamic lung volumes, gas exchanges, ventilatory efficiency, and airway resistance. Measured parameters were expressed as a percentage of the pre-treatment value (PTV) in the individual patient. In the whole group, only small variations in the functional indices were observed at 3 to 6 months after mantle irradiation. In patients with normal PTVs a greater variation in static and dynamic volumes was observed at 3 to 6 months after mantle irradiation, with complete recovery thereafter. The gas exchange parameters also showed a similar variation at 3 to 6 months but no recovery was demonstrated in the subsequent examinations. No changes in ventilatory efficiency and airway resistance were observed. In patients with abnormal PTVs, usually presenting large mediastinal adenopathy, all parameters improved after mantle irradiation, and the favourable effect of tumour regression was probably more important than the radiation damage on the pulmonary parenchyma.

  16. Irradiation experiments and materials testing capabilities in High Flux Reactor in Petten

    International Nuclear Information System (INIS)

    Luzginova, N.; Blagoeva, D.; Hegeman, H.; Van der Laan, J.

    2011-01-01

    The text of publication follows: The High Flux Reactor (HFR) in Petten is a powerful multi-purpose research and materials testing reactor operating for about 280 Full Power Days per year. In combination with hot cells facilities, HFR provides irradiation and post-irradiation examination services requested by nuclear energy research and development programs, as well as by industry and research organizations. Using a variety of the custom developed irradiation devices and a large experience in executing irradiation experiments, the HFR is suitable for fuel, materials and components testing for different reactor types. Irradiation experiments carried out at the HFR are mainly focused on the understanding of the irradiation effects on materials; and providing databases for irradiation behavior of materials to feed into safety cases. The irradiation experiments and materials testing at the HFR include the following issues. First, materials irradiation to support the nuclear plant life extensions, for instance, characterization of the reactor pressure vessel stainless steel claddings to insure structural integrity of the vessel, as well as irradiation of the weld material coupons to neutron fluence levels that are representative for Light Water Reactors (LWR) internals applications. Secondly, development and qualification of the structural materials for next generation nuclear fission reactors as well as thermo-nuclear fusion machines. The main areas of interest are in both conventional stainless steel and advanced reduced activation steels and special alloys such as Ni-base alloys. For instance safety-relevant aspects of High Temperature Reactors (HTR) such as the integrity of fuel and structural materials with increasing neutron fluence at typical HTR operating conditions has been recently assessed. Thirdly, support of the fuel safety through several fuel irradiation experiments including testing of pre-irradiated LWR fuel rods containing UO 2 or MOX fuel. Fourthly

  17. Project accent: graphite irradiated creep in a materials test reactor

    International Nuclear Information System (INIS)

    Brooking, M.

    2014-01-01

    Atkins manages a pioneering programme of irradiation experiments for EDF Energy. One of these projects is Project ACCENT, designed to obtain evidence of a beneficial physical property of the graphite, which may extend the life of the Advanced Gas-cooled Reactors (AGRs). The project team combines the in-house experience of EDF Energy with two supplier organisations (providing the material test reactors and testing facilities) and supporting consultancies (Atkins and an independent technical expert). This paper describes: - Brief summary of the Project; - Discussion of the challenges faced by the Project; and - Conclusion elaborating on the aims of the Project. These challenging experiments use bespoke technology and both un-irradiated (virgin) and irradiated AGR graphite. The results will help to better understand graphite irradiation-induced creep (or stress modified dimensional change) properties and therefore more accurately determine lifetime and safe operating envelopes of the AGRs. The first round of irradiation has been completed, with a second round about to commence. This is a key step to realising the full lifetime ambition for AGRs, demonstrating the relaxation of stresses within the graphite bricks. (authors)

  18. Fabrication, inspection, and test plan for the Advanced Test Reactor (ATR) Mixed-Oxide (MOX) fuel irradiation project

    International Nuclear Information System (INIS)

    Wachs, G.W.

    1997-11-01

    The Department of Energy (DOE) Fissile Materials Disposition Materials Disposition Program (FMDP) has announced that reactor irradiation of MOX fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The MOX fuel test will be irradiated in the ATR to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. In addition, the test will contribute experience with irradiation of gallium-containing fuel to the data base required for resolution of generic CLWR fuel design issues (ORNL/MD/LTR-76). This Fabrication, Inspection, and Test Plan (FITP) is a level 2 document as defined in the FMDP LWR MOX Fuel Irradiation Test Project Plan (ORNL/MD/LTR-78)

  19. Response of unirradiated and irradiated PWR fuel rods tested under power-cooling-mismatch conditions

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Quapp, W.J.; Martinson, Z.R.; McCardell, R.K.; Mehner, A.S.

    1978-01-01

    This report summarizes the results from the single-rod power-cooling-mismatch (PCM) and irradiation effects (IE) tests conducted to date in the Power Burst Facility (PBF) at the U.S. DOE Idaho National Engineering Laboratory. This work was performed for the U.S. NRC under contact to the Department of Energy. These tests are part of the NRC Fuel Behavior Program, which is designed to provide data for the development and verification of analytical fuel behavior models that are used to predict fuel response to abnormal or postulated accident conditions in commercial LWRs. The mechanical, chemical and thermal response of both previously unirradiated and previously irradiated LWR-type fuel rods tested under power-cooling-mismatch condition is discussed. A brief description of the test designs is presented. The results of the PCM thermal-hydraulic studies are summarized. Primary emphasis is placed on the behavior of the fuel and cladding during and after stable film boiling. (orig.) [de

  20. Irradiated diets and its effect on testes and adrenal gland of rats

    International Nuclear Information System (INIS)

    Kushwaha, A.K.S.; Hasan, S.S.

    1988-01-01

    The present investigation was undertaken to study the feeding effects of irradiated normal diet (consisting of equal parts of gram and wheat) and irradiated low protein diet (consisting one part of normal diet and three parts of wheat) on male rats for various periods starting from weaning time. Rats maintained on irradiated low protein diets showed decrease in the activity of androgen sensitive enzymes i.e., alkaline and acid phosphatase while an increase in the cholesterol content of the testes compared with irradiated normal controls. Diminution in androgen sensitive enzymes and accumulation of cholesterol in the rat testes suggest non-conversion of cholesterol into steriod hormones after feeding of irradiated low protein. Besides, rats fed on irradiated low protein diet showed increased cellular activity in the adrenal cortex and medulla as compared to rats fed on the irradiated normal diet. (author). 12 refs., 4 tabs

  1. 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

  2. Results of crack-arrest tests on irradiated a 508 class 3 steel

    International Nuclear Information System (INIS)

    Iskander, S.K.; Milella, P.P.; Pini, M.A.

    1998-02-01

    Ten crack-arrest toughness values for irradiated specimens of A 508 class 3 forging steel have been obtained. The tests were performed according to the American Society for Testing and Materials (ASTM) Standard Test Method for Determining Plane-Strain Crack-Arrest Fracture Toughness, K la of Ferritic Steels, E 1221-88. None of these values are strictly valid in all five ASTM E 1221-88 validity criteria. However, they are useful when compared to unirradiated crack-arrest specimen toughness values since they show the small (averaging approximately 10 degrees C) shifts in the mean and lower-bound crack-arrest toughness curves. This confirms that a low copper content in ASTM A 508 class 3 forging material can be expected to result in small shifts of the transition toughness curve. The shifts due to neutron irradiation of the lower bound and mean toughness curves are approximately the same as the Charpy V-notch (CVN) 41-J temperature shift. The nine crack-arrest specimens were irradiated at temperatures varying from 243 to 280 degrees C, and to a fluence varying from 1.7 to 2.7 x 10 19 neutrons/cm 2 (> 1 MeV). The test results were normalized to reference values that correspond to those of CVN specimens irradiated at 284 degrees C to a fluence of 3.2 x 10 19 neutrons/cm 2 (> 1 MeV) in the same capsule as the crack-arrest specimens. This adjustment resulted in a shift to lower temperatures of all the data, and in particular moved two data points that appeared to lie close to or lower than the American Society of Mechanical Engineers K la curve to positions that seemed more reasonable with respect to the remaining data. A special fixture was designed, fabricated, and successfully used in the testing. For reasons explained in the text, special blocks to receive the Oak Ridge National Laboratory clip gage were designed, and greater-than-standard crack-mouth opening displacements measured were accounted for. 24 refs., 13 figs., 12 tabs

  3. Design and fabrication of water control unit for IASCC irradiation test

    International Nuclear Information System (INIS)

    Mori, Yuichiro; Takeuchi, Yutaka; Matsunami, Kiyotaka; Kosaki, Kazuhiko; Suzuki, Tomio; Hayashi, Motomitsu; Ide, Kiyoshi

    2004-01-01

    In relation to the aging of LWR, the Irradiation Assisted Stress Corrosion Cracking (IASCC) has been regarded as a significant and urgent issue for the reliability of in-core components of LWR, therefore the irradiation research project which was planned by Nuclear and Industrial Safety Agency is now being done under the cooperation of Industry-Government-Academia such as Japan Nuclear Energy Safety Organization, Institute of Research and Innovation (IRI), Central Research Institute of Electric Power Industry, Japan Atomic Energy Research Institute (JAERI), power companies, makers of LWR, and universities. Then at Japan Material Testing Reactor (JMTR) of JAERI, the irradiation test of the material for BWR is being carried out. This paper describes the introduction about the Water Control Unit (WCU) for IASCC irradiation test. The WCU was designed and installed into JMTR by Kawasaki Heavy Industries, LTD, based on the order from JAERI, IRI, and so on. (author)

  4. Microstructure and elemental distribution of americium containing MOX fuel under the short term irradiation tests

    International Nuclear Information System (INIS)

    Tanaka, Kosuke; Hirosawa, Takashi; Obayashi, Hiroshi; Koyama, Shin Ichi; Yoshimochi, Hiroshi; Tanaka, Kenya

    2008-01-01

    In order to investigate the effect of americium addition to MOX fuels on the irradiation behavior, the 'Am-1' program is being conducted in JAEA. The Am-1 program consists of two short term irradiation tests of 10-minute and 24 hour irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post irradiation examinations (PIEs) are in progress. The PIEs for Am-containing MOX fuels focused on the microstructural evolution and redistribution behavior of Am at the initial stage of irradiation and the results to date are reported

  5. Status of Wrought FeCrAl-UO2 Capsules Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harp, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Core, G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-07-01

    Candidate cladding materials for accident tolerant fuel applications require extensive testing and validation prior to commercial deployment within the nuclear power industry. One class of cladding materials, FeCrAl alloys, is currently undergoing such effort. Within these activities is a series of irradiation programs within the Advanced Test Reactor. These programs are developed to aid in commercial maturation and understand the fundamental mechanisms controlling the cladding performance during normal operation of a typical light water reactor. Three different irradiation programs are on-going; one designed as a simple proof-of-principle concept, the other to evaluate the susceptibility of FeCrAl to fuel-cladding chemical interaction, and the last to fully simulate the conditions of a pressurized water reactor experimentally. To date, nondestructive post-irradiation examination has been completed on the rodlet deemed FCA-L3 from the simple proof-of-concept irradiation program. Initial results show possible breach of the rodlet under irradiation but further studies are needed to conclusively determine whether breach has occurred and the underlying reasons for such a possible failure. Further work includes characterizing additional rodlets following irradiation.

  6. Preliminary Analysis of the General Performance and Mechanical Behavior of Irradiated FeCrAl Base Alloys and Weldments

    Energy Technology Data Exchange (ETDEWEB)

    Gussev, Maxim N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Briggs, Samuel A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-30

    The iron-based, iron-chromium-aluminum (FeCrAl) alloys are promising, robust materials for deployment in current and future nuclear power plants. This class of alloys demonstrates excellent performance in a range of environments and conditions, including high-temperature steam (>1000°C). Furthermore, these alloys have the potential to have prolonged survival under loss-of-coolant accident (LOCA) conditions compared to the more traditional cladding materials that are either Zr-based alloys or austenitic steels. However, one of the issues associated with FeCrAl alloys is cracking during welding. The present project investigates the possibility of mitigating welding-induced cracking via alloying and precise structure control of the weldments; in the frame work of the project, several advanced alloys were developed and are being investigated prior to and after neutron irradiation to provide insight into the radiation tolerance and mechanical performance of the weldments. The present report provides preliminary results on the post-irradiation characterization and mechanical tests performed during United States Fiscal Year (FY) 2016. Chapter 1 provides a general introduction, and Chapter 2 describes the alloy compositions, welding procedure, specimen geometry and manufacturing parameters. Also, a brief discussion of the irradiation at the High Flux Isotope Reactor (HFIR) is provided. Chapter 3 is devoted to the analysis of mechanical tests performed at the hot cell facility; tensile curves and mechanical properties are discussed in detail focusing on the irradiation temperature. Limited fractography results are also presented and analyzed. The discussion highlights the limitations of the testing within a hot cell. Chapter 4 underlines the advantages of in-situ testing and discusses the preliminary results obtained with newly developed miniature specimens. Specimens were moved to the Low Activation Materials Development and Analysis (LAMDA) laboratory and prepared for

  7. Postirradiation examination results for the Irradiation Effects Test 2

    International Nuclear Information System (INIS)

    Ploger, S.A.; Kerwin, D.K.; Croucher, D.W.

    1978-01-01

    This report presents the postirradiation examination results of Test IE-2 in the Irradiation Effects Test Series conducted under the Thermal Fuels Behavior Program. The objectives of this test were to evaluate the influence of previous cladding irradiation and fuel-cladding diametral gap on fuel rod behavior during a power ramp and during film boiling operation. Test IE-2, conducted in the Power Burst Facility at the Idaho National Engineering Laboratory, employed two 0.97-m-long pressurized water reactor type fuel rods fabricated from previously irradiated zircaloy-4 cladding and two similar rods fabricated from unirradiated cladding. The four rods were subjected to a preconditioning period, followed by a power ramp to an average peak rod power of 68 kW/m and steady state operation for one hour at an individual rod coolant mass flux of 4880 kg/s . m 2 . After a flow reduction to 2550 kg/s . m 2 , film boiling occurred on three rods. An additional flow reduction to 2245 kg/s . m 2 produced film boiling on the remaining fuel rod. Maximum time in film boiling was 90 s. None of the four fuel rods failed during the test. Damage caused by film boiling, as characterized by oxidation, oxide spalling, and collapse at fuel pellet interfaces, was found on all four rods. Film boiling regions on these rods showed evidence of fuel melting, fuel centerline void formation, and internal cladding oxidation resulting from fuel-cladding reaction. Effects of fuel-cladding diametral gap and cladding irradiation are summarized. Measured temperatures and metallographically estimated temperatures are compared at several axial fuel rod locations

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-01

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

  9. A neutron irradiator to perform nuclear activation

    International Nuclear Information System (INIS)

    Zamboni, C. B.; Zahn, G.S.; Figueredo, A. M. G.; Madi, T. F.; Yoriyaz, H.; Lima, R. B.; Shtejer, K.; Dalaqua Jr, L.

    2001-01-01

    The development of appropriate nuclear instrumentation to perform neutron activation analyze (NAA), using thermal and fast neutrons, can be useful to investigate materials outside the reactor premises. Considering this fact, a small size neutron irradiator prototype was developed at IPEN facilities (Instituto de Pesquisas Energeticas e Nucleares - Brazil). Basically, this prototype consists of a cylinder of 1200 mm long and 985 mm diameter (filled with paraffin) with two Am-Be sources (600GBq each) arranged in the longitudinal direction of its geometric center. The material to be irradiated is positioned at a radial direction of the cylinder between the two Am-Be sources. The main advantage of this irradiator is a very stable neutron flux eliminating the use of standard material (measure of the induced activity in the sample by comparative method). This way the process became agile, practical and economic, but quantities at mg levels of samples are necessary to achieve good sensitivity, when the material has a low microscopy neutron cross section. As fast and thermal neutron can be used, the flux distribution, for both, were calculated and the prototype performance is discussed

  10. An investigation of high-temperature irradiation test program of new ceramic materials

    International Nuclear Information System (INIS)

    Ishino, Shiori; Terai, Takayuki; Oku, Tatsuo

    1999-08-01

    The Japan Atomic Energy Research Institute entrusted the Atomic Energy Society of Japan with an investigation into the trend of irradiation processing/damage research on new ceramic materials. The present report describes the result of the investigation, which was aimed at effective execution of irradiation programs using the High Temperature Engineering Test Reactor (HTTR) by examining preferential research subjects and their concrete research methods. Objects of the investigation were currently on-going preliminary tests of functional materials (high-temperature oxide superconductor and high-temperature semiconductor) and structural materials (carbon/carbon and SiC/SiC composite materials), together with newly proposed subjects of, e.g., radiation effects on ceramics-coated materials and super-plastic ceramic materials as well as microscopic computer simulation of deformation and fracture of ceramics. These works have revealed 1) the background of each research subject, 2) its objective and significance from viewpoints of science and engineering, 3) research methodology in stages from preliminary tests to real HTTR irradiation, and 4) concrete HTTR-irradiation methods which include main specifications of test specimens, irradiation facilities and post-irradiation examination facilities and apparatuses. The present efforts have constructed the important fundamentals in the new ceramic materials field for further planning and execution of the innovative basic research on high-temperature engineering. (author)

  11. Effects of gamma-irradiation of plastics on migration of constituents into test foods

    International Nuclear Information System (INIS)

    Figge, K.; Freytag, W.

    1977-01-01

    Test films prepared from polyethylene (LD- and HD-PE), polypropylene (PP), polystyrene (St- and HI-PS) and polyvinylchloride (rigid PVC) compositions under addition of 2,6-di-tert-butyl-4-methyl[ 14 C]-phenol(I), 3-(3,5-di-tert-butyl -4-hydroxyphenyl)- stearyl-propionate[3- 14 C] (II), n-butyl-stearate[1- 14 C] (III) di-n-octyl[1- 14 C]-tin-2 -ethyl-hexyl-di-thioglycolate inclusive of the corresponding mono-n-octyl[1- 14 C]-tin compound (IV-Oc), di-n-octyl-tin-2-ethylhexyl -di-thioglycolate[2- 14 C] inclusive of the corresponding 2-ethylhexyl-tri -thioglycolate[2- 14 C] (IV-S) or of stearyl alcohol [1- 14 C] (V) respectively, were sterilized in a 60 Co irradiation unit with a radiation dosage of 2,5 Mrad. Then, the irradiated films as well as non-treated reference films were kept in one-sided contact with the test foodstuffs 'dist. Water' and HB 307 (test fat) for 10 days at 40 0 C. Under these conditions the additives I and II migrated from the PP, HD-PE and LD-PE test films into the test fat HB 307 in amounts of 10 to 50%. Migration into distilled water was only 0.05 to 4.6%. The migration of the additives I and II from the irradiated polyolefin test films into the test fat HB 307 was by 8 to 38% lower than that from the corresponding non-irradiated films. In contrast to this, both additives migrated distinctly more strongly from the irradiated polyolefin test films into distilled water, i.e. 1.9 to 8.7 times stronger than from the non-irradiated films. The migration of the additives I to V from the HI-PS St-PS and rigid PVC test films into the two test foodstuffs was very low, in most cases below 0.1%. Generally lower amounts of additive migrated from the irradiated films than from the non-irradiated samples. (orig.) [de

  12. Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

    Energy Technology Data Exchange (ETDEWEB)

    Venkiteswaran, C.N., E-mail: cnv@igcar.gov.in; Jayaraj, V.V.; Ojha, B.K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B.P.C.; Kasiviswanathan, K.V.; Jayakumar, T.

    2014-06-01

    The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel–clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel–clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

  13. Irradiation performance of HTGR Biso fertile particles in HFIR experiments HT-17, -18, and -19

    International Nuclear Information System (INIS)

    Long, E.L. Jr.; Beatty, R.L.; Robbins, J.M.; Kania, M.J.; Eatherly, W.P.

    1978-11-01

    A series of Biso-coated fertile particles was irradiated in the target facility of the High-Flux Isotope Reactor. The primary objectives of this experiment were to relate the fast-neutron stability of dense propylene-derived pyrocarbons to preferred orientation and to relate irradiation performance to preirradiation characterization values. Coating characterization included x-ray BAF, optical anisotropy, gaseous permeability, small-angle x-ray scattering, and thickness and density determinations. Other objectives were to test Biso-coated large-diameter ThO 2 kernels and coatings derived from propylene diluted with CO 2 rather than argon. Visual examination of the irradiated particles showed that the majority had failed as a result of fast-neutron damage

  14. Mutagenicity tests on irradiated food

    International Nuclear Information System (INIS)

    Johnston-Arthur, T.

    1979-01-01

    The mutagenicity of ''standard'' food pellets from three different suppliers was tested after radappertization and after sterilization with steam, respectively. The histidine-deficient mutants G-46 and TA-1530 of salmonella typhimurium were used as indicators in a hostmediated assay. The mutant TA-1530 showed a highly sighificant increase of the back-mutation frequency after feeding with pellets irradiated with 3 Mrad gamma radiation. There were, however, large quantitative differences between the products of different suppliers. (G.G.)

  15. Evaluation of the Centerline Temperature for the Irradiated DUPIC Pellet

    International Nuclear Information System (INIS)

    Park, Chang Je; Lee, Cheol Yong; Kang, Kweon Ho; Song, Kee Chan

    2007-01-01

    The DUPIC (Direct Use of spent PWR fuels In a CANDU reactor) fuel has a proliferation-resistant property and provides an efficient utilization of a spent fuel through a direct fabrication with the OREOX process in which most of the fission products remain and some volatile elements such as Xe, Kr, Cs, and I are reduced significantly. It is expected that the performance of the DUPIC fuel exhibits different behavior when compared with the fresh uranium oxide fuel. To evaluate the performance of the DUPIC fuel, total five irradiation tests have been performed in the HANARO reactor since May 2000. Recently, the fifth irradiation test of the DUPIC fuel was successfully completed for a total of three cycles from March 2006 to July 2006. The important characteristics of the first irradiation test are a high power test and a validation of a remote assembly of an irradiation rig. The second irradiation test was instrumented with a SPND (self-powered neutron detector) first for a typical CANDU burnup test. The third test was an extensive irradiation test of the second test and the total burnup was estimated as 6,700 MWd/tU. The forth test was a remote instrumented test of the pellet centerline temperature and the inlet and outlet coolant temperatures. The first remote instrumentation test was achieved with our own technology. The fifth test was a remote-instrumented test of the pellet centerline temperature by extending the technology of the forth irradiation test. In this paper, a DUPIC fuel performance code (KAOS, KAERI Advanced Oxide fuel performance code System) was used to compare the main simulation results of the irradiation tests in the HANARO

  16. Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Robinson; G. S. Chang; D. D. Keiser, Jr.; D. M. Wachs; D. L. Porter

    2009-08-01

    A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.

  17. KEY RESULTS FROM IRRADIATION AND POST-IRRADIATION EXAMINATION OF AGR-1 UCO TRISO FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A.; Hunn, John D.; Petti, David A.; Morris, Robert N.

    2016-11-01

    The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3×105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time average, volume average irradiation temperatures of the individual compacts ranged from 955 to 1136°C. This paper focuses on key results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior within the US program. The fuel exhibited a very low incidence of TRISO coating failure during irradiation and post-irradiation safety testing at temperatures up to 1800°C. Advanced PIE methods have allowed particles with SiC coating failure to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of

  18. Experimental Breeder Reactor II (EBR-II) Fuel-Performance Test Facility (FPTF)

    International Nuclear Information System (INIS)

    Pardini, J.A.; Brubaker, R.C.; Veith, D.J.; Giorgis, G.C.; Walker, D.E.; Seim, O.S.

    1982-01-01

    The Fuel-Performance Test Facility (FPTF) is the latest in a series of special EBR-II instrumented in-core test facilities. A flow control valve in the facility is programmed to vary the coolant flow, and thus the temperature, in an experimental-irradiation subassembly beneath it and coupled to it. In this way, thermal transients can be simulated in that subassembly without changing the temperatures in surrounding subassemblies. The FPTF also monitors sodium flow and temperature, and detects delayed neutrons in the sodium effluent from the experimental-irradiation subassembly beneath it. This facility also has an acoustical detector (high-temperature microphone) for detecting sodium boiling

  19. Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

    Science.gov (United States)

    Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

    2017-12-01

    A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

  20. Evaluation of ductile-brittle transition behavior with neutron irradiation in nuclear reactor pressure vessel steels using small punch test

    International Nuclear Information System (INIS)

    Kim, M. C.; Lee, B. S.; Oh, Y. J.

    2003-01-01

    A Small Punch (SP) test was performed to evaluate the ductile-brittle transition temperature before and after neutron irradiation in Reactor Pressure Vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the Charpy test and Master Curve fracture toughness test in accordance with the ASTM standard E1921. The samples were taken from 1/4t location of the vessel thickness and machined into a 10x10x0.5mm dimension. Irradiation of the samples was carried out in the research reactor at KAERI (HANARO) at about 290 .deg. C of the different fluence levels respectively. SP tests were performed in the temperature range of RT to -196 .deg. C using a 2.4mm diameter ball. For the materials before and after irradiation, SP transition temperatures (T sp ), which are determined at the middle of the upper and lower SP energies, showed a linear correlation with the Charpy index temperature, T 41J . T sp from the irradiated samples was increased as the fluence level increased and was well within the deviation range of the unirradiated data. The TSP had a correlation with the reference temperature (T 0 ) from the master curve method using a pre-cracked Charpy V-notched (PCVN) specimen

  1. Progress in food irradiation: Bangladesh

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, M

    1982-11-01

    The Bangladesh contribution deals with fish preservation by irradiation and, in this context, with the radiosensitivity of mesophilic and psychophilic microorganisms. Sprouting inhibition is studied with potatoes and onions. A further part deals with irradiation of spices. Mutagenicity tests were carried on rats and mice fed with irradiated fish. The tests were performed at the Institute for Food and Radiation Biology, near Dacca in December 1981.

  2. Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

    Energy Technology Data Exchange (ETDEWEB)

    A. J. Palmer; DC Haggard; J. W. Herter; M. Scervini; W. D. Swank; D. L. Knudson; R. S. Cherry

    2011-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. 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. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature 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 these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired alumina

  3. Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

    Energy Technology Data Exchange (ETDEWEB)

    Palmer, A. J.; Haggard, DC; Herter, J. W.; Swank, W. D.; Knudson, D. L.; Cherry, R. S. [Idaho National Laboratory, P.O. Box 1625, MS 4112, Idaho Falls, ID, (United States); Scervini, M. [University of Cambridge, Department of Material Science and Metallurgy, 27 Charles Babbage Road, CB3 0FS, Cambridge, (United Kingdom)

    2015-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. 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. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature 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 fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly

  4. Neutron irradiation characteristic tests of oxygen sensors using zirconia solid electrolyte

    International Nuclear Information System (INIS)

    Hiura, Nobuo; Endou, Yasuichi; Yamaura, Takayuki; Niimi, Motoji; Hoshiya, Taiji; Saito, Junichi; Souzawa, Shizuo; Ooka, Norikazu; Kobiyama, Mamoru.

    1997-03-01

    In the Department of JMTR of Japan Atomic Energy Research Institute (JAERI), the in-situ measuring technique of oxygen potential has been being developed to study the chemical behavior of high burn-up fuel base-irradiated in the Light Water Reactor. In this test for development of the technique, oxygen sensors using zirconia solid electrolyte stabilized by MgO, CaO and Y 2 O 3 , named MSZ, CSZ and YSZ, respectively, were irradiated by neutrons in the Japan Materials Testing Reactor (JMTR) of JAERI and the characteristics of electromotive force of these sensors under and after irradiation were discussed. From the experimental results, the electromotive force of YSZ sample under irradiation decreased with an increase in irradiation fluence within a range of neutron fluence (E>1 MeV) up to 1 x 10 23 m -2 . The electromotive force of MSZ sensor irradiated with neutron fluences (E>1 MeV) up to 9 x 10 21 m -2 was almost equal to the theoretical value of the electromotive force. It was shown that after irradiation, a decrease in the electromotive force of CSZ sensor was smaller than those of MSZ and YSZ sensors, although the electromotive forces of MSZ, CSZ and YSZ sensors were smaller than the theoretical value. (author)

  5. Assessment of cognitive functions after prophylactic and therapeutic whole brain irradiation using neuropsychological testing

    International Nuclear Information System (INIS)

    Penitzka, S.; Wannenmacher, M.; Steinvorth, S.; MIT, Cambridge, MT; Sehlleier, S.; Universitaetsklinikum Wuerzburg; Fuss, M.; Texas Univ., San Antonio, TX; Wenz, F.; Universitaetsklinikum Mannheim

    2002-01-01

    Purpose: Aim of this study was the assessment of neuropsychological changes after whole brain irradiation. Patients and Method: 64 patients were tested before, and 29 after whole brain irradiation, including 28 patients with small cell lung cancer (SCLC) before prophylactic cranial irradiation (PCI) and 36 patients with cerebral metastases before therapeutic cranial irradiation (TCI), as well as 14 patients after PCI and 15 after TCI (Table 1). Intelligence, attention and memory were assessed applying a 90-minute test battery of standardized, neuropsychological tests (Table 3). Results: Patients with SCLC showed test results significantly below average before PCI (n=28, mean IQ=83, SD=17). Neither after PCI, nor after TCI the tested neuropsychological functions decreased significantly (Tables 4, 5). A comparison between SCLC-patients with and without cerebral metastases before whole brain irradiation showed better test-results in patients with cerebral metastases and fewer cycles of preceding chemotherapy (Table 7). Conclusion: Neuropsychological capacity in patients with SCLC was impaired even before PCI. Possible reason is the preceding chemotherapy. Whole brain irradiation did not induce a significant decline of cognitive functions in patients with PCI or TCI. A decline in a longer follow-up nevertheless seems possible. (orig.) [de

  6. Improvement of the half-embryo test for detection of gamma-irradiated grapefruit and its application to irradiated oranges and lemons

    International Nuclear Information System (INIS)

    Kawamura, Y.; Uchiyama, S.; Saito, Y.

    1989-01-01

    The duration of the half-embryo test used for identification of gamma-irradiated grapefruit was shortened by increasing germination temperature to 35 degrees C. Maximum shooting percentages were reached within 3 days. Gibberellin application reduced the required incubation time to 2 days. Half-embryos extracted from irradiated orange and lemon gave similar results to those of grapefruit. This half-embryo test is proposed as an identification method for irradiated citrus. Assessment can be made after 3 to 4 days when shooting percentage is greater than 50%

  7. Aspheric surface testing by irradiance transport equation

    Science.gov (United States)

    Shomali, Ramin; Darudi, Ahmad; Nasiri, Sadollah; Asgharsharghi Bonab, Armir

    2010-10-01

    In this paper a method for aspheric surface testing is presented. The method is based on solving the Irradiance Transport Equation (ITE).The accuracy of ITE normally depends on the amount of the pick to valley of the phase distribution. This subject is investigated by a simulation procedure.

  8. Updated FY12 Ceramic Fuels Irradiation Test Plan

    International Nuclear Information System (INIS)

    Nelson, Andrew T.

    2012-01-01

    The Fuel Cycle Research and Development program is currently devoting resources to study of numerous fuel types with the aim of furthering understanding applicable to a range of reactors and fuel cycles. In FY11, effort within the ceramic fuels campaign focused on planning and preparation for a series of rabbit irradiations to be conducted at the High Flux Isotope Reactor located at Oak Ridge National Laboratory. The emphasis of these planned tests was to study the evolution of thermal conductivity in uranium dioxide and derivative compositions as a function of damage induced by neutron damage. Current fiscal realities have resulted in a scenario where completion of the planned rabbit irradiations is unlikely. Possibilities for execution of irradiation testing within the ceramic fuels campaign in the next several years will thus likely be restricted to avenues where strong synergies exist both within and outside the Fuel Cycle Research and Development program. Opportunities to augment the interests and needs of modeling, advanced characterization, and other campaigns present the most likely avenues for further work. These possibilities will be pursued with the hope of securing future funding. Utilization of synthetic microstructures prepared to better understand the most relevant actors encountered during irradiation of ceramic fuels thus represents the ceramic fuel campaign's most efficient means to enhance understanding of fuel response to burnup. This approach offers many of the favorable attributes embraced by the Separate Effects Testing paradigm, namely production of samples suitable to study specific, isolated phenomena. The recent success of xenon-imbedded thick films is representative of this approach. In the coming years, this strategy will be expanded to address a wider range of problems in conjunction with use of national user facilities novel characterization techniques to best utilize programmatic resources to support a science-based research program.

  9. Mechanical and irradiation properties of zirconium alloys irradiated in HANARO

    International Nuclear Information System (INIS)

    Kwon, Oh Hyun; Eom, Kyong Bo; Kim, Jae Ik; Suh, Jung Min; Jeon, Kyeong Lak

    2011-01-01

    These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, 1.1 10 21 n/cm 2 ). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed

  10. The Assembly and Test of Pressure Vessel for Irradiation

    International Nuclear Information System (INIS)

    Park, Kook Nam; Lee, Jong Min; Youn, Young Jung; June, Hyung Kil; Ahn, Sung Ho; Lee, Kee Hong; Kim, Young Ki; Kennedy, Timothy C.

    2009-01-01

    The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts: the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature

  11. The Assembly and Test of Pressure Vessel for Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kook Nam; Lee, Jong Min; Youn, Young Jung; June, Hyung Kil; Ahn, Sung Ho; Lee, Kee Hong; Kim, Young Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kennedy, Timothy C. [Oregon State University, Corvallis (United States)

    2009-02-15

    The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts: the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.

  12. AGR-1 Post Irradiation Examination Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    The post-irradiation examination (PIE) of the Advanced Gas Reactor (AGR)-1 experiment was a multi-year, collaborative effort between Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) to study the performance of UCO (uranium carbide, uranium oxide) tristructural isotropic (TRISO) coated particle fuel fabricated in the U.S. and irradiated at the Advanced Test Reactor at INL to a peak burnup of 19.6% fissions per initial metal atom. This work involved a broad array of experiments and analyses to evaluate the level of fission product retention by the fuel particles and compacts (both during irradiation and during post-irradiation heating tests to simulate reactor accident conditions), investigate the kernel and coating layer morphology evolution and the causes of coating failure, and explore the migration of fission products through the coating layers. The results have generally confirmed the excellent performance of the AGR-1 fuel, first indicated during the irradiation by the observation of zero TRISO coated particle failures out of 298,000 particles in the experiment. Overall release of fission products was determined by PIE to have been relatively low during the irradiation. A significant finding was the extremely low levels of cesium released through intact coatings. This was true both during the irradiation and during post-irradiation heating tests to temperatures as high as 1800°C. Post-irradiation safety test fuel performance was generally excellent. Silver release from the particles and compacts during irradiation was often very high. Extensive microanalysis of fuel particles was performed after irradiation and after high-temperature safety testing. The results of particle microanalysis indicate that the UCO fuel is effective at controlling the oxygen partial pressure within the particle and limiting kernel migration. Post-irradiation examination has provided the final body of data that speaks to the quality of the AGR-1 fuel, building

  13. Determination of the mechanical characteristics of irradiated metals from the results of microhardness tests

    International Nuclear Information System (INIS)

    Hofman, A.

    1999-01-01

    To predict the possibilities of using structural materials in nuclear and thermonuclear reactors, it is important to have data on changes of the mechanical characteristics and irradiation obtained from full-scale or simulation tests. Materials are irradiated in nuclear reactors with fast neutrons, the sources of high-energy neutrons with an energy of 14 MeV and the accelerators of charged particles. The restricted volumes for irradiation of these specimens in the systems and also the need to test large numbers of specimens under the same conditions make it necessary to reduce the size of irradiated specimens. To solve this problem, work is being carried out to develop various methods of testing miniature specimens, including tension extrusion of disc-shaped micro-specimens, microhardness, and the Charpy Method. In examination of the irradiation hardening of the materials, the main advantage of the microhardness method is that it makes it possible to examine small specimens. In single microhardness tests, only a small area of the irradiated specimens is examined. This makes it possible to increase the radiation dose and carry out subsequent tests of microhardness on the same specimens. The aim of this work was to determine the possibilities of using the microhardness measurement method for evaluating the mechanical characteristics of metallic materials. The comparison of the data, obtained in microhardness tests and in tensile loading specimens of 0Kh18N10Tsteel, irradiated with neutrons, shows the efficiency of the microhardness method as a tool for investigating the irradiation hardening of reactor materials

  14. Fission gas induced deformation model for FRAP-T6 and NSRR irradiated fuel test simulations

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Takehiko; Sasajima, Hideo; Fuketa, Toyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hosoyamada, Ryuji; Mori, Yukihide

    1996-11-01

    Pulse irradiation tests of irradiated fuels under simulated reactivity initiated accidents (RIAs) have been carried out at the Nuclear Safety Research Reactor (NSRR). Larger cladding diameter increase was observed in the irradiated fuel tests than in the previous fresh fuel tests. A fission gas induced cladding deformation model was developed and installed in a fuel behavior analysis code, FRAP-T6. The irradiated fuel tests were analyzed with the model in combination with modified material properties and fuel cracking models. In Test JM-4, where the cladding temperature rose to higher temperatures and grain boundary separation by the pulse irradiation was significant, the fission gas model described the cladding deformation reasonably well. The fuel had relatively flat radial power distribution and the grain boundary gas from the whole radius was calculated to contribute to the deformation. On the other hand, the power density in the irradiated LWR fuel rods in the pulse irradiation tests was remarkably higher at the fuel periphery than the center. A fuel thermal expansion model, GAPCON, which took account of the effect of fuel cracking by the temperature profile, was found to reproduce well the LWR fuel behavior with the fission gas deformation model. This report present details of the models and their NSRR test simulations. (author)

  15. Irradiation tests of THTR fuel elements in the DRAGON reactor (irradiation experiment DR-K3)

    International Nuclear Information System (INIS)

    Burck, W.; Duwe, R.; Groos, E.; Mueller, H.

    1977-03-01

    Within the scope of the program 'Development of Spherical Fuel Elements for HTR', similar fuel elements (f.e.) have been irradiated in the DRAGON reactor. The f.e. were fabricated by NUKEM and were to be tested under HTR conditions to scrutinize their employability in the THTR. The fuel was in the form of coated particles moulded into A3 matrix. The kernels of the particles were made of mixed oxide of uranium and thorium with an U 235 enrichment of 90%. One aim of the post irradiation examination was the investigation of irradiation induced changes of mechanical properties (dimensional stability and elastic behaviour) and of the corrosion behaviour which were compared with the properties determined with unirradiated f.e. The measurement of the fission gas release in annealing tests and ceramografic examinations exhibited no damage of the coated particles. The measured concentration distribution of fission metals led to conclusions about their release. All results showed, that neither the coated particles nor the integral fuel spheres experienced any significant changes that could impair their utilization in the THTR. (orig./UA) [de

  16. Measurements of integrated components' parameters versus irradiation doses gamma radiation (60Co) dosimetry-methodology-tests

    International Nuclear Information System (INIS)

    Fuan, J.

    1991-01-01

    This paper describes the methodology used for the irradiation of the integrated components and the measurements of their parameters, using Quality Insurance of dosimetry: - Measurement of the integrated dose using the competences of the Laboratoire Central des Industries Electriques (LCIE): - Measurement of irradiation dose versus source/component distance, using a calibrated equipment. - Use of ALANINE dosimeters, placed on the support of the irradiated components. - Assembly and polarization of components during the irradiations. Selection of the irradiator. - Measurement of the irradiated components's parameters, using the competences of the societies: - GenRad: GR130 tests equipement placed in the DEIN/SIR-CEN SACLAY. - Laboratoire Central des Industries Electriques (LCIE): GR125 tests equipment and this associated programmes test [fr

  17. TEM in situ micropillar compression tests of ion irradiated oxide dispersion strengthened alloy

    Energy Technology Data Exchange (ETDEWEB)

    Yano, K.H., E-mail: kaylayano@u.boisestate.edu [Boise State University, 1910 University Drive, Boise, ID, 83725 (United States); Swenson, M.J. [Boise State University, 1910 University Drive, Boise, ID, 83725 (United States); Wu, Y. [Boise State University, 1910 University Drive, Boise, ID, 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd, Idaho Falls, ID, 83401 (United States); Wharry, J.P. [Boise State University, 1910 University Drive, Boise, ID, 83725 (United States); Purdue University, 400 Central Drive, West Lafayette, IN 47907 (United States)

    2017-01-15

    The growing role of charged particle irradiation in the evaluation of nuclear reactor candidate materials requires the development of novel methods to assess mechanical properties in near-surface irradiation damage layers just a few micrometers thick. In situ transmission electron microscopic (TEM) mechanical testing is one such promising method. In this work, microcompression pillars are fabricated from a Fe{sup 2+} ion irradiated bulk specimen of a model Fe-9%Cr oxide dispersion strengthened (ODS) alloy. Yield strengths measured directly from TEM in situ compression tests are within expected values, and are consistent with predictions based on the irradiated microstructure. Measured elastic modulus values, once adjusted for the amount of deformation and deflection in the base material, are also within the expected range. A pillar size effect is only observed in samples with minimum dimension ≤100 nm due to the low inter-obstacle spacing in the as received and irradiated material. TEM in situ micropillar compression tests hold great promise for quantitatively determining mechanical properties of shallow ion-irradiated layers.

  18. Irradiation facilities in JRR-3M

    International Nuclear Information System (INIS)

    Ohtomo, Akitoshi; Sigemoto, Masamitsu; Takahashi, Hidetake

    1992-01-01

    Irradiation facilities have been installed in the upgraded JRR-3 (JRR-3M) in Japan Atomic Energy Research Institute (JAERI). There are hydraulic rabbit facilities (HR), pneumatic rabbit facilities (PN), neutron activation analysis facility (PN3), uniform irradiation facility (SI), rotating irradiation facility and capsule irradiation facilities to carry out the neutron irradiation in the JRR-3M. These facilities are operated using a process control computer system to centerize the process information. Some of the characteristics for the facilities were satisfactorily measured at the same time of reactor performance test in 1990. During reactor operation, some of the tests are continued to confirm the basic characteristics on facilities, for example, PN3 was confirmed to have enough performance for activation analysis. Measurement of neutron flux at all irradiation positions has been carried out for the equilibrium core. (author)

  19. Market testing of irradiated food

    International Nuclear Information System (INIS)

    Duc, Ho Minh

    2001-01-01

    Viet Nam has emerged as one of the three top producers and exporters of rice in the world. Tropical climate and poor infrastructure of preservation and storage lead to huge losses of food grains, onions, dried fish and fishery products. Based on demonstration irradiation facility pilot scale studies and marketing of irradiated rice, onions, mushrooms and litchi were successfully undertaken in Viet Nam during 1992-1998. Irradiation technology is being used commercially in Viet Nam since 1991 for insect control of imported tobacco and mould control of national traditional medicinal herbs by both government and private sectors. About 30 tons of tobacco and 25 tons of herbs are irradiated annually. Hanoi Irradiation Centre has been continuing open house practices for visitors from school, universities and various different organizations and thus contributed in improved public education. Consumers were found to prefer irradiated rice, onions, litchi and mushrooms over those nonirradiated. (author)

  20. Application of half-embryo test to irradiated apples and cherries

    International Nuclear Information System (INIS)

    Kawamura, Yoko; Miura, Aya; Sugita, Takiko; Yamada, Takashi; Saito, Yukio

    1995-01-01

    The half-embryo test was applied to irradiated apples and cherries. The optimum incubation temperature for apples and cherries was 30 o C and 25 o C, respectively. Benzyladenine stimulated the shooting of cherry half-embryos, therefore, they were incubated with 10 μM benzyladenine. The irradiation of apples and cherries caused obvious changes in the growth of the half-embryos. A dose of 0.15 kGy or more almost totally retarded shoot elongation. If shooting is less than 50%, the apples and cherries are identified as ''irradiated''. An assessment could be made after 1 to 4 days and the detection limit of the irradiation dose is 0.15 kGy. (author)

  1. Creep test under irradiation with thermal gradient for the cylindrical carbon fiber reinforced carbon composite. Interim report on irradiation examinations: 03M-47AS

    International Nuclear Information System (INIS)

    Baba, Shin-ichi; Sawa, Kazuhiro; Yamaji, Masatoshi; Matsui, Yoshinori; Ishihara, Masahiro

    2007-03-01

    The creep test under irradiation with thermal gradient for the cylindrical carbon fiber reinforced carbon composites (c/c composite) are carried out in the Japan Material Testing Reactor (JMTR). This report described 4-items; first item is design/evaluation of the capsule for the irradiation test, second is before irradiation measurements for the residual strain due to manufactured cylindrical c/c composite, and third is also before irradiation measurements of the distance between 2-holes of predrilled in the specimen and last item is examination of analysis for the irradiation creep with thermal gradient by VIENUS Code. The normal creep test is static mechanical load on the specimen in thermal condition, but this creep test under irradiation capsule is thermal stress due to thermal gradient at inside specimen in the thermal condition. Consequently, it is necessary as large as possible thermal gradient in the narrow space of the capsule inside volume. In which the tungsten rod (W-rod) was inserted to the cylindrical c/c composite specimen, for γ-ray heat generation density occurred highly and so maximize the difference temperatures of surface wall between inside and outside wall of the specimen. The measurement method of the deflection due to the irradiation creep of cylindrical c/c composite was adopted as way of ruptured the specimen among the predrilled distance of 2-holes before/after irradiation. Accordingly as the laser dimensional apparatus used to measure the distance between the 2-holes of specimen exactly, easy and untouchable. And also before irradiation measurement of the residual stress due to the manufactured process was estimated by neutron diffraction used Residual Stress Analyzer (RESA) at JRR-3M in JAEA. The irradiation test was finished as total irradiation time, average temperature and neutron dose showed 4189 hours, 873 K and 8.2x10 24 (E>1.0MeV:m -2 ) respectively. The thermal stress was estimated by the difference temperatures of 4

  2. Use of EBR-II as a principal fast breeder reactor irradiation test facility in the U.S

    International Nuclear Information System (INIS)

    Staker, R.G.; Seim, O.S.; Beck, W.N.; Golden, G.H.; Walters, L.C.

    1975-01-01

    The EBR-II as originally designed and operated by the Argonne National Laboratory was successful in demonstrating the operation of a sodium-cooled fast breeder power plant with a closed fuel reprocessing cycle. Subsequent operation has been as an experimental facility where thousands of irradiation tests have been performed. Conversion to this application entailed the design and fabrication of special irradiation subassemblies for in-core irradiations, additions to existing facilities for out-of-core irradiations, and additions to existing facilities for out-of-core experiments. Experimental subassemblies now constitute about one third of the core, and changes in the core configuration occur about monthly, requiring neutronic and thermal-hydraulics analyses and monitoring of the reactor dynamic behavior. The surveillance programs provided a wealth of information on irradiation induced swelling and creep, in-reactor fracture behavior, and the compatibility of materials with liquid sodium. (U.S.)

  3. Consumer acceptance, market test and market development of irradiated rice, dehydrated vegetables and spices

    International Nuclear Information System (INIS)

    Shi Peixin; Lin Yin

    2001-01-01

    Establishment of irradiation processing parameters, a quality assurance system, consumer acceptance, market test and market development of irradiated rice, dehydrated vegetables and spices were the activities carried out in this project by the Chinese Agricultural Irradiation Center. The results of the studies showed that the process dose for rice was 0.2-0.5 kGy when the non-uniformity was lower than 2.5, dose range for dehydrated vegetables was 5-7 kGy, dose for spices was 7-8 kGy. The system for quality assurance was established. The processing standards for several irradiated food items were set up. Market test showed that more than 70-80% of consumers accepted irradiated food. Industrial companies also accepted irradiated dehydrated vegetables and spices. The latter were successfully introduced to the markets and successful commercialization of irradiated garlic was followed. The economic benefit of operating the Chinese Agricultural Irradiation Center was analyzed and found attractive, especially for low dose irradiation of foods in sufficient supply. (author)

  4. Consumer acceptance, market test and market development of irradiated rice, dehydrated vegetables and spices

    Energy Technology Data Exchange (ETDEWEB)

    Peixin, Shi; Yin, Lin [Chinese Agricultural Irradiation Center, Institute for Application of Atomic Energy, Chinese Academy of Agricultural Sciences, Beijing (China)

    2001-05-01

    Establishment of irradiation processing parameters, a quality assurance system, consumer acceptance, market test and market development of irradiated rice, dehydrated vegetables and spices were the activities carried out in this project by the Chinese Agricultural Irradiation Center. The results of the studies showed that the process dose for rice was 0.2-0.5 kGy when the non-uniformity was lower than 2.5, dose range for dehydrated vegetables was 5-7 kGy, dose for spices was 7-8 kGy. The system for quality assurance was established. The processing standards for several irradiated food items were set up. Market test showed that more than 70-80% of consumers accepted irradiated food. Industrial companies also accepted irradiated dehydrated vegetables and spices. The latter were successfully introduced to the markets and successful commercialization of irradiated garlic was followed. The economic benefit of operating the Chinese Agricultural Irradiation Center was analyzed and found attractive, especially for low dose irradiation of foods in sufficient supply. (author)

  5. Test marketing and consumer acceptance of irradiated meat products

    International Nuclear Information System (INIS)

    Xu Zhicheng; Feng Zhixiong; Jiang Peizhen

    2001-01-01

    This study consists of two parts: irradiation processing of cooked meat and irradiation preservation of prepackaged chilled fresh cut meats. Irradiation of prepackaged pickled meat products dipped in grains stillage at a dose 6-8 kGy eliminated common food-borne microorganisms, such as E. Coli and other microbial pathogens and extended the shelf life of the product to 10 days at 5 deg. C. Test marketing of 40,000 bags (about 10,000 kg) of the product in more than 100 supermarkets in the city of Shanghai showed no untoward problem with consumer acceptance. Irradiation of prepackaged chilled fresh cut pork at a dose 3 kGy led to inactivation of microbial pathogens and parasites with a concomitant reduction in numbers of common spoilage microorganisms and extension of shelf life of the product for 30 days at 5 deg. C. The cost benefit and marketing applications were evaluated. (author)

  6. Thermal shock testing of ceramics with pulsed laser irradiation

    International Nuclear Information System (INIS)

    Benz, R.; Naoumidis, A.; Nickel, H.

    1986-04-01

    Arguments are presented showing that the resistance to thermal stressing (''thermal shock'') under pulsed thermal energy deposition by various kinds of beam irradiations is approximately proportional to Φ a √tp, where Φ a is the absorbed power density and tp is the pulse length, under conditions of diffusivity controlled spreading of heat. In practical beam irradiation testing, incident power density, Φ, is reported. To evaluate the usefulness of Φ√tp as an approximation to Φ a √tp, damage threshold values are reviewed for different kinds of beams (electron, proton, and laser) for a range of tp values 5x10 -6 to 2 s. Ruby laser beam irradiation tests were made on the following ceramics: AlN, BN, graphite, αSiC, β-SiC coated graphites, (α+β)Si 3 N 4 , CVD (chemical vapor deposition) TiC coated graphite, CVD TiC coated Mo, and CVD TiN coated IN 625. The identified failure mechanisms are: 1. plastic flow followed by tensile and bend fracturing, 2. chemical decomposition, 3. melting, and 4. loss by thermal spallation. In view of the theoretical approximations and the neglect of reflection losses there is reasonable accord between the damage threshold Φ√tp values from the laser, electron, and proton beam tests. (orig./IHOE)

  7. Irradiation facilitates at the advanced test reactor

    International Nuclear Information System (INIS)

    Grover, Blaine S.

    2006-01-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC - formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950's with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens. The paper has the following contents: ATR description and capabilities; ATR operations, quality and safety requirements; Static capsule experiments; Lead experiments; Irradiation test vehicle; In-pile loop experiments; Gas test loop; Future testing; Support facilities at RTC; Conclusions. To summarize, the ATR has a long history in fuel and material irradiations, and will be fulfilling a critical role in the future fuel and material testing necessary to develop the next generation reactor systems and advanced fuel cycles. The

  8. Irradiation tests of ATLAS liquid argon forward calorimeter (FCAL) electronics components

    CERN Document Server

    Leroy, C; Golikov, V; Golubyh, S M; Kukhtin, V; Kulagin, E; Luschikov, V; Merkulov, L; Minashkin, V F; Shalyugin, A N

    2002-01-01

    FCAL resistors, capacitors, and transformers together with capacitors and sintimid disks of the purity monitor have been irradiated in liquid argon to study their possible lead to argon pollution at a maximal neutron fluence of 1016 n cm-2 at the IBR-2 reactor of JINR, Dubna. The results of charge collection measurements before and after irradiation are reported. Electrical measurement on these FCAL capacitors, resistors and transformers were also performed after irradiation. In general, the results of resistance, capacitance, impedance, leakage current and high voltage breakdown measurements after irradiation show minor changes of value only for some parameters from nominal values or values measured before irradiation.

  9. The irradiation performance of austenitic stainless steel clade PWR fuel rods

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Esteves, A.M.

    1988-01-01

    The steady state irradiation performance of austenitic stainless steel clad pressurized water reactor fuel rods is modeled with fuel performance codes of the FRAP series. These codes, originally developed to model the thermal-mechanical behavior of zircaloy clad fuel rods, are modified to model stainless steel clad fuel rods. The irradiation thermal-mechanical behavior of type 348 stainless steel and zircaloy fuel rods is compared. (author) [pt

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  11. Evaluation of Candidate Linear Variable Displacement Transducers for High Temperature Irradiations in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Knudson, D.L.; Rempe, J.L.; Daw, J.E.

    2009-01-01

    in performance that may be introduced as a result of high temperature operation in a radiation environment. The performance assessment focuses on the potential for any changes or degradation in sensitivity and/or electrical resistance. After initial testing, additional testing was completed for the more promising design to detect the changes that occur when some of its components are subjected to their Curie temperature. Results from this study are provided as the first step in qualifying these sensors for possible ATR use. This report will be submitted to the Institute for Energy at the Halden Reactor Project (IFE-HRP) for confirmation of appropriate test procedures. Furthermore, this report will be updated during fiscal year 2010 to include results from the evaluation of 'developmental' LVDTs to be provided by IFE-HRP. This document presents results from laboratory evaluations and results from scoping calculations completed to assess the possible effects of gamma heating during irradiation for the most promising LVDT design. Section 2 of this document provides background information related to LVDT operation and both designs evaluated in this effort. Section 3 contains a description of tests needed to ensure that LVDT outputs will not be adversely affected by fixturing and/or the proximity of other LVDTs during this evaluation. Results from calibration evaluations are provided in Section 4 and results from long duration tests are presented in Section 5. Curie temperature evaluation results are documented in Section 6. Scoping calculations to assess the impact of gamma heating during LVDT irradiation are presented in Section 7. Recommendations from these evaluations are documented in Section 8. References cited in this report are listed in Section 9.

  12. Testing of neutron-irradiated ceramic-to-metal seals

    International Nuclear Information System (INIS)

    Brown, R.D.; Clinard, F.W. Jr.; Lopez, M.R.; Martinez, H.; Romero, T.J.; Cook, J.H.; Barr, H.N.; Hittman, F.

    1990-01-01

    This paper reports on ceramic-to-metal seals prepared by sputtering a titanium metallizing layer onto ceramic disks and then brazing to metal tubes. The ceramics used were alumina, MACOR, spinel, AlON, and a mixture of Al 2 O 3 and Si 3 N 4 . Except for the MACOR, which was brazed to a titanium tube, the ceramics were brazed to niobium tubes. The seals were leak tested and then sent to Los Alamos National Laboratory, where they were irradiated using the spallation neutron source at the Los Alamos Meson Physics Facility. Following irradiation for ∼ 90 days to a fluence of 2.8 x 10 23 n/m 2 , the samples were moved to hot cells and again leak tested. Only the MACOR samples showed any measurable leaks. One set of samples was then pressurized to 6.9 MPa (1000 psi) and subsequently leak tested. No leaks were found. Bursting the seals required hydrostatic pressures of at least 34 MPa (5000 psi). The high seal strength and few leaks indicate that ceramic-to-metal seals can resist radiation-induced degradation

  13. An investigation of the genetic toxicology of irradiated foodstuffs using short-term test systems

    International Nuclear Information System (INIS)

    Phillips, B.J.; Kranz, E.; Elias, P.S.

    1980-01-01

    As part of a programme of short-term tests used to detect possible genetic toxicity in irradiated foodstuffs, cultured Chinese hamster ovary cells were exposed to extracts and digests of irradiated and unirradiated dates, fish and chicken and subjected to tests for cytotoxicity, sister chromatid exchange induction and mutation to thioguanine resistance. The results showed no evidence of genetic toxicity induced in food by irradiation. The general applicability of cell culture tests to the detection of mutagens in food is discussed. (author)

  14. Irradiation test OF-2: high-temperature irradiation behavior of LASL-made fuel rods and LASL-made coated particles

    International Nuclear Information System (INIS)

    Wagner, P.; Reiswig, R.D.; Hollabaugh, C.M.; White, R.W.; O'Rourke, J.A.; Davidson, K.V.; Schell, D.H.

    1977-10-01

    Three LASL-made, substoichiometric ZrC-coated particles with inert kernels, and two high-density molded graphite fuel rods that contained LASL-made, ZrC-coated fissile particles were irradiated in the Oak Ridge Research Reactor test OF-2. The severest test conditions were 8.36 x 10 21 nvt (E greater than 0.18 MeV) at 1350 0 C. The graphite matrix showed no effect of the irradiation. There was no interaction between the matrix and any of the particle coats. The loose ZrC coated particles with inert kernels showed no irradiation effects. The graded ZrC-C coats on the fissile particles were cracked. It is postulated that the cracking is associated with the low LTI deposition rate and is not related to the ZrC

  15. Monte Carlo studies for irradiation process planning at the Portuguese gamma irradiation facility

    International Nuclear Information System (INIS)

    Oliveira, C.; Salgado, J.; Botelho, M.L.M. Luisa; Ferreira, L.M.

    2000-01-01

    The paper describes a Monte Carlo study for planning the irradiation of test samples for microbiological validation of distinct products in the Portuguese Gamma Irradiation Facility. Three different irradiation geometries have been used. Simulated and experimental results are compared and good agreement is observed. It is shown that Monte Carlo simulation improves process understanding, predicts absorbed dose distributions and calculates dose uniformity in different products. Based on these results, irradiation planning of the product can be performed

  16. Detrimental Effects of Helium Ion Irradiation on Cognitive Performance and Cortical Levels of MAP-2 in B6D2F1 Mice.

    Science.gov (United States)

    Raber, Jacob; Torres, Eileen Ruth S; Akinyeke, Tunde; Lee, Joanne; Weber Boutros, Sydney J; Turker, Mitchell S; Kronenberg, Amy

    2018-04-20

    The space radiation environment includes helium (⁴He) ions that may impact brain function. As little is known about the effects of exposures to ⁴He ions on the brain, we assessed the behavioral and cognitive performance of C57BL/6J × DBA2/J F1 (B6D2F1) mice three months following irradiation with ⁴He ions (250 MeV/n; linear energy transfer (LET) = 1.6 keV/μm; 0, 21, 42 or 168 cGy). Sham-irradiated mice and mice irradiated with 21 or 168 cGy showed novel object recognition, but mice irradiated with 42 cGy did not. In the passive avoidance test, mice received a slight foot shock in a dark compartment, and latency to re-enter that compartment was assessed 24 h later. Sham-irradiated mice and mice irradiated with 21 or 42 cGy showed a higher latency on Day 2 than Day 1, but the latency to enter the dark compartment in mice irradiated with 168 cGy was comparable on both days. ⁴He ion irradiation, at 42 and 168 cGy, reduced the levels of the dendritic marker microtubule-associated protein-2 (MAP-2) in the cortex. There was an effect of radiation on apolipoprotein E (apoE) levels in the hippocampus and cortex, with higher apoE levels in mice irradiated at 42 cGy than 168 cGy and a trend towards higher apoE levels in mice irradiated at 21 than 168 cGy. In addition, in the hippocampus, there was a trend towards a negative correlation between MAP-2 and apoE levels. While reduced levels of MAP-2 in the cortex might have contributed to the altered performance in the passive avoidance test, it does not seem sufficient to do so. The higher hippocampal and cortical apoE levels in mice irradiated at 42 than 168 cGy might have served as a compensatory protective response preserving their passive avoidance memory. Thus, there were no alterations in behavioral performance in the open filed or depressive-like behavior in the forced swim test, while cognitive impairments were seen in the object recognition and passive avoidance tests, but not in the contextual or cued fear

  17. Detrimental Effects of Helium Ion Irradiation on Cognitive Performance and Cortical Levels of MAP-2 in B6D2F1 Mice

    Directory of Open Access Journals (Sweden)

    Jacob Raber

    2018-04-01

    Full Text Available The space radiation environment includes helium (4He ions that may impact brain function. As little is known about the effects of exposures to 4He ions on the brain, we assessed the behavioral and cognitive performance of C57BL/6J × DBA2/J F1 (B6D2F1 mice three months following irradiation with 4He ions (250 MeV/n; linear energy transfer (LET = 1.6 keV/μm; 0, 21, 42 or 168 cGy. Sham-irradiated mice and mice irradiated with 21 or 168 cGy showed novel object recognition, but mice irradiated with 42 cGy did not. In the passive avoidance test, mice received a slight foot shock in a dark compartment, and latency to re-enter that compartment was assessed 24 h later. Sham-irradiated mice and mice irradiated with 21 or 42 cGy showed a higher latency on Day 2 than Day 1, but the latency to enter the dark compartment in mice irradiated with 168 cGy was comparable on both days. 4He ion irradiation, at 42 and 168 cGy, reduced the levels of the dendritic marker microtubule-associated protein-2 (MAP-2 in the cortex. There was an effect of radiation on apolipoprotein E (apoE levels in the hippocampus and cortex, with higher apoE levels in mice irradiated at 42 cGy than 168 cGy and a trend towards higher apoE levels in mice irradiated at 21 than 168 cGy. In addition, in the hippocampus, there was a trend towards a negative correlation between MAP-2 and apoE levels. While reduced levels of MAP-2 in the cortex might have contributed to the altered performance in the passive avoidance test, it does not seem sufficient to do so. The higher hippocampal and cortical apoE levels in mice irradiated at 42 than 168 cGy might have served as a compensatory protective response preserving their passive avoidance memory. Thus, there were no alterations in behavioral performance in the open filed or depressive-like behavior in the forced swim test, while cognitive impairments were seen in the object recognition and passive avoidance tests, but not in the contextual or cued

  18. Study for improvement of performance of the test and research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Fumio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    Current utilization needs for the test and research reactors become more advanced and diversified along with the advance of nuclear science and technology. Besides, the requested safety for the research and test reactors grows strictly every year as well as a case of the power reactors. Under this circumstance, every effort to improve reactor performance including its safety is necessary to be sustained for allowing more effective utilization of the test and research reactors as experimental apparatus for advanced researches. In this study, the following three themes i.e., JMTR high-performance fuel element, evaluation method of fast neutron irradiation dose in the JMTR, evaluation method of performance of siphon break valve as core covering system for water-cooled test and research reactors, were investigated respectively from the views of improvement of core performance as a neutron source, utilization performance as an experimental apparatus, and safety as a reactor plant. (author)

  19. 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

  20. MOX fuel irradiation behavior in steady state (irradiation test in HBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Kohno, S; Kamimura, K [Power Reactor and Nuclear Fuel Development Corp., Naka, Ibaraki (Japan)

    1997-08-01

    Two rigs of plutonium-uranium oxide (MOX) fuel rods have been irradiated in Halden boiling water reactor (HBWR) to investigate high burnup MOX fuel behavior for thermal reactor. The objective of irradiation tests is to investigate fuel behavior as influenced by pellet shape, pellet surface treatment, pellet-cladding gap size and MOX fuel powder preparations process. The two rigs have instrumentations for in-pile measurements of the fuel center-line temperature, plenum pressure, cladding elongation and fuel stack length change. The data, taken through in-operation instrumentation, have been analysed and compared with those from post-irradiation examination. The following observations are made: 1) PNC MOX fuels have achieved high burn-up as 59GWd/tMOX (67GWd/tM) at pellet peak without failure; 2) there was no significant difference in fission gas release fraction between PNC MOX fuels and UO{sub 2} fuels; 3) fission gas release from the co-converted fuel was lower than that from the mechanically blended fuel; 4) gap conductance was evaluated to decrease gradually with burn-up and to get stable in high burn-up region. 5) no evident difference of onset LHR for PCMI in experimental parameters (pellet shape and pellet-cladding gap size) was observed, but it decreased with burn-up. (author). 13 refs, 15 figs, 3 tabs.

  1. Fusion Materials Irradiation Test Facility

    International Nuclear Information System (INIS)

    Kemp, E.L.; Trego, A.L.

    1979-01-01

    A Fusion Materials Irradiation Test Facility is being designed to be constructed at Hanford, Washington, The system is designed to produce about 10 15 n/cm-s in a volume of approx. 10 cc and 10 14 n/cm-s in a volume of 500 cc. The lithium and target systems are being developed and designed by HEDL while the 35-MeV, 100-mA cw accelerator is being designed by LASL. The accelerator components will be fabricated by US industry. The total estimated cost of the FMIT is $105 million. The facility is scheduled to begin operation in September 1984

  2. Transfer of test samples and wastes between post-irradiation test facilities (FMF, AGF, MMF)

    International Nuclear Information System (INIS)

    Ishida, Yasukazu; Suzuki, Kazuhisa; Ebihara, Hikoe; Matsushima, Yasuyoshi; Kashiwabara, Hidechiyo

    1975-02-01

    Wide review is given on the problems associated with the transfer of test samples and wastes between post-irradiation test facilities, FMF (Fuel Monitoring Facility), AGF (Alpha Gamma Facility), and MMF (Material Monitoring Facility) at the Oarai Engineering Center, PNC. The test facilities are connected with the JOYO plant, an experimental fast reactor being constructed at Oarai. As introductory remarks, some special features of transferring irradiated materials are described. In the second part, problems on the management of nuclear materials and radio isotopes are described item by item. In the third part, the specific materials that are envisaged to be transported between JOYO and the test facilities are listed together with their geometrical shapes, dimensions, etc. In the fourth part, various routes and methods of transportation are explained with many block charts and figures. Brief explanation with lists and drawings is also given to transportation casks and vessels. Finally, some future problems are discussed, such as the prevention of diffusive contamination, ease of decontamination, and the identification of test samples. (Aoki, K.)

  3. Small Punch Test Techniques for Irradiated Materials in Hot Cell

    International Nuclear Information System (INIS)

    Kim, Do Sik; Ahn, S. B.; Oh, W. H.; Yoo, B. O.; Choo, Y. S.

    2006-06-01

    Detailed procedures of the small punch test including the apparatus, the definition of small punch-related parameters, and the interpretation of results were presented. The testing machine should have a capability of the compressive loading and unloading at a given deflection level. The small punch specimen holder consists of an upper and lower die and clamping screws. The clamped specimen is deformed by using ball or spherical head punch. Two type of specimens with a circular and a square shape were used. The irradiated small punch specimen is made from the undamaged portion of the broken CVN bars or prepared by the irradiation of the specimen fabricated from the fresh materials. The heating and cooling devices should have the capability of the temperature control within ±2 .deg. C for the target value during the test. Based on the load-deflection data obtained from the small punch test. the empirical correlation between the small punch related parameters and a tensile properties such as 0.2% yield strength and ultimate tensile strength, fracture toughness, ductile-brittle transition temperature and creep properties determined from the standard test method is established and used to evaluate the mechanical properties of an irradiated materials. In addition, from the quantitative fractographic assessment of small punch test specimens, the relationship between the small punch energy and the quantity of ductile crack growth is obtained. Analytical formulations demonstrated good agreement with experimental load-deflection curves

  4. Physics calculations for the RIA 1-3 irradiated rod test

    International Nuclear Information System (INIS)

    Young, T.E.

    1981-06-01

    The RIA 1-3 test would employ a square array of four pre-irradiated BWR rods to provide information on fuel failure modes and consequences of postulated Reactivity Initiated Accidents in power reactors. Calculations were done to: (1) predict R-O power distributions in the test rods for thermal-hydraulic and fuel-failure analysis; and (2) predict the steady-state and transient ratios of test fuel energy deposition to core energy deposition (Figures of Merit). Fission distributions for the test were computed with the RAFFL Monte Carlo code using an external neutron current source from a complete-reactor radial calculation with the SCAMP S/sub n/ code. Energies per fission for the rods were computed using the SINBAD buildup and depletion code, the GAMSOR gamma ray source code, and the QAD-BSA point-kernel shielding code. The calculated rod average-to-test average energy deposition ratios are 0.99, 0.99, and 0.97 for the rods irradiated to approximately 12 CWd/tu, and 1.04 for the rod irradiated to 4.8 GWd/tu. The maximum deviation of the power density of 1/12-rod azimuthal segments from the rod average is 4%. For an estimated control rod position of 0.591 m withdrawn the predicted radial average energy deposition at the axial peak in an average test rod is 1.71 (kW/m)/MW during preconditioning, and 1.84 (kJ/kg UO 2 ) MW.S during the burst. 16 figures, 7 tables

  5. Fusion Materials Irradiation Test Facility: a facility for fusion-materials qualification

    International Nuclear Information System (INIS)

    Trego, A.L.; Hagan, J.W.; Opperman, E.K.; Burke, R.J.

    1983-01-01

    The Fusion Materials Irradiation Test Facility will provide a unique testing environment for irradiation of structural and special purpose materials in support of fusion power systems. The neutron source will be produced by a deuteron-lithium stripping reaction to generate high energy neutrons to ensure damage similar to that of a deuterium-tritium neutron spectrum. The facility design is now ready for the start of construction and much of the supporting lithium system research has been completed. Major testing of key low energy end components of the accelerator is about to commence. The facility, its testing role, and the status and major aspects of its design and supporting system development are described

  6. AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Phil [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ploger, Scott A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-12-01

    The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance, and fission product transport (INL 2015). A series of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously.

  7. AGR-1 Compact 1-3-1 Post-Irradiation Examination Results

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-12-01

    The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport (INL 2015). A series of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously (Grover, Petti, and Maki 2010, Maki 2009).

  8. Metal fuel manufacturing and irradiation performance

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Walters, L.C.

    1992-01-01

    The advances in metal fuel by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, and improved passive safety. The goals and the safety philosophy of the Integral Fast Reactor Program are stressed

  9. Water chemistry of the JMTR IASCC irradiation loop system

    International Nuclear Information System (INIS)

    Hanawa, Satoshi; Oogiyanagi, Jin; Mori, Yuichiro; Saito, Junichi; Tsukada, Takashi

    2006-01-01

    Irradiation assisted stress corrosion cracking (IASCC) is recognized as an important degradation issue of the core-internal material for aged Boiling Water Reactors (BWRs). Therefore, irradiation loop system has been developed and installed in the Japan Materials Testing Reactor to perform the IASCC irradiation test. In the IASCC irradiation test, water chemistry of irradiation field is one of the most important key parameters because it affects initiation and propagation of cracks. This paper summarizes the measurement and evaluation method of water chemistry of IASCC irradiation loop system. (author)

  10. Performance of HTGR fertile particles irradiated in HFIR capsule HT-32

    International Nuclear Information System (INIS)

    Long, E.L. Jr.; Robbins, J.M.; Tiegs, T.N.; Kania, M.J.

    1980-04-01

    The HT-32 experiment was an uninstrumented capsule irradiated for four cycles in the target position of the High-Flux Isotope Reactor (HFIR). The experiment was designed to: provide supplemental simulated fuel rods for thermal transport and expansion measurements; test fertile kernels with Al 2 O 3 and SiO 2 additives for improved fission product retention; study the stability and permeability of low-temperature isotropic (LTI) pyrocarbon coatings; test Biso- and Triso-coatings derived in a large (0.24-m-dia) coating furnace with a frit distributor; investigate the performance of particles with an outer layer of SiC both as loose particles and as resin-bonded fuel rods; and evaluate high-density alumina as a potential high-temperature thermometry sheathing material

  11. Utilization of DNA comet assay and half embryo test to identify irradiated lentil

    International Nuclear Information System (INIS)

    Romanelli, Maria Fernanda; Villavicencio, Anna Lucia C.H.

    2001-01-01

    Full text: Legumes make an important contribution to human nutrition on a worldwide basis. Insect infestation cause extensive damage to stored grains. Over the last few decades some countries adopted food irradiation as a safe food process. Radiation's processing on foods improves hygienic quality and extends their shelf life. The use of radiation treatment to reduce the microbial population and thereby extend the shelf life in legumes has been reported in many papers. Irradiation has been shown to be an effective pest control method for these commodities and a good alternative to prohibited methyl bromide. Radiation disinfestation can facilitate trade in foods that often harbor insect pests of quarantine importance. Although the wholesomeness of irradiated food is no longer a question there is a need for irradiation control in the international trade of foods, in order to enhance the consumer confidence in the regulation. As a screening methods to identify irradiated lentils, processed by e-beam as a food treatment to disinfestation, the DNA Comet Assay and Half Embryo tests were performed. The methodologies used in this work are based upon biological changes that occur in Brazilian lentils. The samples were irradiated in an electron beam accelerator facility of Radiation Dynamics Inc., USA (E=1,5 MeV, l=25 mA). The irradiation doses were 0,7; 1,4 and 3,0 kGy at dry conditions. The thickness of samples was less than 0,5 cm. A sensitive technique to detect DNA fragmentation is the microgel electrophoresis of single cells or nuclei, also called 'comet assay'. Since the large molecule of DNA is an easy target for ionizing radiation, changes in DNA offer potential as a detection method. It is restricted to foods that have not been subjected to heat or other treatments, which also cause DNA fragmentation. Lentil samples were crushed with a mortar and pestle and was transferred to 3ml ice-cold PBS. This suspension was stirred for 5 minutes and filtered. 100μl cell

  12. Irradiation of inert matrix and mixed oxide fuel in the Halden test reactor

    International Nuclear Information System (INIS)

    Hellwig, Ch.; Kasemeyer, U.

    2001-01-01

    In a new type of fuel, called Inert Matrix Fuel (IMF), plutonium is embedded in a U-free matrix. This offers advantages for more efficient plutonium consumption, higher proliferation resistance, and for inert behaviour later in a waste repository. In the fuel type investigated at PSI, plutonium is dissolved in yttrium-stabilized zirconium oxide (YSZ), a highly radiation-resistant cubic phase, with addition of erbium as burnable poison for reactivity control. A first irradiation experiment of YSZ-based IMF is ongoing in the OECD Material Test Reactor in Halden (HBWR), together with MOX fuel (Rig IFA-651.1). The experiment is described herein and results are presented of the first 120 days of irradiation with an average assembly burnup of 47 kWd/cm 3 . The results are compared with neutronic calculations performed before the experiment, and are used to model the fuel behaviour with the PSI-modified TRANSURANUS code. The measured fuel temperatures are within the expected range. An unexpectedly strong densification of the IMF during the first irradiation cycle does not alter the fuel temperatures. An explanation for this behaviour is proposed. The irradiation at higher linear heat rates during forthcoming cycles will deliver information about the fission gas release behaviour of the IMF. (author)

  13. Irradiation of inert matrix and mixed oxide fuel in the Halden test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hellwig, Ch.; Kasemeyer, U

    2001-03-01

    In a new type of fuel, called Inert Matrix Fuel (IMF), plutonium is embedded in a U-free matrix. This offers advantages for more efficient plutonium consumption, higher proliferation resistance, and for inert behaviour later in a waste repository. In the fuel type investigated at PSI, plutonium is dissolved in yttrium-stabilized zirconium oxide (YSZ), a highly radiation-resistant cubic phase, with addition of erbium as burnable poison for reactivity control. A first irradiation experiment of YSZ-based IMF is ongoing in the OECD Material Test Reactor in Halden (HBWR), together with MOX fuel (Rig IFA-651.1). The experiment is described herein and results are presented of the first 120 days of irradiation with an average assembly burnup of 47 kWd/cm{sup 3}. The results are compared with neutronic calculations performed before the experiment, and are used to model the fuel behaviour with the PSI-modified TRANSURANUS code. The measured fuel temperatures are within the expected range. An unexpectedly strong densification of the IMF during the first irradiation cycle does not alter the fuel temperatures. An explanation for this behaviour is proposed. The irradiation at higher linear heat rates during forthcoming cycles will deliver information about the fission gas release behaviour of the IMF. (author)

  14. Non-destructive test for irradiated fuels using X-ray CT system in hot-laboratory

    International Nuclear Information System (INIS)

    Kim, Heemoon; Kim, Gil-Soo; Yoo, Boung-Ok; Tahk, Young-Wook; Cho, Moon-Sung; Ahn, Sang-Bok

    2015-01-01

    To inspect inside of irradiated fuel rod for PIE in hotcell, neutron beam and X-ray have been used. Many hot laboratories in the world have shown the results for NDT by 2-D film data. Currently, computed image processing technology instead of film has been developed and CT was applied to the X-ray and neutron beam system. In this trend, our facility needed to set up X-ray system for irradiated fuel inspection and installed in hotcell with consideration of radiation damage. In this study, X-ray system was tested to be operated with radioactive samples and was performed to inspect fuel rods and observe internal damage and dimensional change. 450kV X-ray CT system was installed in hotcell with modification and tested to check image resolution and radiation damage. The image data were analyzed by 3-D computer software. 8 fuel plates and VHTR rods were inspected and measured internal shape and dimension

  15. Minaturized disk bend tests of neutron-irradiated path A type alloys

    International Nuclear Information System (INIS)

    Lee, M.; Sohn, D.S.; Grant, N.J.; Harling, O.K.

    1983-01-01

    Path A Prime Candidate Alloy (PCA) has been rapidly solidified and consoliated by extrusion. Twenty percent CW samples, precision TEM disks, 3 phi x 0.254 mm, were irradiated in the mixed flux of the Oak Ridge HFIR reactor up to approx. 8.5 dpa (360 appm He) and approx. 34 dpa (3100 appm He) at 300, 400, 500 and 600 0 C. Similar samples of conventionally processed PCA were also irradiated for comparison. Mechanical properties were characterized using a minaturized disk bend test (MDBT) developed at MIT. These tests indicate major decreases in strength and ductility especially for the 500 and 600 0 C irradiations. No major differences were found between this first version of a rapidly solidified and extruded PCA type alloy and conventionally processed PCA

  16. Irradiation tests of readout chain components of the ATLAS liquid argon calorimeters

    International Nuclear Information System (INIS)

    Leroy, C.; Cheplakov, A.; Golikov, V.; Golubykh, S.; Kukhtin, V.; Kulagin, E.; Lushchikov, V.; Minashkin, V.; Shalyugin, A.

    2000-01-01

    Various readout chain components of the ATLAS liquid argon calorimeters have been exposed to high neutron fluences and γ doses at the irradiation test facility of the IBR-2 reactor of JINR, Dubna. Results of the capacitance and impedance measurements of coaxial cables are presented. Results of peeling tests of PC board samples (carton and copper strips) as a measure of the bonding agent irradiation hardness are also reported

  17. Irradiation tests of readout chain components of the ATLAS liquid argon calorimeters

    CERN Document Server

    Leroy, C; Golikov, V; Golubyh, S M; Kukhtin, V; Kulagin, E; Luschikov, V; Minashkin, V F; Shalyugin, A N

    1999-01-01

    Various readout chain components of the ATLAS liquid argon calorimeters have been exposed to high neutron fluences and $gamma$-doses at the irradiation test facility of the IBR-2 reactor of JINR, Dubna. Results of the capacitance and impedance measurements of coaxial cables are presented. Results of peeling tests of PC board samples (kapton and copper strips) as a measure of the bonding agent irradiation hardness are also reported.

  18. Performance of Broiler Chicks Fed Irradiated Sorghum Grains

    International Nuclear Information System (INIS)

    Farag, M.D.D.; Farag, M.F. S. El-D.; Afify, A.S.

    2003-01-01

    Substitution of yellow corn with raw sorghum grains in chick diets resulted in decreases in live body weight, accumulative feed consumption and efficiency of feed utilization as compared with reference diet. Relative to raw sorghum diet, inclusion of sorghum grains irradiated at 60 and 100 kGy and/or supplemented with PEG in chick diets resulted in increases in accumulative feed consumption an efficiency feed utilization. The study suggested that irradiation treatment up to 100 kGy up grade broiler chicks performance and the combinations between radiation and PEG treatments sustain the effect of each other

  19. Comet assay in the detection of irradiated garlic; Teste do cometa na deteccao de alho irradiado

    Energy Technology Data Exchange (ETDEWEB)

    Villavicencio, Anna Lucia C.H.; Marin-Huachaca, Nelida Simona; Romanelli, Maria Fernanda [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: villavic@net.ipen.br; Delincee, Henry [Federal Research Centre for Nutrition - BFE, Karlsruhe (Germany)]. E-mail: henry.delincee@bfe.uni-karlsruhe.de

    2002-07-01

    The increased claim for fresh produce has forced a consensus between nations to pay more attention to the phytosanitary regulations. Inhibition of sprouting of bulbs and tubers by applying ionising radiation is authorised by the National Food Codes in Brazil. The availability of methods for detection of irradiated food will contribute to increase consumers' confidence. A quick and simple screening test to indicate whether a food product has been irradiated or not was utilised in this study. The DNA comet assay was applied to verify whether garlic imported from China had been irradiated or not. This test has already been adopted as a European Standard (EN 13784), for detection of irradiated food. Non-irradiated control samples of garlic and garlic treated with maleic hydrazide were compared with garlic samples irradiated in our department. The unirradiated samples exhibited only limited DNA migration. If samples were irradiated, an increased DNA fragmentation was observed which permitted the discrimination between non-irradiated and irradiated samples. Since the garlic samples from China showed only very limited DNA fragmentation, they were deemed non-irradiated. Thus, this simple screening test was shown to be successful for identification of an irradiation treatment. (author)

  20. Performance Test for Neutron Detector and Associated System using Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seongwoo; Park, Sung Jae; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Oh, Se Hyun [USERS, Daejeon (Korea, Republic of); Shin, Ho Cheol [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    SPND (Self-Powered Neutron Detector) has been developed to extend its lifespan. ENFMS (Ex-Core Flux Monitoring System) of pressurized water reactor has been also improved. After the development and improvement, their performance must be verified under the neutron irradiation environment. We used a research reactor for the performance verification of neutron detector and associated system because the research reactor can meet the neutron flux level of commercial nuclear reactor. In this paper, we report the performance verification method and result for the SPND and ENFMS using the research reactor. The performance tests for the SPND and ENFMS were conducted using UCI TRIGA reactor. The test environment of commercial reactor’s neutron flux level must be required. However, it is difficult to perform the test in the commercial rector due to the constraint of time and space. The research reactor can be good alternative neutron source for the test of neutron detectors and associated system.

  1. Facilities and procedures used for the performance testing of DOE personnel dosimetry systems

    Energy Technology Data Exchange (ETDEWEB)

    Roberson, P.L.; Fox, R.A.; Hogan, R.T.; Holbrook, K.L.; Hooker, C.D.; Yoder, R.C.

    1983-04-01

    Radiological calibration facilities for personnel dosimeter testing were developed at the Pacific Northwest Laboratory (PNL) for the Department of Energy (DOE) to provide a capability for evaluating the performance of DOE personnel dosimetry systems. This report includes the testing methodology used. The informational presented here meets requirements specified in draft ANSI N13.11 for the testing laboratory. The capabilities of these facilities include sealed source irradiations for /sup 137/Cs, several beta-particle emitters, /sup 252/Cf, and machine-generated x-ray beams. The x-ray beam capabilities include filtered techniques maintained by the National Bureau of Standards (NBS) and K-fluorescent techniques. The calibration techniques, dosimeter irradiation procedures, and dose-equivalent calculation methods follow techniques specified by draft ANSI N13.11 where appropriate.

  2. Facilities and procedures used for the performance testing of DOE personnel-dosimetry systems

    International Nuclear Information System (INIS)

    Roberson, P.L.; Fox, R.A.; Hogan, R.T.; Holbrook, K.L.; Hooker, C.D.; Yoder, R.C.

    1983-04-01

    Radiological calibration facilities for personnel dosimeter testing were developed at the Pacific Northwest Laboratory (PNL) for the Department of Energy (DOE) to provide a capability for evaluating the performance of DOE personnel dosimetry systems. This report includes the testing methodology used. The informational presented here meets requirements specified in draft ANSI N13.11 for the testing laboratory. The capabilities of these facilities include sealed source irradiations for 137 Cs, several beta-particle emitters, 252 Cf, and machine-generated x-ray beams. The x-ray beam capabilities include filtered techniques maintained by the National Bureau of Standards (NBS) and K-fluorescent techniques. The calibration techniques, dosimeter irradiation procedures, and dose-equivalent calculation methods follow techniques specified by draft ANSI N13.11 where appropriate

  3. Irradiation test of fuel containing minor actinides in the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    Soga, Tomonori; Sekine, Takashi; Wootan, David; Tanaka, Kosuke; Kitamura, Ryoichi; Aoyama, Takafumi

    2007-01-01

    The mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast reactor Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted in the Joyo MK-III 3rd operational cycle. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX), and reference MOX fuel. The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes in order to confirm whether or not fuel melting occurred. After 10 minutes irradiation in May 2006, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins including neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. Linear heat rates for each pin were calculated using MCNP, accounting for both prompt and delayed heating components, and then adjusted using E/C for 10 B (n, α) reaction rates measured in the MK-III core neutron field characterization test. Post irradiation examination of these pins to confirm the fuel melting and the local concentration under irradiation of NpO 2-x or AmO 2-x in the (U, Pu)O 2-x fuel are underway. The test results are expected to reduce uncertainties on the design margin in the thermal design for MA-MOX fuel. (author)

  4. Intense neutron irradiation facility for fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji; Oyama, Yukio; Kato, Yoshio; Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    Technical R and D of d-Li stripping type neutron irradiation facilities for development of fusion reactor materials was carried out in Fusion Materials Irradiation Test Facility (FMIT) project and Energy Selective Neutron Irradiation Test Facility (ESNIT) program. Conceptual design activity (CDA) of International Fusion Materials Irradiation Facility (IFMIF), of which concept is an advanced version of FMIT and ESNIT concepts, are being performed. Progress of users` requirements and characteristics of irradiation fields in such neutron irradiation facilities, and outline of baseline conceptual design of IFMIF were described. (author)

  5. Radionuclides release from re-irradiated fuel under high temperature and pressure conditions. Gamma-ray measurements of VEGA-5 test

    Energy Technology Data Exchange (ETDEWEB)

    Hidaka, Akihide; Kudo, Tamotsu; Nakamura, Takehiko; Kanazawa, Toru; Kiuchi, Toshio; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The VEGA (Verification Experiments of radionuclides Gas/Aerosol release) program is being performed at JAERI to clarify mechanisms of radionuclides release from irradiated fuel during severe accidents and to improve source term predictability. The fifth VEGA-5 test was conducted in January 2002 to confirm the reproducibility of decrease in cesium release under elevated pressure that was observed in the VEGA-2 test and to investigate the release behavior of short-life radionuclides. The PWR fuel of 47 GWd/tU after about 8.2 years of cooling was re-irradiated at Nuclear Safety Research Reactor (NSRR) for 8 hours before the heat-up test. After that, the two pellets of 10.9 g without cladding were heated up to about 2,900 K at 1.0 MPa under the inert He condition. The experiment reconfirmed the decrease in cesium release rate under the elevated pressure. The release data on short-life radionuclides such as Ru-103, Ba-140 and Xe-133 that have never been observed in the previous VEGA tests without re-irradiation was obtained using the {gamma} ray measurement. (author)

  6. The fabrication and performance of Canadian silicide dispersion fuel for test reactors

    International Nuclear Information System (INIS)

    Sears, D.F.; Wood, J.C.; Berthiaume, L.C.; Herbert, L.N.; Schaefer, J.D.

    1985-01-01

    Fuel fabrication effort is now concentrated on the commissioning of large-scale process equipment, defining product specifications, developing a quality assurance plan, and setting up a mini-computer material accountancy system. In the irradiation testing program, full-size NRU assemblies containing 20% enriched silicide dispersion fuel have been Irradiated successfully to burnups in the range 65-80 atomic percent. Irradiations have also been conducted on mini-elements having 1.2 mm diameter holes In their mid-sections, some drilled before irradiation and others after irradiation to 22-83 atomic percent burnup. Uranium was lost to the coolant in direct proportion to the surface area of exposed core material. Pre-irradiation in the intact condition appeared to reduce in-reactor corrosion. Fuel cores developed for the NRU reactor are dimensionally very stable, swelling by only 6-8% at the very high burnup of 93 atomic percent. Two important factors contributing to this good performance are cylindrical clad restraint and coarse silicide particles. Thermal ramping tests were conducted on irradiated silicide aspersion fuels. Small segments of fuel cores released 85 Kr starting at about 520 deg. C and peaking at about 680 deg C. After a holding period of 1 hour at 720 deg. C a secondary 85 Kr peak occurred during cooling (at about 330 deg. C) probably due to thermal contraction cracking. Whole mini-elements irradiated to 93 atomic percent burnup were also ramped thermally, with encouraging results. After about 0.25 h at 530 deg. C the aluminum cladding developed very localized small blisters, some with penetrating pin-hole cracks preventing gross pillowing or ballooning. (author)

  7. The fabrication and performance of Canadian silicide dispersion fuel for test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sears, D F; Wood, J C; Berthiaume, L C; Herbert, L N; Schaefer, J D

    1985-07-01

    Fuel fabrication effort is now concentrated on the commissioning of large-scale process equipment, defining product specifications, developing a quality assurance plan, and setting up a mini-computer material accountancy system. In the irradiation testing program, full-size NRU assemblies containing 20% enriched silicide dispersion fuel have been Irradiated successfully to burnups in the range 65-80 atomic percent. Irradiations have also been conducted on mini-elements having 1.2 mm diameter holes In their mid-sections, some drilled before irradiation and others after irradiation to 22-83 atomic percent burnup. Uranium was lost to the coolant in direct proportion to the surface area of exposed core material. Pre-irradiation in the intact condition appeared to reduce in-reactor corrosion. Fuel cores developed for the NRU reactor are dimensionally very stable, swelling by only 6-8% at the very high burnup of 93 atomic percent. Two important factors contributing to this good performance are cylindrical clad restraint and coarse silicide particles. Thermal ramping tests were conducted on irradiated silicide aspersion fuels. Small segments of fuel cores released {sup 85}Kr starting at about 520 deg. C and peaking at about 680 deg C. After a holding period of 1 hour at 720 deg. C a secondary {sup 85}Kr peak occurred during cooling (at about 330 deg. C) probably due to thermal contraction cracking. Whole mini-elements irradiated to 93 atomic percent burnup were also ramped thermally, with encouraging results. After about 0.25 h at 530 deg. C the aluminum cladding developed very localized small blisters, some with penetrating pin-hole cracks preventing gross pillowing or ballooning. (author)

  8. Improving the AGR fuel testing power density profile versus irradiation-time in the advanced test reactor

    International Nuclear Information System (INIS)

    Chang, Gray S.; Lillo, Misti A.; Maki, John T.; Petti, David A.

    2009-01-01

    The Very High Temperature gas-cooled Reactor (VHTR), which is currently being developed, achieves simplification of safety through reliance on ceramic-coated fuel particles. Each TRISO-coated fuel particle has its own containment which serves as the principal barrier against radionuclide release under normal operating and accident conditions. These fuel particles, in the form of graphite fuel compacts, are currently undergoing a series of irradiation tests in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) to support the Advanced Gas-Cooled Reactor (AGR) fuel qualification program. A representive coated fuel particle with an 235 U enrichment of 19.8 wt% was used in this analysis. The fuel burnup analysis tool used to perform the neutronics study reported herein, couples the Monte Carlo transport code MCNP, with the radioactive decay and burnup code ORIGEN2. The fuel burnup methodology known as Monte-Carlo with ORIGEN2 (MCWO) was used to evaluate the AGR experiment assembly and demonstrate compliance with ATR safety requirements. For the AGR graphite fuel compacts, the MCWO-calculated fission power density (FPD) due to neutron fission in 235 U is an important design parameter. One of the more important AGR fuel testing requirements is to maintain the peak fuel compact temperature close to 1250degC throughout the proposed irradiation campaign of 550 effective full power days (EFPDs). Based on the MCWO-calculated FPD, a fixed gas gap size was designed to allow regulation of the fuel compact temperatures throughout the entire fuel irradiation campaign by filling the gap with a mixture of helium and neon gases. The chosen fixed gas gap can only regulate the peak fuel compact temperature in the desired range during the irradiation test if the ratio of the peak power density to the time-dependent low power density (P/T) at 550 EFPDs is less than 2.5. However, given the near constant neutron flux within the ATR driver core and the depletion of 235 U

  9. Gemstone dedicated gamma irradiation development

    Energy Technology Data Exchange (ETDEWEB)

    Omi, Nelson M.; Rela, Paulo R. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: nminoru@ipen.br; prela@ipen.br

    2007-07-01

    The gemstones gamma irradiation process to enhance the color is widely accepted for the jewelry industry. These gems are processed in conventional industrial gamma irradiation plant which are optimized for other purposes, using underwater irradiation devices with high rejection rate due to its poor dose uniformity. A new conception design, which states the working principles and manufacturing ways of the device, was developed in this work. The suggested device's design is based on the rotation of cylindrical baskets and their translation in circular paths inside and outside a cylindrical source rack as a planetary system. The device is meant to perform the irradiation in the bottom of the source storage pool, where the sources remain always shielded by the water layer. The irradiator matches the Category III IAEA classification. To verify the physical viability of the basic principle, tests with rotating cylindrical baskets were performed in the Multipurpose Irradiator constructed in the CTR, IPEN. Also, simulations using the CADGAMMA software, adapted to simulate underwater irradiations, were performed. With the definitive optimized irradiator, the irradiation quality will be enhanced with better dose control and the production costs will be significantly lower than market prices due to the intended treatment device's optimization. This work presents some optimization parameters and the expected performance of the irradiator. (author)

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

    International Nuclear Information System (INIS)

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

    1976-06-01

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

  11. 14C-lactose breath tests during pelvic radiotherapy: the effect of the amount of small bowel irradiated

    International Nuclear Information System (INIS)

    Weiss, R.G.; Stryker, J.A.

    1982-01-01

    Thirty patients who were undergoing pelvic radiotherapy had 14 C-lactose breath tests performed in the first and fifth weeks of treatment. In Group I (21 patients), a significant portion of the small intestine was irradiated, and in Group II (9 patients), only a small portion of the small intestine was irradiated. In Group I, the average reductions in the excretion of ingested 14 C between the first- and fifth-week tests were 41.5% at 1/2 hour postingestion (p 0.05). The data suggest that lactose malabsorption is a factor in the etiology of the nausea, vomiting, and diarrhea experienced by patients who are undergoing pelvic radiotherapy, and that the amount of bowel included in the treatment volume significantly influences the degree of malabsorption

  12. Effect of Enzyme Supplementation and Irradiation of Barley on Broiler Chicks Performance

    International Nuclear Information System (INIS)

    Farag, D.H.M.; Abd El-Hakeim, N.F.

    1999-01-01

    The experiments were conducted to study the influence of irradiation treatment at dose levels of 0.20 and 60 kGy on barley beta-glucan and the effect of enzyme supplementation and irradiation of barley on broiler chicks performance. The amount of total and water-soluble beta-glucan in raw barley was 36 kg -1 , respectively. The effect of irradiation treatment on total beta-glucan was insignificant while the level of soluble beta-glucan was increased with increasing the dose levels of irradiation. The effect of irradiation treatment and enzyme supplementation of barley diets on growth and conversion performance of broiler chicks indicated that birds fed raw barley diet had lower body weight, body weight gain and feed conversion than those fed control diet throughout the experimental period. Irradiation of barley at dose of 20 kGy did not affect the chick performance (feed consumption, weight gain feed-gain ratio) that received the B 20 diet from 7 to 21 days of age, but when bird maintained on B 20 diet from 7 28 days of age, only feed-gain ratio was improved by 14.4%. The results indicate that there was a significant effect of irradiation of barley at 60 kGy (B 60) on feed -gain ratio of chicks when were fed B 60 diet from 7 to 21 days of age. The corresponding improvement in feed-gain ratio was 16.4%. When birds were fed B 60 diet from 7-28 days of age, the improvement in body weight and feed-gain ratio was 25.5 and 19.6%, respectively

  13. 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

  14. Fuel Performance Modeling of U-Mo Dispersion Fuel: The thermal conductivity of the interaction layers of the irradiated U-Mo dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mistarhi, Qusai M.; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    U-Mo/Al dispersion fuel performed well at a low burn-up. However, higher burn-up and higher fission rate irradiation testing showed enhanced fuel meat swelling which was caused by high interaction layer growth and pore formation. The performance of the dispersion type fuel in the irradiation and un-irradiation environment is very important. During the fabrication of the dispersion type fuel an Interaction Layer (IL) is formed due to the inter-diffusion between the U-Mo fuel particles and the Al matrix which is an intermetallic compound (U,Mo)Alx. During irradiation, the IL becomes amorphous causing a further decrease in the thermal conductivity and an increase in the centerline temperature of the fuel meat. Several analytical models and numerical methods were developed to study the performance of the unirradiated U-Mo/Al dispersion fuel. Two analytical models were developed to study the performance of the irradiated U-Mo/Al dispersion fuel. In these models, the thermal conductivity of the IL was assumed to be constant. The properties of the irradiated U-Mo dispersion fuel have been investigated recently by Huber et al. The objective of this study is to develop a correlation for IL thermal conductivity during irradiation as a function of the temperature and fission density from the experimentally measured thermal conductivity of the irradiated U-Mo/Al dispersion fuel. The thermal conductivity of IL during irradiation was calculated from the experimentally measured data and a correlation was developed from the thermal conductivity of IL as a function of T and fission density.

  15. Standard Test Method for Electrical Performance of Photovoltaic Cells Using Reference Cells Under Simulated Sunlight

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This test method covers the determination of the electrical performance of a photovoltaic cell under simulated sunlight by means of a calibrated reference cell procedure. 1.2 Electrical performance measurements are reported with respect to a select set of standard reporting conditions (SRC) (see Table 1) or to user-specified conditions. 1.2.1 The SRC or user-specified conditions include the cell temperature, the total irradiance, and the reference spectral irradiance distribution. 1.3 This test method is applicable only to photovoltaic cells with a linear response over the range of interest. 1.4 The cell parameters determined by this test method apply only at the time of test, and imply no past or future performance level. 1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this s...

  16. New trends in nuclear fuel experimental irradiation. Modern control and acquisition of the irradiation data

    International Nuclear Information System (INIS)

    Preda, M.; Ciocanescu, M.; Ana, E.M.

    2010-01-01

    With the irradiation devices used in the irradiation tests, the following experiments have been performed in TRIGA-SCN reactor: a) In capsule-type irradiation devices: - fission gases composition determination; - dimensional measurements; - fission gases pressure measurement; - power pre-ramp and ramp; - power cycling; - structural materials testing. b) In loop-type irradiation device: - power ramp; - multiple power ramps; - overpower. Aiming to develop irradiation tests for advanced nuclear fuel elements, it is mandatory to increase the maximum neutron flux in the core with about 20%. This will lead to reactor power increase up to 21 MW. This objective can be reached through: - increasing the number of fuel clusters in the reactor core; - using the 6x6 fuel cluster to replace the present 5x5 clusters; - relocation of the control rods. In this context, the new control system and the data acquisition system operates online and allows real-time data evaluation. (author)

  17. Recent status of development and irradiation performance for plate type fuel elements with reduced 235U enrichment at NUKEM

    International Nuclear Information System (INIS)

    Hrovat, M.F.; Hassel, H.W.

    1984-01-01

    According to the present state of development full size test fuel elements with the maximum uranium densities of 2,2 g U/cm 3 meat for UAlsub(x), 3,2 g U/cm 3 meat for U 3 O 8 and 4,8 g U/cm 3 meat for U 3 Si 2 can be fabricated at NUKEM in production scale. Special chemical procedures for the uranium recovery were developed ensuring an economic fuel fabrication process. The post irradiation examinations (PIE) of 12 UAlsub(x) (U density 2,2 g U/cm 3 meat) and U 3 O 8 (up to 3,1 g U/cm 3 meat) test plates irradiated in the ORR, Oak Ridge research reactor, were terminated. All 12 test plates show unobjectionable irradiation behavior. Extensive irradiation tests on full size fuel elements were performed. All inserted elements show perfect irradiation behavior. The PIE of the first HFR Petten U 3 O 8 fuel elements are in progress. The full size ORR U 3 Si 2 fuel elements with so far highest uranium density of 4,76 g U/cm 3 meat achieved a burnup of 50 % loss of 235 U up to May 1983. One element was withdrawn from the reactor for PIE, the second will be irradiated to a burnup of 75 % loss of 235 U. The further development is concentrated on Usub(x)Sisub(y) fuel with highest uranium density. U 3 Si miniplates with up to 6,1 g U/cm 3 meat are supplied meeting the required specification, U 3 Si miniplates with 6,7 g U/cm 3 are in fabrication. (author)

  18. The irradiance and temperature dependent mathematical model for estimation of photovoltaic panel performances

    International Nuclear Information System (INIS)

    Barukčić, M.; Ćorluka, V.; Miklošević, K.

    2015-01-01

    Highlights: • The temperature and irradiance dependent model for the I–V curve estimation is presented. • The purely mathematical model based on the analysis of the I–V curve shape is presented. • The model includes the Gompertz function with temperature and irradiance dependent parameters. • The input data are extracted from the data sheet I–V curves. - Abstract: The temperature and irradiance dependent mathematical model for photovoltaic panel performances estimation is proposed in the paper. The base of the model is the mathematical function of the photovoltaic panel current–voltage curve. The model of the current–voltage curve is based on the sigmoid function with temperature and irradiance dependent parameters. The temperature and irradiance dependencies of the parameters are proposed in the form of analytic functions. The constant parameters are involved in the analytical functions. The constant parameters need to be estimated to get the temperature and irradiance dependent current–voltage curve. The mathematical model contains 12 constant parameters and they are estimated by using the evolutionary algorithm. The optimization problem is defined for this purpose. The optimization problem objective function is based on estimated and extracted (measured) current and voltage values. The current and voltage values are extracted from current–voltage curves given in datasheet of the photovoltaic panels. The new procedure for estimation of open circuit voltage value at any temperature and irradiance is proposed in the model. The performance of the proposed mathematical model is presented for three different photovoltaic panel technologies. The simulation results indicate that the proposed mathematical model is acceptable for estimation of temperature and irradiance dependent current–voltage curve and photovoltaic panel performances within temperature and irradiance ranges

  19. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Laug, David V.; Scates, Dawn M.; Reber, Edward L.; Roybal, Lyle G.; Walter, John B.; Harp, Jason M. [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Morris, Robert N. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer A system has been developed for safety testing of irradiated coated particle fuel. Black-Right-Pointing-Pointer FACS system is designed to facilitate remote operation in a shielded hot cell. Black-Right-Pointing-Pointer System will measure release of fission gases and condensable fission products. Black-Right-Pointing-Pointer Fuel performance can be evaluated at temperatures as high as 2000 Degree-Sign C in flowing helium. - Abstract: The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 Degree-Sign C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated

  20. Irradiation performance of uranium-molybdenum alloy dispersion fuels; Desempenho sob irradiacao de elementos combustiveis do tipo U-Mo

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, Cirila Tacconi de

    2005-07-01

    The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm{sup 3} were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm{sup 3} showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

  1. Performance testing of refractory alloy-clad fuel elements for space reactors

    International Nuclear Information System (INIS)

    Dutt, D.S.; Cox, C.M.; Karnesky, R.A.; Millhollen, M.K.

    1985-01-01

    Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO 2 ) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at. % burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts

  2. Development of a miniaturized bulge test (small punch test) for post-irradiation mechanical property evaluation

    International Nuclear Information System (INIS)

    Eto, Motokuni; Suzuki, Masahide; Nishiyama, Yutaka; Fukaya, Kiyoshi; Jitsukawa, Shiro; Misawa, Toshihei

    1993-01-01

    To examine the effectiveness of the small punch test for evaluating strength and toughness of irradiated ferritic steels, detailed procedures are described aiming at standardization of the test. The statistical approach to analysis of the SP energy as a function of temperature for evaluation of DBTT was also reviewed. The method was then applied to neutron-irradiated ferritic steels, which included F-82, F-82H, HT-9, and 2 1/4 Cr-1Mo steel. Fluence and irradiation temperatures ranged from 2 to 12 x 10 23 n/m 2 (E ≥ 1 MeV) and from 573 to 673 K, respectively. Comparison of parameters obtained from the small punch test with the properties measured by the conventional method indicated that: (a) the 0.2% offset stress and the ultimate tensile strength at room temperature can be correlated well with the parameters, P y /(t 0 ) 2 and P max /(t 0 ) 2 , respectively. Here, P y and P max are the loads corresponding to the yield and the maximum, and t 0 is the initial thickness of a specimen; (b) fracture toughness, J IC , can be evaluated using equivalent fracture strain, anti ε qf , and the previously established relationship between these values; and (c) DBTT measured by a Charpy test can be predicted from the results of temperature dependence of SP energy determined from the area under the load-deflection curve using a statistical analysis based on a Weibull distribution

  3. The insulation irradiation test program for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    McManamy, T.J.; Kanemoto, G.; Snook, P.

    1990-01-01

    The electrical insulation for the toroidal field coils of the Compact Ignition Tokamak (CIT) is expected to be exposed to radiation doses on the order of 10 10 rad with ∼90% of the dose from neutrons. The coils are cooled to liquid nitrogen temperature and then heated during the pulse to a peak temperature >300 K. In a program to evaluate the effects of radiation exposure on the insulators, three types of boron-free insulation were irradiated at room temperature in the Advanced Technology Reactor (ATR) and tested at the Idaho National Engineering Laboratory. The materials were Spaulrad-S, Shikishima PG5-1, and Shikishima PG3-1. The first two use a bismaleimide resin and the third an aromatic amine hardened epoxy. Spaulrad-S is a two-dimensional (2-D) weave of S-glass, while the others are 3-D weaves of T-glass. Flexure and shear/compression samples were irradiated to approximately 5 x 10 9 rad and 3 x 10 10 rad with 35 to 40% of the total dose from neutrons. The shear/compression samples were tested in pairs by applying an average compression of 345 MPa and then a shear load. After static tests were completed, fatigue testing was done by cycling the shear load for up to 30,000 cycles with a constant compression. The static shear strength of the samples that did not fail was then determined. Generally, shear strengths on the order of 120 MPa were measured. The behavior of the flexure and shear/compression samples was significantly different; large reductions in the flexure strength were observed, while the shear strength stayed the same or increased slightly. The 3-D weave material demonstrated higher strength and significantly less radiation damage than the 2-D material in flexure but performed nearly identically when tested with combined shear and compression. The epoxy system was much more sensitive to fatigue damage than the bismaleimide materials. 9 refs., 5 figs

  4. An investigation of the genetic toxicology of irradiated food-stuffs using short-term test systems

    International Nuclear Information System (INIS)

    Renner, H.W.; Altmann, H.; Asquith, J.C.; Elias, P.S.

    1982-01-01

    Six in vivo genetic toxicity tests were carried out on irradiated or unirradiated cooked chicken, dried dates and cooked fish. The tests were as follows: sex-linked recessive lethal mutations in Drosophila melanogaster (dried dates only), chromosome aberrations in bone marrow of Chinese hamsters, micronucleus test in rats, mice and Chinese hamsters, sister-chromatid exchange in bone marrow of mice and Chinese hamsters and in spermatogonia of mice, and DNA metabolism in spleen cells of Chinese hamsters. None of the tests provided any evidence of genetic toxicity induced by irradiation. However, dried dates, whether irradiated or not, showed evidence of some genetic toxicity in their effect on DNA metabolism in spleen cells and SCE induction in bone marrow. Feeding irradiated fish affected DNA metabolism in the spleen cells of Chinese hamsters. This effect could be interpreted as an induction of an immunoactive compound, although it could also be explained by the persistence of an immunoactive compound due to the removal by irradiation of spoilage organisms that would normally degrade it. (author)

  5. High-dose irradiation of food

    International Nuclear Information System (INIS)

    Diehl, J.F.

    1999-01-01

    Studies performed on behalf of the International Project on Food Irradiation in the period from 1971 until 1980 resulted in the concluding statement that ''.the irradiation of any food commodity up to an overall average dose of 10 kGy presents no toxicological hazard; hence, toxicological testing of foods so treated is no longer required.'' Since then, licenses for food irradiation have been restricted to this maximum dose in any country applying this technology. Further testing programmes have been carried out investigating the wholesomeness or hazards of high-dose irradiation, but there has been little demand so far by the food industry for licensing of high-dose irradiation, as there is only a small range of products whose irradiation at higher doses offers advantages for given, intended use. These include eg. spices, dried herbs, meat products in flexible pouch packagings for astronauts, or patients with immune deficiencies. (orig./CB) [de

  6. Investigation of high flux test module for the international fusion materials irradiation facilities (IFMIF)

    International Nuclear Information System (INIS)

    Miyashita, Makoto; Sugimoto, Masayoshi; Yutani, Toshiaki

    2007-03-01

    This report describes investigation on structure of a high neutron flux test module (HFTM) for the International Fusion Materials Irradiation Facilities (IFMIF). The HFTM is aimed for neutron irradiation of a specimen in a high neutron flux domain of the test cell for irradiation ground of IFMIF. We investigated the overall structure of the HFTM that was able to include specimens in a rig and thermocouple arrangement, an interface of control signal and support structure. Moreover, pressure and the amount of the bend in the module vessel (a rectangular section pressure vessel) were calculated. The module vessel did a rectangular section from limitation of a high neutron flux domain. Also, we investigated damage of thermocouples under neutron irradiation, which was a temperature sensor of irradiation materials temperature control demanded high precision. Based on these results, drawings on the HTFM structure. (author)

  7. Technological tests at the preindustrial level on irradiated potatoes. Prospects for the practical introduction of irradiated foods in Italy

    International Nuclear Information System (INIS)

    Baraldi, D.

    1978-01-01

    To confirm the technological feasibility of potato irradiation in large pallet boxes for a period up to 150 days' plant operation, a pilot-scale technological study was carried out in Italy during 1975-76. Potatoes (14t, cultivar Tonda di Berlino) were received from the Avezzano area on a commercial truck, irradiated at the Casaccia gamma plant and transported back for storage. The irradiation was carried out in pallet boxes (500kg) using a rotating platform at an average dose of 11.7krad. The radiation treatment was carried out at 3-week intervals for a total of 9 treatments. During 1976 15t of irradiated potatoes were put on the Italian market. Irradiation was carried out again at the gamma irradiation plant of the Applied Radiation Division, Casaccia Nuclear Center. After irradiation, the product was transported back to the Fucino area and stored in warehouses of the Fucino Agency at environmental conditions. Two months later the material was taken to Bologna, Milan, Rome and Pescara and put on the market there. At the end of the marketing test and upon receipt of the consumers' opinions by means of distributed postcards, it was concluded that the majority of the consumers expressed a preference for irradiated potatoes with respect to both quality and storage. (author)

  8. Fabrication of Non-instrumented capsule for DUPIC simulated fuel irradiation test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.G.; Kang, Y.H.; Park, S.J.; Shin, Y.T. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    In order to develope DUPIC nuclear fuel, the irradiation test for simulated DUPIC fuel was planed using a non-instrumented capsule in HANARO. Because DUPIC fuel is highly radioactive material the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO was designed to remotely assemble and disassemble in hot cell. And then, according to the design requirements the non-instrumented DUPIC capsule was successfully manufactured. Also, the manufacturing technologies of the non-instrumented capsule for irradiating the nuclear fuel in HANARO were established, and the basic technology for the development of the instrumented capsule technology was accumulated. This report describes the manufacturing of the non-instrumented capsule for simulated DUPIC fuel. And, this report will be based to develope the instrumented capsule, which will be utilized to irradiate the nuclear fuel in HANARO. 26 refs., 4 figs. (Author)

  9. Testing of camera performance standards at steady and local overloading

    International Nuclear Information System (INIS)

    Keszthelyine Landori, S.; Adorjanne Farkas, M.; Csirik, J.

    1983-01-01

    Camera performance standards are usually given for low count rates and uniform irradiation. A conventional analog gamma camera system (Gamma MB 9100 manufactured under the know-how of Picker DC 4/12 in Hungary) was studied on the basis of the Picker test procedure and the NEMA standard system. Uniformity, linearity, spatial and energy resolution were measured at high count rates and uniform irradiation (steady overloading). Linearity was studied at local overloading. Linearity, spatial and energy resolution were measured by a 1024-channel analyzer-computer system of KFKI, Hungary. The data were evaluated on the basis of NEMA standards, while uniformity was measured by the Gamma data processing system and evaluated by special SEGAMS programs. Performance variations were studied between 7500 cps and 75.000 cps pulse rates. Spatial and energy resolution were influenced strongly, uniformity slightly by pulse rates, while linearity did not change at all. Linearity was not influenced even by local overloading. (author)

  10. Technology development on production of test specimens from irradiated capsule outer-tube and mechanical evaluation test of stainless steel with high dose carried out by the technology

    International Nuclear Information System (INIS)

    Hayashi, Koji; Shibata, Akira; Iwamatsu, Shigemi; Sozawa, Shizuo; Takada, Fumiki; Ohmi, Masao; Nakagawa, Tetsuya

    2008-03-01

    The irradiation capsule 74M-52J was irradiated during total 136 cycles at reactor core of JMTR and the maximum neutron dose reached on 3.9x10 26 n/m 2 at the capsule outer-tube made of a type 304 stainless steel. In order to produce mechanical test specimens from the outer-tube, a punching technique was developed as a simple remote-handling method in a hot-cell. From comparison between the punching and the mechanical cutting methods, it was clarified that the punching technique was applicable to practical use. Moreover, an evaluation test of mechanical properties using specimens sampled from the 74M-52 was performed with in-water high temperature condition, less than 288degC. The result shows that the residual elongation is 18% at 150degC and 13% at 288degC. It was confirmed that the type 304 stainless steel irradiated up to such high dose shows enough ductility. (author)

  11. Newly developed non-destructive testing method for evaluation of irradiation brittleness of structural materials using ultrasonic

    International Nuclear Information System (INIS)

    Ishii, Toshimitsu; Ooka, Norikazu; Kato, Yoshiaki; Saito, Junichi; Hoshiya, Taiji; Shibata, Saburo; Kobayashi, Hideo

    1999-01-01

    Surveillance testing is important to evaluate neutron irradiation embrittlement of reactor pressure vessel material for long life operation. An alternative test method for evaluating the irradiation embrittlement of the pressure vessel material will have to be proposed to support the limited number of surveillance test specimens in order to manage the plant life to be extended. In this study, ultrasonic testing for irradiated A533B-1 steel and weld metal was applied to examine material degradation nondestructively. With increasing the shift of Charpy 41 J transition temperature, ultrasonic velocity decreased and attenuation coefficient of ultrasonic wave increased. Especially, the difference of ultrasonic velocity for 5 MHz shear wave between as-received and irradiated material is corresponding to the shift of transition temperature showing material degradation. (author)

  12. Developments, characterization and proton irradiation damage tests of AlN detectors for VUV solar observations

    Energy Technology Data Exchange (ETDEWEB)

    BenMoussa, A., E-mail: ali.benmoussa@stce.be [Solar Terrestrial Center of Excellence (STCE), Royal Observatory of Belgium, Circular Avenue 3, B-1180 Brussels (Belgium); Soltani, A.; Gerbedoen, J.-C [Institut d’Electronique, de Microélectronique et de Nanotechnologie (IEMN), F-59652 Villeneuve d’Ascq (France); Saito, T. [Department of Environment and Energy, Tohoku Institute of Technology, 35-1, Yagiyama-Kasumi-cho, Taihaku-ku, Sendai, Miyagi 982-8577 (Japan); Averin, S. [Fryazino Branch of the Kotel’nikov Institute of Radioengineering and Electronics of Russian Academy of Sciences, 141190 Square Vvedenski 1, Fryazino, Moscow Region (Russian Federation); Gissot, S.; Giordanengo, B. [Solar Terrestrial Center of Excellence (STCE), Royal Observatory of Belgium, Circular Avenue 3, B-1180 Brussels (Belgium); Berger, G. [Catholic University of Louvain-la-Neuve, Chemin du Cyclotron 2, B-1348 Louvain la Neuve (Belgium); Kroth, U. [Physikalisch-Technische Bundesanstalt (PTB), Abbestr. 2-12, D-10587 Berlin (Germany); De Jaeger, J.-C. [Institut d’Electronique, de Microélectronique et de Nanotechnologie (IEMN), F-59652 Villeneuve d’Ascq (France); Gottwald, A. [Physikalisch-Technische Bundesanstalt (PTB), Abbestr. 2-12, D-10587 Berlin (Germany)

    2013-10-01

    For next generation spaceborne solar ultraviolet radiometers, innovative metal–semiconductor–metal detectors based on wurtzite aluminum nitride are being developed and characterized. A set of measurement campaigns and proton irradiation damage tests was carried out to obtain their ultraviolet-to-visible characterization and degradation mechanisms. First results on large area prototypes up to 4.3 mm diameter are presented here. In the wavelength range of interest, this detector is reasonably sensitive and stable under brief irradiation with a negligible low dark current (3–6 pA/cm{sup 2}). No significant degradation of the detector performance was observed after exposure to protons of 14.4 MeV energy, showing a good radiation tolerance up to fluences of 1 × 10{sup 11} protons/cm{sup 2}.

  13. Irradiation Behavior and Post-Irradiation Examinations of an Acoustic Sensor Using a Piezoelectric Transducer

    International Nuclear Information System (INIS)

    Lambert, T.; Zacharie-Aubrun, I.; Hanifi, K.; Valot, Ch.; Fayette, L.; Rosenkantz, E.; Ferrandis, J.Y.; Tiratay, X.

    2013-06-01

    The development of advanced instrumentation for in-pile experiments in Material Testing Reactor constitutes a main goal for the improvement of the nuclear fuel behavior knowledge. In the framework of high burn-up fuel experiments under transient operating conditions, an innovative sensor based on acoustic method was developed by CEA and IES (Southern Electronic Institute).This sensor is used to determine the on-line composition of the gases located in fuel rodlet free volume and thus, allows calculating the molar fractions of fission gases and helium. The main principle of the composition determination by acoustic method consists in measuring the time of flight of an acoustic signal emitted and reflected in a specific cavity. A piezoelectric transducer, driven by a pulse generator, generates the acoustic wave in the cavity. The piezoelectric transducer is a PZT ceramic disk, mainly consisting of lead, zirconium and titanium. This acoustic method was tested with success during a first experiment called REMORA 3, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. However, during the irradiation test, acoustic signal degradation was observed, mainly due to irradiation effect but also due to the increasing of the gas temperature. Despite this acoustic signal degradation, the time of flight measurements were carried out with good accuracy throughout the test, thanks to the development of a more efficient signal processing. After experiment, neutronic calculations were performed in order to determine neutron fluence at the level of the piezoelectric transducer. In order to have a better understanding of the acoustic sensor behavior under irradiation, Post Irradiation Examination program was done on piezoelectric transducer and on acoustic coupling material too. These examinations were also realized on a non-irradiated acoustic sensor built in the same conditions and with the same materials and the same

  14. Influence of e-Beam Irradiation on the Performance of Energy Storage and Conversion Electrode

    Energy Technology Data Exchange (ETDEWEB)

    Baeok, Sung Hyeon; Jo, Won Jun; Lee, Duwon; Lee, Myung An [Inha Univ., Incheon (Korea, Republic of); Shin, Joong Hyeok; Lee, Byung Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    Electron beam irradiation was known as an effective method to improve the stability and performance of electrodes by varying the chemical and physical properties. It has been reported that surface morphology, oxidation state, optical properties, and electrochemical properties can be modified by e-beam irradiation. In this work, influence of electron beam irradiation on the performance of electrode was studied for the applications in energy storage and conversion, such as secondary battery, supercapacitor, and fuel cell. Changes in physical and chemical properties of electrodes before and after e-beam irradiation were investigated. The crystallinity of the synthesized materials was investigated by X-ray diffraction, and the oxidation states were determined by X-ray photoelectron spectroscopy. Scanning electron microscopy was utilized to examine surface morphology. Crystallinity, surface morphology, and oxidation state were significantly changed by electron beam irradiation, and were found to be strongly dependent on irradiation time.

  15. Performance studies under high irradiation of resistive bulk-micromegas chambers at the CERN Gamma Irradiation Facility

    CERN Document Server

    Sidiropoulou, Ourania; Bortfeldt, J; Farina, E; Iengo, P; Longo, L; Sidiropoulou, O; Wotschack, J

    2017-01-01

    Radiation studies on several resistive bulk-Micromegas chambers produced at CERN will be viewed in this document. Two resistive bulk-Micromegas chambers have been installed at the CERN Gamma Irradiation Facility (GIF++) exposed to an intense gamma irradiation with the aim of evaluating the detector behaviour under high irradiation and carrying out a long-term age- ing study. The chambers under study have an active area of 10 x 10 cm 2 , a strip pitch of 400 m m , an ampli- fication gap of 128 m m , and a drift gap of 5 mm. The results on the detector performance as a function of the photon flux up to 44 MHz/cm 2 will be shown as well as the ageing properties as function of the integrated charge and the current intensity and its stability with time. In addition, the results of the efficiency measurements before, during, and after the irradiation will also be presented as a function of the amplification voltage at which the chambers are operated.

  16. Design of a high-flux test assembly for the Fusion Materials Irradiation Test Facility

    International Nuclear Information System (INIS)

    Opperman, E.K.; Vogel, M.A.

    1982-01-01

    The Fusion Material Test Facility (FMIT) will provide a high flux fusion-like neutron environment in which a variety of structural and non-structural materials irradiations can be conducted. The FMIT experiments, called test assemblies, that are subjected to the highest neutron flux magnitudes and associated heating rates will require forced convection liquid metal cooling systems to remove the neutron deposited power and maintain test specimens at uniform temperatures. A brief description of the FMIT facility and experimental areas is given with emphasis on the design, capabilities and handling of the high flux test assembly

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

    Science.gov (United States)

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

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

  18. Study on sensory test of irradiated spices

    International Nuclear Information System (INIS)

    Chiba, Etsuko; Iizuka, Tomoko; Ichikawa, Mariko; Kobayashi, Yasuhiko; Ukai, Mitsuko; Kikuchi, Masahiro

    2016-01-01

    For the spices used in curry dishes and the spices used except for curry dishes, the effects of irradiation sterilization and conventional superheated-steam sterilization were compared with sensory test. As for spices, superheated-steam sterilization reduces aroma and changes color tone compared with irradiation sterilization. Even with cooked curry, radiologically sterilized products were stronger in 'flavor before sample tasting' or 'spicy taste during sample tasting' with statistically significant difference compared with superheated-steam sterilized products. As for the comparison with spices themselves, red pepper and white/black pepper tended to be stronger in taste and pungent taste than radiologically sterilized products. In addition, superheated-steam sterilized products of red hot pepper and turmeric were very different in color from untreated products, while radiologically sterilized products showed a little difference. When comparing color and flavor in a 2D map, it was found at a glance that the radiologically sterilized product was close to the untreated product. Thia map can easily convince the merit of irradiation sterilization, and it was found to be effective for promoting risk communication. In the case of white pepper, the radiologically sterilized product showed more strong pungent than the superheated-steam sterilized product with statistically significant difference. However, not only the strength difference but also qualitative difference was perceived in flavor. (A.O.)

  19. Vegetative and reproductive performances of irradiated strawberry plants

    International Nuclear Information System (INIS)

    Velchev, V.; Stoyanov, S.

    1982-01-01

    The effect of radioactive gamma-irradiation (Cs 137 source with 600 rad/min intensity) at 250 to 1500 rad of the vegetative and reproductive performances of the strawberry was studied, under field conditions with the cvs. Gorella, Cambridge Favourite, and Zenga Zengana. It was found that in all cvs. the highest stimulating effect in respect to number of leaves was produced by rates of 1000 and 1500 rad. Irradiation with 750 and 1000 rad had a stimulating effect on the formation of more flower buds and flowers in Gorella and Cambridge Favourite, while similar effect was evident in Zenga Zengana only at 1000 rad. Radioactive irradiation had different effect on the yield of the cultivars studied. A positive after-effect was observed in the second year in Gorella at 1000 rad - 117512 kg/ha; by Zenga Zengana at 750 rad - 337237 kg/ha; and by Cambridge Favourite at 250 rad - 199766 kg/ha, which represented 113,02; 111,25 and 105.00% respectively as compared to the control

  20. Short term mutagenicity tests and their application to irradiated foods

    International Nuclear Information System (INIS)

    Phillips, B.J.; Elias, P.S.

    1980-01-01

    Although traditional long-term animal tests are likely to continue to be required, these are not only extremely costly but are coming more and more to be recognised as an imprecise and unsatisfactory method of testing the safety of irradiated foods for human consumption. It is therefore clearly advisable to include a selection of quicker and more direct testing methods in any toxicological assessment procedures. The International Project has therefore undertaken a study of the feasibility of using the newer systems for investigation of irradiated foodstuffs. Although some work in this field has already been carried out, some shortcomings in the published work can be identified which justify a more detailed and intensive research programme. As expected, little difficulty has been encountered in testing food by methods involving mammals, but considerable effort has been required to adapt in vitro systems. The use of enzymatic digestion in vitro to provide food samples for testing in mammalian cell cultures has never been attempted before and the procedures developed by the Project represent a positive contribution to methodology in this field. A series of foodstuffs is being tested by a wide spectrum of short-term tests and the first results are now being obtained. (orig./MG) [de

  1. Irradiation tests of critical components for remote handling system in gamma radiation environment

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi

    1996-03-01

    This report covers the gamma ray irradiation tests according to the Agreement of ITER R and D Task (T35) in 1994 and describes radiation hardness of the standard components for the ITER remote handling system which are categorized into the robotics (Subtask-1), the viewing system (Subtask-2) and the common components (Subtask-3). The gamma ray irradiation tests have been conducted using No.2 and No.3 cells at the cobalt building of Takasaki Establishment in JAERI. The radiation source is cobalt sixty (Co-60), and the maximum dose rate of No.2 and No.3 cells is about 1x10 6 R/h and 2x10 6 R/h, respectively. The environmental conditions of the irradiation tests are described below and all of components excepting electrical wires have been tested in the No.2 cell. [No.2 cell : Atmosphere and ambient temperature No.3 cell : Nitrogen gas and 250degC] As a whole, many of components have been irradiated up to the rated dose of around 1x10 10 rads and the following main results are obtained. The developed AC servo motor and periscope for radiation use have shown excellent durability with the radiation hardness tolerable for more than 10 9 rads. An electrical connector compatible with remote operation has also shown no degradation of electrical characteristics after the irradiation of 10 10 rads. As for polyimide insulated wires, the mechanical and electrical characteristics are not degradated after the irradiation of 10 9 rads and more radiation hardness can be expected than the anticipation. On the contrary, standard position sensors such as rotary encoder show extremely low radiation hardness and further efforts have to be made for improvements. (J.P.N.)

  2. Behavior of pre-irradiated fuel under a simulated RIA condition. Results of NSRR Test JM-5

    International Nuclear Information System (INIS)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Tanzawa, Sadamitsu; Ishijima, Kiyomi; Kobayashi, Shinsho; Kamata, Hiroshi; Homma, Kozo; Sakai, Haruyuki.

    1995-11-01

    This report presents results from the power burst experiment with pre-irradiated fuel rod, Test JM-5, conducted in the Nuclear Safety Research Reactor (NSRR). The data concerning test method, pre-irradiation, pre-pulse fuel examination, pulse irradiation, transient records and post-pulse fuel examination are described, and interpretations and discussions of the results are presented. Preceding to the pulse irradiation in the NSRR, test fuel rod was irradiated in the Japan Materials Testing Reactor (JMTR) up to a fuel burnup of 25.7 MWd/kgU with average linear heat rate of 33.4 kW/m. The fuel rod was subjected to the pulse irradiation resulting in a desposited energy of 223 ± 7 cal/g·fuel (0.93 ± 0.03 kJ/g·fuel) and a peak fuel enthalpy of 167 ± 5 cal/g·fuel (0.70 ± 0.02 kJ/g·fuel) under stagnant water cooling condition at atmospheric pressure and ambient temperature. Test fuel rod behavior was assessed from pre- and post-pulse fuel examinations and transient records during the pulse. The Test JM-5 resulted in cladding failure. More than twenty small cracks were found in the post-test cladding, and most of the defects located in pre-existing locally hydrided region. The result indicates an occurrence of fuel failure by PCMI (pellet/cladding mechanical interaction) in combination with decreased integrity of hydrided cladding. (author)

  3. Market testing and consumer acceptance of irradiated rice (Oryza sativa indica Linn.)

    International Nuclear Information System (INIS)

    Ungsunantwiwat, Ampai; Sophonsa, Sombut

    2001-01-01

    Special grade A fragrant rice (Jasmine rice) of 13% moisture content was obtained from a local miller in Bangkok. Low density polyethylene, 29.5 cm in width x 45 cm in length and 200 micron in thickness, was used to pack the rice with a net weight of 5 kg. The irradiated food label was printed on one side of the bag to comply with food control regulations. The color and the ink for marking were tested for gamma radiation compatibility. A total of 800 bags of rice, with a total gross weight of 4,000 kg, were irradiated at a minimum absorbed dose at 0.5 kGy for insect disinfestation. Radiation treatment was carried out using a multi-purpose, carrier type gamma irradiator (Model JS-8900, Serial No. IR-155) located at the Thai Irradiation Center. Irradiated rice was distributed on a weekly basis to food stores in Bangkok and Pathum Thani, as well as to various governmental organizations and interested individuals. The product was sold at 60 bahts per bag (approx. US$ 2.4) to retailers. Various commercial brands of non-irradiated rice of 5 kg size, were available in the market at 52 to 78 bahts per bag (approx. US $ 2.08 to 3.12), depending on quality and brand name. During the distribution, a leaflet of educational information was given to the consumer. A simple questionnaire used in the marketing trial indicated that 72% of the consumers bought irradiated rice because of the good quality of the product based on visual inspection, and 28% of them were willing to try the new product. Most consumers preferred irradiated rice to chemical treatment (fumigation) for insect disinfestation. However, most consumers were not sure if they would like to buy irradiated rice again unless its cooking quality was acceptable. Market testing of irradiated rice in the upper-class market or supermarket was unsuccessful because of limitations in the sale and service conditions. To meet the requirement of the supermarket retailer, irradiated rice had to be supplied on a monthly basis, with

  4. Alloy development for irradiation performance: program strategy

    International Nuclear Information System (INIS)

    Bloom, E.E.; Stiegler, J.O.; Wiffen, F.W.; Dalder, E.N.C.; Reuther, T.C.; Gold, R.E.; Holmes, J.J.; Kummer, D.L.; Nolfi, F.V.

    1978-01-01

    The objective of the Alloy Development for Irradiation Performance Program is the development of structural materials for use in the first wall and blanket region of fusion reactors. The goal of the program is a material that will survive an exposure of 40 MWyr/m 2 at a temperature which will allow use of a liquid-H 2 O heat transport system. Although the ultimate aim of the program is development of materials for commercial reactors by the end of this century, activities are organized to provide materials data for the relatively low performance interim machines that will precede commercial reactors

  5. Establishment of experimental equipments in irradiation technology development building

    International Nuclear Information System (INIS)

    Ishida, Takuya; Tanimoto, Masataka; Shibata, Akira; Kitagishi, Shigeru; Saito, Takashi; Ohmi, Masao; Nakamura, Jinichi; Tsuchiya, Kunihiko

    2011-06-01

    The Neutron Irradiation and Testing Reactor Center has developed new irradiation technologies to provide irradiation data with high technical value for the resume of the Japan Materials Testing Reactor (JMTR). For the purpose to perform assembling of capsules, materials tests, materials inspection and analysis of irradiation specimens for the development of irradiation capsules, improvement and maintenance of facilities were performed. From the viewpoint of effective use of existing buildings in the Oarai research and development center, the RI application development building was refurbished and maintained for above-mentioned purpose. The RI application development building is a released controlled area, and was used as storage of experimental equipments and stationeries. The building was named 'Irradiation Technology Development Building' after it refurbished and maintained. Eight laboratories were maintained based on the purpose of use, and the installation of the experimental apparatuses was started. A basic management procedure of the Irradiation Technology Development Building was established and has been operated. This report describes the refurbish work of the RI application development building, the installation and operation method of the experimental apparatuses and the basic management procedure of the Irradiation Technology Development Building. (author)

  6. Oogenesis, follicular development and reproductive performance in the prenatally irradiated bovine

    International Nuclear Information System (INIS)

    Erickson, B.H.; Reynolds, R.A.

    1978-01-01

    To determine the radioresponse of the various developmental stages of the prenatal bovine germ cell, cows bearing foetuses varying in age from 40+-5 (point of gonadal sex differentiation) to 270+-10 (approximately 13 days before parturition) days of gestation were irradiated with 300 R of 60 Co gamma radiation at 25-50 R/min, a level just below that likely to cause maternal deaths and gross foetal abnormalities at some developmental stages. Dose to the foetal gonad was approximately 100 rad. When the prenatally irradiated heifers were approximately 10 months old their ovaries were recovered at slaughter, serially sectioned, and prepared for microscopic analysis. A complete quantitative analysis of oocytes in primordial, growing and vesicular follicles was effected. Follicular development (reflected by counts of growing and vesicular follicles) was apparently unaffected at all ages tested (P>0.25), whereas oogenesis (reflected by counts of oocytes in primordial follicles) was significantly impeded (64% of control) only between 70 and 90 days of gestation (P<0.05). The germ-cell population at this period differs from later developmental stages principally by the presence of a high proportion of mitotically active oogonia, hence it appears that the oogonium is the most vulnerable cell type. Neither the viability nor reproductive performance of 60 cows that were prenatally irradiated (100 rad) at either 80+-10 or 130+-20 days of gestation was affected through a minimum of 5 years of postnatal life and three pregnancies. (author)

  7. Effects of gamma irradiation on the performance of Jatropha (Jatropha curcas L.) accessions

    Science.gov (United States)

    Surahman, M.; Santosa, E.; Agusta, H.; Aisyah, S. I.; Nisya, F. N.

    2018-03-01

    This study aimed to assess the effects of mutation by using gamma ray on the performance of jatropha plants. The study was conducted at PAIR BATAN. Jatropha seeds obtained from the collection farm of SBRC LPPM IPB and PT Indocement Tunggal Prakarsa Tbk in Gunung Putri, Bogor, were irradiated. The irradiated seeds were grown in Jonggol Trial Farm of IPB. Gamma irradiation was conducted by using a GCM 4000A device. Treatments consisted of irradiation doses, irradiation methods, and accessions. Irradiation doses given were 175, 200, 225 Gy, and no irradiation (control). Irradiation methods consisted of acute, intermittent, and split-dose. Accessions used in this study were Dompu, Medan, Bima, Lombok, ITP II, IP2P, and Thailand. Results of the study were analysed until 5 months after planting showed that gamma ray mutation gave stimulating and inhibiting effects on similar traits. Irradiation dose of 225 Gy was good to be given in acute, intermittent, and split-dose methods. Irradiation effects were found to be significant in jatropha accessions. Effects of irradiation on production will be published soon.

  8. Photopatch and UV-irradiated patch testing in photosensitive dermatitis

    Directory of Open Access Journals (Sweden)

    Reena Rai

    2016-01-01

    Full Text Available Background: The photopatch test is used to detect photoallergic reactions to various antigens such as sunscreens and drugs. Photosensitive dermatitis can be caused due to antigens like parthenium, fragrances, rubbers and metals. The photopatch test does not contain these antigens. Therefore, the Indian Standard Series (ISS along with the Standard photopatch series from Chemotechnique Diagnostics, Sweden was used to detect light induced antigens. Aim: To detect light induced antigens in patients with photosensitive dermatitis. Methods: This study was done in a descriptive, observer blinded manner. Photopatch test and ISS were applied in duplicate on the patient's back by the standard method. After 24 hours, readings were recorded according to ICDRG criteria. One side was closed and other side irradiated with 14 J/cm2 of UVA and a second set of readings were recorded after 48 hrs. Result: The highest positivity was obtained with parthenium, with 18 out of 35 (51% patients showing a positive patch test reaction with both photoallergic contact dermatitis and photoaggravation. Four patients (11% showed positive patch test reaction suggestive of contact dermatitis to potassium dichromate and fragrance mix. Six patients had contact dermatitis to numerous antigens such as nickel, cobalt, chinoform and para-phenylenediamine. None of these patients showed photoaggravation on patch testing. Conclusion: Parthenium was found to cause photoallergy, contact dermatitis with photoaggravation and contact allergy. Hence, photopatch test and UV irradiated patch test can be an important tool to detect light induced antigens in patients with photosensitive dermatitis.

  9. Use of hot set and gel content test in QA/QC for irradiation of commercial products

    International Nuclear Information System (INIS)

    Ruzalina Baharin; Siti Aiasah Hashim; Zulkafli Ghazali; Zarina Mohd Noor

    2006-01-01

    Currently, irradiation of polyolefin heat shrinkable tube is the main commercial activity at Alurtron. Other products include medical devices and cosmetic products. As required by the quality management system, ISO 9001:2000, Alurtron is responsible to ensure that customer received their irradiation dose as requested. At present two test methods are employed namely, hot set and gel content measurement to verify received dose . Hot set test is a physical test to determine longitudinal shrinkage of the irradiated products. Whereas gel content measurement determine to certain extend the degree of crosslinking occurred in polymeric material upon irradiation, in the customer products. Both test are routinely used to cross-check the dose required by customer. This paper outlines the correlation between hot set and gel content measurement of customer products. (Author)

  10. Irradiation of the testes in radiotherapy of Hodgkin's disease and testicular tumors

    International Nuclear Information System (INIS)

    Bakyrdzhiev, S.; Ganchev, M.; Milchev, V.; Naumova, Ts.

    1983-01-01

    Direct measurements using TLD dosimeters permitted to calculate radiation doses delivered to the testes in radiotherapy for supradiaphragmatic forms of Hodgkin's disease, stages I and II; they were found to constitute from 1 to 2% of focal dose, that is, to amount to 40-80 cGy given in 20 nonuniform fractions. In radiotherapy for subdiaphragmatic forms of the disease, the dose to the testes varied from 360 to 400 cGy. Anthropomorphic, phantom measurements with ionization chambers placed within the phantom testes showed contributions to total testicular dose to vary with individual irradiation fields in these cases. Thus, for instance, 82% of total dose was due to irradiation of both iliac fields; shielding of the testes with lead caps (5mm) reduced this irradiation to one-half. Doses to the testes were relatively large in radiotherapy for testicular tumors (seminomas and teratocarcinomas). By combining radiotherapy with chemotherapy or surgical intervention, permanent cure or long-lasting remissions may be achieved in most cases. In this connection, there arises the question as to the potential risk of patients - mostly young males with maintained reproductive capacity - transmitting radiation - induced genetic damage to their progeny. An attempt was made to appraise such genetic risk from additional above-background exposure. (authors)

  11. Dry Process Fuel Performance Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Song, K. C.; Moon, J. S. and others

    2005-04-15

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

  12. Dry Process Fuel Performance Evaluation

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Song, K. C.; Moon, J. S. and others

    2005-04-01

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

  13. Performance testing of beta dosimeters used at Department of Energy facilities

    International Nuclear Information System (INIS)

    Roberson, P.L.; Holbrook, K.L.; Pappin, J.L.

    1983-01-01

    A performance test based on the American national draft standard N13.11 was conducted for dosimeter systems in use at Department of Energy facilities. The large differences in dosimeter response found were due to use of different calibration source standards and different dosimeter designs. Differences in 90 Sr/ 90 Y calibrations were approximately 20% or less for all but one participant. The differences observed were attributed to variable thicknesses of dosimeter elements and variable source irradiation geometries. Improved beta calibration standards will result if irradiation specifications include acceptable ranges from the depth-dose characteristics. The low-energy beta responses observed were consistent with the thicknesses of dosimeter sensitive elements and overlying filtration

  14. Analysis of excess reactivity of JOYO MK-III performance test core

    International Nuclear Information System (INIS)

    Maeda, Shigetaka; Yokoyama, Kenji

    2003-10-01

    JOYO is currently being upgraded to the high performance irradiation bed JOYO MK-III core'. The MK-III core is divided into two fuel regions with different plutonium contents. To obtain a higher neutron flux, the active core height was reduced from 55 cm to 50 cm. The reflector subassemblies were replaced by shielding subassemblies in the outer two rows. Twenty of the MK-III outer core fuel subassemblies in the performance test core were partially burned in the transition core. Four irradiation test rigs, which do not contain any fuel material, were loaded in the center of the performance test core. In order to evaluate the excess reactivity of MK-III performance test core accurately, we evaluated it by applying not only the JOYO MK-II core management code system MAGI, but also the MK-III core management code system HESTIA, the JUPITER standard analysis method and the Monte Carlo method with JFS-3-J3.2R content set. The excess reactivity evaluations obtained by the JUPITER standard analysis method were corrected to results based on transport theory with zero mesh-size in space and angle. A bias factor based on the MK-II 35th core, which sensitivity was similar to MK-III performance test core's, was also applied, except in the case where an adjusted nuclear cross-section library was used. Exact three-dimensional, pin-by-pin geometry and continuous-energy cross sections were used in the Monte Carlo calculation. The estimated error components associated with cross-sections, methods correction factors and the bias factor were combined based on Takeda's theory. Those independently calculated values agree well and range from 2.8 to 3.4%Δk/kk'. The calculation result of the MK-III core management code system HESTLA was 3.13% Δk/kk'. The estimated errors for bias method range from 0.1 to 0.2%Δk/kk'. The error in the case using adjusted cross-section was 0.3%Δk/kk'. (author)

  15. Tests on irradiated magnet-insulator materials

    International Nuclear Information System (INIS)

    Schmunk, R.E.; Miller, L.G.; Becker, H.

    1983-01-01

    Fusion-reactor coils, located in areas where they will be only partially shielded, must be fabricated from materials which are as resistant to radiation as possible. They will probably incorporate resistive conductors with either water or cryogenic cooling. Inorganic insulators have been recommended for these situations, but the possibility exists that some organic insulators may be usuable as well. Results were previously reported for irradiation and testing of three glass reinforced epoxies: G-7, G-10, and G-11. Thin disks of these materials, nominally 0.5 mm thick by 11.1 mm diameter, were tested in compressive fatigue, a configuration and loading which represents reasonably well the magnet environment. In that work G-10 was shown to withstand repeated loading to moderately high stress levels without failure, and the material survived better at liquid nitrogen temperature than at room temperature

  16. PRELIMINARY RESULTS OF THE AGC-4 IRRADIATION IN THE ADVANCED TEST REACTOR AND DESIGN OF AGC-5 (HTR16-18469)

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Very High Temperature Gas Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. The third experiment, AGC-3, started its irradiation in late November 2012 and completed in the April of 2014. AGC-4 is currently being irradiated in the ATR. This paper will briefly discuss the preliminary irradiation results

  17. Irradiation effect study on dynamic performance of grease

    International Nuclear Information System (INIS)

    Arakawa, Kazuo; Sasuga, Tsuneo; Hagiwara, Miyuki; Yajima, Toshio; Nakanishi, Hiroshi; Soda, Takao; Iwamoto, Takeshi.

    1986-03-01

    The radiation resistance of newly developed grease, MORESCO HIRAD RG-42-0, consisting of polyphenyl ether as base oil and inorganic compound as gelling agent was evaluated by conducting measurements of frictional resistance force on mechanical snubbers. After the irradiation of 60 Co γ-rays up to a predetermined dose [max. 100 MGy (10000 Mrad)], the grease was applied to mechanical snubbers. The rated load of mechanical snubbers used (standard stroke of 125 mm) was set at 0.16 ton and 1.0 ton. The frictional resistance force was measured at a velocity of 0.001, 0.01, 0.1 and 2.0 mm/sec and at temperature of 60 deg C. When the rated load was 0.16 ton, the frictional resistance force of irradiated grease was kept less than a criterion, a greater value of either 5 kgf or 1 % of the rated load, at every tested dose up to 90 MGy. In case of 1.0 ton, the frictional resistance force of the irradiated grease was less than a greater value of either 10 kgf or 1 % of the rated load at doses up to 70 MGy. (author)

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

    International Nuclear Information System (INIS)

    Ozawa, Takayuki; Abe, Tomoyuki

    2006-03-01

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

  19. Test of irradiation of tellurium oxide for obtaining iodine-131 by dry distillation

    International Nuclear Information System (INIS)

    Alanis M, J.

    2003-07-01

    With the purpose of optimizing to the maximum independently the work of the reactor of those mathematical calculations of irradiation that are already optimized, now it corresponds to carry out irradiation tests in the different positions with their respective neutron fluxes that it counts the reactor for samples irradiation. Then, it is necessary to carry out the irradiation of the tellurium dioxide through cycles, with the purpose of observing the activity that it goes accumulating in each cycle and this way to obtain an activity of the Iodine-131 obtained when finishing the last cycle. (Author)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  1. Results from the CDE phase activity on neutron dosimetry for the international fusion materials irradiation facility test cell

    Energy Technology Data Exchange (ETDEWEB)

    Esposito, B. E-mail: esposito@frascati.enea.it; Bertalot, L.; Maruccia, G.; Petrizzi, L.; Bignan, G.; Blandin, C.; Chauffriat, S.; Lebrun, A.; Recroix, H.; Trapp, J.P.; Kaschuck, Y

    2000-11-01

    The international fusion materials irradiation facility (IFMIF) project deals with the study of an accelerator-based, deuterium-lithium source, producing high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials for fusion energy reactors. IFMIF would also provide calibration and validation of data from fission reactor and other accelerator based irradiation tests. This paper describes the activity on neutron/gamma dosimetry (necessary for the characterization of the specimens' irradiation) performed in the frame of the IFMIF conceptual design evaluation (CDE) neutronics tasks. During the previous phase (conceptual design activity (CDA)) the multifoil activation method was proposed for the measurement of the neutron fluence and spectrum and a set of suitable foils was defined. The cross section variances and covariances of this set of foils have now been used for tests on the sensitivity of the IFMIF neutron spectrum determination to cross section uncertainties. The analysis has been carried out using the LSL-M2 code, which optimizes the neutron spectrum by means of a least-squares technique taking into account the variance and covariance files. In the second part of the activity, the possibility of extending to IFMIF the use of existing on-line in-core neutron/gamma monitors (to be located at several positions inside the IFMIF test cell for beam control, safety and diagnostic purposes) has been studied. A feasibility analysis of the modifications required to adapt sub-miniature fission chambers (recently developed by CEA-Cadarache) to the high flux test module of the test cell has been carried out. The verification of this application pertinence and a gross definition of the in-core detector characteristics are described. The option of using self-powered neutron detectors (SPNDs) is also discussed.

  2. Review Paper: Review of Instrumentation for Irradiation Testing of Nuclear Fuels and Materials

    International Nuclear Information System (INIS)

    Kim, Bong Goo; Rempe, Joy L.; Villard, Jean-Francois; Solstadd, Steinar

    2011-01-01

    Over 50 years of nuclear fuels and materials irradiation testing has led to many countries developing significant improvements in instrumentation to monitor physical parameters and to control the test conditions in material test reactors (MTRs). Recently, there is increased interest to irradiate new materials and reactor fuels for advanced pressurized water reactors and Gen-IV reactor systems, such as sodium-cooled fast reactors, very high temperature reactors, supercritical water-cooled reactors, and gas-cooled fast reactors. This review paper documents the current state of instrumentation technologies in MTRs in the world and summarizes ongoing research efforts to deploy new sensors. As described in this paper, a wide range of sensors is available to measure key parameters of interest during fuels and materials irradiations in MTRs. Ongoing development efforts focus on providing MTR users a wider range of parameter measurements with smaller, higher accuracy sensors.

  3. Experimental study associated to irradiation of FBR structural material, (4)

    International Nuclear Information System (INIS)

    1976-01-01

    The study presents one of the bases to evaluate the results of the post-irradiation tests to conduct the thermal control tests related to the second JMTR irradiation (70M-61P) of the demestic austenitic stainless steels for the structural material of the FBR performed by Power Reactor and Nuclear Fuel Development Corporation. The thermal control specimens were given the temperature history which simulated that of the irradiation temperature in vacuum by the electrical furnance, and then the tensile, fatigue and Charpy impact tests were performed. The changes of the material properties caused by the thermal history were investigated. (auth.)

  4. Irradiation performance of AGR-1 high temperature reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Hunn, John D. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Ploger, Scott A. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Morris, Robert N.; Baldwin, Charles A. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Harp, Jason M.; Winston, Philip L. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Gerczak, Tyler J. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Rooyen, Isabella J. van [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Montgomery, Fred C.; Silva, Chinthaka M. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States)

    2016-09-15

    Highlights: • Post-irradiation examination was performed on AGR-1 coated particle fuel. • Cesium release from the particles was very low in the absence of failed SiC layers. • Silver release was often substantial, and varied considerably with temperature. • Buffer and IPyC layers were found to play a key role in TRISO coating behavior. • Fission products palladium and silver were found in the SiC layer of particles. - Abstract: The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of {sup 110m}Ag from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10{sup −4} to 5 × 10{sup −4} for {sup 154}Eu and 8 × 10{sup −7} to 3 × 10{sup −5} for {sup 90}Sr. The average {sup 134}Cs fractional release from compacts was <3 × 10{sup −6} when all particles maintained intact SiC. An estimated four particles out of 2.98 × 10{sup 5} in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving {sup 134}Cs fractional release in two capsules to approximately 10{sup −5}. Identification and characterization of these particles has provided unprecedented insight into

  5. Present status of ESNIT (energy selective neutron irradiation test facility) program

    International Nuclear Information System (INIS)

    Noda, K.; Ohno, H.; Sugimoto, M.; Kato, Y.; Matsuo, H.; Watanabe, K.; Kikuchi, T.; Sawai, T.; Usui, T.; Oyama, Y.; Kondo, T.

    1994-01-01

    The present status of technical studies of a high energy neutron irradiation facility, ESNIT (energy selective neutron irradiation test facility), is summarized. Technological survey and feasibility studies of ESNIT have continued since 1988. The results of technical studies of the accelerator, the target and the experimental systems in ESNIT program were reviewed by an International Advisory Committee in February 1993. Recommendations for future R and D on ESNIT program are also summarized in this paper. ((orig.))

  6. Gemstone enhancing dedicated gamma irradiator development

    International Nuclear Information System (INIS)

    Omi, Nelson Minoru

    2006-01-01

    The gemstones gamma irradiation process to enhance the color is widely accepted for the jewelry industry. These gems are processed in conventional industrial gamma irradiation plant which are optimized for other purposes, using underwater irradiation devices with high rejection rate due to it's poor dose uniformity. A new conception design, which states the working principles and manufacturing ways of the device, was developed in this work. The suggested device's design is based on the rotation of cylindrical baskets and their translation in circular paths inside and outside a cylindrical source rack as a planetary system. The device is meant to perform the irradiation in the bottom of the source storage pool, where the sources remain always shielded by the water layer. The irradiator matches the Category III IAEA classification. To verify the physical viability of the basic principle, tests with rotating cylindrical baskets were performed in the Multipurpose Irradiator raised in the CTR, IPEN. Also, simulations using the CADGAMMA software, adapted to simulate underwater irradiations were performed. With the definitive irradiator, the irradiation quality will be enhanced with better dose control and the production costs will be significantly lower than market prices due to the intended treatment device's optimization. (author)

  7. Status report on the irradiation testing and post-irradiation examination of low-enriched U3O8-Al and UAlsub(x)-Al fuel element by the Netherlands Energy Research Foundation (ECN)

    International Nuclear Information System (INIS)

    Pruimboom, H.; Lijbrink, E.; Otterdijk, K. von; Swanenburg de Veye, R.J.

    1984-01-01

    Within the framework of the RERTR-programme four low-enriched (20%) MTR-type fuel elements have been irradiated in the High Flux Reactor at Petten (The Netherlands) and are presently subjected to postirradiation examination. Two of the elements contain UAlsub(x)-Al and two contain U 3 O 8 -Al fuel. The test irradiation has been completed up to the target burn-up values of 50% and 75% respectively. An extensive surveillance programme carried out during the test period has confirmed the excellent in-reactor behaviour of both types. Post-irradiation examination of the 50% burn-up test elements, comprising of dimensional measurements, burn-up determination, fuel metallography and blister testing, has sofar confirmed the irradiation experiences. Good agreement between calculated and measured power and burn-up characteristics has been found. A survey of the test element characteristics, their irradiation history, the irradiation tests and the preliminary PIE results is given in the paper. (author)

  8. The efficacy testing of irradiated shrimp paste

    International Nuclear Information System (INIS)

    Nouchpramool, Kovit; Eamsiri, Jaruratana; Sujjabut, Surusak

    2005-10-01

    Two lots of shrimp paste from commercial source in Samutsakhon were irradiated at a recommended minimum dose of 6 kGy using a J S 8900 cobalt-60 carrier gamma irradiator of Thai Irradiation Center in Patum Thani. Red Perspex dosimeter were used to measure the absorbed dose throughout the product with emphasis on the region of minimum and maximum absorbed dose. This way, it was aimed to compare the dose effects of gamma irradiation on the microbiological, chemical and sensory quality of shrimp paste. The results indicated that the shrimp paste received minimum and maximum absorbed dose of 6.85 and 12.83 kGy with dose uniformity ratio of 1.87 . Throughput rate is 468 kilogram per hour. The microbiological load of shrimp paste was rather high resulting in not compliance with Thai industrial standard 1080-2535. Irradiation at 6.8 kGy reduced total viable bacterial count by one log cycle. Although the irradiated product was organoleptic ally acceptable and could be kept for 16 months at room temperature, mold and Clostridium perfringens were still present in some samples after irradiation and during prolonged storage in amount that exceeds the limitation of Thai industrial standard. Chemical properties such as p H, moisture and sodium chloride content of irradiated shrimp paste were not significantly changed after irradiation

  9. Effects of environmentally differential rearing upon maze performance in prenatally irradiated microcephalic rats

    International Nuclear Information System (INIS)

    Seo, M.L.; Inouye, M.; Kiyono, S.; Shibagaki, M.

    1982-01-01

    Pregnant rats received 100 rads of X-irradiation on day 17 of gestation. Control pregnant rats were sham-irradiated on the same gestation day. The male offspring were reared under environmentally enriched, standard colony, and impoverished conditions for 30 days after weaning. Then the Hebb-Williams maze test was carried out. All the prenatally X-irradiated rats were microcephalic: their mean cerebral wet weight was 15.5% less than controls. The effect of X-irradiation was not significant in error scores and running times, whereas the effect of environment was significant in these items; initial and total error scores and running times were decreased in enriched groups compared to impoverished groups in controls as well as in X-irradiated animals

  10. Fatigue performance of HFIR-irradiated Nimonic PE-16 at 4300C

    International Nuclear Information System (INIS)

    Grossbeck, M.L.; Liu, K.C.

    1983-01-01

    Nimonic PE-16 was irradiated in the HFIR to 6 to 9 dpa and 560 to 1000 at. ppM He at 430 0 C. Postirradiation fatigue tests revealed a reduction in fatigue life by about a factor of 10 at 430 0 C. In contrast to AISI type 316 stainless steel, no endurance limit was observed. All irradiated specimens exhibited some intergranular fracture with an increasing tendency toward cleavage-like intragranular fracture for low strain ranges

  11. Development of oxygen sensing technology in an irradiated fuel rod. Characteristic test of oxygen sensor

    International Nuclear Information System (INIS)

    Saito, Junichi; Hoshiya, Taiji; Sakurai, Fumio; Sakai, Haruyuki

    1996-03-01

    At the Department of JMTR (Japan Materials Test Reactor), the re-instrumentation technologies to a high burnup fuel rod irradiated in an LWR have been developed to study irradiation behavior of the fuel during power transient. It has been progressed developing a chemical sensor as one of the re-instrumentation technologies. This report summarizes the results of characteristic tests of an oxygen sensor made of Yttria Stabilized Zirconia (YSZ) as a solid electrolyte. Several kinds of experiments were carried out to evaluate the electromotive force (emf) performance, stability and lifetime of the oxygen sensor with Ni/NiO, Cr/Cr 2 O 3 and Fe/FeO, respectively as a reference electrode. From the experimental data, it is suggested that the reference electrode of Ni/NiO reveals the most appropriate characteristic of the sensor to measure the partial oxygen pressure in a fuel rod. It is the final goal of this development to clarify the change of oxygen chemical potential in a fuel rod during power transient. (author)

  12. Irradiation performance of experimental fast reactor 'JOYO' MK-1 driver fuel assemblies

    International Nuclear Information System (INIS)

    Itaki, Toshiyuki; Kono, Keiichi; Tachi, Hirokatsu; Yamanouchi, Sadamu; Yuhara, Shunichi; Shibahara, Itaru

    1985-01-01

    The experimental fast reactor ''JOYO'' completed it's breeder core (MK-I) operation in January 1982. The MK-I driver fuel assemblies were removed from the core sequencially in order of burnup increase and have been under postirradiation examination (PIE). The PIE has almost been completed for 30 assemblies including the highest burnup assemblies of 48,000 MWD/MTM. It has been confirmed that all fuel assemblies have exhibited satisfactory performance without detrimental assembly deformation or without any indications of fuel pin breach. The irradiation conditions of the MK-I core were somewhat more moderate than those conditions envisioned for prototypic reactor. However the results of the examination revealed the typical irradiation behavior of LMFBR fuels, although such characteristics were benign as compared with those anticipated in high burnup fuels. Systematic performance data have been accumulated through the fuel fabrication, irradiation and postirradiation examination processes. Based on these data, the MK-I fuel designing and fabrication techniques were totally confirmed. This technical experience and the associated insight into irradiation behavior have established a milestone to the next step of fast reactor fuel development. (author)

  13. NSRR experiment with un-irradiated uranium-zirconium hydride fuel. Design, fabrication process and inspection data of test fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kuroha, Hiroshi; Ikeda, Yoshikazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Aizawa, Keiichi

    1998-08-01

    An experiment plan is progressing in the Nuclear Safety Research Reactor (NSRR) to perform pulse-irradiation with uranium-zirconium hydride (U-ZrH{sub x}) fuel. This fuel is widely used in the training research and isotope production reactor of GA (TRIGA). The objectives of the experiment are to determine the fuel rod failure threshold and to investigate fuel behavior under simulated reactivity initiated accident (RIA) conditions. This report summarizes design, fabrication process and inspection data of the test fuel rods before pulse-irradiation. The experiment with U-ZrH{sub x} fuel will realize precise safety evaluation, and improve the TRIGA reactor performance. The data to be obtained in this program will also contribute development of next-generation TRIGA reactor and its safety evaluation. (author)

  14. Differential expression of growth factors in irradiated mouse testes

    International Nuclear Information System (INIS)

    Mauduit, Claire; Siah, Ahmed; Foch, Marie; Chapet, Olivier; Clippe, Sebastien; Gerard, Jean-Pierre; Benahmed, Mohamed

    2001-01-01

    Purpose: By using as an experimental model the male mouse gonad, which contains both radiosensitive (germ) and radioresistant (somatic) cells, we have studied the growth factor (and/or receptor) expression of transforming growth factor-β receptor (TGFβ RI), stem cell factor (SCF), c-kit, Fas-L, Fas, tumor necrosis factor receptor (TNF R55), and leukemia inhibiting factor receptor (LIF-R) after local irradiation. Methods and Materials: Adult male mice were locally irradiated on the testes. Induction of apoptosis in the different testicular cell types following X-ray radiation was identified by the TdT-mediated dUTP Nick End Labeling (TUNEL) approach. Growth factor expression was evidenced by semiquantitative RT-PCR and Western blot analyses. Results: Apoptosis, identified through the TUNEL approach, occurred in radiosensitive testicular (premeotic) germ cells with the following kinetics: the number of apoptotic cells increased after 24 h (p<0.001) and was maximal 48 h after a 2-Gy ionizing radiation (p<0.001). Apoptotic cells were no longer observed 72 h after a 2-Gy irradiation. The number of apoptotic cells increased with the dose of irradiation (1-4 Gy). In the seminiferous tubules, the growth factor expression in premeiotic radiosensitive germ cells was modulated by irradiation. Indeed Fas, c-kit, and LIF-R expression, which occurs in (radiosensitive) germ cells, decreased 24 h after a 2-Gy irradiation, and the maximal decrease was observed with a 4-Gy irradiation. The decrease in Stra8 expression occurred earlier, at 4 h after a 2-Gy irradiation. In addition, a significant (p<0.03) decrease in Stra8 mRNA levels was observed at the lowest dose used (0.5 Gy, 48 h). Moreover, concerning a growth factor receptor, such as TGFβ RI, which is expressed both in radiosensitive and radioresistant cells, we observed a differential expression depending on the cell radiosensitivity after irradiation. Indeed, TGFβ RI expression was increased after irradiation in

  15. Application of fiber optic grating strain sensor for measurement of strain under irradiation environment

    International Nuclear Information System (INIS)

    Kaji, Y.; Matsui, Y.; Kita, S.; Ide, H.; Tsukada, T.; Tsuji, H.

    2001-01-01

    In Japan Atomic Energy Research Institute (JAERI), in-pile strain measurement techniques have been developed using Japan Materials Testing Reactor (JMTR). In order to evaluate the performance of fiber optic grating sensor under irradiation environment, heat-up and performance tests at elevated temperature before irradiation and in-pile tests were performed in JMTR. (author)

  16. Development of a Low Temperature Irradiation Capsule for Research Reactor Materials

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  17. Results of crack-arrest tests on two irradiated high-copper welds

    International Nuclear Information System (INIS)

    Iskander, S.K.; Corwin, W.R.; Nanstead, R.K.

    1990-12-01

    The objective of this study was to determine the effect of neutron irradiation on the shift and shape of the lower-bound curve to crack-arrest data. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plate. Crack-arrest specimens fabricated from these welds were irradiated at a nominal temperature of 288 degree C to an average fluence of 1.9 x 10 19 neutrons/cm 2 (>1 MeV). Evaluation of the results shows that the neutron-irradiation-induced crack-arrest toughness temperature shift is about the same as the Charpy V-notch impact temperature shift at the 41-J energy level. The shape of the lower-bound curves (for the range of test temperatures covered) did not seem to have been altered by irradiation compared to those of the ASME K Ia curve. 9 refs., 21 figs., 10 tabs

  18. Apparatus of irradiation of steel test pieces in the Marcoule pile G 1

    International Nuclear Information System (INIS)

    Marinot, R.; Wallet, Ph.

    1960-01-01

    Test pieces of steel were irradiated in the reactor G1 at Marcoule, in convectors replacing fuel elements, and in vertical channels in furnace-heated containers. The apparatus designed for this irradiation is described: containers, converter-rods, suspension fixtures and clamps, temperature measurement devices, lead castles and unloading set-ups. (author) [fr

  19. Postirradiation examination results for the Irradiation Effects Test IE-5

    International Nuclear Information System (INIS)

    Cook, T.F.; Ploger, S.A.; Hobbins, R.R.

    1978-03-01

    The results are presented of the postirradiation examination of four pressurized water reactor type fuel rods which were tested in-pile under a fast power ramp and film boiling operation during Irradiation Effects (IE) Test 5. The major objectives of this test were to evaluate the effects of simulated fission products on fuel rod behavior during a fast power ramp, to determine the effects of high initial internal pressure on a fuel rod during film boiling, and to assess fuel rod property changes that occur during film boiling in a fuel rod with previously irradiated cladding. The overall condition of the rods and changes that occurred in fuel and cladding as a result of the power ramp and film boiling operation, as determined from the postirradiation examination, are reported and analyzed. Effects of the simulated fission products on fuel rod behavior during a power ramp are discussed. The effect of high internal pressure on rod behavior during film boiling is evaluated. Cladding temperatures are estimated at various axial and circumferential locations. Cladding embrittlement by oxidation is also assessed

  20. Rape seed glucosinolate: radiation inactivation and physiological performance of broiler fed irradiated rapeseed meal

    International Nuclear Information System (INIS)

    Farag, M.Diaa El-Din H.

    1994-01-01

    Rape seeds meal (RSM) is a high quality protein supplement suitable for all classes of livestock. The major area of concern in animal nutrition has been glucosinolates and their derivative products which cause depressed performance in poultry or may be even toxic. Therefore, these substances must be removed or inactivated before the meal can be used as potential protein source for food or feed. I the current study, RSM has been used to test whether gamma radiation processing can inactivate glucosinolates as a step towards detoxication. Samples were exposed to gamma rays of 10, 50, 100 and 250 kGy. Approximated analysis showed that RSM was not affected by irradiation processing up to 250 kGy. However, the crude fiber content decreased at the highest dose while at doses of 10, 50 100 and 250 kGy the available lysine decreased by 6.76%, 9.46%, 17.84% and 22.43%, respectively. Radiation processing at 250 kGy significantly inactivated glucosinolate by 85% from its initial value. In a 8-week chick-feeding study, raw and irradiated RSM were applied at 30%. The diets containing raw and irradiated (at 10, 50 and 100 kGy) RSM had somewhat low growth and thyroid, liver and kidney enlargement compared to the basal control group. No significant difference was observed between chicks fed on RSM irradiated at 250 kGy and those fed on basal diet. No significant differences were observed in the serum protein, albumin, GPT, uric acid, creatine and basal diet groups. Those kept on raw and irradiated at 10, 50 and 100 kGy RSM had higher GOT than those kept on irradiated at 250 kGy RSM and basal diet. Radiation treatment of RSM up to 250 kGy improved its nutritional quality by decreasing the glucosinolate and consequently maintained the chicks in a better health condition. (author)

  1. Development of a capsule assembly machine for the re-irradiation tests in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Y. H.; Choi, M. H.; Sohn, J. M.; Choo, K. N.; Cho, M. S.; Kim, B. G. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-11-15

    A capsule assembly machine (CAM) for the long term irradiation tests in the HANARO reactor has been designed, developed and demonstrated at the Korea Atomic Energy Reasearch Institute (KAERI). The CAM will provide a technical base for viable re-irradiation servives. This machine will be installed in the reactor service pool of the HANARO reactor. The new assembly technique by using a mockup of the CAM in air demonstrated its suitability for an assembly operation, and for an application of this technique to a reactor. The technique will be upgraded after a commissioning test under water environments. This would be expected to be recommended for a country where an under water canal for transporting irradiated devices and enough space of a hot cell for assembling capsule components are not available.

  2. Summary report of the experimental fast reactor JOYO MK-III performance test

    International Nuclear Information System (INIS)

    Maeda, Yukimoto; Aoyama, Takafumi; Yoshida, Akihiro

    2004-03-01

    An upgrading project (MK-III project) was started to improve the irradiation capability of the experimental fast reactor JOYO. In this project, core replacement and increase of the reactor thermal power by the factor 1.4 were necessary for increasing the maximum fast neutron flux by the factor 1.3 and doubling the capacity for irradiation rigs. The modification of the cooling system that included the replacement of the main intermediate heat exchangers and the dump heat exchangers was completed in September 2000. After a series of system function tests, the performance test, of which objective is to fully characterize the upgraded core and heat transfer system, was started in June 2003. Twenty eight tests were selected and carried out as performance test, in order to confirm that the whole plant satisfy the design criteria and have sufficient characteristics (data necessary for safe and steady operation, core management, reactor control and monitoring) as an irradiation bed. After attaining the initial criticality of the core on 2nd July 2003, core characteristics (the excess reactivity, the isotherm temperature reactivity coefficient, the power reactivity coefficient and so on), plant characteristics (the plant heat balance, the adjustment of the temperature control system, the plant behavior at transient), shielding characteristics (dose rate distribution). As the result, it was confirmed that all the criteria regulated was satisfied and the core and plant have sufficient margins for full power operation, which was increased by the factor 1.4. Especially, nuclear analysis accuracy was verified by comparing the calculation with measured core characteristics of the initial core which consists of fifty five fresh fuel subassemblies. The operational data which is supposed to be useful for developing in-core anomaly detection system were also obtained. The operation manual and training simulator and design of next reactor development were revised based on the results

  3. RERTR-12 Insertion 1 Irradiation Summary Report

    International Nuclear Information System (INIS)

    Perez, D.M.; Lillo, M.A.; Chang, G.S.; Woolstenhulme, N.E.; Roth, G.A.; Wachs, D.M.

    2012-01-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications. RERTR-12 insertion 1 includes the capsules irradiated during the first two irradiation cycles. These capsules include Z, X1, X2 and X3 capsules. The following report summarizes the life of the RERTR-12 insertion 1 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

  4. RERTR-12 Insertion 2 Irradiation Summary Report

    International Nuclear Information System (INIS)

    Perez, D.M.; Chang, G.S.; Wachs, D.M.; Roth, G.A.; Woolstenhulme, N.E.

    2012-01-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications.1 RERTR-12 insertion 2 includes the capsules irradiated during the last three irradiation cycles. These capsules include Z, Y1, Y2 and Y3 type capsules. The following report summarizes the life of the RERTR-12 insertion 2 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

  5. Stability of the Hall sensors performance under neutron irradiation

    International Nuclear Information System (INIS)

    Duran, I.; Hron, M.; Stockel, J.; Viererbl, L.; Vsolak, R.; Cerva, V.; Bolshakova, I.; Holyaka, R.; Vayakis, G.

    2004-01-01

    A principally new diagnostic method must be developed for magnetic measurements in steady state regime of operation of fusion reactor. One of the options is the use of transducers based on Hall effect. The use of Hall sensors in ITER is presently limited by their questionable radiation and thermal stability. Issues of reliable operation in ITER like radiation and thermal environment are addressed in the paper. The results of irradiation tests of candidate Hall sensors in LVR-15 and IBR-2 experimental fission reactors are presented. Stable operation (deterioration of sensitivity below one percent) of the specially prepared sensors was demonstrated during irradiation by the total fluence of 3.10 16 n/cm 2 in IBR-2 reactor. Increasing the total neutron fluence up to 3.10 17 n/cm 2 resulted in deterioration of the best sensor's output still below 10% as demonstrated during irradiation in LVR-15 fission reactor. This level of neutron is already higher than the expected ITER life time neutron fluence for a sensor location just outside the ITER vessel. (authors)

  6. Development of irradiation technique with satured temperature capsule in the JMTR

    International Nuclear Information System (INIS)

    Ohtaka, Kimihiro

    1999-01-01

    The irradiation assisted stress corrosion cracking (IASCC) of in-core structural materials caused by the simultaneous effects of neutron irradiation and high temperature water environments has been pointed out as one of the major concerns not only for the light water reactors (LWRs) but also for the water-cooled fusion reactor, i.e,. ITER. The IASCC of the austenitic stainless steels or nickel base alloys has been studied for more than ten years under international efforts in the various projects for the plant life assessment and extension of LWRs. However its mechanism has not been clarified yet in spite of the extensive post-irradiation examinations. Under this situation, it is desired to perform irradiation tests under specially controlled conditions so that the effect of irradiation and high temperature water can be separately evaluated. In the Japan Materials Testing Reactor (JMTR), irradiation technique with the saturation temperature capsule (SATCAP) was developed for irradiation of the materials in the water with high, but constant, temperature and applied to study the IASCC. The capability of the SATCAP was improved by enhancing the temperature controllability to irradiate materials even in a low gamma region in the JMTR core. The performance tests of the improved SATCAP carried out in the JMTR have proven its capabilities. Based on experiences of the SATCAP, preliminary design study for the upgraded in-pile test facility are now underway in the JMTR. The test facility has a new test loop to achieve irradiate test simulated water environment of LWRs. The design, test results of the SATCAP and the design study of upgraded in-pile test facility are described in this paper

  7. Doses from external irradiation to Marshall Islanders from Bikini and Enewetak nuclear weapons tests.

    Science.gov (United States)

    Bouville, André; Beck, Harold L; Simon, Steven L

    2010-08-01

    Annual doses from external irradiation resulting from exposure to fallout from the 65 atmospheric nuclear weapons tests conducted in the Marshall Islands at Bikini and Enewetak between 1946 and 1958 have been estimated for the first time for Marshallese living on all inhabited atolls. All tests that deposited fallout on any of the 23 inhabited atolls or separate reef islands have been considered. The methodology used to estimate the radiation doses at the inhabited atolls is based on test- and location-specific radiation survey data, deposition density estimates of 137Cs, and fallout times-of-arrival provided in a companion paper (Beck et al.), combined with information on the radionuclide composition of the fallout at various times after each test. These estimates of doses from external irradiation have been combined with corresponding estimates of doses from internal irradiation, given in a companion paper (Simon et al.), to assess the cancer risks among the Marshallese population (Land et al.) resulting from exposure to radiation from the nuclear weapons tests.

  8. A proton irradiation test facility for space research in Ankara, Turkey

    Science.gov (United States)

    Gencer, Ayşenur; Yiǧitoǧlu, Merve; Bilge Demirköz, Melahat; Efthymiopoulos, Ilias

    2016-07-01

    Space radiation often affects the electronic components' performance during the mission duration. In order to ensure reliable performance, the components must be tested to at least the expected dose that will be received in space, before the mission. Accelerator facilities are widely used for such irradiation tests around the world. Turkish Atomic Energy Authority (TAEA) has a 15MeV to 30MeV variable proton cyclotron in Ankara and the facility's main purpose is to produce radioisotopes in three different rooms for different target systems. There is also an R&D room which can be used for research purposes. This paper will detail the design and current state of the construction of a beamline to perform Single Event Effect (SEE) tests in Ankara for the first time. ESA ESCC No.25100 Standard Single Event Effect Test Method and Guidelines is being considered for these SEE tests. The proton beam kinetic energy must be between 20MeV and 200MeV according to the standard. While the proton energy is suitable for SEE tests, the beam size must be 15.40cm x 21.55cm and the flux must be between 10 ^{5} p/cm ^{2}/s to at least 10 ^{8} p/cm ^{2}/s according to the standard. The beam size at the entrance of the R&D room is mm-sized and the current is variable between 10μA and 1.2mA. Therefore, a defocusing beam line has been designed to enlarge the beam size and reduce the flux value. The beam line has quadrupole magnets to enlarge the beam size and the collimators and scattering foils are used for flux reduction. This facility will provide proton fluxes between 10 ^{7} p/cm ^{2}/s and 10 ^{10} p/cm ^{2}/s for the area defined in the standard when completed. Also for testing solar cells developed for space, the proton beam energy will be lowered below 10MeV. This project has been funded by Ministry of Development in Turkey and the beam line construction will finish in two years and SEE tests will be performed for the first time in Turkey.

  9. Evaluation of strain-rate sensitivity of ion-irradiated austenitic steel using strain-rate jump nanoindentation tests

    Energy Technology Data Exchange (ETDEWEB)

    Kasada, Ryuta, E-mail: r-kasada@iae.ky