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Sample records for charpy specimens irradiated

  1. Standard Guide for Reconstitution of Irradiated Charpy-Sized Specimens

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This guide covers procedures for the reconstitution of ferritic pressure boundary steels used in nuclear power plant applications, Type A Charpy (Test Methods E 23) specimens and specimens suitable for testing in three point bending in accordance with Test Methods E 1921 or E 1820. Materials from irradiation programs (principally broken specimens) are reconstituted by welding end tabs of similar material onto remachined specimen sections that were unaffected by the initial test. Guidelines are given for the selection of suitable specimen halves and end tab materials, for dimensional control, and for avoidance of overheating the notch area. A comprehensive overview of the reconstitution methodologies can be found in Ref (1). 1.2 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only. 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 standard...

  2. Fractographic examination of reduced activation ferritic/martensitic steel charpy specimens irradiated to 30 dpa at 370{degrees}C

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, D.S.; Hamilton, M.L. [Pacific Northwest National Lab., Richland, WA (United States); Schubert, L.E. [Univ. of Missouri, Rolla, MO (United States)

    1996-10-01

    Fractographic examinations are reported for a series of reduced activation ferritic/Martensitic steel Charpy impact specimens tested following irradiation to 30 dpa at 370{degrees}C in FFTF. One-third size specimens of six low activation steels developed for potential application as structural materials in fusion reactors were examined. A shift in brittle fracture appearance from cleavage to grain boundary failure was noted with increasing manganese content. The results are interpreted in light of transmutation induced composition changes in a fusion environment.

  3. Reconstituted Charpy impact specimens. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Perrin, J.S.; Wullaert, R.A.; McConnell, P.; Server, W.L.; Fromm, E.O.

    1982-12-01

    The arc stud welding process was used to produce new, full size Charpy V-notch impact specimens from halves of Charpy specimens which had been previously tested. The apparatus was developed such that it could be used not only for unirradiated specimens, but also so that it could be adapted for in-cell use to produce new reconstituted specimens of irradiated material. The materials studied are of interest in nuclear applications. They include A533B, A36, A516-80, submerged arc weld metal (A508 base metal), HY80, cast duplex stainless steel, irradiated A533B, and irradiated submerged arc weld metal (A508 base metal). Both unirradiated and irradiated specimens were successfully produced and subsequently impact tested. In general, there was excellent agreement when comparing the original curves to the subsequent curves generated with reconstituted specimens. This program has shown that the arc stud welding process is well suited for producing reconstituted specimens at a reasonable cost using either unirradiated or irradiated material.

  4. Correlations between Standard and Miniaturised Charpy-V Specimens

    Energy Technology Data Exchange (ETDEWEB)

    Lucon, E.; Van Walle, E.; Fabry, A.; Puzzolante, J.-L.; Verstrepen, A.; Vosch, R.; Van de Velde, L

    1998-12-01

    A total of 565 instrumented impact tests (232 performed on full-size and 333 on sub-size Charpy-V specimens) have been analysed in order to derive meaningful assumptions on the correlations existing between test results obtained on specimens of different size. Nine materials (pressure vessel steels) have been considered, in both as-received and irradiated state, for a total of 19 conditions examined. For the analysis of data, conventional as well novel approaches have been investigated; former ones, based on a review of the existing literature, include predictions of USE values by the use of normalization factors (NF), shifts of index temperatures related to energy/lateral expansion/shear fracture levels, and a combination of both approaches (scaling and shifting of energy curves). More original and recent proposals have also been verified, available in the literature but also proposed by SCK-CEN in the frame of enhanced surveillance of nuclear reactor pressure vessels. Conclusions have been drawn regarding the applicability and reliability of these methodologies, and recommendations have been given for future developments of the activities on this topic.

  5. Tensile and charpy impact properties of irradiated reduced-activation ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Alexander, D.J. [Oak Ridge National Lab., TN (United States)

    1996-10-01

    Tensile tests were conducted on eight reduced-activation Cr-W steels after irradiation to 15-17 and 26-29 dpa, and Charpy impact tests were conducted on the steels irradiated to 26-29 dpa. Irradiation was in the Fast Flux Test Facility at 365{degrees}C on steels containing 2.25-12% Cr, varying amounts of W, V, and Ta, and 0.1%C. Previously, tensile specimens were irradiated to 6-8 dpa and Charpy specimens to 6-8, 15-17, and 20-24 dpa. Tensile and Charpy specimens were also thermally aged to 20000 h at 365{degrees}C. Thermal aging had little effect on the tensile behavior or the ductile-brittle transition temperature (DBTT), but several steels showed a slight increase in the upper-shelf energy (USE). After {approx}7 dpa, the strength of the steels increased and then remained relatively unchanged through 26-29 dpa (i.e., the strength saturated with fluence). Post-irradiation Charpy impact tests after 26-29 dpa showed that the loss of impact toughness, as measured by an increase in DBTT and a decrease in the USE, remained relatively unchanged from the values after 20-24 dpa, which had been relatively unchanged from the earlier irradiations. As before, the two 9Cr steels were the most irradiation resistant.

  6. Fracture toughness and Charpy impact properties of several RAFMS before and after irradiation in HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, M.A. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6151 (United States)]. E-mail: sokolovm@ornl.gov; Tanigawa, H. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Odette, G.R. [University of California-Santa Barbara, Santa Barbara, CA 93106-5080 (United States); Shiba, K. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Klueh, R.L. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6151 (United States)

    2007-08-01

    As part of the development of candidate reduced-activation ferritic steels for fusion applications, several steels, namely F82H, 9Cr-2WVTa steels and F82H weld metal, are being investigated in the joint DOE-JAEA collaboration program. Within this program, three capsules containing a variety of specimen designs were irradiated at two design temperatures in the ORNL High Flux Isotope Reactor (HFIR). Two capsules, RB-11J and RB-12J, were irradiated in the HFIR removable beryllium positions with europium oxide (Eu{sub 2}O{sub 3}) thermal neutron shields in place. Specimens were irradiated up to 5 dpa. Capsule JP25 was irradiated in the HFIR target position to 20 dpa. The design temperatures were 300 {sup o}C and 500 {sup o}C. Precracked third-sized V-notch Charpy (3.3 x 3.3 x 25.4 mm) and 0.18 T DC(T) specimens were tested to determine transition and ductile shelf fracture toughness before and after irradiation. The master curve methodology was applied to evaluate the fracture toughness transition temperature, T {sub 0}. Irradiation induced shifts of T {sub 0} and reductions of J {sub Q} were compared with Charpy V-notch impact properties. Fracture toughness and Charpy shifts were also compared to hardening results.

  7. Superior Charpy impact properties of ODS ferritic steel irradiated in JOYO

    Science.gov (United States)

    Kuwabara, T.; Kurishita, H.; Ukai, S.; Narui, M.; Mizuta, S.; Yamazaki, M.; Kayano, H.

    1998-10-01

    The effect of neutron irradiation on Charpy impact properties of an ODS ferritic steel developed by PNC was studied. The miniaturized Charpy V-notch (MCVN) specimens (1.5 × 1.5 × 20 mm) of two orientations (longitudinal, called 1DS-L, and transverse, 1DS-T) were irradiated to fluence levels of (0.3-3.8) × 10 26 n/m 2 ( E n > 0.1 MeV) between 646 and 845 K in JOYO. MCVN specimens before and after the irradiation were subjected to instrumented Charpy impact tests. The test results and fracture surface observations showed that in the unirradiated state the steel showed no ductile-to-brittle transition behavior until 153 K regardless of orientation and the upper shelf energy of the steel was as high as that of a high-strength ferritic steel without dispersed oxide. Such excellent impact properties were essentially maintained after the irradiation although an appreciable decrease in absorbed energy occurred by higher temperature irradiations at and above 793 K.

  8. Weld investigations by 3D analyses of Charpy V-notch specimens

    DEFF Research Database (Denmark)

    Tvergaard, Viggo; Needleman, Allan

    2005-01-01

    The Charpy impact test is a standard procedure for determining the ductile-brittle transition in welds. The predictions of such tests have been investigated by full three dimensional transient analyses of Charpy V-notch specimens. The material response is characterised by an elastic-viscoplastic ......The Charpy impact test is a standard procedure for determining the ductile-brittle transition in welds. The predictions of such tests have been investigated by full three dimensional transient analyses of Charpy V-notch specimens. The material response is characterised by an elastic...... parameters in the weld material differ from those in the base material, and the heat a®ected zone (HAZ) tends to be more brittle than the other material regions. The effect of weld strength undermatch or overmatch is an important issue. Some specimens, for which the notched surface is rotated relative...

  9. Certification of NIST Room Temperature Low-Energy and High-Energy Charpy Verification Specimens

    OpenAIRE

    Lucon, Enrico; McCowan, Chris N.; Santoyo, Ray L.

    2015-01-01

    The possibility for NIST to certify Charpy reference specimens for testing at room temperature (21 °C ± 1 °C) instead of −40 °C was investigated by performing 130 room-temperature tests from five low-energy and four high-energy lots of steel on the three master Charpy machines located in Boulder, CO. The statistical analyses performed show that in most cases the variability of results (i.e., the experimental scatter) is reduced when testing at room temperature. For eight out of the nine lots ...

  10. An improved correlation procedure for subsize and full-size Charpy impact specimen data

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, M.A.; Alexander, D.J. [Oak Ridge National Lab., TN (United States)

    1997-03-01

    The possibility of using subsize specimens to monitor the properties of reactor pressure vessel steels is receiving increasing attention for light-water reactor plant life extension. This potential results from the possibility of cutting samples of small volume form the internal surface of the pressure vessel for determination of the actual properties of the operating pressure vessel. In addition, plant life extension will require supplemental data that cannot be provided by existing surveillance programs. Testing of subsize specimens manufactured from broken halves of previously tested surveillance Charpy specimens offers an attractive means of extending existing surveillance programs. Using subsize Charpy V-notch-type specimens requires the establishment of a specimen geometry that is adequate to obtain a ductile-to-brittle transition curve similar to that obtained from full-size specimens, and the development of correlations for transition temperature and upper-shelf energy (USE) level between subsize and full-size specimens. Five different geometries of subsize specimens were selected for testing and evaluation. The specimens were made from several types of pressure vessel steels with a wide range of yield strengths, transition temperatures, and USEs. The effects of specimen dimensions, including notch depth, angle, and radius, have been studied. The correlations of transition temperatures determined from different types of subsize specimens and the full-size specimens are presented. A new procedure for transforming data from subsize specimens is developed. The transformed data are in good agreement with data from full-size specimens for materials that have USE levels less than 200 J.

  11. Certification of NIST Room Temperature Low-Energy and High-Energy Charpy Verification Specimens.

    Science.gov (United States)

    Lucon, Enrico; McCowan, Chris N; Santoyo, Ray L

    2015-01-01

    The possibility for NIST to certify Charpy reference specimens for testing at room temperature (21 °C ± 1 °C) instead of -40 °C was investigated by performing 130 room-temperature tests from five low-energy and four high-energy lots of steel on the three master Charpy machines located in Boulder, CO. The statistical analyses performed show that in most cases the variability of results (i.e., the experimental scatter) is reduced when testing at room temperature. For eight out of the nine lots considered, the observed variability was lower at 21 °C than at -40 °C. The results of this study will allow NIST to satisfy requests for room-temperature Charpy verification specimens that have been received from customers for several years: testing at 21 °C removes from the verification process the operator's skill in transferring the specimen in a timely fashion from the cooling bath to the impact position, and puts the focus back on the machine performance. For NIST, it also reduces the time and cost for certifying new verification lots. For one of the low-energy lots tested with a C-shaped hammer, we experienced two specimens jamming, which yielded unusually high values of absorbed energy. For both specimens, the signs of jamming were clearly visible. For all the low-energy lots investigated, jamming is slightly more likely to occur at 21 °C than at -40 °C, since at room temperature low-energy samples tend to remain in the test area after impact rather than exiting in the opposite direction of the pendulum swing. In the evaluation of a verification set, any jammed specimen should be removed from the analyses.

  12. Fracture behaviors of neutron-irradiated ferritic steels studied by the instrumented charpy impact test

    Science.gov (United States)

    Yoshida, H.; Miyata, K.; Narui, M.; Kayano, H.

    1989-12-01

    The instrumented Charpy impact test for quarter-size specimens was developed and applied to study fracture behavior of ferritic steels and a ferritic-martensitic steel (JFMS) before and after neutron irradiation. The load-deflection curves obtained for U- and V-notched specimens showed typical characteristics of fracture properties of these steels. The temperature dependence of the fracture energy ( Ef) and the failure deflection ( Df) clearly indicates ductile-brittle transition and the DBTT can be determined from the Ef and Df versus temperature curves. The V-notched specimens showed sharper transition at higher temperatures for the JFMS than the U-notched ones, where the former were sensitive to brittle fracture and the latter well demonstrated the behavior of crack propagation. For the ferritic steels the DBTTs showed low values at compositions containing approximate 8-10% Cr and the increase of the DBTT (Δ DBTT) due to irradiation also showed a similar tendency. The Δ DBTT appeared to be relatively larger for the JFMS than the ferritic steels.

  13. Experimental study on variations in Charpy impact energies of low carbon steel, depending on welding and specimen cutting method

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Zhaorui; Kang, Hansaem; Lee, Young Seog [Chung-Ang University, Seoul (Korea, Republic of)

    2016-05-15

    This paper presents an experimental study that examines variations of Charpy impact energy of a welded steel plate, depending upon the welding method and the method for obtaining the Charpy specimens. Flux cored arc welding (FCAW) and Gas tungsten arc welding (GTAW) were employed to weld an SA516 Gr. 70 steel plate. The methods of wire cutting and water-jet cutting were adopted to take samples from the welded plate. The samples were machined according to the recommendations of ASTM SEC. II SA370, in order to fit the specimen dimension that the Charpy impact test requires. An X-ray diffraction (XRD) method was used to measure the as-weld residual stress and its redistribution after the samples were cut. The Charpy impact energy of specimens was considerably dependent on the cutting methods and locations in the welded plate where the specimens were taken. The specimens that were cut by water jet followed by FCAW have the greatest resistance-to-fracture (Charpy impact energy). Regardless of which welding method was used, redistributed transverse residual stress becomes compressive when the specimens are prepared using water-jet cutting. Meanwhile, redistributed transverse residual stress becomes tensile when the specimens are prepared using wire cutting.

  14. On the Use of the Master Curve based on the Precracked Charpy Specimen

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R.; Scibetta, M.; Van Walle, E.; Gerard, R

    1999-08-01

    Recently, worldwide interest has been demonstrated in the evaluation of the use of the Master Curve approach to characterize fracture toughness of ferritic steels in the transition regime. This was acknowledged by the recent release of the ASTM Standard Test Method for Determination of Reference Temperature, T{sub 0}, for Ferritic Steels in the Transition Range (E1921). The present work aims to investigate the use of the Charpy specimen along with the Master Curve approach to derive the fracture toughness behaviour of reactor pressure vessel steels. Therefore, four well characterized and documented reactor pressure vessel steels were selected. A large experimental program to measure fracture toughness with Charpy size specimens was carried out. Four important aspects were investigated: (1) the T0 determination as a function of test temperature; (2) the E1921 specimen size requirement (factor M=30); (3) the censoring procedure for specimens not satisfying the E1921 size requirements; (4) the estimation of the fracture toughness lower bound, and its comparison to the ASME KIC curve. It is found that within the experimental and statistical uncertainties, the reference temperature T0 is not affected by the test temperature, even when data are not valid according to E1921 requirements. By application of the censoring procedure, the determination of the reference temperature may lead to non conservative results. Comparison to larger specimen size suggests the use of M=60 rather than 30 to limit the loss of constraint, in agreement with finite element calculations. Nevertheless, the differences are not large enough to be statistically significant. The lower bound based on the Master Curve is very close to the experimental lower bound, while the ASME K{sub IC} curve trends to be over conservative. Replacing RT{sub NDT} by the new index, RT{sub To}, in the ASME KIC equation reduces this over conservatism.

  15. Identification of neutron irradiation induced strain rate sensitivity change using inverse FEM analysis of Charpy test

    Science.gov (United States)

    Haušild, Petr; Materna, Aleš; Kytka, Miloš

    2015-04-01

    A simple methodology how to obtain additional information about the mechanical behaviour of neutron-irradiated WWER 440 reactor pressure vessel steel was developed. Using inverse identification, the instrumented Charpy test data records were compared with the finite element computations in order to estimate the strain rate sensitivity of 15Ch2MFA steel irradiated with different neutron fluences. The results are interpreted in terms of activation volume change.

  16. Irradiation programme MANITU: Results of pre-examinations and Charpy tests with unirradiated materials; Bestrahlungsprogramm MANITU. Ergebnisse der Voruntersuchungen und der Kerbschlagbiegeversuche mit den unbestrahlten Werkstoffen

    Energy Technology Data Exchange (ETDEWEB)

    Rieth, M.; Dafferner, B.; Ries, H.; Romer, O.

    1995-04-01

    The irradiation project MANITU was planned in the frame of the European Long-term Fusion Materials Development Programme. The results of MANITU will have a lasting influence on the future actions within the materials development programme. The problem of the irradiation induced embrittlement of possible martensitic alloy candidates is still unsolved. But after the evaluation of sub-size Charpy tests with the unirradiated refrence specimens of MANITU a first tendency is recognizable. The Charpy properties of the newly developed low activation 7-10% Cr-WVTa alloys are clearly better compared with the modified commerical 10-11% Cr-NiMoVNb steels. In the present report the pre-examinations are documented and the Charpy test results with unirradiated reference specimens are analysed and assessed. (orig.) [Deutsch] Das Bestrahlungsprojekt MANITU wurde im Rahmen des europaeischen Langzeitprogramms fuer Materialentwicklung fuer die Kernfusion geplant. Die daraus gewonnenen Ergebnisse werden das weitere Vorgehen bei der Materialentwicklung entscheidend beeinflussen. Das Problem der bestrahlungsinduzierten Versproedung bei den in Frage kommenden martensitischen Werkstoffen ist nach wie vor ungeloest. Eine erste Tendenz zeichnet sich nach der Auswertung der Kerbschlagbiegeversuche an den unbestrahlten miniaturisierten Referenzproben des MANITU-Programms ab. Die neu entwickelten niedrig aktivierbaren 7-10% Cr-WVTa-Legierungen weisen gegenueber den modifizierten kommerziellen 10-11% Cr-NiMoVNb-Staehlen deutlich bessere Kerbschlageigenschaften auf. Im vorliegenden Bericht werden die Voruntersuchungen dokumentiert und die Ergebnisse aus den Kerbschlagbiegeversuchen der unbestrahlten Referenzproben analysiert und bewertet. (orig.)

  17. Test Technique Development on the Irradiated Reconstituted PCVN Specimen in Hot Cell

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sangbok; Oh, Wanho; Choo, Yongsun; Kim, Minchul; Lee, Bongsang [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    The degradation of fracture toughness is the important factor to restrict the life of nuclear pressure vessel in PWR reactors. A pressure vessel is operated in conformity with the fracture analysis based on ASME codes to ensure safety margins from the unstable fracture. A fracture analysis is performed based on the result from the Charpy impact tests in PWR reactor, but it has the questions to be exact solutions because the test results give indirect and excessively conservative values. Therefore the research to find an exact toughness parameter is undergoing to use the pre-cracked Charpy v-notch (PCVN). As results the master curve method is proposed in ASTM E1921 to be supposed an appropriate tool to evaluate the fracture toughness for the irradiated, or the operated pressure vessel materials. The surveillance test program to evaluate toughness degradation on existing commercial PWR reactor is performed through the impact test on Charpy specimens. It gives the lack of the specimen to evaluate the safety in toughness for on-going operation beyond design life. To overcome the shortage of specimen, the test method to use a reconstituted PCVN specimen fabricated from the broken half of Charpy specimen is proposed and adopted in foreign reactors. In this paper techniques developed for the reconstituted specimen from the domestic commercial PWR reactor in hot cell are described.

  18. Dynamic Toughness Testing of Pre-Cracked Charpy V-Notch Specimens. Convention ELECTRABEL - SCK-CEN

    Energy Technology Data Exchange (ETDEWEB)

    Lucon, E

    1999-04-01

    This document describes the experimental and analytical procedures which have been adopted at the laboratories of the Belgian Nuclear Research Centre SCK-CEN for performing dynamic toughness tests on pre-cracked Charpy-V specimens. Such procedures were chosen on the basis of the existing literature on the subject, with several updates in the data analysis stages which reflect more recent developments in fracture toughness testing. Qualification tests have been carried out on PCCv specimens of JRQ steel, in order to assess the reliability of the results obtained; straightforward comparisons with reference data have been performed, as well as more advanced analyses using the Master Curve approach. Aspects related to machine compliance and dynamic tup calibration have also been addressed.

  19. Miniature Precracked Charpy Specimens for Measuring the Master Curve Reference Temperature of RPV Steels at Impact Loading Rates

    Energy Technology Data Exchange (ETDEWEB)

    Lucon, E.; Scibetta, M.; Puzzolante, L.

    2008-10-15

    In the framework of the 2006 Convention, we investigated the applicability of fatigue precracked miniature Charpy specimens of KLST type (MPCC - B = 3 mm, W = 4 mm and L = 27 mm) for impact toughness measurements, using the well-characterized JRQ RPV steel. In the ductile to-brittle transition region, MPCC tests analyzed using the Master Curve approach and compared to data previously obtained from PCC specimens had shown a more ductile behavior and therefore un conservative results. In the investigation presented in this report, two additional RPV steels have been used to compare the performance of impact-tested MPCC and PCC specimens in the transition regime: the low-toughness JSPS steel and the high-toughness 20MnMoNi55 steel. The results obtained (excellent agreement for 20MnMoNi55 and considerable differences between T0 values for JSPS) are contradictory and do not presently allow qualifying the MPCC specimens as a reliable alternative to PCC samples for impact toughness measurements.

  20. 3D analyses of the effect of weld orientation in Charpy specimens

    DEFF Research Database (Denmark)

    Tvergaard, Viggo; Needleman, A.

    2004-01-01

    of failure in the weld material, base material and heat affected zone (HAZ). For these rotated specimens the location where the notch crosses the thin layer of HAZ, i.e. whether this location is near the center of the specimen or near the free specimen edge, makes a large difference in the response. (C) 2004...

  1. Charpy impact test results for low activation ferritic alloys irradiated to 30 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S. [Pacific Northwest National Laboratory, Richland, WA (United States)

    1996-04-01

    Miniature specimens of six low activation ferritic alloys have been impact field tested following irradiation at 370{degrees}C to 30 dpa. Comparison of the results with those of control specimens and specimens irradiated to 10 dpa indicates that degradation in the impact behavior appears to have saturated by {approx}10 dpa in at least four of these alloys. The 7.5Cr-2W alloy referred to as GA3X appears most promising for further consideration as a candidate structural material in fusion reactor applications, although the 9Cr-1V alloy may also warrant further investigation.

  2. Experimental study on the material dynamic fracture properties by Instrumented Charpy Impact test with single specimen method

    Science.gov (United States)

    Jian, F.; Fulian, D.; Chengzhong, W.

    2003-09-01

    With the determination of load-time curve recorded by Amsler/Roell RKP 450 Instrumented Charpy Impact test and based on the Newton's Second Law, Impact character of a single standard V-notch specimen of X70 pipeline steel under the low temperature -70 ^{circ}C was investigated by studying the impact energy distribution. It was revealed that maximum load point (Fm point) was not exact the dynamic crack initiation, which was detected somewhere prior and very close to Fm point by using Compliance Changing Rate method. This fact was also confirmed by Dynamic CTOD method. That is to say, Impact energy related to the Fm point (i.e. Em) consists not only the crack initiation energy Ei, but a small part of crack extension energy as well. Ratio of Ei/Em was found to be 0.90 just applicable to the material used here. Dynamic fracture toughness JJd was then estimated by modified Rice equation. Crack extension behavior and dynamic crack growth resistance curve (J-Δa) during stable crack propagation period was carefully analyzed by Key Curve method. Finally, methods for evaluating tearing module Tmat, and CTOD curve under the impact test were also briefly introduced in the paper.

  3. Dynamic finite element analysis of third size charpy specimens of V-4Cr-4Ti

    Energy Technology Data Exchange (ETDEWEB)

    Lansberry, M.R.; Kumar, A.S.; Mueller, G.E. [Univ. of Missouri, Rolla, MO (United States); Kurtz, R.J. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-04-01

    A 2-D finite element analysis was performed on precracked, one third scale CVN specimens to investigate the sensitivity of model results to key material parameters such as yield strength, failure strain and work hardening characteristics. Calculations were carried out at temperatures of -196{degree}C and 50{degree}C. The dynamic finite element analyses were conducted using ABAQUS/Explicit V5.4. The finite element results were compared to experimental results for the production-scale heat of V-4Cr-4Ti (ANL Heat No. 832665) as a benchmark. Agreement between the finite element model and experimental data was very good at -196{degree}C, whereas at 50{degree}C the model predicted a slightly lower absorbed energy than actually measured.

  4. Effect of specimen size on the impact properties of neutron irradiated A533B steel

    Energy Technology Data Exchange (ETDEWEB)

    Schubert, L.E. [Missouri Univ., Rolla, MO (United States). Dept. of Nuclear Engineering; Kumar, A.S. [Missouri Univ., Rolla, MO (United States). Dept. of Nuclear Engineering; Rosinski, S.T. [Sandia National Laboratories, MS-0741, Albuquerque, NM 87185 (United States); Hamilton, M.L. [Pacific Northwest Laboratory, P.O. Box 999, Richland, WA 99352 (United States)

    1995-08-01

    A new methodology is proposed to correlate the upper shelf energy (USE) of full-size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material. The methodology appears to be more satisfactory than those methodologies proposed earlier. The USE was normalized by a normalization factor involving the dimensions of the Charpy specimen, the elastic stress concentration factor, and the plastic constraint at the notch root. The normalized values of the USE were found to be invariant with specimen size. In addition, it was also found that the ratio of the USE of unirradiated to that of irradiated materials was approximately the same for full-, half-, and third-size specimens. The ductile-to-brittle transition temperature (DBTT) increased due to irradiation at 150 C to a nominal fluence of 1.0x10{sup 19} n/cm{sup 2} (E>1 MeV) by 78, 83, and 70 C for full-, half-, and third-size specimens, respectively. These shifts in DBTT appeared to be independent of specimen size and notch geometry. (orig.).

  5. A study of the fracture process and factors that control toughness variability in Charpy V-notch specimens

    Science.gov (United States)

    Bouchard, Real

    La presente etude a ete initiee pour developper une comprehension quantitative du processus de rupture avec les facteurs qui controlent la dispersion des mesures de tenacite lorsque des eprouvettes Charpy entaillees en V sont utilisees. Un grand nombre d'essais ont ete realises pour un acier C-Mn: eprouvettes Charpy testees sous impact, eprouvettes Charpy testees en flexion lente, eprouvettes axisymetriques entaillees et sollicitees en traction et essais de tenacite sur eprouvettes prefissurees. Base sur le concept de la statistique de Weibull, l'approche locale developpee par le groupe Beremin a ete utilisee pour decrire la probabilite de rupture par clivage en fonction de la contrainte appliquee aussi bien qu'en fonction de l'energie Charpy obtenue. Le calcul par elements finis a ete realise pour determiner la distribution de la deformation et des contraintes en pointe d'entaille et de fissure. La nouvelle approche introduite decrit bien les resultats experimentaux. Les points d'initiation du clivage ont ete identifies au MEB et par la suite, avec la technique de faisceau d'ions focalise, sectionnes, polis et examines. L'examen de la microstructure sous le point d'initiation revele clairement que le clivage s'initie par un mecanisme d'empilement de dislocations ou les dislocations sont arretees aux joints de grain, aux interfaces de perlite/ferrite ou de perlite qui agissent comme barrieres physiques.

  6. Applicability of the Modified Ritchie-Knott-Rice Failure Criterion to Examine the Feasibility of Miniaturized Charpy Type SE(B Specimens

    Directory of Open Access Journals (Sweden)

    Toshiyuki Meshii

    2016-01-01

    Full Text Available This paper examined whether the modified Ritchie-Knott-Rice (RKR failure criterion can be applied to examine the feasibility of miniaturized Charpy type SE(B specimens of thickness-to-width ratio B/W=1. The modified RKR failure criterion considered in this paper is the (4δt,σ22c criterion which predicts the onset of cleavage fracture when the midplane crack-opening stress measured at a distance equal to four times the crack-tip opening displacement, denoted as σ22d, exceeds a critical stress σ22c. Specimens with B values of 25, 10, 3, and 2 mm (denoted as 25t, 10t, 3t, and 2t specimens, resp. manufactured with 0.55% carbon steel were tested at 20°C. The results showed that the modified RKR criterion could appropriately predict the occurrence of cleavage fracture accompanied by negligibly small stable crack extension (denoted as KJc fracture naturally for the 25t and 10t specimens. The modified RKR criterion could also predict that KJc fracture does not occur for the 2t specimen. The σ22c obtained from specimens for the 25t and 10t specimens exhibited only a small difference, indicating that the Jc obtained from the 10t specimens can be used to predict the Jc that will be obtained with the 25t specimens.

  7. Effects of annealing time on the recovery of Charpy V-notch properties of irradiated high-copper weld metal

    Energy Technology Data Exchange (ETDEWEB)

    Iskander, S.K.; Sokolov, M.A.; Nanstad, R.K. [Oak Ridge National Lab., TN (United States). Metals and Ceramics Div.

    1994-12-31

    One of the options to mitigate the effects of irradiation on reactor pressure vessels is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. An important issue to be resolved is the effect on the toughness properties of reirradiating a vessel that has been annealed. This paper describes the annealing response of irradiated high-copper submerged-arc weld HSSI 73W. For this study, the weld has been annealed at 454 C (850 F) for lengths of time varying between 1 and 14 days. The Charpy V-notch 41-J (30-ft-lb) transition temperature (TT{sub 41J}) almost fully recovered for the longest period studied, but recovered to a lesser degree for the shorter periods. No significant recovery of the TT{sub 41J} was observed for a 7-day anneal at 343 C (650 F). At 454 C for the durations studied, the values of the upper-shelf impact energy of irradiated and annealed weld metal exceeded the values in the unirradiated condition. Similar behavior was observed after aging the unirradiated weld metal at 460 and 490 C for 1 week.

  8. Further Charpy impact test results of low activation ferritic alloys, irradiated at 430{degrees}C to 67 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-04-01

    Miniature CVN specimens of four ferritic alloys, GA3X, F82H, GA4X and HT9, have been impact tested following irradiation at 430{degrees}C to 67 dpa. Comparison of the results with those of the previously tested lower dose irradiation condition indicates that the GA3X and F82H alloys, two primary candidate low activation alloys, exhibit virtually identical behavior following irradiation at 430{degrees}C to {approximately}67 dpa and at 370{degrees}C to {approximately}15 dpa. Very little shift is observed in either DBTT or USE relative to the unirradiated condition. The shifts in DBTT and USE observed in both GA4X and HT9 were smaller after irradiation at 430{degrees}C to {approximately}67 dpa than after irradiation at 370{degrees}C to {approximately}15 dpa.

  9. Charpy impact test results of four low activation ferritic alloys irradiated at 370{degrees}C to 15 DPA

    Energy Technology Data Exchange (ETDEWEB)

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-10-01

    Miniature CVN specimens of four low activation ferritic alloys have been impact tested following irradiation at 370{degrees}C to 15 dpa. Comparison of the results with those of control specimens indicates that degradation in the impact behavior occurs in each of these four alloys. The 9Cr-2W alloy referred to as GA3X and the similar alloy F82H with 7.8Cr-2W appear most promising for further consideration as candidate structural materials in fusion energy system applications. These two alloys exhibit a small DBTT shift to higher temperatures but show increased absorbed energy on the upper shelf.

  10. The feasibility of small size specimens for testing of environmentally assisted cracking of irradiated materials and of materials under irradiation in reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Toivonen, A.; Moilanen, P.; Pyykkoenen, M.; Taehtinen, S.; Rintamaa, R.; Saario, T. [Valtion Teknillinen Tutkimuskeskus, Espoo (Finland)

    1998-11-01

    Environmentally assisted cracking (EAC) of core materials has become an increasingly important issue of downtime and maintenance costs in nuclear power plants. Small size specimens are necessary in stress corrosion testing of irradiated materials because of difficulties in handling high dose rate materials and because of restricted availability of the materials. The drawback of using small size specimens is that in some cases they do not fulfil the requirements of the relevant testing standards. Recently VTT has developed J-R testing with irradiated and non-irradiated sub size 3 PB specimens, both in inert and in LWR environments. Also, a new materials testing system which will enable simultaneous multiple specimen testing both in laboratory conditions and in operating reactor core is under development. The new testing system will utilize Charpy and sub size 3 PB specimens. The feasibility study of the system has been carried out using different materials. Fracture resistance curves of a Cu-Zr-Cr alloy are shown to be independent of the specimen geometry and size, to some extent. Results gained from tests in simulated boiling water reactor (BWR) water are presented for sensitized SIS 2333 stainless steel. The experimental results indicate that the size of the plastic zone or stress triaxiality must be further studied although no significant effect on the environmentally assisted crack growth rate was observed. (orig.)

  11. AGC-2 Specimen Post Irradiation Data Package Report

    Energy Technology Data Exchange (ETDEWEB)

    Windes, William Enoch [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rohrbaugh, David T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cottle, David L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    This report documents results of the post-irradiation examination material property testing of the creep, control, and piggyback specimens from the irradiation creep capsule Advanced Graphite Creep (AGC)-2 are reported. This is the second of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C and to a peak dose of 5 dpa (displacements per atom). One-half of the creep specimens were subjected to mechanical stresses (an applied stress of either 13.8, 17.2, or 20.7 MPa) to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. Material property tests were conducted on specimens from 15 nuclear graphite grades using a similar loading configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114. The data reported include specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and volume data necessary to derive density, elastic constants (Young’s modulus, shear modulus, and Poisson’s ratio) from ultrasonic time-of-flight velocity measurements, Young’s modulus from the fundamental frequency of vibration, electrical resistivity, and thermal diffusivity and thermal expansion data from 100–500°C. No data outliers were determined after all measurements were completed. A brief statistical analysis was performed on the irradiated data and a limited comparison between

  12. Application of subsize specimens in nuclear plant life extension

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [Sandia National Labs., Albuquerque, NM (United States); Kumar, A.S. [Missouri Univ., Rolla, MO (United States); Cannon, S.C. [Westinghouse Hanford Co., Richland, WA (United States); Hamilton, M.L. [Pacific Northwest Lab., Richland, WA (United States)

    1991-12-31

    The US Department of Energy is sponsoring a research effort through Sandia National Laboratories and the University of Missour-Rolla to test a correlation for the upper shelf energy (USE) values obtained from the impact testing of subsize Charpy V-notch specimens to those obtained from the testing of full size samples. The program involves the impact testing of unirradiated and irradiated full, half, and third size Charpy V-notch specimens. To verify the applicability of the correlation on LWR materials unirradiated and irradiated full, half, and third size Charpy V-notch specimens of a commercial pressure vessel steel (ASTM A533 Grade B) will be tested. This paper will provide details of the program and present results obtained from the application of the developed correlation methodology to the impact testing of the unirradiated full, half, and third size A533 Grade B Charpy V-notch specimens.

  13. Application of subsize specimens in nuclear plant life extension

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. (Sandia National Labs., Albuquerque, NM (United States)); Kumar, A.S. (Missouri Univ., Rolla, MO (United States)); Cannon, S.C. (Westinghouse Hanford Co., Richland, WA (United States)); Hamilton, M.L. (Pacific Northwest Lab., Richland, WA (United States))

    1991-01-01

    The US Department of Energy is sponsoring a research effort through Sandia National Laboratories and the University of Missour-Rolla to test a correlation for the upper shelf energy (USE) values obtained from the impact testing of subsize Charpy V-notch specimens to those obtained from the testing of full size samples. The program involves the impact testing of unirradiated and irradiated full, half, and third size Charpy V-notch specimens. To verify the applicability of the correlation on LWR materials unirradiated and irradiated full, half, and third size Charpy V-notch specimens of a commercial pressure vessel steel (ASTM A533 Grade B) will be tested. This paper will provide details of the program and present results obtained from the application of the developed correlation methodology to the impact testing of the unirradiated full, half, and third size A533 Grade B Charpy V-notch specimens.

  14. Method for Estimating Percent Shear Fracture Appearance of Charpy-V Specimen of Nuclear Pressure Vessel Steel%核压力容器钢冲击断口剪切面积百分比的估算方法

    Institute of Scientific and Technical Information of China (English)

    伍晓勇; 冯明全; 崔永海

    2001-01-01

    Percent shear fracture appearance is an important factor to evaluate neutron irradiation embrittlement of nuclear pressure vessel steel.But it is difficult to measure the irregular surface directly,especially for post-irradiation speciments.Depending on Charpy-V instrumented impact tests,characteristic load values can be determined according to load-deflection curve which completely represents the impact failure processes,and percent shear fracture appearance can be estimated using different formula.The estimating method is more effective to calculate irradiation-induced changes in transition temperature,which already successfully applied to the irradiation surveillance test of reactor pressure vessel steel of Daya-bay nuclear power station.%在核压力容器钢的中子辐照脆化评价中,断口剪切面积百分比是一个重要的参数。但此参数不易直接测量,对于辐照后的放射性试样其测量更加困难。本文采用 Charpy-V示波冲击试验,并根据计算机采集得到的完整记录冲击过程的载荷 -位移曲线,即可确定相应的载荷特征值,同时估算出断口剪切面积百分比。该估算方法用于计算核压力容器钢因中子辐照引起的脆性转变温度的变化值,其计算结果较为准确且计算方法也简便,现已成功地应用于大亚湾核电站压力容器的辐照监督试验。

  15. Small Specimen Data from a High Temperature HFIR Irradiation Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Burchell, Timothy D [ORNL; McDuffee, Joel Lee [ORNL; Thoms, Kenneth R [ORNL

    2014-01-01

    The HTV capsule is a High Flux Isotope Reactor (HFIR) target-rod capsule designed to operate at very high temperatures. The graphite containing section of the capsule (in core) is approximately 18 inches (457.2 mm) long and is separated into eight temperature zones. The specimen diameters within each zone are set to achieve the desired gas gap and hence design temperature (900 C, 1200 C or 1500 C). The capsule has five zones containing 0.400 inch (10.16 mm) diameter specimens, two zones containing 0.350 inch (8.89 mm) diameter specimens and one zone containing 0.300 inch (7.62 mm) diameter specimens. The zones have been distributed within the experiment to optimize the gamma heating from the HFIR core as well as minimize the axial heat flow in the capsule. Consequently, there are two 900 C zones, three 1200 C zones, and three 1500 C zones within the HTV capsule. Each zone contains nine specimens 0.210 0.002 inches (5.334 mm) in length. The capsule will be irradiated to a peak dose of 3.17 displacements per atom. The HTV specimens include samples of the following graphite grades: SGL Carbon s NBG-17 and NBG-18, GrafTech s PCEA, Toyo Tanso s IG-110, Mersen s 2114 and the reference grade H-451 (SGL Carbon). As part of the pre-irradiation program the specimens were characterized using ASTM Standards C559 for bulk density, and ASTM C769 for approximate Young s modulus from the sonic velocity. The probe frequency used for the determination of time of flight of the ultrasonic signal was 2.25 MHz. Marked volume (specimen diameter) effects were noted for both bulk density (increased with increasing specimen volume or diameter) and Dynamic Young s modulus (decreased with increasing specimen volume or diameter). These trends are extended by adding the property vs. diameter data for unirradiated AGC-1 creep specimens (nominally 12.5 mm-diameter x 25.4 mm-length). The relatively large reduction in Dynamic Young s Modulus was surprising given the trend for increasing density

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-11-08

    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.

  17. Instrumented charpy impact tests of austenitic and ferritic steels

    Science.gov (United States)

    Yoshida, H.; Miyata, K.; Hayashi, Y.; Narui, M.; Kayano, H.

    1985-08-01

    The instrumented Charpy impact test was applied to commercial Mn-steel and ferritic steels before and after JMTR irradiation ( 6.5 × 10 22 n/m 2). The load-deflection curves show typical characteristics of the fracture properties of the specimens; i.e. linear elastic behaviour for the brittle fracture and elastic-plastic behaviour for the ductile fracture. The fracture deflection and the absorption energy (fracture energy) dropped rapidly at the temperature of ductile to brittle transition. The ductile-brittle transition temperatures (DBTTs) showed shifts of about 30 and 40 K due to the irradiation for 9Cr-1Mo and 9Cr-2Mo steels, respectively. In Mn-steel the transition from ductile to brittle did not appear at temperatures higher than 77 K. The lateral expansions measured from the scanning electron micrographs show good correspondence to the above results.

  18. Charpy V, an application in Mat lab; Charpy V, una aplicacion en Matlab

    Energy Technology Data Exchange (ETDEWEB)

    Castillo M, J.A.; Torres V, M. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    The obtained results with the system Charpy V{sub V}1 designed in Mat lab for the estimate of parameters of three mathematical models are shown. The adjustment of data is used to determine the fracture energy, the lateral expansion and the percentage of ductility of steels coming from the reactor vessels of Laguna Verde, Veracruz. The data come from test tubes type Charpy V of irradiated material and not irradiated. To verify our results they were compared with those obtained by General Electric of data coming from the Laguna Verde nuclear power plant. (Author)

  19. Correlation between standard Charpy and sub-size Charpy test results of selected steels in upper shelf region

    Science.gov (United States)

    Konopík, P.; Džugan, J.; Bucki, T.; Rzepa, S.; Rund, M.; Procházka, R.

    2017-02-01

    Absorbed energy obtained from impact Charpy tests is one of the most important values in many applications, for example in residual lifetime assessment of components in service. Minimal absorbed energy is often the value crucial for extending components service life, e.g. turbines, boilers and steam lines. Using a portable electric discharge sampling equipment (EDSE), it is possible to sample experimental material non-destructively and subsequently produce mini-Charpy specimens. This paper presents a new approach in correlation from sub-size to standard Charpy test results.

  20. Disassembly of irradiated lithium-bonded capsules containing vanadium alloy specimens

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Strain, R.V.

    1996-04-01

    Capsules containing vanadium alloy specimens from irradiation experiments in FFTF and EBR-II are being processed to remove the lithium bond and retrieve the specimens for testing. The work has progressed smoothly.

  1. Recent Accomplishments in the Irradiation Testing of Engineering-Scale Monolithic Fuel Specimens

    Energy Technology Data Exchange (ETDEWEB)

    N.E. Woolstenhulme; D.M. Wachs; M.K. Meyer; H.W. Glunz; R.B. Nielson

    2012-10-01

    The US fuel development team is focused on qualification and demonstration of the uranium-molybdenum monolithic fuel including irradiation testing of engineering-scale specimens. The team has recently accomplished the successful irradiation of the first monolithic multi-plate fuel element assembly within the AFIP-7 campaign. The AFIP-6 MKII campaign, while somewhat truncated by hardware challenges, exhibited successful irradiation of a large-scale monolithic specimen under extreme irradiation conditions. The channel gap and ultrasonic data are presented for AFIP-7 and AFIP-6 MKII, respectively. Finally, design concepts are summarized for future irradiations such as the base fuel demonstration and design demonstration experiment campaigns.

  2. Multifunctional TEM-specimen holder equipped with a piezodriving probe and an electron irradiation port.

    Science.gov (United States)

    Shindo, Daisuke; Suzuki, Satoshi; Sato, Kuniaki; Akase, Zentaro; Murakami, Yasukazu; Yamazaki, Kazuya; Ikeda, Yuuta; Fukuda, Tomohisa

    2013-08-01

    The charging effect due to electron irradiation in an electron microscope has been studied so far with incident electrons. Here we report on a new specimen holder to control the charging effect by using electrons emitted from an irradiation port in the holder while maintaining a constant intensity of the incident electron beam. Details of the charging effect, such as electric field variation, are expected to be investigated by electron holography. The new specimen holder was developed by modifying a double-probe piezodriving specimen holder to introduce an electron irradiation port in one of its two arms. As a result, the new modified specimen holder consists of a piezodriving probe and an electron irradiation port, both of which can be controlled in three dimensions, using piezoelectric elements and micrometers. We demonstrate that variations in the charging effect for epoxy resin and surface contamination can be observed by electron holography.

  3. Miniaturized Charpy test for reactor pressure vessel embrittlement characterization

    Energy Technology Data Exchange (ETDEWEB)

    Manahan, M.P. Sr. [MPM Research and Consulting, Lemont, PA (United States)

    1999-10-01

    Modifications were made to a conventional Charpy machine to accommodate the miniaturized Charpy V-Notch (MCVN) specimens which were fabricated from an archived reactor pressure vessel (RPV) steel. Over 100 dynamic MCVN tests were performed and compared to the results from conventional Charpy V-Notch (CVN) tests to demonstrate the efficacy of the miniature specimen test. The optimized sidegrooved MCVN specimens exhibit transitional fracture behavior over essentially the same temperature range as the CVN specimens which indicates that the stress fields in the MCVN specimens reasonably simulate those of the CVN specimens and this fact has been observed in finite element calculations. This result demonstrates a significant breakthrough since it is now possible to measure the ductile-brittle transition temperature (DBTT) using miniature specimens with only small correction factors, and for some materials as in the present study, without the need for any correction factor at all. This development simplifies data interpretation and will facilitate future regulatory acceptance. The non-sidegrooved specimens yield energy-temperature data which is significantly shifted downward in temperature (non-conservative) as a result of the loss of constraint which accompanies size reduction.

  4. Shear Punch Testing of BOR-60 Irradiated TEM Specimens

    Energy Technology Data Exchange (ETDEWEB)

    Saleh, Tarik A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Quintana, Matthew Estevan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Romero, Tobias J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-13

    As a part of the project “High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation” an Integrated Research Program (IRP) project from the U.S. Department of Energy, Nuclear Energy University Programs (NEUP), TEM geometry samples of ferritic cladding alloys, Ni based super alloys and model alloys were irradiated in the BOR-60 reactor to ~16 dpa at ~370°C and ~400°C. Samples were sent to Los Alamos National Laboratory and subjected to shear punch testing. This report presents the results from this testing.

  5. Influence of irradiation on the ductile fracture of a reactor pressure vessel steel

    Science.gov (United States)

    Haušild, Petr; Kytka, Miloš; Karlík, Miroslav; Pešek, Pavel

    2005-05-01

    The mechanical properties of 15Ch2MFA steel were characterised by tensile and instrumented Charpy tests. The fracture surfaces of Charpy specimens broken in the ductile-to-brittle transition temperature range contain a certain proportion of ductile fracture correlated to fracture energy. Measured ductile crack lengths show the same dependence on fracture deflection and/or fracture energy for irradiated and non-irradiated specimens. The decrease of upper shelf energy with increasing neutron fluence could be explained by an increasing amount of shear fracture.

  6. Specimen size effect considerations for irradiation studies of SiC/SiC

    Energy Technology Data Exchange (ETDEWEB)

    Youngblood, G.E.; Henager, C.H. Jr.; Jones, R.H. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-10-01

    For characterization of the irradiation performance of SiC/SiC, limited available irradiation volume generally dictates that tests be conducted on a small number of relatively small specimens. Flexure testing of two groups of bars with different sizes cut from the same SiC/SiC plate suggested the following lower limits for flexure specimen number and size: Six samples at a minimum for each condition and a minimum bar size of 30 x 6.0 x 2.0 mm{sup 3}.

  7. Transferability of Charpy Absorbed Energy to Fracture Toughness Based on Weibull Stress Criterion

    Institute of Scientific and Technical Information of China (English)

    Hongyang JING; Lianyong XU; Lixing HUO; Fumiyoshi Minami

    2005-01-01

    The relationship between Charpy absorbed energy and the fracture toughness by means of the (crack tip opening displacement (CTOD)) method was analyzed based on the Weibull stress criterion. The Charpy absorbed energy and the fracture toughness were measured for the SN490B steel under the ductile-brittle transition temperature region. For the instrumented Charpy impact test, the curves between the loading point displacement and the load against time were recorded. The critical Weibull stress was taken as a fracture controlled parameter, and it could not be affected by the specimen configuration and the loading pattern based on the local approach. The parameters controlled brittle fracture are obtained from the Charpy absorbed energy results, then the fracture toughness for the compact tension (CT) specimen is predicted. It is found that the results predicted are in good agreement with the experimental. The fracture toughness could be evaluated by the Charpy absorbed energy, because the local approach gives a good description for the brittle fracture even though the Charpy impact specimen or the CT specimen is used for the given material.

  8. Effect of neutron irradiation on fracture toughness of metal matrix composites

    Science.gov (United States)

    Sato, Shinji; Hamada, Kenichi; Kohyama, Akira

    1992-09-01

    Based on the recent improvement in mechanical properties of unidirectionally reinforced metal matrix composites (MMCs), SiC/Al and C/Al, impact property change due to neutron irradiation has been investigated. This paper details effects of neutron irradiation on fracture toughness of the MMCs. Materials used were formed sheets of SiC/Al and C/Al. Miniaturized Charpy V-notched specimens were tested by an instrumented Charpy impact tester. Neutron irradiation was performed in JMTR(LWR) at Oarai. The Charpy value was increased with increasing test temperature and with neutron irradiation. SiC/Al was rather more neutron fluence insensitive than C/Al and the insensitivity was correlated to differences in interfacial structure between the two systems.

  9. Influence of specimen size/type on the fracture toughness of five irradiated RPV materials

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Mikhail A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lucon, Enrico [National Inst. of Standards and Technology (NIST), Boulder, CO (United States)

    2015-01-01

    The Heavy-Section Steel Irradiation (HSSI) Program had previously irradiated five reactor pressure vessel (RPV) steels/welds at fast neutron fluxes of about 4 to 8 x 1011 n/cm2/s (>1 MeV) to fluences from 0.5 to 3.4 1019 n/cm2 and at 288 °C. The unirradiated fracture toughness tests were performed by Oak Ridge National Laboratory with 12.7-mm and 25.4-mm thick (0.5T and 1T) compact specimens, while the HSSI Program provided tensile and 5 x 10-mm three-point bend specimens to SCK-CEN for irradiation in the in-pile section of the Belgian Reactor BR2 at fluxes > 1013 n/cm2/s and subsequent testing by SCK-CEN. The BR2 irradiations were conducted at about 2 and 4 x 1013 n/cm2/s with irradiation temperature between 295 °C and 300 °C (water temperature), and to fluences between 6 and 10 x 1019n/cm2. The irradiation-induced shifts of the Master Curve reference temperatures, ΔT0, for most of the materials deviated from the embrittlement correlations much more than expected, motivating the testing of 5 x 10-mm three-point bend specimens of all five materials in the unirradiated condition to eliminate specimen size and geometry as a variable. Tests of the unirradiated small bend specimens resulted in Master Curve reference temperatures, T0, 25 °C to 53 °C lower than those from the larger compact specimens, meaning that the irradiation-induced reference temperature shifts, ΔT0, were larger than the initial measurements, resulting in much improved agreement between the measured and predicted fracture toughness shifts.

  10. Investigation of some problems in developing standards for precracked Charpy slow bend tests

    Science.gov (United States)

    Succop, G.; Bubsey, R. T.; Jones, M. H.; Brown, W. F., Jr.

    1977-01-01

    The reported investigation was undertaken in connection with an attempt to develop procedures which would be useful in standardizing a test method for the precracked Charpy slow bend specimen. A number of alloys was studied for which valid plane-strain fracture toughness values have been established. The investigation shows that useful relations between precracked Charpy slow bend results and crack size factors can be obtained under some circumstances. However, it is not yet known what factors control these circumstances.

  11. Embrittlement of irradiated ferritic/martensitic steels in the absence of irradiation hardening

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L. [Oak Ridge Noational Laboratory, TN (United States); Shiba, K. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Sokolov, M. [Oak Ridge National Laboratory, Materials Science and Technology Div., TN (United States)

    2007-07-01

    Full text of publication follows: Neutron irradiation of 9-12% Cr ferritic/martensitic steels below 425-450 deg. C produces microstructural defects that cause an increase in yield stress and ultimate tensile strength. This irradiation hardening causes embrittlement, which is observed in Charpy impact and toughness tests as an increase in ductile-brittle transition temperature (DBTT). Based on observations that show little change in strength in these steels irradiated above 425-450 deg. C, the general conclusion has been that no embrittlement occurs above this irradiation-hardening temperature regime. In a recent study of F82H steel irradiated at 300, 380, and 500 deg. C, irradiation hardening-an increase in yield stress-was observed in tensile specimens irradiated at the two lower temperatures, but no change was observed for the specimens irradiated at 500 deg. C. As expected, an increase in DBTT occurred for the Charpy specimens irradiated at 300 and 380 deg. C. However, there was an unexpected increase in the DBTT of the specimens irradiated at 500 deg. C. The observed embrittlement was attributed to the irradiation-accelerated precipitation of Laves phase. This conclusion was based on results from a detailed thermal aging study of F82H, in which tensile and Charpy specimens were aged at 500, 550, 600, and 650 deg. C to 30,000 h. These studies indicated that there was a decrease in yield stress at the two highest temperatures and essentially no change at the two lowest temperatures. Despite the strength decrease or no change, the DBTT increased for Charpy specimens irradiated at all four temperatures. Precipitates were extracted from thermally aged specimens, and the amount of precipitate was correlated with the increase in transition temperature. Laves phase was identified in the extracted precipitates by X-ray diffraction. Earlier studies on conventional elevated-temperature steels also showed embrittlement effects above the irradiation-hardening temperature

  12. Development of reconstitution technique of irradiated specimen. 2. Annual report for FY1994 on JAERI-IHI cooperated research program

    Energy Technology Data Exchange (ETDEWEB)

    Nishiyama, Yutaka; Fukaya, Kiyoshi; Onizawa, Kunio; Suzuki, Masahide; Shibata, Katsuyuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaihara, Shoichiro; Nakamura, Terumi; Sato, Akira; Yoshida, Kazuo

    1996-02-01

    A surface-activated joining method to construct Charpy impact specimens from a limited volume of broken specimens is being developed. The method is likely to decrease the thermal input led to annealing and metallurgical changes. This paper describes the technical qualification process of the joining parameters and surface configuration of joined specimens. All tests have been done with A533B cl.1. The joining machine with higher vacuum than that previously used was prepared for the tests. Precise control of joining parameters led to heat-affected zone as small as 1mm in each side. In the case of joining the square shaped (10x10mm) and circular shaped ({phi} 16mm) specimens, overall joining was achieved by an attached envelope to the square shaped specimen. In addition, the grooved surface of the circular shaped specimen brought out uniformly distributed heat-affected zone. The specification of hot-use joining machine which involves the joining sequence and restrictions of the dimension was also examined. (author).

  13. Status Report on Irradiation Capsules Containing Welded FeCrAl Specimens for Radiation Tolerance Evaluation

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-26

    This status report provides the background and current status of a series of irradiation capsules, or “rabbits”, that were designed and built to test the contributions of microstructure, composition, damage dose, and irradiation temperature on the radiation tolerance of candidate FeCrAl alloys being developed to have enhanced weldability and radiation tolerance. These rabbits will also test the validity of using an ultra-miniature tensile specimen to assess the mechanical properties of irradiated FeCrAl base metal and weldments. All rabbits are to be irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) to damage doses up to ≥15 dpa at temperatures between 200-550°C.

  14. Effect of small additional elements on DBTT of V 4Cr 4Ti irradiated at low temperatures

    Science.gov (United States)

    Shibayama, Tamaki; Yamagata, Ichiro; Kayano, Hideo; Namba, Chusei

    1998-10-01

    As a part of a program to screen several V-4Cr-4Ti containing Si, Al and Y alloys and optimize the amounts of Si, Al and Y, the Charpy impact test of five kinds of V-4Cr-4Ti-Si-Al-Y alloys by an instrumented Charpy impact testing machine using miniaturized specimens (1.5 mm × 1.5 mm × 20 mm) have been conducted before and after neutron irradiation. Charpy impact specimens were encapsulated in an aluminum vial filled with high purity He and irradiated up to 1.06 × 10 19 n/cm 2 ( E > 1 MeV, 156 h) at low temperatures (about 150°C) in Japan Materials Testing Reactor (JMTR). The ductile brittle transition temperature (DBTT) of each alloy was determined by various methods on absorbed energy, brittle fracture ratio and lateral expansion from a quantitative analysis of fractography for broken specimens after the Charpy impact test. Almost all specimens were embrittled after low temperature irradiation. Decomposition of primary precipitates could result in migration of interstitial elements to irradiation defects and many precipitates are formed under irradiation. Radiation hardening then caused the substantial degradation of its fracture toughness.

  15. Reactor Materials Program electrochemical potential measurements by ORNL with unirradiated and irradiated stainless steel specimens

    Energy Technology Data Exchange (ETDEWEB)

    Baumann, E.W.; Caskey, G.R. Jr.

    1993-07-01

    Effect of irradiation of stainless steel on electrochemical potential (ECP) was investigated by measurements in dilute HNO{sub 3} and H{sub 2}O{sub 2} solutions, conditions simulating reactor moderator. The electrodes were made from unirradiated/irradiated, unsensitized/sensitized specimens from R-reactor piping. Results were inconclusive because of budgetary restrictions. The dose rate may have been too small to produce a significant radiolytic effect. Neither the earlier CERT corrosion susceptibility tests nor the present ECP measurements showed a pronounced effect of irradiation on susceptibility of the stainless steel to IGSCC; this is confirmed by the absence in the stainless steel of the SRS reactor tanks (except for the C Reactor tank knuckle area).

  16. A reassessment of the effects of helium on Charpy impact properties of ferritic/martensitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, D.S.; Hamilton, M.L. [Pacific Northwest National Lab., Richland, WA (United States); Hankin, G.L. [Loughborough Univ. (United Kingdom)

    1998-03-01

    To test the effect of helium on Charpy impact properties of ferritic/martensitic steels, two approaches are reviewed: quantification of results of tests performed on specimens irradiated in reactors with very different neutron spectra, and isotopic tailoring experiments. Data analysis can show that if the differences in reactor response are indeed due to helium effects, then irradiation in a fusion machine at 400 C to 100 dpa and 1000 appm He will result in a ductile to brittle transition temperature shift of over 500 C. However, the response as a function of dose and helium level is unlikely to be simply due to helium based on physical reasoning. Shear punch tests and microstructural examinations also support this conclusion based on irradiated samples of a series of alloys made by adding various isotopes of nickel in order to vary the production of helium during irradiation in HFIR. The addition of nickel at any isotopic balance to the Fe-12Cr base alloy significantly increased the shear yield and maximum strengths of the alloys. However, helium itself, up to 75 appm at over 7 dpa appears to have little effect on the mechanical properties of the alloys. This behavior is instead understood to result from complex precipitation response. The database for effects of helium on embrittlement based on nickel additions is therefore probably misleading and experiments should be redesigned to avoid nickel precipitation.

  17. Charpy Impact Energy and Microindentation Hardness of 60-NITINOL

    Science.gov (United States)

    Stanford, Malcolm K.

    2012-01-01

    60-NITINOL (60 wt.% Ni 40 wt.% Ti) is being studied as a material for advanced aerospace components. The Charpy impact energy and microindentation hardness has been studied for this material, fabricated by vacuum induction skull melting (casting) and by hot isostatic pressing. Test specimens were prepared in various hardened and annealed heat treatment conditions. The average impact energy ranged from 0.33 to 0.49J for the hardened specimens while the annealed specimens had impact energies ranging from 0.89 to 1.18J. The average hardness values of the hardened specimens ranged from 590 to 676 HV while that of the annealed specimens ranged from 298 to 366 HV, suggesting an inverse relationship between impact energy and hardness. These results are expected to provide guidance in the selection of heat treatment processes for the design of mechanical components.

  18. Evaluation of hydrogen embrittlement and temper embrittlement by key curve method in instrumented Charpy test

    Directory of Open Access Journals (Sweden)

    Makita A.

    2010-06-01

    Full Text Available Instrumented Charpy test was conducted on small sized specimen of 21/4Cr-1Mo steel. In the test the single specimen key curve method was applied to determine the value of fracture toughness for the initiation of crack extension with hydrogen free, KIC, and for hydrogen embrittlement cracking, KIH. Also the tearing modulus as a parameter for resistance to crack extension was determined. The role of these parameters was discussed at an upper shelf temperature and at a transition temperature. Then the key curve method combined with instrumented Charpy test was proven to be used to evaluate not only temper embrittlement but also hydrogen embrittlement.

  19. Evaluation of hydrogen embrittlement and temper embrittlement by key curve method in instrumented Charpy test

    Science.gov (United States)

    Ohtsuka, N.; Shindo, Y.; Makita, A.

    2010-06-01

    Instrumented Charpy test was conducted on small sized specimen of 21/4Cr-1Mo steel. In the test the single specimen key curve method was applied to determine the value of fracture toughness for the initiation of crack extension with hydrogen free, KIC, and for hydrogen embrittlement cracking, KIH. Also the tearing modulus as a parameter for resistance to crack extension was determined. The role of these parameters was discussed at an upper shelf temperature and at a transition temperature. Then the key curve method combined with instrumented Charpy test was proven to be used to evaluate not only temper embrittlement but also hydrogen embrittlement.

  20. Charpy Impact Test on Polymeric Molded Parts

    Directory of Open Access Journals (Sweden)

    Alexandra Raicu

    2012-09-01

    Full Text Available The paper presents the Charpy impact tests on the AcrylonitrileButadiene-Styrene (ABS polymeric material parts. The Charpy impact test, also known as the Charpy V-notch test, is a standardized strain rate test which determines the amount of energy absorbed by a material during fracture. This is a typical method described in ASTM Standard D 6110. We use for testing an Instron - Dynatup equipment which have a fully integrated hardware and software package that let us capture load information at very high speed from the impact tests.

  1. Microstructural characterization of Charpy-impact-tested nanostructured bainite

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, Y.T.; Chang, H.T.; Huang, B.M. [Department of Materials Science and Engineering, National Taiwan University, Taipei 10617, Taiwan, ROC (China); Huang, C.Y. [Iron and Steel R& D Department, China Steel Corporation, Kaohsiung, Taiwan, ROC (China); Yang, J.R., E-mail: jryang@ntu.edu.tw [Department of Materials Science and Engineering, National Taiwan University, Taipei 10617, Taiwan, ROC (China)

    2015-09-15

    In this work, a possible cause of the extraordinary low impact toughness of nanostructured bainite has been investigated. The microstructure of nanostructured bainite consisted chiefly of carbide-free bainitic ferrite with retained austenite films. X-ray diffractometry (XRD) measurement indicated that no retained austenite existed in the fractured surface of the Charpy-impact-tested specimens. Fractographs showed that cracks propagated mainly along bainitic ferrite platelet boundaries. The change in microstructure after impact loading was verified by transmission electron microscopy (TEM) observations, confirming that retained austenite was completely transformed to strain-induced martensite during the Charpy impact test. However, the zone affected by strained-induced martensite was found to be extremely shallow, only to a depth of several micrometers from the fracture surface. It is appropriately concluded that upon impact, as the crack forms and propagates, strain-induced martensitic transformation immediately occurs ahead of the advancing crack tip. The successive martensitic transformation profoundly facilitates the crack propagation, resulting in the extremely low impact toughness of nanostructured bainite. Retained austenite, in contrast to its well-known beneficial role, has a deteriorating effect on toughness during the course of Charpy impact. - Highlights: • The microstructure of nanostructured bainite consisted of nano-sized bainitic ferrite subunits with retained austenite films. • Special sample preparations for SEM, XRD and TEM were made, and the strain-affected structures have been explored. • Retained austenite films were found to transform into martensite after impact loading, as evidenced by XRD and TEM results. • The zone of strain-induced martensite was found to extend to only several micrometers from the fracture surface. • The poor Charpy impact toughness is associated with the fracture of martensite at a high strain rate during

  2. Analysis of stress-induced Burgers vector anisotropy in pressurized tube specimens of irradiated ferritic-martensitic steel: JLF-1

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States); Shibayama, T. [Univ. of Hokkaido, Oarai, Ibaraki (Japan). Inst. for Materials Research

    1998-09-01

    A procedure for determining the Burgers vector anisotropy in irradiated ferritic steels allowing identification of all a<100> and all a/2<111> dislocations in a region of interest is applied to a pressurized tube specimen of JLF-1 irradiated at 430 C to 14.3 {times} 10{sup 22} n/cm{sup 2} (E > 0.1 MeV) or 61 dpa. Analysis of micrographs indicates large anisotropy in Burgers vector populations develop during irradiation creep.

  3. Accelerated 54{degree}C irradiated test of Shippingport neutron shield tank and HFIR vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Hawthorne, J.R. [Materials Engineering Associates, Inc., Lanham, MD (United States); Rosinski, S.T. [Sandia National Labs., Albuquerque, NM (United States)

    1993-01-01

    Charpy V-notch specimens (ASTM Type A) and 5.74-mm diameter tension test specimens of the Shippingport Reactor Neutron Shield Tank (NST) (outer wall material) were irradiated together with Charpy V-notch specimens of the Oak Ridge National Laboratory (ORNI), High,, Flux Isotope Reactor (HFIR) vessel (shell material), to 5.07 {times} 10{sup 17} n/cm{sup 2}, E > 1 MeV. The irradiation was performed in the Ford Nuclear Reactor (FNR), a test reactor, at a controlled temperature of 54{degrees}C (130{degrees}F) selected to approximate the prior service temperatures of the cited reactor structures. Radiation-induced elevations in the Charpy 41-J transition temperature and the ambient temperature yield strength were small and independent of specimen test orientation (ASTM LT vs. TL). The observations are consistent with prior findings for the two materials (A 212-B plate) and other like materials irradiated at low temperature (< 200{degrees}C) to low fluence. The high radiation embrittlement sensitivity observed in HFIR vessel surveillance program tests was not found in the present accelerated irradiation test. Response to 288{degrees}C-168 h postirradiation annealing was explored for the NST material. Notch ductility recovery was found independent of specimen test orientation but dependent on the temperature within the transition region at which the specimens were tested.

  4. Accelerated 54[degree]C irradiated test of Shippingport neutron shield tank and HFIR vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Hawthorne, J.R. (Materials Engineering Associates, Inc., Lanham, MD (United States)); Rosinski, S.T. (Sandia National Labs., Albuquerque, NM (United States))

    1993-01-01

    Charpy V-notch specimens (ASTM Type A) and 5.74-mm diameter tension test specimens of the Shippingport Reactor Neutron Shield Tank (NST) (outer wall material) were irradiated together with Charpy V-notch specimens of the Oak Ridge National Laboratory (ORNI), High,, Flux Isotope Reactor (HFIR) vessel (shell material), to 5.07 [times] 10[sup 17] n/cm[sup 2], E > 1 MeV. The irradiation was performed in the Ford Nuclear Reactor (FNR), a test reactor, at a controlled temperature of 54[degrees]C (130[degrees]F) selected to approximate the prior service temperatures of the cited reactor structures. Radiation-induced elevations in the Charpy 41-J transition temperature and the ambient temperature yield strength were small and independent of specimen test orientation (ASTM LT vs. TL). The observations are consistent with prior findings for the two materials (A 212-B plate) and other like materials irradiated at low temperature (< 200[degrees]C) to low fluence. The high radiation embrittlement sensitivity observed in HFIR vessel surveillance program tests was not found in the present accelerated irradiation test. Response to 288[degrees]C-168 h postirradiation annealing was explored for the NST material. Notch ductility recovery was found independent of specimen test orientation but dependent on the temperature within the transition region at which the specimens were tested.

  5. Summary of the U.S. specimen matrix for the HFIR 13J varying temperature irradiation capsule

    Energy Technology Data Exchange (ETDEWEB)

    Zinkle, S.J. [Oak Ridge National Lab., TN (United States)

    1998-03-01

    The US specimen matrix for the collaborative DOE/Monbusho HFIR 13J varying temperature irradiation capsule contains two ceramics and 29 different metals, including vanadium alloys, ferritic/martensitic steels, pure iron, austenitic stainless steels, nickel alloys, and copper alloys. This experiment is designed to provide fundamental information on the effects of brief low-temperature excursions on the tensile properties and microstructural evolution of a wide range of materials irradiated at nominal temperatures of 350 and 500 C to a dose of {approximately}5 dpa. A total of 340 miniature sheet tensile specimens and 274 TEM disks are included in the US-supplied matrix for the irradiation capsule.

  6. Compatibility of reduced activation ferritic/martensitic steel specimens with liquid Na and NaK in irradiation rig of IFMIF

    OpenAIRE

    2005-01-01

    In the high flux region of the International Fusion Materials Irradiation Facility (IFMIF), the neutron irradiation damage for iron-based alloys will exceed 20 dpa/ year. An accurate specimen temperature measurement under a large amount of nuclear heating is a key issue but the change of heat transfer of gap between irradiation specimens and specimen holder during irradiation test is inevitable, if gap is filled with an inert gas and temperature is monitored by a thermocouple buried in the sp...

  7. Fracture properties of neutron-irradiated martensitic 9Cr-WVTa steels below room temperature

    Science.gov (United States)

    Abe, F.; Narui, M.; Kayano, H.

    1994-09-01

    Fracture properties of the reduced activation martensitic 9Cr-1WVTa and 9Cr-3WVTa steels were investigated by carrying out instrumented Charpy impact tests and tensile tests at temperatures below room temperature after irradiation in the Japan Materials Testing Reactor at 493 and 538 K. Modified 9Cr-1MoVNb steel was also examined for comparison. The irradiation-induced increase in ductile-to-brittle transition temperature was 53, 26 and 40 K for the {1}/{3} size Charpy specimens of 9Cr-1WVTa, 9Cr-3WVTa and 9Cr-1MoVNb steels, respectively, which resulted primarily from the irradiation-induced increase in yield stress. The cleavage fracture stress was 1820-1870 MPa for the three steels in unirradiated conditions, which was scarcely affected by irradiation. The deflections to the maximum load and to the brittle fracture initiation were decreased by irradiation. In the tensile test, quasi-cleavage fracture occurred at 77 K in both unirradiated and irradiated conditions. The cleavage fracture stress was 1320-1380 MPa for the tensile specimens of the three steels, which was about 1.4 times smaller than that for the Charpy specimens.

  8. Dynamic fracture toughness and Charpy impact properties of an AISI 403 martensitic stainless steel

    Science.gov (United States)

    Sreenivasan, P. R.; Ray, S. K.; Mannan, S. L.; Rodriguez, P.

    1996-04-01

    Dynamic fracture toughness and Charpy impact properties of a normalised and tempered AISI 403 martensitic stainless steel obtained from instrumented impact tests are presented. Procedures for estimating dynamic fracture toughness ( KId) from the load-time traces obtained in instrumented tests of unprecracked Charpy V-notch (CVN) specimens are considered. The estimated KId values show reasonable agreement with those obtained from instrumented drop-weight and precracked Charpy tests. Also, except in the upper transition and uppershelf regions, the ASME KIR curve is generally conservative (i.e. gives lower KId values) when compared to the above KId estimates. The conservatism of the ASME KIR at the upper transition and uppershelf temperatures needs verification/validation. The lowest KId values estimated at the lower shelf temperatures for the above steel, namely, 33-42 MPa√m are in good agreement with the reported values of 35-50 MPa√m for the same steel in the literature.

  9. Instrumented impact testing machine with reduced specimen oscillation effects

    Science.gov (United States)

    Rintamaa, R.; Ranka, K.; Wallin, K.; Ikonen, K.; Talja, H.; Kotilainen, H.; Sirkkola, E.

    1984-07-01

    A pendulum-type instrumented Charpy test apparatus based on inverted test geometry was developed. Geometry inversion reduces inertia load and specimen oscillation effects. Initial impact energy is double that of standard (300 J) impact testers, allowing the use of larger (10 x 20 x 110 mm) bend specimens than normal Charpy specimens. The rotation axis in the three point bending is nearly stationary, making COD-measurements possible. Inertia effects and specimen oscillations are compared with the conventional tester, and using an analytical finite element model for Charpy V-notch specimens. Better performance for the inverted geometry is reported.

  10. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    Directory of Open Access Journals (Sweden)

    Panferov Pavel

    2016-01-01

    Full Text Available The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  11. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    Science.gov (United States)

    Panferov, Pavel; Kochkin, Viacheslav; Erak, Dmitry; Makhotin, Denis; Reshetnikov, Alexandr; Timofeev, Andrey

    2016-02-01

    The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  12. ATR-A1 irradiation experiment on vanadium alloys and low activation steels

    Energy Technology Data Exchange (ETDEWEB)

    Tasi, H.; Strain, R.V.; Gomes, I.; Hins, A.G.; Smith, D.L.

    1996-04-01

    To study the mechanical properties of vanadium alloys under neutron irradiation at low temperatures, an experiment was designed and constructed for irradiation in the Advanced Test Reactor (ATR). The experiment contained Charpy, tensile, compact tension, TEM, and creep specimens of vanadium alloys. It also contained limited low-activation ferritic steel specimens as part of the collaborative agreement with Monbusho of Japan. The design irradiation temperatures for the vanadium alloy specimens in the experiment are {approx}200 and 300{degrees}C, achieved with passive gap-gap sizing and fill gas blending. To mitigate vanadium-to-chromium transmutation from the thermal neutron flux, the test specimens are contained inside gadolinium flux filters. All specimens are lithium-bonded. The irradiation started in Cycle 108A (December 3, 1995) and is expected to have a duration of three ATR cycles and a peak influence of 4.4 dpa.

  13. The Assessment and Validation of Mini-Compact Tension Test Specimen Geometry and Progress in Establishing Technique for Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Mikhail A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Nanstad, Randy K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-01

    Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Assessment and validation of mini-CT specimen geometry has been performed on previously well characterized HSST Plate 13B, an A533B class 1 steel. It was shown that the fracture toughness transition temperature measured by these Mini-CT specimens is within the range of To values that were derived from various large fracture toughness specimens. Moreover, the scatter of the fracture toughness values measured by Mini-CT specimens perfectly follows the Weibull distribution function providing additional proof for validation of this geometry for the Master Curve evaluation of rector pressure vessel steels. Moreover, the International collaborative program has been developed to extend the assessment and validation efforts to irradiated weld metal. The program is underway and involves ORNL, CRIEPI, and EPRI.

  14. Tensile and impact properties of vanadium-base alloys irradiated at low temperatures in the ATR-A1 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Nowicki, L.J.; Billone, M.C.; Chung, H.M.; Smith, D.L. [Argonne National Lab., IL (United States)

    1998-03-01

    Subsize tensile and Charpy specimens made from several V-(4-5)Cr-(4-5)Ti alloys were irradiated in the ATR-A1 experiment to study the effects of low-temperature irradiation on mechanical properties. These specimens were contained in lithium-bonded subcapsules and irradiated at temperatures between {approx}200 and 300 C. Peak neutron damage was {approx}4.7 dpa. Postirradiation testing of these specimens has begun. Preliminary results from a limited number of specimens indicate a significant loss of work-hardening capability and dynamic toughness due to the irradiation. These results are consistent with data from previous low-temperature neutron irradiation experiments on these alloys.

  15. Irradiation effects on fracture toughness of two high-copper submerged-arc welds, HSSI Series 5. Volume 1, Main report and Appendices A, B, C, and D

    Energy Technology Data Exchange (ETDEWEB)

    Nanstad, R.K.; Haggag, F.M.; McCabe, D.E.; Iskander, S.K.; Bowman, K.O. [Oak Ridge National Lab., TN (United States); Menke, B.H. [Materials Engineering Associates, Inc., Lanham, MD (United States)

    1992-10-01

    The Fifth Irradiation Series in the Heavy-Section Steel Irradiation Program obtained a statistically significant fracture toughness data base on two high-copper (0.23 and 0.31 wt %) submerged-arc welds to determine the shift and shape of the K{sub Ic} curve as a consequence of irradiation. Compact specimens with thicknesses to 101.6 mm (4 in) in the irradiated condition and 203.2 mm (8 in) in the unirradiated condition were tested, in addition to Charpy impact, tensile, and drop-weight specimens. Irradiations were conducted at a nominal temperature of 288{degree}C and an average fluence of 1.5 {times} 10{sup 19} neutrons/cm{sup 2} (>l MeV). The Charpy 41-J temperature shifts are about the same as the corresponding drop-weight NDT temperature shifts. The irradiated welds exhibited substantial numbers of cleavage pop-ins. Mean curve fits using two-parameter (with fixed intercept) nonlinear and linearized exponential regression analysis revealed that the fracture toughness 100 MPa{lg_bullet}{radical}m shifts exceeded the Charpy 41-J shifts for both welds. Analyses of curve shape changes indicated decreases in the slopes of the fracture toughness curves, especially for the higher copper weld. Weibull analyses were performed to investigate development of lower bound curves to the data, including the use of a variable K{sub min} parameter which affects the curve shape.

  16. Characterization of radiation induced defects in EUROFER 97 after neutron irradiation

    Science.gov (United States)

    Klimenkov, M.; Materna-Morris, E.; Möslang, A.

    2011-10-01

    Specimens of EUROFER 97 prepared for impact tests have been irradiated to an average dose of 16.3 dpa at irradiation temperatures of 250-450 °C. TEM investigations have been performed to study radiation induced changes in the microstructure. The characterization and statistical analysis show the temperature dependant formation of small dislocation loops and He bubbles. The Burgers vector of dislocation loops was ½. A novel feature is that within statistical uncertainty the maximum in the dislocation density observed around 300 °C decreased with decreasing irradiation temperature down to 250 °C. The TEM data are correlated with tensile and instrumented Charpy test results.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-02-15

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

  18. Stability of SARS Coronavirus in Human Specimens and Environment and Its Sensitivity to Heating and UV Irradiation

    Institute of Scientific and Technical Information of China (English)

    SHU-MING DUAN; XIAO-PING DONG; SARS RESEARCH TEAM; XIN-SHENG ZHAO; RUI-FU WEN; JING-JING HUANG; GUO-HUA PI; SU-XIANG ZHANG; JUN HAN; SHENG-LI BI; LI RUAN

    2003-01-01

    The causal agent for SARS is considered as a novel coronavirus that has never been described both in human and animals previously. The stability of SARS coronavirus in human specimens and in environments was studied. Methods Using a SARS coronavirus strain CoV-P9,which was isolated from pharyngeal swab of a probable SARS case in Beijing, its stability in mimic human specimens and in mimic environment including surfaces of commonly used materials or in household conditions, as well as its resistances to temperature and UV irradiation were analyzed. A total of 106 TCID50 viruses were placed in each tested condition, and changes of the viral infectivity in samples after treatments were measured by evaluating cytopathic effect (CPE) in cell line Vero-E6 at 48 h after infectionn. Results The results showed that SARS coronavirus in the testing condition could survive in serum, 1:20 diluted sputum and feces for at least 96 h, whereas it could remain alive in urine for at least 72 h with a low level of infectivity. The survival abilities on the surfaces of eight different materials and in water were quite comparable, revealing reduction of infectivity after 72 to 96 h exposure. Viruses stayed stable at 4℃, at room temperature (20℃) and at 37℃ for at least 2 h without remarkable change in the infectious ability in cells, but were convened to be non-infectious after 90-, 60- and 30-min exposure at 56℃, at 67℃ and at 75℃, respectively. Irradiation of UV for 60 min on the virus in culture medium resulted in the destruction of viral infectivity at an undetectable level. Conclusion The survival ability of SARS coronavirus in human specimens and in environments seems to be relatively strong. Heating and UV irradiation can efficiently eliminate the viral infectivity.

  19. The irradiation test plan and safety analysis of the creep capsule(03S-07K) equipped with double specimen

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Man Soon; Kim, B. G.; Choo, K. N.; Sohn, J. M.; Choi, M. H.; Kim, Y

    2005-04-15

    The irradiation test plan and safety analysis of the creep capsule(03S-07K) equipped with double specimen. In this report, the reactivity effect was reviewed and an analysis for the structural and thermal integrity was performed to review the safety of the creep capsule 03S-07K, which will be irradiated at a temperature higher than 550 .deg. C. The irradiation test will be performed at the in-core IR2 hole for 23 days at the 30 MWth power of HANARO. In the irradiation test, the temperature of the inside parts in the capsule will be measured and compared with the design value for reviewing the design data, and also the integrity of the bellows and LVDT etc. will be confirmed. The reactivity worth by the insertion of the creep capsule is no more than +9.2 mk, and this indicates that the reactivity effect does not exceed +12.5mk as specified in 'the HANARO operation technical specification'. The temperatures of the specimen, LVDT and the center rods of the bellows are less than the melting temperatures of the corresponding materials, therefore, the integrity of the materials are maintained. The center rod is made as a hollow tube shape of {phi}13mm x 2.5mmt of Ti material instead of the STS304 rod to lower the temperature. Thus, the temperature of the center rod of the bellows reaches 332{approx}1,095 .deg. C according to the vacuum condition of the capsules inside. By the structural analysis considering this temperature, the combined stress(the primary membrane and the secondary thermal) on the outer tube is 96.06 MPa for the HANARO 30 MWth power. The results of this stress analysis satisfiy the allowable stress limits.

  20. Correlation between irradiation-induced changes of microstructural parameters and mechanical properties of RPV steels

    Science.gov (United States)

    Böhmert, J.; Viehrig, H.-W.; Ulbricht, A.

    2004-08-01

    Radiation hardening, displayed by the yield stress increase, and irradiation embrittlement, described by the Charpy transition temperature shift, were experimentally determined for a broad variety of irradiation specimens machined from different reactor pressure vessel base and weld materials and irradiated in several VVER-type reactors. Additionally, the same specimens were investigated by small angle neutron scattering. The analysis of the neutron scattering data suggests the presence of nano-scaled irradiation defects. The volume fraction of these defects depends on the neutron fluence and the material. Both irradiation hardening and irradiation embrittlement correlate linearly with the square root of the defect volume fraction. However, a generally valid proportionality is only a rough approximation. In detail, chemical composition and technological pretreatment clearly affect the correlation.

  1. On the Effectiveness of the Dynamic Force Adjustment for Reducing the Scatter of Instrumented Charpy Results

    Energy Technology Data Exchange (ETDEWEB)

    Lucon, E.

    2008-09-15

    One of the key factors for obtaining reliable instrumented Charpy results is the calibration of the instrumented striker. An interesting alternative to the conventional static calibration recommended by the standards is the Dynamic Force Adjustment (DFA), in which forces and displacements are iteratively adjusted until equality is achieved between absorbed energies calculated under the test record (Wt) and measured by the machine encoder (KV). In this study, this procedure has been applied to the instrumented data obtained by 10 international laboratories using notched and precracked Charpy specimens, in the framework of a Coordinated Research Project (CRP8) of IAEA. DFA is extremely effective in reducing the between-laboratory scatter for both general yield and maximum forces. The effect is less significant for dynamic reference temperatures measured from precracked Charpy specimens using the Master Curve procedure, but a moderate reduction of the standard deviation is anyway observed. It is shown that striker calibration is a prominent contribution to the interlaboratory variability of instrumented impact forces, particularly in the case of maximum forces.

  2. Evaluation of dynamic fracture mechanics in the AISI 316 stainless steel using instrumented Charpy impact testing

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Juliano Daniel de [Empresa Brasileira de Aeronautica S.A. (EMBRAER), Sao Jose dos Campos, SP (Brazil)]. E-mail: juliano.daniel@embraer.com.br; Rodrigues, Bruno Jardim Franca [Novo Nordisk, Montes Claros, MG (Brazil)]. E-mail: brro@novonordisk.com; Vilela, Jefferson Jose; Martins, Geraldo de Paula [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)]. E-mail: gpm@cdtn.br; Carneiro, Jose Rubens Goncalves [Pontificia Universidade Catolica de Minas Gerais (PUC Minas), Belo Horizonte, MG (Brazil)]. E-mail: joserub@pucminas.br

    2007-07-01

    The nuclear power plant's surveillance program is based in Charpy test. But, this test could be used to evaluate integrity's secondary circuit. The steel similar to AISI 316 stainless steel could be used in this circuit. Some secondary circuit's components could be failed in dynamic condition. The dynamic fracture mechanics behavior of the AISI 316 was studied by using instrumented Charpy impact testing. The dynamic fracture toughness (J{sub ld}) could be evaluated by four different methods: compliance changing rate, stretching zone, energy revised and maximum load energy. The tests were made in temperature -196 deg C, room and 200 deg C. At each temperature two specimens were tested. The impact energy was 300 J and the impact velocity was 5.12 m/s. The Charpy specimens 10 x 10 x 50 mm were pre-cracked until 5 mm according to ASTM E-23. Stretching zone size was measured and analyzed by observing the fracture surfaces that were obtained in a scanning electron microscope. The dynamic fracture toughness calculated among four different methods showed a large difference. All studied methods did not agree ASTM E1820 (2001) standard that indicated to plane strain did not occurred in the tip crack. (author)

  3. Material inertia and size effects in the Charpy V-notch test

    DEFF Research Database (Denmark)

    Desandre, D. A.; Benzerga, A. A.; Tvergaard, Viggo

    2004-01-01

    The effect of material inertia on the size dependence of the absorbed energy in the Charpy V-notch test is investigated. The material response is characterized by an elastic-viscoplastic constitutive relation for a porous plastic solid, with adiabatic heating due to plastic dissipation and the re......The effect of material inertia on the size dependence of the absorbed energy in the Charpy V-notch test is investigated. The material response is characterized by an elastic-viscoplastic constitutive relation for a porous plastic solid, with adiabatic heating due to plastic dissipation...... and the resulting thermal softening accounted for. The onset of cleavage is taken to occur when a critical value of the maximum principal stress is attained over a critical volume. Plane strain dynamic analyses are carried out for geometrically similar specimens of various sizes with all parameters adjusted so...

  4. Evaluation of ductile-brittle transition temperature before and after neutron irradiation for RPV steels using small punch tests

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min-Chul [Korea Atomic Energy Research Institute, 150 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)]. E-mail: mckim@kaeri.re.kr; Oh, Yong Jun [Hanbat National University, Deogmyeong-dong, Yuseong-gu, Daejeon 305-719 (Korea, Republic of); Lee, Bong Sang [Korea Atomic Energy Research Institute, 150 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2005-08-01

    Small punch (SP) tests were performed to evaluate the ductile-brittle transition temperature before and after a neutron irradiation of reactor pressure vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the conventional Charpy tests and the Master Curve fracture toughness tests in accordance with the American Society for Testing and Materials (ASTM) standard E1921. Small punch specimens were taken from a 1/4t location of the vessel thickness and machined into a 10 mm x 10 mm x 0.5 mm dimension. The specimens were irradiated in the research reactors at Korea Atomic Energy Research Institute Nuclear Research Institute in the Czech Republic at the different fluence levels of about 290 deg C. Small punch tests were performed in the temperature range of RT to -196 deg C using a 2.4 mm diameter ball. For the materials before and after irradiation, the small punch transition temperatures (T {sub SP}), which are determined at the middle of the upper small punch energies, showed a linear correlation with the Charpy index temperature, T {sub 41J}. T {sub SP} from the irradiated samples was increased with the fluence levels and was well within the deviation range of the unirradiated data. However, the transition temperature shift from the Charpy test ({delta}T {sub 41J}) shows a better correlation with the transition temperature shift ({delta}T {sub SP(E)}) when a specific small punch energy level rather than the middle energy level of the small punch curve is used to determine the transition temperature. T {sub SP} also had a correlation with the reference temperature (T {sub 0}) from the Master Curve method using a pre-cracked Charpy V-notched (PCVN) specimen.

  5. Effects of post-irradiation annealing and re-irradiation on microstructure in surveillance test specimens of the Loviisa-1 reactor studied by atom probe tomography and positron annihilation

    Science.gov (United States)

    Toyama, T.; Kuramoto, A.; Nagai, Y.; Inoue, K.; Nozawa, Y.; Shimizu, Y.; Matsukawa, Y.; Hasegawa, M.; Valo, M.

    2014-06-01

    This paper presents a microstructural study of a surveillance test specimen from the Loviisa-1 reactor in Finland, which is a Russian-type pressurized water reactor (VVER-440), after initial irradiation to a neutron fluence of 2.5 × 1019 n/cm2 (E > 1 MeV), post-irradiation annealing at 475 °C for 100 h and re-irradiation to three different fluences up to 2.7 × 1019 n/cm2. Atom probe tomography (APT) and positron annihilation spectroscopy (PAS) were used to characterize the test specimens. APT results showed the formation of Cu-rich solute clusters (SCs) during the initial irradiation and their subsequent coarsening during annealing. After re-irradiation, a small number of SCs formed once again. The hardening due to the SCs was estimated using the Russell-Brown model based on the APT results, and was in good agreement with the measured hardening after the initial irradiation and post-irradiation annealing. In contrast, during the first-step of re-irradiation, the estimated hardening due to the SCs was smaller than the measured hardening. This suggested that the hardening after re-irradiation was due to some microstructure other than the observed SCs. This difference was attributed to newly-formed matrix defects during re-irradiation, which was supported by the PAS results. However in subsequent steps of re-irradiation, the hardening was almost constant.

  6. Effect of Cadmium Plating Thickness on the Charpy Impact Energy of Hydrogen-Charged 4340 Steel

    Science.gov (United States)

    Es-Said, O. S.; Alcisto, J.; Guerra, J.; Jones, E.; Dominguez, A.; Hahn, M.; Ula, N.; Zeng, L.; Ramsey, B.; Mulazimoglu, H.; Li, Yong-Jun; Miller, M.; Alrashid, J.; Papakyriakou, M.; Kalnaus, S.; Lee, E. W.; Frazier, W. E.

    2016-09-01

    Hydrogen was intentionally introduced into ultra-high strength steel by cadmium plating. The purpose was to examine the effect of cadmium plate thickness and hence hydrogen on the impact energy of the steel. The AISI 4340 steel was austenitized at 1000 °C for 1 h, water quenched, and tempered at temperatures between 257 and 593 °C in order to achieve a range of targeted strength levels. The specimens were cadmium plated with 0.00508 mm (0.2 mils), 0.00762 mm (0.3 mils), and 0.0127 mm (0.5 mils). Results demonstrated that the uncharged specimens exhibited higher impact energy values when compared to the plated specimens at all tempering temperatures. The cadmium-plated specimens had very low Charpy impact values irrespective of their ultimate tensile strength values. The model of hydrogen transport by mobile dislocations to the fracture site appears to provide the most suitable explanation of the results.

  7. Subtask 12F4: Effects of neutron irradiation on the impact properties and fracture behavior of vanadium-base alloys

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Loomis, B.A.; Smith, D.L. [Argonne National Lab., IL (United States)

    1995-03-01

    Up-to-date results on the effects of neutron irradiation on the impact properties and fracture behavior of V, V-Ti, V-Cr-Ti and V-Ti-Si alloys are presented in this paper, with an emphasis on the behavior of the U.S. reference alloys V-4Cr-4Ti containing 500-1000 wppm Si. Database on impact energy and cluctile-brittle transition temperature (DBTT) has been established from Charpy impact tests of one-third-size specimens irradiated at 420{degrees}C-600{degrees}C up to {approx}50 dpa in lithium environment in fast fission reactors. To supplement the Charpy impact tests fracture behavior was also characterized by quantitative SEM fractography on miniature tensile and disk specimens that were irradiated to similar conditions and fractured at -196{degrees}C to 200{degrees}C by multiple bending. For similar irradiation conditions irradiation-induced increase in DBTT was influenced most significantly by Cr content, indicating that irradiation-induced clustering of Cr atoms takes place in high-Cr (Cr {ge} 7 wt.%) alloys. When combined contents of Cr and Ti were {le}10 wt.%, effects of neutron irradiation on impact properties and fracture behavior were negligible. For example, from the Charpy-impact and multiple-bend tests there was no indication of irradiation-induced embrittlement for V-5Ti, V-3Ti-1Si and the U.S. reference alloy V-4Cr-4Ti after irradiation to {approx}34 dpa at 420{degrees}C to 600{degrees}C, and only ductile fracture was observed for temperatures as low as -196{degrees}C. 14 refs., 8 figs., 1 tab.

  8. Effect of neutron irradiation on the impact properties of A533B steel

    Energy Technology Data Exchange (ETDEWEB)

    Schubert, L.E.; Kumar, A.S. [Univ. of Missouri, Rolla, MO (United States); Rosinski, S.T. [Sandia National Labs., Albuquerque, NM (United States); Hamilton, M.L. [Pacific Northwest Lab., Richland, WA (United States)

    1994-10-01

    A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, ASTM type A 533 Grade B (A533B) having a low USE (USE < 100 J). The methodology appears to be more satisfactory than those methodologies proposed earlier. The USE was normalized by a normalization factor involving the dimensions of the Charpy specimen, the elastic stress concentration factor, and the plastic constraint at the notch root. The normalized values of the USE were found to be invariant with specimen size. In addition, it was also found that the ratio of the USE of unirradiated to that of irradiated materials was approximately the same for full, half, and third size specimens. The ductile-to-brittle transition temperature (DBTT) increased due to irradiation at 150 C to a nominal fluence of 1.0 {times} 10{sup 19} n/cm{sup 2} (E > 1 MeV) by 78 {degree}, 83{degree}, and 70{degree}C for full, half, and third size specimens, respectively. These shifts in DBTT appeared to be independent of specimen size and notch geometry.

  9. Influence of Stacking Sequence and Notch Angle on the Charpy Impact Behavior of Hybrid Composites

    Science.gov (United States)

    Behnia, S.; Daghigh, V.; Nikbin, K.; Fereidoon, A.; Ghorbani, J.

    2016-09-01

    The low-velocity impact behavior of hybrid composite laminates was investigated. The epoxy matrix was reinforced with aramid, glass, basalt, and carbon fabrics using the hand lay-up technique. Different stacking sequences and notch angles were and notch angles considered and tested using a Charpy impact testing machine to study the hybridization and notch angle effects on the impact response of the hybrid composites. The energy absorption capability of specimens with different stacking sequences and notch angles is compared and discussed. It is shown that the hybridization can enhance the mechanical performance of composite materials.

  10. Irradiation programme MANITU: results of impact tests with the irradiated materials (2,4 dpa); Bestrahlungsprogramm MANITU. Ergebnisse der Kerbschlagbiegeversuche mit den bis 2,4 dpa bestrahlten Werkstoffen

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, H.C.; Dafferner, B.; Ries, H.; Romer, O.

    2001-05-01

    The irradiation project MANITU was planned and carried out in the frame of the European long-term fusion materials development programme. The problem of the irradiation induced embrittlement of possible martensitic alloy candidates is still unsolved. But after the evaluation of sub-size Charpy tests with the unirradiated reference specimens of MANITU a first tendency was recognizable. The mechanical properties of the newly developed low activation 7-10% Cr-W(Ge)VTa alloys are partly better compared to the modified commercial 10-11% Cr-NiMoVNb steels. After the evaluation of subsize Charpy tests with specimens irradiated up to 0.2 and 0.8 dpa in the first phase of the MANITU programme, better mechanical properties of the 7-10% Cr-W(Ge)VTa alloys were obvious. In the present report the results of instrumented impact tests within the second phase of the MANITU programme (irradiation dose 2.4 dpa, irradiation temperatures 250, 300, 350, 400, and 450 C) are analysed and assessed in comparison to the results of the irradiation up to 0.2 and 0.8 dpa in the first phase of the project. Among the examined alloys (MANET-I/II, K-Heat, OPTIFER-Ia/II, F82H, ORNL 3791) the ORNL steel shows the very best embrittlement behaviour after neutron irradiation. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. C.; Lee, B. S. [KAERI, Taejon (Korea, Republic of); Oh, Y. J. [Hanbat National Univ., Taejon (Korea, Republic of)

    2003-10-01

    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{sub sp}), which are determined at the middle of the upper and lower SP energies, showed a linear correlation with the Charpy index temperature, T{sub 41J}. T{sub 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{sub 0}) from the master curve method using a pre-cracked Charpy V-notched (PCVN) specimen.

  12. Effect on fast neutron irradiation to 4 dpa at 400{degrees}C on the properties of V-(4-5)Cr-(4-5)Ti alloys

    Energy Technology Data Exchange (ETDEWEB)

    Zinkle, S.J.; Alexander, D.J.; Robertson, J.P. [Oak Ridge National Lab., TN (United States)] [and others

    1997-04-01

    Tensile, Charpy impact and electrical resistivity measurements have been performed at ORNL on V-4Cr-4Ti and V-5Cr-5Ti specimens that were prepared at ANL and irradiated in the lithium-bonded X530 experiment in the EBR-II fast reactor. All of the specimens were irradiated to a damage level of about 4 dpa at a temperature of {approximately}400{degrees}C. A significant amount of radiation hardening was evident in both the tensile and Charpy impact tests. The irradiated V-4Cr-4Ti yield strength measured at {approximately}390{degrees}C was >800 MPa, which is more than three times as high as the unirradiated value. The uniform elongations of the irradiated tensile specimens were typically {approximately}1%, with corresponding total elongations of 4-6%. The ductile to brittle transition temperature of the irradiated specimens was less than the unirradiated resistivity, which suggests that hardening associated with interstitial solute pickup was minimal.

  13. Validation Study of Unnotched Charpy and Taylor-Anvil Impact Experiments using Kayenta

    Energy Technology Data Exchange (ETDEWEB)

    Kamojjala, Krishna [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lacy, Jeffrey [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chu, Henry S. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Brannon, Rebecca [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    Validation of a single computational model with multiple available strain-to-failure fracture theories is presented through experimental tests and numerical simulations of the standardized unnotched Charpy and Taylor-anvil impact tests, both run using the same material model (Kayenta). Unnotched Charpy tests are performed on rolled homogeneous armor steel. The fracture patterns using Kayenta’s various failure options that include aleatory uncertainty and scale effects are compared against the experiments. Other quantities of interest include the average value of the absorbed energy and bend angle of the specimen. Taylor-anvil impact tests are performed on Ti6Al4V titanium alloy. The impact speeds of the specimen are 321 m/s and 393 m/s. The goal of the numerical work is to reproduce the damage patterns observed in the laboratory. For the numerical study, the Johnson-Cook failure model is used as the ductile fracture criterion, and aleatory uncertainty is applied to rate-dependence parameters to explore its effect on the fracture patterns.

  14. Embrittlement behavior of neutron irradiated RAFM steels

    Energy Technology Data Exchange (ETDEWEB)

    Gaganidze, E. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)]. E-mail: ermile.gaganidze@imf.fzk.de; Schneider, H.-C. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Dafferner, B. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Aktaa, J. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

    2007-08-01

    The effects of neutron irradiation on the embrittlement behavior of reduced activation ferritic/martensitic (RAFM) steel EUROFER97 for different heat treatment conditions have been investigated. The irradiation to 16.3 dpa at different irradiation temperatures (250-450 {sup o}C) was carried out in the Petten High Flux Reactor in the framework of the HFR Phase-IIb (SPICE) irradiation project. Several reference RAFM steels (F82H-mod, OPTIFER-Ia, GA3X) and MANET-I were also irradiated at selected temperatures. The embrittlement behavior and hardening were investigated by instrumented Charpy-V tests with subsize specimens. The neutron irradiation induced embrittlement and hardening of as-delivered EUROFER97 are comparable to those of investigated reference steels, being mostly pronounced for 250 {sup o}C and 300 {sup o}C irradiation temperatures. Heat treatment of EUROFER97 at higher austenization temperature substantially improves the embrittlement behavior at irradiation temperatures of 250 {sup o}C and 350 {sup o}C.

  15. Embrittlement behavior of neutron irradiated RAFM steels

    Science.gov (United States)

    Gaganidze, E.; Schneider, H.-C.; Dafferner, B.; Aktaa, J.

    2007-08-01

    The effects of neutron irradiation on the embrittlement behavior of reduced activation ferritic/martensitic (RAFM) steel EUROFER97 for different heat treatment conditions have been investigated. The irradiation to 16.3 dpa at different irradiation temperatures (250-450 °C) was carried out in the Petten High Flux Reactor in the framework of the HFR Phase-IIb (SPICE) irradiation project. Several reference RAFM steels (F82H-mod, OPTIFER-Ia, GA3X) and MANET-I were also irradiated at selected temperatures. The embrittlement behavior and hardening were investigated by instrumented Charpy-V tests with subsize specimens. The neutron irradiation induced embrittlement and hardening of as-delivered EUROFER97 are comparable to those of investigated reference steels, being mostly pronounced for 250 °C and 300 °C irradiation temperatures. Heat treatment of EUROFER97 at higher austenization temperature substantially improves the embrittlement behavior at irradiation temperatures of 250 °C and 350 °C.

  16. Fracture process of a low carbon low alloy steel relevant to charpy toughness at ductile-brittle fracture transition region

    Science.gov (United States)

    Tani, T.; Nagumo, M.

    1995-02-01

    The fracture process that determines the Charpy energy at the ductile-brittle transition region was investigated by means of the instrumented Charpy test and fractographic analysis with a low carbon low alloy steel subjected to different control-rolling conditions. The decomposition of a Charpy energy into the energies dissipated in the course of the notch-tip blunting, stable crack growth, and brittle crack propagation is unique irrespective of the testing temperatures and specimen series. Toughness level can be divided into four regions according to the pre-dominating fracture process. The temperature dependence of toughness and effects of the an-isotropy of a specimen originates in the brittle fracture initiation stage rather than the resistance against the notch-tip blunting or stable crack growth. From fractographic examination referring to the stress analyses, it is discussed that the brittle fracture initiation is controlled by the local deformation microstructures in the plastic zone together with the stress field ahead of the notch or the stable crack front.

  17. EUROFER 97. Tensile, charpy, creep and structural tests

    Energy Technology Data Exchange (ETDEWEB)

    Rieth, M.; Schirra, M.; Falkenstein, A.; Graf, P.; Heger, S.; Kempe, H.; Lindau, R.; Zimmermann, H.

    2003-10-01

    EUROFER 97 - the European reference material for the first wall of a DEMO fusion reactor - was produced as 3.5 t batch of rods and plates. Following the history of the development activities from conventional martensitic 12% Cr steel, MANET and OPTIFER up to the low or reduced activation (RAFM) EUROFER steel, results obtained from experiments on specimens from rods (diameter 100 mm) and plates (14 mm) are presented for a basic characterization. Physical and mechanical properties are compared with those of OPTIFER-1W and the F82H-mod 2% W steel. The transition behaviour was determined by plotting a continuous TTT (time temperature transition) diagram. In addition, extension coefficients were determined from room temperature up to 1000 C. Hardening tests at temperatures from 850 C to 1120 C illustrated the range of maximum hardness as well as grain size development. Tempering tests and additional annealing experiments from 300 C to 875 C allowed characterizing tempering behaviour and stability. Charpy properties were examined for various heat treatments and specimen types between 60 C and -100 C. Further, ductility criteria like FATT, DBTT and 68 J were determined. Particular attention was paid to the influence of grain size and O{sub 2} content. Tensile strength was measured for several heat treatments between room temperature and 700 C. Long-term ageing was investigated by means of stabilization annealing experiments. These were carried out with various temperature/time combinations including tensile tests. In EUROFER tensile strength was hardly affected by the different heat treatments while the ductility criteria showed only a moderate increase in temperature. Therefore, it can be concluded that EUROFER is not susceptible to ageing. Creep and creep rupture properties were investigated in the temperature range of 450 C to 650 C. So far, creep times of up to 15000 h have been covered by the experiments. The status of the test program allows for an extrapolation of

  18. Quality assurance of absorbed energy in Charpy impact test

    Science.gov (United States)

    Rocha, C. L. F.; Fabricio, D. A. K.; Costa, V. M.; Reguly, A.

    2016-07-01

    In order to ensure the quality assurance and comply with standard requirements, an intralaboratory study has been performed for impact Charpy tests, involving two operators. The results based on ANOVA (Analysis of Variance) and Normalized Error statistical techniques pointed out that the execution of the tests is appropriate, because the implementation of quality assurance methods showed acceptable results.

  19. Investigation of temperature dependence of fracture toughness in high-dose HT9 steel using small-specimen reuse technique

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong-Hyuk [KAERI; Byun, Thak Sang [ORNL; Maloy, S [Los Alamos National Laboratory (LANL); Toloczko, M [Pacific Northwest National Laboratory (PNNL)

    2014-01-01

    The temperature dependence of fracture toughness in HT9 steel irradiated to 3 145 dpa at 380 503 C was investigated using miniature three-point bend (TPB) fracture specimens. A miniature-specimen reuse technique has been established: the tested halves of subsize Charpy impact specimens with dimensions of 27 mm 3mm 4 mm were reused for this fracture test campaign by cutting a notch with a diamond-saw in the middle of each half, and by fatigue-precracking to generate a sharp crack tip. It was confirmed that the fracture toughness of HT9 steel in the dose range depends more strongly on the irradiation temperature than the irradiation dose. At an irradiation temperature <430 C, the fracture toughness of irradiated HT9 increased with the test temperature, reached an upper shelf of 180 200 MPa ffiffiffiffiffi m p at 350 450 C, and then decreased with the test temperature. At an irradiation temperatureP430 C, the fracture toughness was nearly unchanged up to about 450 C and decreased slowly with test temperatures in a higher temperature range. Such a rather monotonic test temperature dependence after high-temperature irradiation is similar to that observed for an archive material generally showing a higher degree of toughness. A brittle fracture without stable crack growth occurred in only a few specimens with relatively lower irradiation and test temperatures. In this discussion, these TPB fracture toughness data are compared with previously published data from 12.7 mm diameter disc compact tension (DCT) specimens.

  20. Investigation of temperature dependence of fracture toughness in high-dose HT9 steel using small-specimen reuse technique

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong-Hyuk, E-mail: jhbaek@kaeri.re.kr [Korea Atomic Energy Research Institute, Daejeon 305-353 (Korea, Republic of); Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Byun, Thak Sang [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Maloy, Start A. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Toloczko, Mychailo B. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States)

    2014-01-15

    The temperature dependence of fracture toughness in HT9 steel irradiated to 3–145 dpa at 380–503 °C was investigated using miniature three-point bend (TPB) fracture specimens. A miniature-specimen reuse technique has been established: the tested halves of subsize Charpy impact specimens with dimensions of 27 mm × 3 mm × 4 mm were reused for this fracture test campaign by cutting a notch with a diamond-saw in the middle of each half, and by fatigue-precracking to generate a sharp crack tip. It was confirmed that the fracture toughness of HT9 steel in the dose range depends more strongly on the irradiation temperature than the irradiation dose. At an irradiation temperature <430 °C, the fracture toughness of irradiated HT9 increased with the test temperature, reached an upper shelf of 180–200MPa√(m) at 350–450 °C, and then decreased with the test temperature. At an irradiation temperature ⩾430 °C, the fracture toughness was nearly unchanged up to about 450 °C and decreased slowly with test temperatures in a higher temperature range. Such a rather monotonic test temperature dependence after high-temperature irradiation is similar to that observed for an archive material generally showing a higher degree of toughness. A brittle fracture without stable crack growth occurred in only a few specimens with relatively lower irradiation and test temperatures. In this discussion, these TPB fracture toughness data are compared with previously published data from 12.7 mm diameter disc compact tension (DCT) specimens.

  1. Investigation of temperature dependence of fracture toughness in high-dose HT9 steel using small-specimen reuse technique

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong-Hyuk; Byun, Thak Sang; Maloy, Stuart A.; Toloczko, Mychailo B.

    2014-01-01

    The temperature dependence of fracture toughness in HT9 steel irradiated to 3–145 dpa at 380–503 degrees*C was investigated using miniature three-point bend (TPB) fracture specimens. A miniature-specimen reuse technique has been established: the tested halves of subsize Charpy impact specimens with dimensions of 27 mm *3mm* 4 mm were reused for this fracture test campaign by cutting a notch with a diamond-saw in the middle of each half, and by fatigue-precracking to generate a sharp crack tip. It was confirmed that the fracture toughness of HT9 steel in the dose range depends more strongly on the irradiation temperature than the irradiation dose. At an irradiation temperature <430 *degreesC, the fracture toughness of irradiated HT9 increased with the test temperature, reached an upper shelf of 180—200 MPa*m^.5 at 350–450 degrees*C, and then decreased with the test temperature. At an irradiation temperature >430 degrees*C, the fracture toughness was nearly unchanged up to about 450 *degreesC and decreased slowly with test temperatures in a higher temperature range. Such a rather monotonic test temperature dependence after high-temperature irradiation is similar to that observed for an archive material generally showing a higher degree of toughness. A brittle fracture without stable crack growth occurred in only a few specimens with relatively lower irradiation and test temperatures. In this discussion, these TPB fracture toughness data are compared with previously published data from 12.7 mm diameter disc compact tension (DCT) specimens.

  2. The evaluation of tempered martensite embrittlement in 4130 steel by instrumented charpy V-notch testing

    Science.gov (United States)

    Zia-Ebrahimi, F.; Krauss, G.

    1983-06-01

    Tempered martensite embrittlement (TME) was studied in vacuum-melted 4130 steel with either 0.002 or 0.02 wt pct P. TME was observed as a severe decrease in Charpy V-notch impact energy, from 46 ft-lb. at 200 °C to 35 ft-lb. at 300 °C in the low P alloy. The impact energy of the high P alloy was consistently lower than that of the low P alloy in all tempered conditions. Fracture was transgranular for all specimens; therefore, segregation of P to the prior austenitic grain boundaries was not a factor in the o°Currence of TME. Analysis of load-time curves obtained by instrumented Charpy testing revealed that the embrittlement is associated with a drop in the pre-maximum-load and post-unstable-fracture energies. In specimens tempered at 400 °C the deleterious effect of phosphorus on impact energy became pronounced, a result more consistent with classical temper embrittlement rather than TME. A constant decrease in pre-maximum-load energy due to phosphorus content was observed. The pre-maximum-load energy decreases with increasing tempering temperature in the range of 200 °C to 400 °C, a result explained by the change in work hardening rate. Carbon extraction replicas of polished and etched as-quenched specimens revealed the presence of Fe2MoC and/or Fe3C carbides retained after austenitizing. Ductile crack extension close to the notch root was related to the formation of fine micro voids at the retained carbides.

  3. Embrittlement of Cr-Mo steels after low fluence irradiation in HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Alexander, D.J.

    1995-04-01

    The goal of this work is the determination of the possible effect of the simultaneous formation of helium and displacement damage during irradiation on the Charpy impact behavior. Subsize Charpy impact specimens of 9Cr-1MoVNb (modified 9Cr-1Mo) and 12Cr-1MoVW (Sandvik HT9) steels and 12Cr-1MoVW with 2%Ni (12Cr-1MOVW-2Ni) were irradiated in the High Flux Isotope Reactor (HFIR) at 300 and 400{degree}C to damage levels up to 2.5 dpa. The objective was to study the effect of the simultaneous formation of displacement damage and transmutation helium on impact toghness. Despite the low fluence relative to previous irradiations of these steels, significant increases in the ductile-brittle transition temperature (DBTT) occurred. The 12Cr-1MoVW-2Ni steel irradiated at 400{degree}C had the largest increase in DBTT and displayed indications of intergranular fracture. A mechanism is proposed to explain how helium can affect the fracture behaviour of this latter steel in the present tests, and how it affected all three steels in previous experiments, where the steels were irradiated to higher fluences.

  4. Isolation and Irradiation-Modification of Lignin Specimens from Black Liquor and Evaluation of Their Effects on Wastewater Purification

    Directory of Open Access Journals (Sweden)

    Ke-Qin Wang

    2014-09-01

    Full Text Available In this study, crude lignin extracted from the black liquor generated by a pulp and paper mill was modified by different doses of irradiation. The crude and irradiation-modified lignins were used to treat wastewater that was generated during the production of starch glucoamylase. Changes to the physical and chemical properties and structure of the irradiation-modified lignins were determined using scanning electron microscopy, solubility analysis, elemental analysis, analysis of phenolic hydroxyl group, ultraviolet–visible spectroscopy, and Fourier transform infrared spectroscopy. Irradiation reduced the phenolic hydroxyl content in the lignin but increased its solubility by about 40%; analysis revealed that irradiation also destroyed the skeletal structure of the benzene ring in the lignin. After four minutes of settling, the total nitrogen (TN and chemical oxygen demand (COD in the wastewater reached 7.0 mg/L and 1573.1 mg/L, respectively. The settled solids content and protein recovery were 1.12 g/L and 98%, respectively. This study suggested that irradiation-modified lignin extracted from black liquor generated in the pulp and paper industry can be used to treat wastewater from the production of starch glucoamylase.

  5. Subsize specimen testing of nuclear reactor pressure vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A.S. [Missouri Univ., Rolla, MO (United States). Materials Research Center; Rosinski, S.T. [Sandia National Labs., Albuquerque, NM (United States); Cannon, N.S. [Westinghouse Hanford Co., Richland, WA (United States); Hamilton, M.L. [Pacific Northwest Lab., Richland, WA (United States)

    1991-12-31

    A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, A533B. The methodology appears to be more satisfactory than the methodologies proposed earlier. USE of a notched-only specimen is partitioned into macro-crack initiation and crack propagation energies. USE of a notched and precracked specimen provides the crack propagation energy. {Delta}USE, the difference between the USE`s of notched-only and precracked specimens, is an estimate of the crack initiation energy. {Delta}USE was normalized by a factor involving the dimensions of the Charpy specimen and the stress concentration factor at the notch root. The normalized values of the {Delta}USE were found to be invariant with specimen size.

  6. Subsize specimen testing of nuclear reactor pressure vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A.S. (Missouri Univ., Rolla, MO (United States). Materials Research Center); Rosinski, S.T. (Sandia National Labs., Albuquerque, NM (United States)); Cannon, N.S. (Westinghouse Hanford Co., Richland, WA (United States)); Hamilton, M.L. (Pacific Northwest Lab., Richland, WA (United States))

    1991-01-01

    A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, A533B. The methodology appears to be more satisfactory than the methodologies proposed earlier. USE of a notched-only specimen is partitioned into macro-crack initiation and crack propagation energies. USE of a notched and precracked specimen provides the crack propagation energy. [Delta]USE, the difference between the USE's of notched-only and precracked specimens, is an estimate of the crack initiation energy. [Delta]USE was normalized by a factor involving the dimensions of the Charpy specimen and the stress concentration factor at the notch root. The normalized values of the [Delta]USE were found to be invariant with specimen size.

  7. Fabrication Control Plan for ORNL RH-LOCA ATF Test Specimens to be Irradiated in the ATR

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Howard, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Teague, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-06-01

    The purpose of this fabrication plan is (1) to summarize the design of a set of rodlets that will be fabricated and then irradiated in the Advanced Test Reactor (ATR) and (2) provide requirements for fabrication and acceptance criteria for inspections of the Light Water Reactor (LWR) – Accident Tolerant Fuels (ATF) rodlet components. The functional and operational (F&OR) requirements for the ATF program are identified in the ATF Test Plan. The scope of this document only covers fabrication and inspections of rodlet components detailed in drawings 604496 and 604497. It does not cover the assembly of these items to form a completed test irradiation assembly or the inspection of the final assembly, which will be included in a separate INL final test assembly specification/inspection document. The controls support the requirements that the test irradiations must be performed safely and that subsequent examinations must provide valid results.

  8. High-dose neutron irradiation embrittlement of RAFM steels

    Energy Technology Data Exchange (ETDEWEB)

    Gaganidze, E. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, P.O. Box 3640, 76021 Karlsruhe (Germany)]. E-mail: ermile.gaganidze@imf.fzk.de; Schneider, H.-C. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, P.O. Box 3640, 76021 Karlsruhe (Germany); Dafferner, B. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, P.O. Box 3640, 76021 Karlsruhe (Germany); Aktaa, J. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2006-09-01

    Neutron irradiation-induced embrittlement of the reduced-activation ferritic/martensitic (RAFM) steel EUROFER97 was studied under different heat treatment conditions. Irradiation was performed in the Petten High Flux Reactor within the HFR Phase-IIb (SPICE) irradiation project up to 16.3 dpa and at different irradiation temperatures (250-450 deg. C). Several reference RAFM steels (F82H-mod, OPTIFER-Ia, GA3X and MANET-I) were also irradiated at selected temperatures. The impact properties were investigated by instrumented Charpy-V tests with subsize specimens. Embrittlement and hardening of as-delivered EUROFER97 steel are comparable to those of reference steels. Heat treatment of EUROFER97 at a higher austenitizing temperature substantially improves the embrittlement behaviour at low irradiation temperatures. Analysis of embrittlement in terms of the parameter C = {delta}DBTT/{delta}{sigma} indicates hardening-dominated embrittlement at irradiation temperatures below 350 deg. C with 0.17 {<=} C {<=} 0.53 deg. C/MPa. Scattering of C at irradiation temperatures above 400 deg. C indicates no hardening embrittlement.

  9. High-dose neutron irradiation embrittlement of RAFM steels

    Science.gov (United States)

    Gaganidze, E.; Schneider, H.-C.; Dafferner, B.; Aktaa, J.

    2006-09-01

    Neutron irradiation-induced embrittlement of the reduced-activation ferritic/martensitic (RAFM) steel EUROFER97 was studied under different heat treatment conditions. Irradiation was performed in the Petten High Flux Reactor within the HFR Phase-IIb (SPICE) irradiation project up to 16.3 dpa and at different irradiation temperatures (250-450 °C). Several reference RAFM steels (F82H-mod, OPTIFER-Ia, GA3X and MANET-I) were also irradiated at selected temperatures. The impact properties were investigated by instrumented Charpy-V tests with subsize specimens. Embrittlement and hardening of as-delivered EUROFER97 steel are comparable to those of reference steels. Heat treatment of EUROFER97 at a higher austenitizing temperature substantially improves the embrittlement behaviour at low irradiation temperatures. Analysis of embrittlement in terms of the parameter C = ΔDBTT/Δ σ indicates hardening-dominated embrittlement at irradiation temperatures below 350 °C with 0.17 ⩽ C ⩽ 0.53 °C/MPa. Scattering of C at irradiation temperatures above 400 °C indicates no hardening embrittlement.

  10. Room-temperature fracture in V-(4-5)Cr-(4-5)Ti tensile specimens irradiated in Fusion-1 BOR-60 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Gazda, J.; Meshii, M. [Northwestern Univ., Evanston, IL (United States); Tsai, H. [Argonne National Lab., IL (United States)

    1998-09-01

    Specimens of V-(4-5)Cr-(4-5)Ti alloys were irradiated to {approx}18 dpa at 320 C in the Fusion-1 capsule inserted into the BOR-60 reactor. Tensile tests at 23 C indicated dramatic yield strength increase (>300%), lack of work hardening, and minimal (<1%) total elongations. SEM analysis of fracture and side surfaces were conducted to determine reduction in are and the mode of fracture. The reduction of area was negligible. All but one specimen failed by a combination of ductile shear deformation and cleavage crack growth. Transgranular cleavage cracks were initiated by stress concentrations at the tips of the shear bands. In side-view observations, evidence was found of slip bands typically associated with dislocation channeling. No differences due to pre-irradiation heat treatment and heat-to-heat composition variations were detected. The only deviation from this behavior was found in V-4Cr-4Ti-B alloy, which failed in the grip portion by complete cleavage cracking.

  11. Embrittlement behaviour of different international low activation alloys after neutron irradiation

    Science.gov (United States)

    Schneider, H.-C.; Dafferner, B.; Aktaa, J.

    2001-05-01

    The embrittlement behaviour of ferritic/martensitic steels after irradiation in the Petten high flux reactor (HFR) was investigated by instrumented Charpy-V tests with subsize specimens. The main objective, apart from studying effects of particularly low doses, was a comparison of low activation alloys (LAA) from various countries with different Cr contents and different types and concentrations of minor alloying elements and impurities. In the present report, the results of another three materials (OPTIMAR, OPTIFER-IV, GA3X) obtained within the second phase of the MANITU programme (0.8 dpa, at 250-450°C) were analysed and assessed in comparison to the results of the first irradiation up to 0.8 dpa. The evaluation clearly showed a reduced embrittlement problem for the advanced reduced-activation alloys. Of the examined alloys, the GA3X steel shows the very best embrittlement behaviour after neutron irradiation.

  12. Irradiation behavior of a submerged arc welding material with different copper content; Bestrahlungsverhalten einer UP-Versuchsschweissnaht mit unterschiedlichen Kupfergehalten

    Energy Technology Data Exchange (ETDEWEB)

    Langer, R. [Siemens AG Energieerzeugung KWU, Erlangen (Germany); Bartsch, R. [Kernkraftwerk Obrigheim GmbH (Germany)

    1998-11-01

    Che report presents results of an irradiation program on specimens of submerged arc weldings with copper contents of 0.14% up to 0.42% and a fluence up to 2.2E19 cm{sup -2} (E>1MeV). Unirradiated and irradiated tensile- Charpy-, K{sub lc}- and Pellini-specimens were tested of material with a copper content of 0.22%. On the other materials Charpy tests and tensile tests were performed. The irradiation of the specimens took place in the KWO - ``RPV, a PWR with low flux and in the VAK - RPV, a small BWR with high flux. - The irradiation induced embrittlemnt shows a copper dependence up to about 30%. The specimens with a copper content higher than 0.30% show no further embrittlement. Irradiation in different reactors with different flux (factor > 33) shows the same state of embrittlement. Determination of a K{sub lc}, T-curve with irradiated specimens is possible. The conservative of the RT{sub NDT} - concept could be confirmed by the results of Charpy-V, drop weight- and K{sub lc}-test results. [Deutsch] Zur zusaetzlichen Absicherung des KWO-RDB wurde Ende 1979 eine UP-Versuchsschweissnaht mit vergleichbarer chemischer Zusammensetzung und vergleibaren mechanisch-technologischen Werkstoffen im unbestrahlten Ausgangszustand wie die RDB Core-Rundnaht hergestellt. Teile der Naht wurden durch Verkupfern der Schweissdraehte auf unterschiedliche Gehalte von Cu=0,14% bis 0,42% eingestellt. Aus dieser Schweissverbindung wurden Proben im VAK und KWO-RDB bestrahlt. Im Rahmen der Aktivitaeten zur Absicherung des KWO-RDBs erfolgte 1995 die Pruefung der bestrahlten Proben. Die mechanisch technologischen Werkstoffwerte vor und nach Bestrahlung werden gegenuebergestellt und praesentiert. Mit dem Ergebnis wurde ein weiterer Nachweis fuer die Konservativitaet des RT{sub NDT}-Konzeptes erbracht. Es wurde nachgewiesen, dass fuer den untersuchten Bereich kein Dose-Rate Effekt bzw. Bestrahlungszeiteinfluss existiert. Fuer UP-Schweissungen mit den vorliegenden Fertigungsparametern und bei

  13. Effect of grain structure on Charpy impact behavior of copper

    Science.gov (United States)

    Liang, Ningning; Zhao, Yonghao; Wang, Jingtao; Zhu, Yuntian

    2017-03-01

    Nanostructured (NS) and ultrafine-grained (UFG) materials have high strength and relatively low ductility. Their toughness has not been comprehensively investigated. Here we report the Charpy impact behavior and the corresponding microstructural evolutions in UFG Cu with equi-axed and elongated grains which were prepared by equal channel angular pressing (ECAP) for 2 and 16 passes at room temperature. It is found that their impact toughness (48 J/cm2) is almost comparable to that of coarse grained (CG) Cu: 55 J/cm2. The high strain rate during the Charpy impact was found to enhance the strain hardening capability of the UFG Cu due to the suppression of dynamic dislocation recovery. The crack in the CG Cu was blunted by dislocation-slip mediated plastic deformation, while the cracks in the UFG Cu were formed at grain boundaries and triple junctions due to their limited plasticity. Near the crack surfaces the elongated grains in ECAP-2 sample were refined by recrystallization, while equi-axed grains in the ECAP-16 sample grew larger.

  14. Effect of grain structure on Charpy impact behavior of copper

    Science.gov (United States)

    Liang, Ningning; Zhao, Yonghao; Wang, Jingtao; Zhu, Yuntian

    2017-01-01

    Nanostructured (NS) and ultrafine-grained (UFG) materials have high strength and relatively low ductility. Their toughness has not been comprehensively investigated. Here we report the Charpy impact behavior and the corresponding microstructural evolutions in UFG Cu with equi-axed and elongated grains which were prepared by equal channel angular pressing (ECAP) for 2 and 16 passes at room temperature. It is found that their impact toughness (48 J/cm2) is almost comparable to that of coarse grained (CG) Cu: 55 J/cm2. The high strain rate during the Charpy impact was found to enhance the strain hardening capability of the UFG Cu due to the suppression of dynamic dislocation recovery. The crack in the CG Cu was blunted by dislocation-slip mediated plastic deformation, while the cracks in the UFG Cu were formed at grain boundaries and triple junctions due to their limited plasticity. Near the crack surfaces the elongated grains in ECAP-2 sample were refined by recrystallization, while equi-axed grains in the ECAP-16 sample grew larger. PMID:28303950

  15. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950`s are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  16. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950's are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  17. A Mechanistically-Guided Charpy Embrittlement Correlation For RPV (reactor pressure vessel) Integrity Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [EPRI, CHARLOTTE, NC (United States); Server, W.L. [ATI Consulting, Pinehurst, NC (United States)

    2002-07-01

    The current neutron irradiation embrittlement trend equation used in the Usa is contained in Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Revision 2. The equivalent equation for estimating the mean shift in irradiated Charpy properties is used also in ASTM Standard Guide E 900-87. The three chemistry and irradiation parameters in this old correlation are copper (Cu) content, nickel (Ni) content, and irradiation fluence; base and weld metals are separated also, due to the enhanced embrittlement in welds. The database used to establish this old correlation was compiled in the late 1980's. Today, the database has increased by a multiple of about 5. Through the EPRI Materials Reliability Program a new and improved transition temperature shift embrittlement correlation has been developed. The recommended model is mechanistically-guided, statistically robust, and stems from earlier work on a mechanistic/statistical model proposed by the NRC. From the independent reviews performed on the NRC proposed correlation, the evaluations of mechanistic understanding and statistical testing were combined to assess the most appropriate form for a mean correlation model. The process of evaluating and reducing the number of fitting parameters was not a simple decision. Engineering judgment, through the development of gating criteria and value/magnitude considerations, led to the development of the proposed correlation. This paper discusses the proposed embrittlement correlation and its mechanistic/statistical bases. Predictions using the proposed correlation are compared using the current version of the embrittlement database; comparisons with the predictions with Regulatory Guide 1.99, Revision 2, are also made. (authors)

  18. 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.

  19. HANARO instrumented capsule development for supporting a study on the irradiation damage of stainless steels for nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Y. H.; Cho, M. S.; Sohn, J. M.; Kim, H. R.; Lee, B. C.; Kim, K. H

    2000-10-01

    As a part of the program for the maximum utilization of HANARO by MOST, Korea, an instrumented capsule (00M-01U) was designed and fabricated for supporting a study on the irradiation damage of stainless steels for nuclear applications. The basic structure of the capsule for the irradiation of Stainless steels was based on that of the 99M-01K capsule irradiated successfully in HANARO. To satisfy the user requirements such as irradiation temperature and neutron fluence, the optimal arrangement of test specimens was done in the axial and circumferential direction. The temperature distribution and thermal stress of a capsule with multi-holes were obtained by a finite element analysis code, ANSYS. From these analyzed data, this capsule was found to be compatible with HANARO design requirement. Various types of specimens such as small tensile, Charpy, TEM and EPMA specimens were inserted in the capsule. The specimens will be irradiated in the IR2 test hole of HANARO at 288, 300 and 350 deg C up to a fast neutron fluence of 1.0x10{sup 20}(n/cm{sup 2})(E>1.0MeV)

  20. Impact energy analysis of HSLA specimens after simulated welding thermal cycle

    Directory of Open Access Journals (Sweden)

    Samarždić, I.

    2008-04-01

    Full Text Available This paper presents impact energy results of specimens made from high strength fine grained steel TStE 420 after thermal cycle simulation. These results are obtained by examining Charpy specimens. Metallographic analysis is performed, hardness is measured and total impact energy is divided into ductile and brittle components.

  1. Irradiation programme HFR phase 1b. Results of impact tests with the irradiated materials (2.4 dpa); Bestrahlungsprogramm HFR Phase 1b. Ergebnisse der Kerbschlagbiegeversuche mit den bis 2,4 dpa bestrahlten Werkstoffen

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, H.C.; Dafferner, B.; Ries, H.; Lautensack, S.; Romer, O.

    2004-04-01

    The irradiation project HFR phase 1b was planned and carried out in the frame of the European Long-term Fusion Materials Development Programme. It represents the continuation of the former High-Flux-Reactor irradiation programmes which are documented in detail in former FZK-reports. The problem of the irradiation induced embrittlement of possible martensitic alloy candidates is still not finally solved. After evaluation of sub-size Charpy tests with the 2.4 dpa irradiated specimens, the low activation 7-10%-Cr-W(Ge)VTa alloys showed outstandingly better characteristics compared to the modified commercial 10-11%-Cr-NiMoVNb steels. The emphasis of the phase 1b project is now on the investigation of low activating OPTIFER-alloys, which exhibit in contrast to the previously irradiated OPTIFER-steels a clearly reduced boron content in order to reduce the irradiation embrittlement. In the present report the results of instrumented impact tests within the phase 1b programme (irradiation dose 2.4 dpa, irradiation temperatures 250, 300, 350, 400, and 450 C) are analysed and assessed in comparison to the results of the MANITU- and HFR phase 1a- irradiation programmes. Herein, the OPTIFER-V steel, which is nearly identical to the industrial heat EUROFER97 (which became available after the irradiation), supplied the best results. (orig.)

  2. Irradiation embrittlement of reactor pressure vessel steel outside the astm specification A508 CL2

    Science.gov (United States)

    Pachur, D.; Krawczynski, S. J.; Derz, H.; Pott, G.

    1990-04-01

    Radiation embrittlement of reactor pressure vessel steels is of considerable significance for safety engineering. Steel manufacturers must therefore comply with specifications defined by national design codes. The extent to which a steel deviating from the specification is influenced by irradiation is being examined under the German Research Programme on the Integrity of Reactor Components. Charpy-V specimens were taken from a forged steel block longitudinally and vertically to the direction of main deformation and irradiated in the FRJ-1 research reactor at a temperature of 288 °C corresponding to the operating temperature of power reactors. The neutron fluences obtained ranged between 0.8 × 10 19 and 8 × 10 19n/ cm2. Instrumented pendulum impact tests have been evaluated and the load signals measured were analysed, fitting and calculating transition temperature curves and trend curves.

  3. Mechanical properties of neutron-irradiated nickel-containing martensitic steels: I. Experimental study

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L. [Oak Ridge National Laboratory, P.O. Box 2008, MS 6151, Oak Ridge, Tennessee 37831-6151 (United States)]. E-mail: kluehrl@ornl.gov; Hashimoto, N. [Oak Ridge National Laboratory, P.O. Box 2008, MS 6151, Oak Ridge, Tennessee 37831-6151 (United States); Sokolov, M.A. [Oak Ridge National Laboratory, P.O. Box 2008, MS 6151, Oak Ridge, Tennessee 37831-6151 (United States); Shiba, K. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Tokai, Ibaraki 319-1195 (Japan); Jitsukawa, S. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Tokai, Ibaraki 319-1195 (Japan)

    2006-10-15

    Tensile and Charpy specimens of 9Cr-1MoVNb (modified 9Cr-1Mo) and 12Cr-1MoVW (Sandvik HT9) steels and these steels doped with 2% Ni were irradiated at 300 and 400 deg. C in the High Flux Isotope Reactor (HFIR) up to {approx}12 dpa and at 393 deg. C in the Fast Flux Test Facility (FFTF) to {approx}15 dpa. In HFIR, a mixed-spectrum reactor (n, {alpha}) reactions of thermal neutrons with {sup 58}Ni produce helium in the steels. Little helium is produced during irradiation in FFTF. After HFIR irradiation, the yield stress of all steels increased, with the largest increases occurring for nickel-doped steels. The ductile-brittle transition temperature (DBTT) increased up to two times and 1.7 times more in steels with 2% Ni than in those without the nickel addition after HFIR irradiation at 300 and 400 deg. C, respectively. Much smaller differences occurred between these steels after irradiation in FFTF. The DBTT increases for steels with 2% Ni after HFIR irradiation were 2-4 times greater than after FFTF irradiation. Results indicated there was hardening due to helium in addition to hardening by displacement damage and irradiation-induced precipitation.

  4. MACHINING TEST SPECIMENS FROM HARVESTED ZION RPV SEGMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Nanstad, Randy K [ORNL; Rosseel, Thomas M [ORNL; Sokolov, Mikhail A [ORNL

    2015-01-01

    The decommissioning of the Zion Nuclear Generating Station (NGS) in Zion, Illinois, presents a special and timely opportunity for developing a better understanding of materials degradation and other issues associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, is coordinating and contracting with Zion Solutions, LLC, a subsidiary of Energy Solutions, an international nuclear services company, the selective procurement of materials, structures, components, and other items of interest from the decommissioned reactors. In this paper, we will discuss the acquisition of segments of the Zion Unit 2 Reactor Pressure Vessel (RPV), cutting these segments into blocks from the beltline and upper vertical welds and plate material and machining those blocks into mechanical (Charpy, compact tension, and tensile) test specimens and coupons for microstructural (TEM, SEM, APT, SANS and nano indention) characterization. Access to service-irradiated RPV welds and plate sections will allow through wall attenuation studies to be performed, which will be used to assess current radiation damage models [1].

  5. Fracture toughness of irradiated modified 9Cr-lMo steel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.H.; Yoon, J.H.; Ryu, W.S.; Lee, C.B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Hong, J.H. [KAERI - Korea Atomic Energy Research Institute, Nuclear Materials Technology Development Div., Daejon (Korea, Republic of)

    2007-07-01

    Full text of publication follows: Ferritic/martensitic steels have been used for a long time in the power generation industry as boiler and turbine materials. These steels are the proposed candidates for the crosscutting materials of the advanced nuclear power system. It is important to realize the change of mechanical properties by neutron irradiation for application these materials to nuclear power system. Irradiation effect on the fracture toughness of the structural materials is one of the concerns for the designing of the fusion devices. The test material was a 16 mm thick commercial Modified 9Cr-1Mo plate which was normalized at 1050 deg. C and tempered at 770 deg. C. The half sized pre-cracked Charpy specimens were irradiated at CT test hole in HANARO. Irradiation test was conducted at 340 deg. C and 400 deg. C to investigate the irradiation temperature effect on the degradation of the fracture toughness. And the irradiation fluence was 1.2x10{sup 21} n/cm{sup 2} (E>0.1 MeV). Toughness tests for the irradiated specimens will be performed in the hot cell at KAERI. The fracture toughness of the unirradiated condition was carried out in order to assess the changes in the materials properties caused by neutron irradiation. The K{sub JC} values in accordance at the ASTM E1921- 05 standard were obtained by three-point bending tests. Tests have been carried out at several temperatures within transition region. The multi-temperature method was used to determine reference temperature, T{sub o}. The applicability of the Master Curve method for irradiated and unirradiated ferritic/martensitic steel is another focus of this study. The reference temperature of the unirradiated specimen was -72.4 deg. C. And the Master Curve successfully expressed the trend of the fracture toughness change with temperature for unirradiated Modified 9Cr-1Mo steel. (authors)

  6. Subtask 12H1: Vanadium alloy irradiation experiment X530 in EBR-II

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Strain, R.V.; Hins, A.G.; Chung, H.M.; Nowicki, L.J.; Smith, D.L. [Argonne National Lab., IL (United States)

    1995-03-01

    The objective of the X530 experiment in EBR-II was to obtain early irradiation performance data, particularly the fracture properties, on the new 500-kg production heat of V-4Cr-4Ti material before the scheduled reactor shutdown at the end of September 1994. To obtain early irradiation performance data on the new 500-kg production heat of the V-4Cr-4Ti material before the scheduled EBR-II shutdown, an experiment, X530, was expeditiously designed and assembled. Charpy, compact tension, tensile and TEM specimens with different thermal mechanical treatments (TMTs), were enclosed in two capsules and irradiated in the last run of EBR-II, Run 170, from August 9 through September 27. For comparison, specimens from some of the previous heats were also included in the test. The accrued exposure was 35 effective full power days, yielding a peak damage of {approx}4 dpa in the specimens. The irradiation is now complete and the vehicle is awaiting to be discharged from EBR-II for postirradiation disassembly. 4 figs., 2 tabs.

  7. Impact energy analysis of quenched and tempered fine grain structural steel specimens after weld thermal cycle simulation

    Directory of Open Access Journals (Sweden)

    M. Dunđer

    2014-10-01

    Full Text Available The paper presents impact energy results of thermal cycle simulated specimens of quenched and tempered fine grain structural steel S960QL. These results are obtained by examining notched Charpy specimens. Upon performed metallographic analysis and measured hardness, total impact energy is separated into ductile and brittle components.

  8. Confocal microscopy-fracture reconstruction and finite element modeling characterization of local cleavage toughness in a ferritic/martensitic steel in subsized Charpy V-notch impact tests

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, T. E-mail: yamataku@fusion.imr.tohoku.ac.jp; Odette, G.R.; Lucas, G.E.; Matsui, H

    2000-12-01

    The confocal microscopy (CM)-fracture reconstruction (FR) method, coupled with scanning electron microscopy (SEM) fractography, was used to measure the critical notch deformation conditions at cleavage initiation for two subsized Charpy V-notch (CVN) specimen geometries of Japan ferritic/martensitic steel (JFMS). A new method was developed to permit FR of notched specimens. Three-dimensional finite element analysis (FEA) simulations of the notch and specimen deformation were used to estimate values of critical micro-cleavage fracture stress, {sigma}{sup *}, and critical stressed area, A{sup *}. Since {sigma}{sup *}-A{sup *} is independent of size and geometry, it provides a fundamental local measure of cleavage toughness.

  9. Irradiation programme MANITU. Results of impact tests with the irradiated materials of the second irradiation phase (0.8 dpa); Bestrahlungsprogramm MANITU. Ergebnisse der Kerbschlagbiegeversuche mit den bis 0,8 dpa bestrahlten Werkstoffen der zweiten Bestrahlungsphase

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, H.C.; Rieth, M.; Dafferner, B.; Ries, H.; Romer, O.

    2000-09-01

    The irradiation project MANITU was planned and carried out in the frame of the European Longterm Fusion Materials Development Programme. The problem of the irradiation induced embrittlement of possible martensitic alloy candidates is still unsolved. But after the evaluation of sub-size Charpy tests with the unirradiated reference specimens of MANITU a first tendency was recognizable. The mechanical properties of the newly developed low activation 7-10% Cr-W(Ge)VTa alloys are partly better compared to the modified commercial 10-11% Cr-NiMoVNb steels. After the evaluation of subsize Charpy tests with specimens irradiated up to 0.2 and 0.8 dpa in the first phase of the MANITU programme, better mechanical properties of the 7-10% Cr-W(Ge)VTa alloys were obvious. In the present report the results of instrumented impact tests within the second phase of the MANITU programme (irradiation dose 0.8 dpa, irradiation temperatures 250, 300, 350, 400, and 450 C) are analysed and assessed in comparison to the results of the irradiation up to 0.8 dpa in the first phase of the project. Among the examined alloys (OPTIMAR, OPTIFER-IV, GA3X) the GA3X steel shows the very best embrittlement behaviour after neutron irradiation. (orig.) [German] Das Bestrahlungsprojekt MANITU wurde im Rahmen des europaeischen Langzeitprogramms fuer Materialentwicklung fuer die Kernfusion geplant und durchgefuehrt. Das Problem der bestrahlungsinduzierten Versproedung bei den in Frage kommenden martensitischen Werkstoffen ist nach wie vor ungeloest. Eine erste Tendenz zeichnete sich jedoch nach der Auswertung der Kerbschlagbiegeversuche an den unbestrahlten miniaturisierten Referenzproben des MANITU-Programms ab. Die neu entwickelten niedrig aktivierbaren 7-10% Cr-W(Ge)VTa-Legierungen weisen gegenueber den modifizierten kommerziellen 10-11% Cr-NiMoVNb-Staehlen teilweise bessere mechanische Eigenschaften auf. Nach Auswertung der Kerbschlagbiegeversuche an den in der ersten Phase des MANITU-Programms mit 0

  10. Hardness and microstructural response to thermal annealing of irradiated ASTM A533B class 1 plate steel

    Energy Technology Data Exchange (ETDEWEB)

    Reinhart, D.E. [SMS Concast, Inc., Pittsburgh, PA (United States); Kumar, A.S. [Univ. of Missouri, Rolla, MO (United States); Gelles, D.S.; Hamilton, M.L. [Pacific Northwest Lab., Richland, WA (United States); Rosinski, S.T. [Electric Power Research Inst., Charlotte, NC (United States)

    1999-10-01

    Hardness measurements were used to determine the post-irradiation annealing response of A533B class 1 plate steel irradiated to a fluence of 1 {times} 10{sup 19} n/cm{sup 2} (E > 1 MeV) at 150 C. Rockwell hardness measurements indicated that the material had hardened by 6.6 points on the B scale after irradiation. The irradiation induced hardness increase was associated with a decrease in upper shelf energy from 63.4 J to 5-1.8 J and a temperature shift in the Charpy curve at the 41 J level from 115 C to 215 C. Specimens were annealed after irradiation at temperatures of 343 C (650 F), 399 C (750 F), and 454 C (850 F) for durations of up to one week (168 h). Hardness measurements were made to chart recovery of hardness as a function of time and temperature. Specimens annealed at the highest temperature 454 C recovered the fastest, fully recovering within 144 h. Specimens annealed at 399 C recovered completely within 168 h. Specimens annealed at the lowest temperature, 343 C recovered only {approximately}70% after 168 h of annealing. After neutron irradiation, a new feature of black spot damage was found to be superimposed on the unirradiated microstructure. The density of black spots was found to vary from 2.3 {times} 10{sup 15}/cm{sup 3} to 1.1 {times} 10{sup 16}/cm{sup 3} with an average diameter of 2.85 nm. Following annealing at 454 C for 24 h the black spot damage was completely annealed out. It was concluded that the black spot damage was responsible for 70% of the irradiation-induced hardness.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-09-15

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

  12. Comparison of Impact Duration Between Experiment and Theory From Charpy Impact Test

    Directory of Open Access Journals (Sweden)

    Muhammad Said N.B.

    2016-01-01

    Full Text Available This study presents the comparison of impact duration between experiment and theory from impact signal through a Charpy test. Recently, the number of accidents on the highway has been increased and it depends on the impact duration of material that have the ability to provide adequate protection to passengers from harmful and improve occupant survivability during crash event. Charpy impact test was implemented on different material and thickness but at the same striker velocity. Impact signal is obtained through the strain gauge that has been installed to striker hammer and connected to frequency data acquisition system. Collected signal is then analysed to identify the time period during impact before fractured. Result from both experiment and theory shows an increment to the impact duration as thickness is increased. Charpy test shows that aluminium 6061-T6 has a higher impact duration compared to carbon steel 1050.

  13. Failure Behaviors Depending on the Notch Location of the Impact Test Specimens on the HAZ

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Yun Chan; Kim, Dong Wook; Lee, Young Suk [Chungang Univ., Seoul (Korea, Republic of); Hong, Jae Keun; Park, Ji Hong [Korea Institute of Machinery and Materials, Daejeon (Korea, Republic of)

    2007-07-01

    Numerical studies were performed to examine the effects of notch location of impact specimens on the failure behavior of HAZ (Heat Affected Zone) when Charpy V-notch impact test were made at a low temperature (1 .deg. C). Carbon steel plate (SA-516 Gr. 70) with thickness of 25mm for pressure vessel was welded by SMAW (Shielded Metal-Arc Welding) and specimens were fabricated from the welded plate. Charpy tests were then performed with specimens having different notch positions of specimens varying from the fusion line through HAZ to base metal. A series of finite element analysis which simulates the Charpy test and crack propagation initiating at the tip of V-notch was carried out as well. The finite element analysis takes into account the irregular fusion line and non-homogenous material properties due to the notch location of the specimen in HAZ. Results reveals that the energies absorbed during impact test depend significantly on the notch location and direction of specimen. Finite element analysis also demonstrates that the notch location of specimens, to a great extent, influences the reliability and consistency of the test.

  14. Charpy Impact Response of the Cracked Aluminum Plates Repaired with FML Patches using the Response Surface Methodology

    Directory of Open Access Journals (Sweden)

    Faramarz Ashenai Ghasemi

    2016-09-01

    Full Text Available Here, the effect of fiber metal laminate (FMLs patches was studied for repairing of single-sided cracked aluminum plates experimentally to see their response to Charpy impact tests. The main desired parameters were composite patch lay-up, crack length, and crack angle each one in three levels. All experimental attempts generated and followed based on the design of experiments method by using of response surface methodology. The predicted energy absorption values obtained from the model were in good agreement with the experimental results. No matter the specimens were repaired or not, as the crack length was increased the energy absorption of the structure was decreased. The experimental results also showed that for lengthen cracks, increasing of the crack angle had more effect on energy absorption. Also it was observed that the patch lay-up effective on the impact response of the specimens. The more the metal layer was departed from the aluminum plate and the FML patches interfacial surface, the less energy was absorbed in the structure.

  15. Numerical study of the Notch Location of the Impact Test Specimens on the HAZ of SA516 Steels

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Yun Chan; Kim, Dong Wook; Lee, Young Seog [Chungang Univ., Seoul (Korea, Republic of); Hong, Jae Keun; Park, Ji Hong [Korea Institute of Machinery and Materials, Daejeon (Korea, Republic of)

    2007-07-01

    Experimental and numerical studies were performed to examine the effects of notch position on the failure behavior and energy absorption when the Charpy V-notch impact test is made at 1 .deg. C. For this purpose, carbon steel plate (SA-516 Gr. 70) with thickness of 25mm usually used for pressure vessel was welded by SMAW (Shielded Metal-Arc Welding) method and specimens were fabricated from the welded plate. The Charpy impact tests were then performed with specimens having different notch positions varying within HAZ. A series of three-dimensional FE analysis which simulates the Charpy test and crack propagation are carried out as well to examine the reproducibility of test results. The FE analysis takes into account the heterogeneous mechanical properties with complex microstructures in HAZ. Results reveal that the absorbed energies during impact test depend significantly on the notch position.

  16. Instrumentação de um pendulo para ensaio de impacto Charpy

    OpenAIRE

    1994-01-01

    Resumo: O ensaio de impacto Charpy convencional, possibilita que se determine a energia total ('E IND. T') absorvida para fratura de um corpo de prova entalhado, de dimensões padronizadas. Entretanto, sabe-se que essa energia total, resulta da soma de duas componentes: a energia necessária para iniciar a fratura ('E IND.I') mais a energia necessária para a propagação dessa fratura ('E IND. P'). No presente trabalho, é instrumentado um pêndulo de ensaio de impacto Charpy, utilizando-se técnica...

  17. Evaluation of neutron irradiation embrittlement in the Korean reactor pressure vessel steels(I) (1st progress report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jun Hwa; Lee, Bong Sang; Park, Duck Gun; Byun, Tak Sang; Kim, Joo Hag; Oh, Yong Jun; Yoon, Ji Hyun; Chi, Sei Hwan; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    The SA508-3 reactor pressure vessel materials degrade due to the application at high temperature, high pressure, and neutron irradiation. In the present study it is planned to examine the effects of neutron irradiation on the properties for assessing the integrity of domestic reactors. The key tests are the Charpy impact test, tensile test, static and dynamic fracture toughness test, J-R test. The additional tests for obtaining basic material properties, such as micro-hardness, microstructural properties, small punch energy etc., are also performed. The irradiation tests are being performed at HANARO of KAERI through the instrumented capsules designed by KAERI and the post-irradiation tests are being performed at IMEF(Irradiated Material Evaluation Facility) of material (UCN-4), Si+Al (YGN-5), UCN-4 weld metal, and UCN-4 HAZ. In the irradiation test the temperature should be controlled in the range of 290 {+-} 10 deg C and the test materials would be irradiated to 2 to 3 neutron fluence levels including the end-of-life fluence. The status of performing this project is that (1) the key data on mechanical properties, mainly related to the fracture toughness, of the unirradiated materials have been obtained, (2) the irradiation of the 1st instrumented capsule, a preliminary test capsule containing miniature specimens, has been completed and is being stored for testing in IMEF, and (3) the 2nd instrumented capsule is being manufactured and will be irradiated in the beginning or 1999. This report includes mainly the experimental methods and results. The status of the design and manufacturing of the instrumented capsules and specimens was also briefly described. (author). 13 refs., 15 figs., 10 tabs.

  18. Virtual Specimens

    Science.gov (United States)

    de Paor, D. G.

    2009-12-01

    Virtual Field Trips have been around almost as long as the Worldwide Web itself yet virtual explorers do not generally return to their desktops with folders full of virtual hand specimens. Collection of real specimens on fields trips for later analysis in the lab (or at least in the pub) has been an important part of classical field geoscience education and research for generations but concern for the landscape and for preservation of key outcrops from wanton destruction has lead to many restrictions. One of the author’s favorite outcrops was recently vandalized presumably by a geologist who felt the need to bash some of the world’s most spectacular buckle folds with a rock sledge. It is not surprising, therefore, that geologists sometimes leave fragile localities out of field trip itineraries. Once analyzed, most specimens repose in drawers or bins, never to be seen again. Some end up in teaching collections but recent pedagogical research shows that undergraduate students have difficulty relating specimens both to their collection location and ultimate provenance in the lithosphere. Virtual specimens can be created using 3D modeling software and imported into virtual globes such as Google Earth (GE) where, they may be linked to virtual field trip stops or restored to their source localities on the paleo-globe. Sensitive localities may be protected by placemark approximation. The GE application program interface (API) has a distinct advantage over the stand-alone GE application when it comes to viewing and manipulating virtual specimens. When instances of the virtual globe are embedded in web pages using the GE plug-in, Collada models of specimens can be manipulated with javascript controls residing in the enclosing HTML, permitting specimens to be magnified, rotated in 3D, and sliced. Associated analytical data may be linked into javascript and localities for comparison at various points on the globe referenced by ‘fetching’ KML. Virtual specimens open up

  19. Machining Test Specimens from Harvested Zion RPV Segments for Through Wall Attenuation Studies

    Energy Technology Data Exchange (ETDEWEB)

    Rosseel, Thomas M [ORNL; Sokolov, Mikhail A [ORNL; Nanstad, Randy K [ORNL

    2015-01-01

    The decommissioning of the Zion Units 1 and 2 Nuclear Generating Station (NGS) in Zion, Illinois presents a special opportunity for developing a better understanding of materials degradation and other issues associated with extending the lifetime of existing Nuclear Power Plants (NPPs) beyond 60 years of service. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, is coordinating and contracting with Zion Solutions, LLC, a subsidiary of Energy Solutions, the selective procurement of materials, structures, and components from the decommissioned reactors. In this paper, we will discuss the acquisition of segments of the Zion Unit 2 Reactor Pressure Vessel (RPV), the cutting of these segments into sections and blocks from the beltline and upper vertical welds and plate material, the current status of machining those blocks into mechanical (Charpy, compact tension, and tensile) test specimens and coupons for chemical and microstructural (TEM, APT, SANS, and nano indention) characterization, as well as the current test plans and possible collaborative projects. Access to service-irradiated RPV welds and plate sections will allow through wall attenuation studies to be performed, which will be used to assess current radiation damage models (Rosseel et al. (2012) and Rosseel et al. (2015)).

  20. Results of crack-arrest tests on two irradiated high-copper welds

    Energy Technology Data Exchange (ETDEWEB)

    Iskander, S.K.; Corwin, W.R.; Nanstead, R.K. (Oak Ridge National Lab., TN (USA))

    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 {times} 10{sup 19} neutrons/cm{sup 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{sub Ia} curve. 9 refs., 21 figs., 10 tabs.

  1. Application of Charpy Impact Absorbed Energy to the Safety Assessment Based on SINTAP

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    The European Structural Integrity Assessment Procedure(SINTAP) was applied to the assessment of welded joints of the APl 5L X65 pipeline steel with an assumed embedded flaw and surface flaw at the weld toe. At default level( level 0), the assessment point was established by estimating fracture toughness value KIc conservatively from Charpy energy test data. At the same time, the analysis level 1 (basic level)was applied based on the fracture toughness CTOD. Then the two assessment levels were compared. The assessment results show that all assessment points are located within the failure lines of analysis levels 0 and 1. So the welded joint of the pipeline is safe. It can be concluded that the assessment based on Charpy absorbed energy is practicable when other fracture toughness data are not available, or cannot be easily obtained. The results are conservative.

  2. A Calibration of the Wierzbicki-Xue Damage Model Using Charpy Test Results

    Directory of Open Access Journals (Sweden)

    Kim Jong-Bong

    2015-01-01

    Full Text Available Damage models are frequently used to predict fractures in large deformation problems such as penetration of a projectile into a target. Though many damage models have been proposed so far, coefficients of each model have been provided for only a few materials. In this study, the coefficients of the Wierzbicki-Xue (2005 damage model for tungsten heavy alloy (DX2HCMF are determined using the Charpy impact test. The Wierzbicki-Xue fracture criterion is implemented into NET3D code in which a node-split algorithm is built in. By comparing the energy absorbed in the Charpy test with the results of finite element analysis, the fracture model coefficients are determined.

  3. Apportion of Charpy energy in API 5L grade X70 pipeline steel

    Energy Technology Data Exchange (ETDEWEB)

    Hashemi, Sayyed H. [Department of Mechanical Engineering, The University of Birjand, PO Box 97175 /615, Birjand (Iran, Islamic Republic of)], E-mail: shhashemi@birjand.ac.ir

    2008-12-15

    Conventional Charpy based failure models for gas transportation pipelines recommend the minimum fracture energy for safe performance of these structures. In recent years however, full-scale burst experiments have shown that such models cannot fully guarantee the safety of higher grade pipeline steels. One possible reason for this discrepancy, which is further investigated in this research, is that Charpy energy inherently contains both fracture and non-fracture related energy. To separate this, energy partitioning analysis was used. First, the overall fracture energy of X70 steel is measured experimentally on an instrumented Charpy rig. Next, the measured energy is divided into fracture initiation and propagation parts using load-displacement data. It appeared from test results that a significant amount of energy was consumed in non-fracture related processes. From this, correction factors were suggested for possible use in current industry failure models. Interestingly, these corrections factors agreed well with those reported from full-thickness burst tests for tough pipeline steels.

  4. Instrumented Impact Testing: Influence of Machine Variables and Specimen Position

    Energy Technology Data Exchange (ETDEWEB)

    Lucon, E.; McCowan, C. N.; Santoyo, R. A.

    2008-09-15

    An investigation has been conducted on the influence of impact machine variables and specimen positioning on characteristic forces and absorbed energies from instrumented Charpy tests. Brittle and ductile fracture behavior has been investigated by testing NIST reference samples of low, high and super-high energy levels. Test machine variables included tightness of foundation, anvil and striker bolts, and the position of the center of percussion with respect to the center of strike. For specimen positioning, we tested samples which had been moved away or sideways with respect to the anvils. In order to assess the influence of the various factors, we compared mean values in the reference (unaltered) and altered conditions; for machine variables, t-test analyses were also performed in order to evaluate the statistical significance of the observed differences. Our results indicate that the only circumstance which resulted in variations larger than 5 percent for both brittle and ductile specimens is when the sample is not in contact with the anvils. These findings should be taken into account in future revisions of instrumented Charpy test standards.

  5. Irradiation embrittlement of neutron-irradiated ferritic steel

    Science.gov (United States)

    Kayano, H.; Narui, M.; Ohta, S.; Morozumi, S.

    1985-08-01

    In this study three kinds of Fe-Cr ferritic steels were examined by the instrumented Charpy test and tensile test before and after JMTR irradiation ( 2.2×10 23 f.n./m 2). In the unirradiated samples, 100%-martensite 5Cr-2Mo steel showed the highest adsorbed energy and the highest toughness at low temperatures, follewed by the 9Cr-2Mo steel, and the 20%-martensite 5Cr-2Mo steel showed the third highest toughness. In the irradiated samples, however, thoughness was low as a whole, especially in 20%-martensite 5Cr-2Mo steel. It was clarified that 100%-martensite 5Cr-2Mo steel had the lowest Ductile-to-Brittle Transition Temperature (DBTT) and the highest fracture toughness, and that its DBTT and fracture toughness changed a little upon irradiation, showing excellent irradiation characteristics. The general equations were considered for correlation among strength, ductillity, DBTT and fracture toughness ( J value)

  6. Irradiation programme MANITU. Results of impact tests with the irradiated materials of the first irradiation phase (0.8 dpa); Bestrahlungsprogramm MANITU. Ergebnisse der Kerbschlagbiegeversuche mit den bis 0,8 dpa bestrahlten Werkstoffen der ersten Bestrahlungsphase

    Energy Technology Data Exchange (ETDEWEB)

    Rieth, M.; Dafferner, B.; Ries, H.; Romer, O.

    1995-09-01

    The irradiation project MANITU was planned and carried out in the frame of the European Longterm Fusion Materials Development Programme. The problem of the irradiation induced embrittlement of possible martensitic alloy candidates is still unsolved. But after the evaluation of subsize Charpy tests with the unirradiated reference specimens of MANITU a first tendency was recognizable. The mechanical properties of the newly developed low activation 7-10% Cr-W(Ge)VTa alloys are partly better compared to the modified commercial 10-11% Cr-NiMoVNb steels. In the present report the results of instrumented impact tests within the first phase of the MANITU programme (irradiation dose 0.8 dpa, irradiation temperatures 250, 300, 350, 400, and 450 C) are analysed and assessed. Among all examined alloys (MANET-I, MANET-II, K-heat, OPTIFER-Ia, OPTIFER-II, F82H, 9Cr-2WVTa ORNL 3791) the ORNL steel shows the very best embrittlement behaviour after neutron irradiation. (orig.) [Deutsch] Das Bestrahlungsprojekt MANITU wurde im Rahmen des europaeischen Langzeitprogramms fuer Materialentwicklung fuer die Kernfusion geplant und durchgefuehrt. Das Problem der bestrahlungsinduzierten Versproedung bei den in Frage kommenden martensitischen Werkstoffen ist nach wie vor ungeloest. Eine erste Tendenz zeichnete sich jedoch nach der Auswertung der Kerbschlagbiegeversuche an den unbestrahlten miniaturisierten Referenzproben des MANITU-Programms ab. Die neu entwickelten niedrig aktivierbaren 7-10% Cr-W(Ge)VTa-Legierungen weisen gegenueber den modifizierten kommerziellen 10-11% Cr-NiMoVNb-Staehlen teilweise bessere mechanische Eigenschaften auf. Im vorliegenden Bericht werden die Ergebnisse aus den instrumentierten Kerbschlagbiegeversuchen der ersten Phase des MANITU-Programms (Bestrahlungsdosis 0,8 dpa, Bestrahlungstemperaturen 250, 300, 350, 400 und 450 C) analysiert und bewertet. Von den untersuchten Legierungen (MANET-I, MANET-II, Kastencharge, OPTIFER-Ia, OPTIFER-II, F82H, 9Cr-2WVTa ORNL 3791

  7. A New Analytical Expression for the Relationship Between the Charpy Impact Energy and Notch Tip Position for Functionally Graded Steels

    Institute of Scientific and Technical Information of China (English)

    H.Samareh Salavati Pour; F.Berto; Y.Alizadeh

    2013-01-01

    The effect of the distance between the notch tip and the position of the middle phase in the FGSs on the Charpy impact energy is investigated in the present paper.The results show that when the notch apex is close to the middle layer,the Charpy impact energy reaches its maximum value.This is due to the increment of the absorbed energy by plastic deformation ahead of the notch tip.On the other hand,when the notch tip is far from the middle layer,the Charpy impact energy strongly decreases.Another fundamental motivation of the present work is that for crack arrester configuration,no accurate mathematical or analytical modelling is available up to now.By considering the relationship between the Charpy impact energy and the plastic volume size,a new theoretical model has been developed to link the Charpy impact energy with the distance from the notch apex to the middle phase.This model is a simplified one and the effect of different shapes of the layers and the effect of microstructure on the mechanical properties and plastic region size will be considered in further investigation.The results of the new developed closed form expression show a sound agreement with some recent experimental results taken from the literature.

  8. Embrittlement of irradiated F82H in the absence of irradiation hardening

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L. [Oak Ridge National Laboratory, Oak Ridge, Tennessee (United States)], E-mail: kluehrl@ornl.gov; Shiba, K. [Japan Atomic Energy Agency, Toki-Mura, Ibaraki (Japan); Sokolov, M.A. [Oak Ridge National Laboratory, Oak Ridge, Tennessee (United States)

    2009-04-30

    Neutron irradiation of 7-12% Cr ferritic/martensitic steels below 425-450 deg. C produces microstructural defects and precipitation that cause an increase in yield stress. This irradiation hardening causes embrittlement, which is observed in a Charpy impact or fracture toughness test as an increase in the ductile-brittle transition temperature. Based on observations that show little change in strength in steels irradiated above 425-450 deg. C, the general conclusion has been that no embrittlement occurs above these temperatures. In a recent study of F82H steel, significant embrittlement was observed after irradiation at 500 deg. C, but no hardening occurred. This embrittlement is apparently due to irradiation-accelerated Laves-phase precipitation. Observations of the embrittlement of F82H in the absence of irradiation hardening have been examined and analyzed with thermal-aging studies and computational thermodynamics calculations to illuminate and understand the embrittlement during irradiation.

  9. Elastic-plastic analysis of the SS-3 tensile specimen

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, S. [Argonne National Lab., IL (United States)

    1998-09-01

    Tensile tests of most irradiated specimens of vanadium alloys are conducted using the miniature SS-3 specimen which is not ASTM approved. Detailed elastic-plastic finite element analysis of the specimen was conducted to show that, as long as the ultimate to yield strength ratio is less than or equal to 1.25 (which is satisfied by many irradiated materials), the stress-plastic strain curve obtained by using such a specimen is representative of the true material behavior.

  10. Influence of Loading Rate on the Calibration of Instrumented Charpy Strikers

    Energy Technology Data Exchange (ETDEWEB)

    Lucon, E.; Scibetta, M.; McColskey, D.; McCowan, C.

    2009-01-15

    One of the key factors for obtaining reliable instrumented Charpy results is the calibration of the instrumented striker. The conventional approach for establishing an analytical relationship between strain gage output and force applied to the transducer is the static calibration, which is preferably performed with the striker installed in the pendulum assembly. However, the response of an instrumented striker under static force application may sometimes differ significantly from its dynamic performance during an actual Charpy test. This is typically reflected in a large difference between absorbed energy returned by the pendulum encoder (KV) and calculated under the instrumented force/displacement test record (Wt). Such difference can be either minimized by optimizing the striker design or analytically removed by adjusting forces and displacements until KV = Wt (the so-called 'Dynamic Force Adjustment'). This study investigates the influence of increasing force application rates on the force/voltage characteristics of two instrumented strikers, one at NIST in Boulder, CO and one at SCK-CEN in Mol, Belgium.

  11. Effects of neutron irradiation on microstructure and mechanical properties of pure iron

    DEFF Research Database (Denmark)

    Singh, B.N.; Horsewell, Andy; Toft, P.

    1999-01-01

    tensile tested at the irradiation temperatures. Microstructures of the as-irradiated and irradiated and tensile tested specimens were investigated by transmission electron microscopy. Fracture surfaces of tensile tested specimens in unirradiated and irradiated conditions were examined in a scanning...

  12. Embrittlement behaviour of low-activation alloys with reduced boron content after neutron irradiation

    Science.gov (United States)

    Schneider, H.-C.; Dafferner, B.; Aktaa, J.

    2003-09-01

    Ferritic/martensitic steels for fusion applications have been irradiated up to 2.4 dpa in the Petten high flux reactor (HFR); their embrittlement behaviour was investigated by instrumented Charpy-V tests with subsize specimens. The aim of this mid-dose range programme was a comparison of low-activation alloys subjected to different heat treatments and with reduced B contents (down to 2 wt ppm). In the present report, the results of different OPTIFER alloys (Ia, II, IV, V, VI), as obtained in Phases IA and IB of the HFR-irradiation programme (2.4 dpa, at 250-450 °C), are analysed and assessed in comparison to the results of the former MANITU programme. The evaluation clearly shows the eliminated embrittlement problem for the advanced European reduced-activation alloys in comparison to international reference steels. This improvement can be clearly correlated to the reduction of the boron content. Furthermore, the influence of different heat treatments on the impact properties is presented.

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

    Energy Technology Data Exchange (ETDEWEB)

    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{sub 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{sup 19} neutrons/cm{sup 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{sup 19} neutrons/cm{sup 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{sub 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.

  14. Embrittlement of reduced-activation ferritic/martensitic steels irradiated in HFIR at 300 deg. C and 400 deg. C

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L. E-mail: ku2@ornl.gov; Sokolov, M.A.; Shiba, K.; Miwa, Y.; Robertson, J.P

    2000-12-01

    Miniature tensile and Charpy specimens of four ferritic/martensitic steels were irradiated at 300 deg. C and 400 deg. C in the high flux isotope reactor (HFIR) to a maximum dose of {approx}12 dpa. The steels were standard F82H (F82H-Std), a modified F82H (F82H-Mod), ORNL 9Cr-2WVTa, and 9Cr-2WVTa-2Ni, the 9Cr-2WVTa containing 2% Ni to produce helium by (n,{alpha}) reactions with thermal neutrons. More helium was produced in the F82H-Std than the F82H-Mod because of the presence of boron. Irradiation embrittlement in the form of an increase in the ductile-brittle transition temperature ({delta}DBTT) and a decrease in the upper-shelf energy (USE) occurred for all the steels. The two F82H steels had similar {delta}DBTTs after irradiation at 300 deg. C, but after irradiation at 400 deg. C, the {delta}DBTT for F82H-Std was less than for F82H-Mod. Under these irradiation conditions, little effect of the extra helium in the F82H-Std could be discerned. Less embrittlement was observed for 9Cr-2WVTa steel irradiated at 400 deg. C than for the two F82H steels. The 9Cr-2WVTa-2Ni steel with {approx}115 appm He had a larger {delta}DBTT than the 9Cr-2WVTa with {approx}5 appm He, indicating a possible helium effect.

  15. Correction of constraint loss in fracture toughness measurement of PCVN specimens based on fracture toughness diagram

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Shin Beom; Kim, Young Jin [Sungkyunkwan University, Suwon (Korea, Republic of); Chang, Yoon-Suk [Kyung Hee University, Yongin (Korea, Republic of); Kim, Min Chul; Lee, Bong Sang [Korea Atomic Energy Reserch Institute, Daejeon (Korea, Republic of)

    2010-03-15

    The aim of this paper is to suggest an approach to generate master curves by using miniature specimens, especially pre-cracked Charpy V-notched (PCVN) specimen, made of SA508 carbon steel. Firstly, fracture toughness diagram is derived from comparing finite element analyses results with the fixed mesh size at crack tip between standard compact tension and PCVN specimens. To compensate the constraint effects from different geometry, further examination based on the fracture toughness diagram was performed. In this context, a scale factor to deal with specimen size effects is proposed by statistically manipulating the numerical analysis data. Finally, the proposed scale factor is applied to calculate reference temperature which affects on the master curve. We expect that the approach can be applicable to compensate the geometrical constraint effects on fracture toughness of SA508 carbon steel when the PCVN specimen is used

  16. Influence of Striking Edge Radius (2 mm versus 8 mm) on Instrumented Charpy Data and Absorbed Energies

    Energy Technology Data Exchange (ETDEWEB)

    Lucon, E.

    2008-08-15

    The most commonly used test standards for performing Charpy impact tests (ISO 148 and ASTM E 23) envisage the use of strikers having different radii of the striking edge, i.e. 2 mm (ISO) and 8 mm (ASTM). The effect of striker geometry on Charpy results was extensively studied in the past in terms of absorbed energy measured by the machine encoder, but few investigations are available on the influence of striker configuration on the results of instrumented Charpy tests (characteristic forces, displacements and integrated energy). In this paper, these effects are investigated based on the analysis of published results from three interlaboratory studies and some unpublished Charpy data obtained at SCK-CEN. The instrumented variables which are the most sensitive to the radius of the striking edge are the maximum force and its corresponding displacement, with 8mm-strikers providing systematically higher values. Absorbed energies, obtained both from the instrumented trace and from the pendulum encoder, are almost insensitive to the type of striker up to 200 J. For higher energy levels, the values obtained from 8mm strikers become progressively larger. Data scatter is generally higher for 2mm-strikers.

  17. Unnotched Charpy Impact Energy Transition Behavior of Austempered Engineering Grade Ductile Iron Castings

    Science.gov (United States)

    Kisakurek, Sukru Ergin; Ozel, Ahmet

    2014-04-01

    Unnotched Charpy impact energy transition behavior of five different engineering grade ductile iron castings, as specified by EN 1563 Standards, were examined in as-cast, as well as in austempered states. ADIs were produced with the maximum impact energy values permissible for the grades. Austempering treatment detrimented the sub-zero impact properties of the ferritic castings, but considerably enhanced those of the pearlitic-ferritic irons. The impact energy transition behavior of the austempered states of all the grades examined were noted to be determined by the progressive transformation of the unavoidable carbon-unsaturated and untransformed regions of the austenite remaining in the matrix of the austempered ductile iron to martensite with decreasing temperature.

  18. Urine culture - catheterized specimen

    Science.gov (United States)

    Culture - urine - catheterized specimen; Urine culture - catheterization; Catheterized urine specimen culture ... urinary tract infections may be found in the culture. This is called a contaminant. You may not ...

  19. Certification of Charpy V-notch Reference Test Pieces of 80 J Nominal Absorbed Energy (ERM®-FA015x and ERM®-FA015y)

    OpenAIRE

    LAMBERTY MARIE ANDREE; Dean, Alan; Roebben, Gert

    2011-01-01

    This certification report describes the processing and characterisation of ERM®-FA015x and ERM®-FA015y, two batches of Charpy V-notch certified reference test pieces. Sets of five of these test pieces are used for the verification of pendulum impact test machines according to EN 10045-2 (Charpy impact test on metallic materials, Part 2. Method for the verification of impact testing machines) or according to ISO 148-2 (Metallic materials - Charpy pendulum impact test – Part 2: Verification of ...

  20. Embrittlernent of irradiated F82H in the absence of irradiation hardening

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, Ronald L [ORNL; Shiba, Kiyoyuki [ORNL; Sokolov, Mikhail A [ORNL

    2009-01-01

    Neutron irradiation of 7-12% Cr ferritic/martensitic steels below 425-450 C produces microstructural defects and precipitation that cause an increase in yield stress. This irradiation hardening causes embrittlement, which is observed in a Charpy impact or fracture toughness test as an increase in the ductile-brittle transition temperature. Based on observations that show little change in strength in steels irradiated above 425-450 C, the general conclusion has been that no embrittlement occurs above these temperatures. In a recent study of F82H steel, significant embrittlement was observed after irradiation at 500 C. This embrittlement is apparently due to irradiation-accelerated Laves-phase precipitation. Observations of the embrittlement in the absence of hardening has been examined and analyzed with thermal-aging studies and computational thermodynamics calculations to illuminate and understand the effect.

  1. Effect of neutron irradiation on the mechanical properties of weld overlay cladding for reactor pressure vessel

    Science.gov (United States)

    Tobita, Tohru; Udagawa, Makoto; Chimi, Yasuhiro; Nishiyama, Yutaka; Onizawa, Kunio

    2014-09-01

    This study investigates the effects of high fluence neutron irradiation on the mechanical properties of two types of cladding materials fabricated using the submerged-arc welding and electroslag welding methods. The tensile tests, Charpy impact tests, and fracture toughness tests were conducted before and after the neutron irradiation with a fluence of 1 × 1024 n/m2 at 290 °C. With neutron irradiation, we could observe an increase in the yield strength and ultimate strength, and a decrease in the total elongation. All cladding materials exhibited ductile-to-brittle transition behavior during the Charpy impact tests. A reduction in the Charpy upper-shelf energy and an increase in the ductile-to-brittle transition temperature was observed with neutron irradiation. There was no obvious decrease in the elastic-plastic fracture toughness (JIc) of the cladding materials upon irradiation with high neutron fluence. The tearing modulus was found to decrease with neutron irradiation; the submerged-arc-welded cladding materials exhibited low JIc values at high temperatures.

  2. Effect of neutron irradiation on the mechanical properties of weld overlay cladding for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Tobita, Tohru, E-mail: tobita.tohru@jaea.go.jp; Udagawa, Makoto; Chimi, Yasuhiro; Nishiyama, Yutaka; Onizawa, Kunio

    2014-09-15

    This study investigates the effects of high fluence neutron irradiation on the mechanical properties of two types of cladding materials fabricated using the submerged-arc welding and electroslag welding methods. The tensile tests, Charpy impact tests, and fracture toughness tests were conducted before and after the neutron irradiation with a fluence of 1 × 10{sup 24} n/m{sup 2} at 290 °C. With neutron irradiation, we could observe an increase in the yield strength and ultimate strength, and a decrease in the total elongation. All cladding materials exhibited ductile-to-brittle transition behavior during the Charpy impact tests. A reduction in the Charpy upper-shelf energy and an increase in the ductile-to-brittle transition temperature was observed with neutron irradiation. There was no obvious decrease in the elastic–plastic fracture toughness (J{sub Ic}) of the cladding materials upon irradiation with high neutron fluence. The tearing modulus was found to decrease with neutron irradiation; the submerged-arc-welded cladding materials exhibited low J{sub Ic} values at high temperatures.

  3. Vickers Microhardness Testing with Miniaturized Disk Specimens

    OpenAIRE

    Kurishita, Hiroaki; Kayano, Hideo

    1991-01-01

    The microhardness technique has been increasingly important for testing irradiated materials because of the necessity of small-scale specimen technology. In order to establish Vickers microhardness testing over a wide temperature range using miniaturized specimens such as transmission electron microscopy (TEM) disks, an apparatus that permits the measurements in the temperature range of well below liquid nitrogen temperature to well above room temperature is developed. Effects of indentation ...

  4. A master curve-mechanism based approach to modeling the effects of constraint, loading rate and irradiation on the toughness-temperature behavior of a V-4Cr-4Ti alloy

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G.R.; Donahue, E.; Lucas, G.E.; Sheckherd, J.W. [Univ. of California, Santa Barbara, CA (United States)

    1996-10-01

    The influence of loading rate and constraint on the effective fracture toughness as a function of temperature [K{sub e}(T)] of the fusion program heat of V-4Cr-4Ti was measured using subsized, three point bend specimens. The constitutive behavior was characterized as a function of temperature and strain rate using small tensile specimens. Data in the literature on this alloy was also analysed to determine the effect of irradiation on K{sub e}(T) and the energy temperature (E-T) curves measured in subsized Charpy V-notch tests. It was found that V-4Cr-4Ti undergoes {open_quotes}normal{close_quotes} stress-controlled cleavage fracture below a temperature marking a sharp ductile-to-brittle transition. The transition temperature is increased by higher loading rates, irradiation hardening and triaxial constraint. Shifts in a reference transition temperature due to higher loading rates and irradiation can be reasonably predicted by a simple equivalent yield stress model. These results also suggest that size and geometry effects, which mediate constraint, can be modeled by combining local critical stressed area {sigma}*/A* fracture criteria with finite element method simulations of crack tip stress fields. The fundamental understanding reflected in these models will be needed to develop K{sub e}(T) curves for a range of loading rates, irradiation conditions, structural size scales and geometries relying (in large part) on small specimen tests. Indeed, it may be possible to develop a master K{sub e}(T) curve-shift method to account for these variables. Such reliable and flexible failure assessment methods are critical to the design and safe operation of defect tolerant vanadium structures.

  5. Behavior of Aramid Fiber/Ultrahigh Molecular Weight Polyethylene Fiber Hybrid Composites under Charpy Impact and Ballistic Impact

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    The aramid fiber/UHMWPE (ultrahigh molecular weight polyethylene) fiber hybrid composites (AF/DF) were manufactured. By Charpy impact, the low velocity impact behavior of AF/DF composite was studied. And the high velocity impact behavior under ballistic impact was also investigated. The influence of hybrid ratio on the performances of low and high velocity impact was analyzed, and hybrid structures with good impact properties under low velocity impact and high velocity were optimized. For Charpy impact, the maximal impact load increased with the accretion of the AF layers for AF/DF hybrid composites. The total impact power was reduced with the decrease of DF layers and the delamination can result in the increase of total impact power. For ballistic impact, the DF ballistic performance was better than that of the AF and the hybrid ratio had a crucial influence. The failure morphology of AF/DF hybrid composite under Charpy impact and ballistic impact was analyzed. The AF/DF hybrid composites in suitable hybrid ratio could attain better performance than AF or DF composites.

  6. Ultrahigh Charpy impact toughness (~450J) achieved in high strength ferrite/martensite laminated steels.

    Science.gov (United States)

    Cao, Wenquan; Zhang, Mingda; Huang, Chongxiang; Xiao, Shuyang; Dong, Han; Weng, Yuqing

    2017-02-02

    Strength and toughness are a couple of paradox as similar as strength-ductility trade-off in homogenous materials, body-centered-cubic steels in particular. Here we report a simple way to get ultrahigh toughness without sacrificing strength. By simple alloying design and hot rolling the 5Mn3Al steels in ferrite/austenite dual phase temperature region, we obtain a series of ferrite/martensite laminated steels that show up-to 400-450J Charpy V-notch impact energy combined with a tensile strength as high as 1.0-1.2 GPa at room temperature, which is nearly 3-5 times higher than that of conventional low alloy steels at similar strength level. This remarkably enhanced toughness is mainly attributed to the delamination between ferrite and martensite lamellae. The current finding gives us a promising way to produce high strength steel with ultrahigh impact toughness by simple alloying design and hot rolling in industry.

  7. Ultrahigh Charpy impact toughness (~450J) achieved in high strength ferrite/martensite laminated steels

    Science.gov (United States)

    Cao, Wenquan; Zhang, Mingda; Huang, Chongxiang; Xiao, Shuyang; Dong, Han; Weng, Yuqing

    2017-01-01

    Strength and toughness are a couple of paradox as similar as strength-ductility trade-off in homogenous materials, body-centered-cubic steels in particular. Here we report a simple way to get ultrahigh toughness without sacrificing strength. By simple alloying design and hot rolling the 5Mn3Al steels in ferrite/austenite dual phase temperature region, we obtain a series of ferrite/martensite laminated steels that show up-to 400–450J Charpy V-notch impact energy combined with a tensile strength as high as 1.0–1.2 GPa at room temperature, which is nearly 3–5 times higher than that of conventional low alloy steels at similar strength level. This remarkably enhanced toughness is mainly attributed to the delamination between ferrite and martensite lamellae. The current finding gives us a promising way to produce high strength steel with ultrahigh impact toughness by simple alloying design and hot rolling in industry. PMID:28150692

  8. Fracture behavior of neutron-irradiated high-manganese austenitic steels

    Science.gov (United States)

    Yoshida, H.; Miyata, K.; Narui, M.; Kayano, H.

    1991-03-01

    The instrumented Charpy impact test was applied to study the fracture behavior of high-manganese austenitic steels before and after neutron irradiations. Quarter-size specimens of a commercial high-manganese steel (18% Mn-5% Ni-16% Cr), three reference steels (21% Mn-1% Ni-9% Cr, 20% Mn-1% Ni-11% Cr, 15% Mn-1% Ni-13% Cr) and two model steels (17% Mn-4.5% Si-6.5% Cr, 22% Mn-4.5% Si-6.5% Cr-0.2% N) were used for the impact tests at temperatures between 77 and 523 K. The load-deflection curves showed typical features corresponding to characteristics of the fracture properties. The temperature dependences of fracture energy and failure deflection obtained from the curves clearly demonstrate only small effects up to 2 × 10 23 n/m 2 ( E > 0.1 MeV) and brittleness at room temperature in 17% Mn-Si-Cr steel at 1.6 × 10 25 n/m 2 ( E > 0.1 MeV), while ductility still remains in 22%Mn-Si-Cr steel.

  9. Irradiation programme MANITU. Results of impact tests with the irradiated materials (0.2 dpa); Bestrahlungsprogramm MANITU. Ergebnisse der Kerbschlagbiegeversuche mit den bis 0,2 dpa bestrahlten Werkstoffen

    Energy Technology Data Exchange (ETDEWEB)

    Rieth, M.; Dafferner, B.; Kunisch, W.; Ries, H.; Romer, O.

    1997-04-01

    The irradiation project MANITU was planned and carried out in the frame of the European Longterm Fusion Materials Development Programme. The problem of the irradiation induced embrittlement of possible martensitic alloy candidates is still unsolved. But after the evaluation of subsize Charpy tests with specimens of MANITU (0.8 dpa) a first tendency was recognizable. The mechanical properties of the newly developed low activation 7-10% Cr-W(Ge)VTa alloys are better compared to the modified commercial 10-11% Cr-NiMoVNb steels. In the present report the results of instrumented impact tests of the MANITU programme (irradiation dose 0.2 dpa, irradiation temperatures 250, 300, 350, 400, and 450 C) are analysed and assessed. Among all examined alloys (MANET-I, MANET-II, K-heat, OPTIFER-Ia, OPTIFER-II, F82H, 9Cr-2WVTa ORNL 3791) the ORNL steel shows the best embrittlement behaviour after neutron irradation. (orig.) [Deutsch] Das Bestrahlungsprojekt MANITU wurde im Rahmen des europaeischen Langzeitprogramms fuer Materialentwicklung fuer die Kernfusion geplant und durchgefuehrt. Das Problem der bestrahlungsinduzierten Versproedung bei den in Frage kommenden martensitischen Werkstoffen ist nach wie vor ungeloest. Eine erste Tendenz zeichnete sich jedoch nach der Auswertung der Kerbschlagbiegeversuche an den bis 0,8 dpa bestrahlten miniaturisierten Proben des MANITU-Programms ab. Die neu entwickelten niedrig aktivierbaren 7-10% Cr-W(Ge)VTa-Legierungen weisen gegenueber den modifizierten kommerziellen 10-11% Cr-NiMoVNb-Staehlen bessere mechanische Eigenschaften auf. Im vorliegenden Bericht werden die Ergebnisse aus den instrumentierten Kerbschlagbiegeversuchen des MANITU-Programms (Bestrahlungsdosis 0,2 dpa, Bestrahlungstemperaturen 250, 300, 350, 400 und 450 C) analysiert und bewertet. Von den untersuchten Legierungen (MANET-I, MANET-II, Kastencharge, OPTIFER-Ia, OPTIFER-II, F82H, 9Cr-2WVTa ORNL 3791) zeigt der ORNL-Stahl das beste Versproedungsverhalten nach

  10. PVRC/MPC Round Robin Tests for the Low Toughness High-Copper 72W Weld Using Master Curve Methodology of PCVN Specimens

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Bong-Sang; Hong, Jun Hwa; Yang, Won Jon

    2000-06-01

    This report summarizes the results obtained from the Korean contribution the PVRC/MPC cooperative program on {sup R}ound Robin Tests for Low Toughness High-Copper 72W Weld Using Master Curve Methodology of PCVN Specimens. The mandatory part of this program is to perform fracture toughness (K{sub jc}) tests on the low toughness 72W weld at three different temperatures using pre-cracked Charpy specimens. The purpose of the tests is to verify the specimen size requirements in the ASTM E 1921, 'Standard test method for determination of reference temperature, T{sub o}, for ferritic steels in the transition range'.

  11. Experimental tests of irradiation-anneal-reirradiation effects on mechanical properties of RPV plate and weld materials

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-01-01

    The Charpy-V (C{sub V}) notch ductility and tension test properties of three reactor pressure vessel (RPV) steel materials were determined for the 288{degree}C (550{degree}F) irradiated (I), 288{degree}C (550{degree}F) irradiated + 454{degree}C (850{degree}F)-168 h postirradiation annealed (IA), and 288{degree}C (550{degree}F) reirradiated (IAR) conditions. Total fluences of the I condition and the IAR condition were, respectively, 3.33 {times} 10{sup 19} n/cm{sup 2} and 4.18 {times} 10{sup 19} n/cm{sup 2}, E > 1 MeV. The irradiation portion of the IAR condition represents an incremental fluence increase of 1. 05 {times} 10{sup 19} n/cm{sup 2}, E > 1 MeV, over the I-condition fluence. The materials (specimens) were supplied by the Yankee Atomic Electric Company and represented high and low nickel content plates and a high nickel, high copper content weld deposit prototypical of the Yankee-Rowe reactor vessel. The promise of the IAR method for extending the fluence tolerance of radiation-sensitive steels and welds is clearly shown by the results. The annealing treatment produced full C{sub V} upper shelf recovery and full or nearly full recovery in the C{sub V} 41 J (30 ft-lb) transition temperature. The C{sub V} transition temperature increases produced by the reirradiation exposure were 22% to 43% of the increase produced by the first cycle irradiation exposure. A somewhat greater radiation embrittlement sensitivity and a somewhat greater reirradiation embrittlement sensitivity was exhibited by the low nickel content plate than the high nickel content plate. Its high phosphorus content is believed to be responsible. The IAR-condition properties of the surface vs. interior regions of the low nickel content plate are also compared.

  12. STEM tomography for thick biological specimens

    Energy Technology Data Exchange (ETDEWEB)

    Aoyama, Kazuhiro [FEI Company Japan Ltd., Application Laboratory, NSS-II Building, 2-13-34 Kohnan, Minato-ku, Tokyo 108-0075 (Japan)], E-mail: kazuhiro.aoyama@fei.com; Takagi, Tomoko [FEI Company Japan Ltd., Application Laboratory, NSS-II Building, 2-13-34 Kohnan, Minato-ku, Tokyo 108-0075 (Japan); Laboratory of Electron Microscopy, Japan Women' s University, 2-8-1 Mejirodai, Bunkyo-ku, Tokyo 112-8681 (Japan); Hirase, Ai; Miyazawa, Atsuo [Bio-multisome Research Team, RIKEN SPring-8 Center, Harima Institute, 1-1-1 Kouto, Sayo, Hyogo 679-5148 (Japan); CREST, JST (Japan)

    2008-12-15

    Scanning transmission electron microscopy (STEM) tomography was applied to biological specimens such as yeast cells, HEK293 cells and primary culture neurons. These cells, which were embedded in a resin, were cut into 1-{mu}m-thick sections. STEM tomography offers several important advantages including: (1) it is effective even for thick specimens, (2) 'dynamic focusing', (3) ease of using an annular dark field (ADF) mode and (4) linear contrasts. It has become evident that STEM tomography offers significant advantages for the observation of thick specimens. By employing STEM tomography, even a 1-{mu}m-thick specimen (which is difficult to observe by conventional transmission electron microscopy (TEM)) was successfully analyzed in three dimensions. The specimen was tilted up to 73 deg. during data acquisition. At a large tilt angle, the specimen thicknesses increase dramatically. In order to observe such thick specimens, we introduced a special small condenser aperture that reduces the collection angle of the STEM probe. The specimen damage caused by the convergent electron beam was expected to be the most serious problem; however, the damage in STEM was actually smaller than that in TEM. In this study, the irradiation damage caused by TEM- and STEM-tomography in biological specimens was quantitatively compared.

  13. Heat treatment effects on impact toughness of 9Cr-1MoVNb and 12Cr-1MoVW steels irradiated to 100 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Alexander, D.J. [Oak Ridge National Lab., TN (United States)

    1997-08-01

    Plates of 9Cr-1MoVNb and 12Cr-1MoVW steels were given four different heat treatments: two normalizing treatments were used and for each normalizing treatment two tempers were used. Miniature Charpy specimens from each heat treatment were irradiated to {approx}19.5 dpa at 365{degrees}C and to {approx}100 dpa at 420{degrees}C in the Fast Flux Test Facility (FFTF). In previous work, the same materials were irradiated to 4-5 dpa at 365{degrees}C and 35-36 dpa at 420{degrees}C in FFTF. The tests indicated that prior austenite grain size, which was varied by the different normalizing treatments, had a significant effect on impact behavior of the 9Cr-1MoVNb but not on the 12Cr-1MoVW. Tempering treatment had relatively little effect on the shift in DBTT for both steels. Conclusions are presented on how heat treatment can be used to optimize impact properties.

  14. Embrittlement of irradiated ferritic/martensitic steels in the absence of irradiation hardening

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L. [Oak Ridge National Laboratory, Materials Science and Technology Division, P.O. 2008 MS6138, Oak Ridge, TN 37831-6138 (United States)], E-mail: kluehrl@ornl.gov; Shiba, K. [Japan Atomic Energy Agency, Toki-Mura, Ibaraki (Japan); Sokolov, M.A. [Oak Ridge National Laboratory, Materials Science and Technology Division, P.O. 2008 MS6138, Oak Ridge, TN 37831-6138 (United States)

    2008-07-15

    Irradiation damage caused by neutron irradiation below 425-450 deg. C of 9-12% Cr ferritic/martensitic steels produces microstructural defects that cause an increase in yield stress. This irradiation hardening causes embrittlement observed in a Charpy impact test as an increase in the ductile-brittle transition temperature. Little or no change in strength is observed in steels irradiated above 425-450 deg. C. Therefore, the general conclusion has been that no embrittlement occurs above these temperatures. In a recent study, significant embrittlement was observed in F82H steel irradiated at 500 deg. C to 5 and 20 dpa without any change in strength. Earlier studies on several conventional steels also showed embrittlement effects above the irradiation-hardening temperature regime. Indications are that this embrittlement is caused by irradiation-accelerated or irradiation-induced precipitation. Observations of embrittlement in the absence of irradiation hardening that were previously reported in the literature have been examined and analyzed with computational thermodynamics calculations to illuminate and understand the effect.

  15. 37 CFR 2.59 - Filing substitute specimen(s).

    Science.gov (United States)

    2010-07-01

    ... 37 Patents, Trademarks, and Copyrights 1 2010-07-01 2010-07-01 false Filing substitute specimen(s..., DEPARTMENT OF COMMERCE RULES OF PRACTICE IN TRADEMARK CASES Drawing § 2.59 Filing substitute specimen(s). (a... specimen(s), the applicant must: (1) For an amendment to allege use under § 2.76, verify by affidavit...

  16. Specimen Machining for the Study of the Effect of Swelling on CGR in PWR Environment.

    Energy Technology Data Exchange (ETDEWEB)

    Teysseyre, Sebastien Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-06-01

    This report describes the preparation of ten specimens to be used for the study of the effect of swelling on the propagation of irradiation assisted stress corrosion cracking cracks. Four compact tension specimens, four microscopy plates and two tensile specimens were machined from a AISI 304 material that was irradiated up to 33 dpa. The specimens had been machined such as to represent the behavior of materials with 3.7%swelling and <2% swelling.

  17. Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

    Science.gov (United States)

    Keiser, Dennis D.; Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W.; Ross Finlay, M.; Moore, Glenn; Medvedev, Pavel; Meyer, Mitch

    2017-05-01

    The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U-Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

  18. Assessment of Ductile-to-Brittle Transition Behavior of Localized Microstructural Regions in a Friction-Stir Welded X80 Pipeline Steel with Miniaturized Charpy V-Notch Testing

    Science.gov (United States)

    Avila, Julian A.; Lucon, Enrico; Sowards, Jeffrey; Mei, Paulo Roberto; Ramirez, Antonio J.

    2016-06-01

    Friction-stir welding (FSW) is an alternative welding process for pipelines. This technology offers sound welds, good repeatability, and excellent mechanical properties. However, it is of paramount importance to determine the toughness of the welds at low temperatures in order to establish the limits of this technology. Ductile-to-brittle transition curves were generated in the present study by using a small-scale instrumented Charpy machine and miniaturized V-notch specimens (Kleinstprobe, KLST); notches were located in base metal, heat-affected, stirred, and hard zones within a FSW joint of API-5L X80 Pipeline Steel. Specimens were tested at temperatures between 77 K (-196 °C) and 298 K (25 °C). Based on the results obtained, the transition temperatures for the base material and heat-affected zone were below 173 K (-100 °C); conversely, for the stirred and hard zones, it was located around 213 K (-60 °C). Fracture surfaces were characterized and showed a ductile fracture mechanism at high impact energies and a mixture of ductile and brittle mechanisms at low impact energies.

  19. Effects of thermal aging on fracture toughness and Charpy-impact strength of stainless steel pipe welds

    Energy Technology Data Exchange (ETDEWEB)

    Gavenda, D.J.; Michaud, W.F.; Galvin, T.M.; Burke, W.F.; Chopra, O.K. [Argonne National Lab., IL (United States)

    1996-05-01

    Degradation of fracture toughness, tensile, and Charpy-impact properties of Type 304 and 304/308 SS pipe welds due to thermal aging was studied at room temperature and 290 C. Thermal aging of SS welds results in moderate decreases in charpy-impact strength and fracture toughness. Upper-shelf energy decreased by 50-80 J/cm{sup 2}. Decrease in fracture toughness J-R curve or J{sub IC} is relatively small. Thermal aging had no or little effect on tensile strength of the welds. Fracture properties of SS welds are controlled by the distribution and morphology of second-phase particles. Failure occurs by formation and growth of microvoids near hard inclusions; such processes are relatively insensitive to thermal aging. The ferrite phase has little or no effect on fracture properties of the welds. Differences in fracture resistance of the welds arise from differences in the density and size of inclusions. Mechanical-property data from the present study are consistent with results from other investigations. The existing data have been used to establish minimum expected fracture properties for SS welds.

  20. Controlled Environment Specimen Transfer

    DEFF Research Database (Denmark)

    Damsgaard, Christian Danvad; Zandbergen, Henny W.; Hansen, Thomas Willum

    2014-01-01

    Specimen transfer under controlled environment conditions, such as temperature, pressure, and gas composition, is necessary to conduct successive complementary in situ characterization of materials sensitive to ambient conditions. The in situ transfer concept is introduced by linking an environme...

  1. Development of PIE techniques for irradiated LWR pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-09-01

    For the evaluation of safety and integrity of light water reactors (LWRs), various post irradiation examinations (PIEs) of reactor pressure vessel (RPV) steels and fuel claddings have been carried out in the Research Hot Laboratory (RHL). In recent years, the instrumented Charpy impact testing machine was remodeled aiming at the improvement of accuracy and reliability. By this remodeling, absorbed energy and other useful information on impact properties can be delivered from the force-displacement curve for the evaluation of neutron irradiation embrittlement behavior of LWR-RPV steels at one-time striking. In addition, two advanced PIE technologies are now under development. One is the remote machining of mechanical test pieces from actual irradiated pressure vessel steels. The other is development of low-cycle and high-cycle fatigue test technology in order to clarify the post-irradiation fatigue characteristics of structural and fuel cladding materials. (author)

  2. Fracture toughness evaluation using small specimens for assuring structural integrity of PRV's

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Bong Sang; Hong, J. H.; Chi, S. H.; Kim, J. H.; Yang, W. J

    1999-08-01

    This report summarizes the results obtained from the three year contribution of KAERI to the IAEA-CRP on ''Assuring Structural Integrity of Reactor Pressure Vessels''. The mandatory part of this programme is to perform fracture toughness K{sub jc} tests using pre-cracked Charpy specimens on the IAEA reference material JRQ (ASTM A533-B1 steel). The results will be used to validate the small specimens for surveillance tests. In this report, three different heats of reactor pressure vessel materials are characterized by the ASTM E 1921-97 'standard test method for determination of reference temperature, T{sub o}, for ferritic steels in the transition range'. The materials are the IAEA reference plate (JRQ), a Japanese forging (JEL), and a Korean forging (KFY5). 6 refs., 7 tabs., 20 figs.

  3. Determinación de la tenacidad a la fractura de muestras de Acero 45 fundido, empleando las correlaciones entre el KIC y la energía de impacto medida en el ensayo de Charpy. // Determination of the fracture tenacity of cast Steel grade 45 samples, using th

    Directory of Open Access Journals (Sweden)

    F. Ramos Morales

    2005-05-01

    Full Text Available En el presente trabajo se determinan los valores de tenacidad a la fractura (KIC de muestras de Acero 45 fundido,empleando las correlaciones entre la tenacidad a la fractura y la energía de impacto (CVN obtenida del ensayo de Charpy.Se hace una discusión sobre las correlaciones que más se ajustan en la región de transición y en upper shelf. Se comparanlos valores obtenidos de estas correlaciones a valores de tenacidad a la fractura establecidos en la literatura.Palabras claves: Fractura, energía de impacto, acero fundido.______________________________________________________________________________Abstract.In this paper, the values of fracture toughness (KIC are determined on specimens of cast steel grade 45, using thecorrelations among the fracture toughness (KIC and the impact energy (CVN obtained from a Charpy test. A discussion ismade on the correlations that are better adjusted in the transition region and in upper shelf region. The obtained values arecompared from these correlations to values of fracture toughness (KIC settled down in the literature.Key words. Fracture, impact energy, cast steel.

  4. Genomics and museum specimens.

    Science.gov (United States)

    Nachman, Michael W

    2013-12-01

    Nearly 25 years ago, Allan Wilson and colleagues isolated DNA sequences from museum specimens of kangaroo rats (Dipodomys panamintinus) and compared these sequences with those from freshly collected animals (Thomas et al. 1990). The museum specimens had been collected up to 78 years earlier, so the two samples provided a direct temporal comparison of patterns of genetic variation. This was not the first time DNA sequences had been isolated from preserved material, but it was the first time it had been carried out with a population sample. Population geneticists often try to make inferences about the influence of historical processes such as selection, drift, mutation and migration on patterns of genetic variation in the present. The work of Wilson and colleagues was important in part because it suggested a way in which population geneticists could actually study genetic change in natural populations through time, much the same way that experimentalists can do with artificial populations in the laboratory. Indeed, the work of Thomas et al. (1990) spawned dozens of studies in which museum specimens were used to compare historical and present-day genetic diversity (reviewed in Wandeler et al. 2007). All of these studies, however, were limited by the same fundamental problem: old DNA is degraded into short fragments. As a consequence, these studies mostly involved PCR amplification of short templates, usually short stretches of mitochondrial DNA or microsatellites. In this issue, Bi et al. (2013) report a breakthrough that should open the door to studies of genomic variation in museum specimens. They used target enrichment (exon capture) and next-generation (Illumina) sequencing to compare patterns of genetic variation in historic and present-day population samples of alpine chipmunks (Tamias alpinus) (Fig. 1). The historic samples came from specimens collected in 1915, so the temporal span of this comparison is nearly 100 years.

  5. Small Punch Test on Before and Post Irradiated Domestic Reactor Pressure Steel

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    Problems may be caused when applying the standard specimen to study the properties of irradiated reactor materials, because of its big dimension, e.g.: The inner temperature gradient of the specimen is high when irradiated, the radiation

  6. Numerical modelling of Charpy-V notch test by local approach to fracture. Application to an A508 steel in the ductile-brittle transition range; Modelisation de l'essai Charpy par l'approche locale de la rupture. Application au cas de l'acier 16MND5 dans le domaine de transition

    Energy Technology Data Exchange (ETDEWEB)

    Tanguy, B

    2001-07-15

    Ferritic steels present a transition of the rupture mode which goes progressively of a brittle rupture (cleavage) to a ductile rupture when the temperature increases. The following of the difference of the transition temperature of the PWR vessel steel by the establishment of toughness curves makes of the Charpy test an integrating part of the monitoring of the French PWR reactors. In spite of the advantages which are adapted to it in particular its cost, the Charpy test does not allow to obtain directly a variable which characterizes a crack propagation resistance as for instance the toughness used for qualifying the mechanical integrity of a structure. This work deals with the establishment of the through impact strength-toughness in the transition range of the vessel steel: 16MND5 from a non-empirical approach based on the local approach of the rupture. The brittle rupture is described by the Beremin model (1983), which allows to describe the dispersion inherent in this rupture mode. The description of the brittle fissure is carried out by the GTN model (1984) and by the Rousselier model (1986). This last model has been modified in order to obtain a realistic description of the brittle damage in the case of fast solicitations and of local heating. The method proposed to determine the parameters of the damage models depends only of tests on notched specimens and of the inclusion data of the material. The behaviour is described by an original formulation parametrized in temperature which allows to describe all the tests carried out in this study. Before using this methodology, an experimental study of the behaviour and of the rupture modes of the steel 16MND5 has been carried out. From the toughness tests carried out in quasi-static and dynamical conditions, it has been revealed that this steel does not present important unwedging of its toughness curve due to the velocity effect. In the transition range, local heating of about 150 C have been measured in the root

  7. Evaluation of irradiation hardening of proton irradiated stainless steels by nanoindentation

    Energy Technology Data Exchange (ETDEWEB)

    Yabuuchi, Kiyohiro, E-mail: kiyohiro.yabuuchi@qse.tohoku.ac.jp [Graduate School of Engineering, Tohoku University, 6-6-01-2 Aramaki-Aza-Aoba, Aobaku, Sendai, Miyagi 980-8579 (Japan); Kuribayashi, Yutaka [Graduate School of Engineering, Tohoku University, 6-6-01-2 Aramaki-Aza-Aoba, Aobaku, Sendai, Miyagi 980-8579 (Japan); Nogami, Shuhei, E-mail: shuhei.nogami@qse.tohoku.ac.jp [Graduate School of Engineering, Tohoku University, 6-6-01-2 Aramaki-Aza-Aoba, Aobaku, Sendai, Miyagi 980-8579 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hasegawa, Akira, E-mail: akira.hasegawa@qse.tohoku.ac.jp [Graduate School of Engineering, Tohoku University, 6-6-01-2 Aramaki-Aza-Aoba, Aobaku, Sendai, Miyagi 980-8579 (Japan)

    2014-03-15

    Ion irradiation experiments are useful for investigating irradiation damage. However, estimating the irradiation hardening of ion-irradiated materials is challenging because of the shallow damage induced region. Therefore, the purpose of this study is to prove usefulness of nanoindentation technique for estimation of irradiation hardening for ion-irradiated materials. SUS316L austenitic stainless steel was used and it was irradiated by 1 MeV H{sup +} ions to a nominal displacement damage of 0.1, 0.3, 1, and 8 dpa at 573 K. The irradiation hardness of the irradiated specimens were measured and analyzed by Nix–Gao model. The indentation size effect was observed in both unirradiated and irradiated specimens. The hardness of the irradiated specimens changed significantly at certain indentation depths. The depth at which the hardness varied indicated that the region deformed by the indenter had reached the boundary between the irradiated and unirradiated regions. The hardness of the irradiated region was proportional to the inverse of the indentation depth in the Nix–Gao plot. The bulk hardness of the irradiated region, H{sub 0}, estimated by the Nix–Gao plot and Vickers hardness were found to be related to each other, and the relationship could be described by the equation, HV = 0.76H{sub 0}. Thus, the nanoindentation technique demonstrated in this study is valuable for measuring irradiation hardening in ion-irradiated materials.

  8. Certification of Charpy V-Notch Reference Test Pieces of 30 J Nominal Absorbed Energy - Certified Reference Materials ERM®-FA013bg and ERM®-FA013bh

    OpenAIRE

    LAMBERTY MARIE ANDREE; Dean, Alan; Roebben, Gert

    2011-01-01

    This certification report describes the processing and characterisation of ERM®-FA013bg and ERM®-FA013bh, two batches of Charpy V-notch certified reference test pieces. Sets of five of these test pieces are used for the verification of pendulum impact test machines according to EN 10045-2 (Charpy impact test on metallic materials, Part 2. Method for the verification of impact testing machines [1]) or according to ISO 148-2 (Metallic materials - Charpy pendulum impact test - Part 2: Verificati...

  9. Biaxial Creep Specimen Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    JL Bump; RF Luther

    2006-02-09

    This report documents the results of the weld development and abbreviated weld qualification efforts performed by Pacific Northwest National Laboratory (PNNL) for refractory metal and superalloy biaxial creep specimens. Biaxial creep specimens were to be assembled, electron beam welded, laser-seal welded, and pressurized at PNNL for both in-pile (JOYO reactor, O-arai, Japan) and out-of-pile creep testing. The objective of this test campaign was to evaluate the creep behavior of primary cladding and structural alloys under consideration for the Prometheus space reactor. PNNL successfully developed electron beam weld parameters for six of these materials prior to the termination of the Naval Reactors program effort to deliver a space reactor for Project Prometheus. These materials were FS-85, ASTAR-811C, T-111, Alloy 617, Haynes 230, and Nirnonic PE16. Early termination of the NR space program precluded the development of laser welding parameters for post-pressurization seal weldments.

  10. Neutron irradiation of beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, D.S.; Ermi, R.M. [Pacific Northwest National Lab., Richland, WA (United States); Tsai, H. [Argonne National Lab., IL (United States)

    1998-03-01

    Seven subcapsules from the FFTF/MOTA 2B irradiation experiment containing 97 or 100% dense sintered beryllium cylindrical specimens in depleted lithium have been opened and the specimens retrieved for postirradiation examination. Irradiation conditions included 370 C to 1.6 {times} 10{sup 22} n/cm{sup 2}, 425 C to 4.8 {times} 10{sup 22} n/cm{sup 2}, and 550 C to 5.0 {times} 10{sup 22} n/cm{sup 2}. TEM specimens contained in these capsules were also retrieved, but many were broken. Density measurements of the cylindrical specimens showed as much as 1.59% swelling following irradiation at 500 C in 100% dense beryllium. Beryllium at 97% density generally gave slightly lower swelling values.

  11. The studies of irradiation hardening of stainless steel reactor internals under proton and xenon irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Chaoliang; Zhang, Lu; Qian, Wangjie; Mei, Jinna; Liu, Xiang Bing [Suzhou Nuclear Power Research Institute, Suzuhou (China)

    2016-06-15

    Specimens of stainless steel reactor internals were irradiated with 240 keV protons and 6 MeV Xe ions at room temperature. Nanoindentation constant stiffness measurement tests were carried out to study the hardness variations. An irradiation hardening effect was observed in proton- and Xe-irradiated specimens and more irradiation damage causes a larger hardness increment. The Nix-Gao model was used to extract the bulk-equivalent hardness of irradiation-damaged region and critical indentation depth. A different hardening level under H and Xe irradiation was obtained and the discrepancies of displacement damage rate and ion species may be the probable reasons. It was observed that the hardness of Xe-irradiated specimens saturate at about 2 displacement/atom (dpa), whereas in the case of proton irradiation, the saturation hardness may be more than 7 dpa. This discrepancy may be due to the different damage distributions.

  12. [Blood Count Specimen].

    Science.gov (United States)

    Tamura, Takako

    2015-12-01

    The circulating blood volume accounts for 8% of the body weight, of which 45% comprises cellular components (blood cells) and 55% liquid components. We can measure the number and morphological features of blood cells (leukocytes, red blood cells, platelets), or count the amount of hemoglobin in a complete blood count: (CBC). Blood counts are often used to detect inflammatory diseases such as infection, anemia, a bleeding tendency, and abnormal cell screening of blood disease. This count is widely used as a basic data item of health examination. In recent years, clinical tests before consultation have become common among outpatient clinics, and the influence of laboratory values on consultation has grown. CBC, which is intended to count the number of raw cells and to check morphological features, is easily influenced by the environment, techniques, etc., during specimen collection procedures and transportation. Therefore, special attention is necessary to read laboratory data. Providing correct test values that accurately reflect a patient's condition from the laboratory to clinical side is crucial. Inappropriate medical treatment caused by erroneous values resulting from altered specimens should be avoided. In order to provide correct test values, the daily management of devices is a matter of course, and comprehending data variables and positively providing information to the clinical side are important. In this chapter, concerning sampling collection, blood collection tubes, dealing with specimens, transportation, and storage, I will discuss their effects on CBC, along with management or handling methods.

  13. Specimen Holder For Flammability Tests

    Science.gov (United States)

    Rucker, Michelle A.

    1992-01-01

    Fixture holds sheet specimens for flammability tests. Frame and clamps designed to minimize local overstress on specimen. Heat capacity of fixture low, interfering less with interpretation of results of test by drawing less heat away from specimen. Accepts films, fabrics, foams, and other sheets, rigid or flexible. Specimens thin or thick, or of variable thickness. Bent to accommodate curved rigid specimens. Also used for such other tests as particle-impact tests.

  14. Irradiation programme HFR phase IIb - SPICE. Impact testing on up to 16.3 dpa irradiated RAFM steels. Final report for task TW2-TTMS 001b-D05

    Energy Technology Data Exchange (ETDEWEB)

    Gaganidze, E.; Dafferner, B.; Ries, H.; Rolli, R.; Schneider, H.C.; Aktaa, J.

    2008-04-15

    The objective of this work is to study the effects of neutron irradiation on the embrittlement behavior of different reduced activation ferritic/martensitic (RAFM) steels. The irradiation was carried out in the Petten High Flux Reactor (HFR) in the framework of the HFR Phase IIb (SPICE) irradiation project at a nominal dose of 16.3 dpa and at different irradiation temperatures (250-450 C). The impact properties are investigated by instrumented Charpy-V tests with subsize specimens (KLST-type). The emphasis is put on the investigation of irradiation induced embrittlement and hardening of the European RAFM reference steel for the first wall of a DEMO fusion reactor, EURO- FER97 under different heat treatment conditions. The mechanical properties of EUROFER97 are compared with the results on international reference steels (F82H-mod, OPTIFER-Ia, GA3X and MANET-I) included in the SPICE project. EUROFER97 irradiated up to 16.3 dpa between 250 and 450 C showed irradiation resistance that is comparable to those of best RAFM steels. Large low temperature (T{sub irr} {<=} 300 C) embrittlement is seen for all investigated RAFM steels. Heat treatment of EUROFER97 at higher austenitizing temperature led to the reduction of embrittlement at low temperatures (T{sub irr} {<=} 350 C). At T{sub irr} {>=} 350 C the DBTTs of the steels remain below -20 C and, hence, are well below the application temperature. Analysis of hardening vs. embrittlement behaviour indicated hardening dominated embrittlement at T{sub irr} {<=} 350 C with 0.17 {<=} C{sub 100} {<=} 0.53 C/MPa. Oxygen dispersion hardened ODS EUROFER with 0.5 wt.% Y{sub 2}O{sub 3} has been irradiated at selected irradiation temperatures. ODS EUROFER showed not satisfying impact properties already in the unirradiated condition characterized by low USE = 2.54 J and large DBTT = 135 C. Furthermore, the increase of USE for irradiation temperatures below T{sub irr} {<=} 350 C indicates not optimized fabrication process. At T{sub irr

  15. NASA Biological Specimen Repository

    Science.gov (United States)

    McMonigal, K. A.; Pietrzyk, R. A.; Sams, C. F.; Johnson, M. A.

    2010-01-01

    The NASA Biological Specimen Repository (NBSR) was established in 2006 to collect, process, preserve and distribute spaceflight-related biological specimens from long duration ISS astronauts. This repository provides unique opportunities to study longitudinal changes in human physiology spanning may missions. The NBSR collects blood and urine samples from all participating ISS crewmembers who have provided informed consent. These biological samples are collected once before flight, during flight scheduled on flight days 15, 30, 60, 120 and within 2 weeks of landing. Postflight sessions are conducted 3 and 30 days after landing. The number of in-flight sessions is dependent on the duration of the mission. Specimens are maintained under optimal storage conditions in a manner that will maximize their integrity and viability for future research The repository operates under the authority of the NASA/JSC Committee for the Protection of Human Subjects to support scientific discovery that contributes to our fundamental knowledge in the area of human physiological changes and adaptation to a microgravity environment. The NBSR will institute guidelines for the solicitation, review and sample distribution process through establishment of the NBSR Advisory Board. The Advisory Board will be composed of representatives of all participating space agencies to evaluate each request from investigators for use of the samples. This process will be consistent with ethical principles, protection of crewmember confidentiality, prevailing laws and regulations, intellectual property policies, and consent form language. Operations supporting the NBSR are scheduled to continue until the end of U.S. presence on the ISS. Sample distribution is proposed to begin with selections on investigations beginning in 2017. The availability of the NBSR will contribute to the body of knowledge about the diverse factors of spaceflight on human physiology.

  16. Modelling of Specimen Fracture

    Science.gov (United States)

    2013-09-23

    the plate center. An end load of 1.0 MPa was applied. 1 2 3 Modelling of Specimen Fracture – Final Report 11 TR-13-47 Figure 2.5: Crack Geometry Figure...Christopher Bayley DRDC Atlantic Dockyard Laboratory Pacific CFB Esquimalt, Building 199 PO Box 17000, Station Forces Victoria, British Columbia Canada...q The weighting function, q , can be any arbitrary function within the J-integral domain, and must be zero on the domain boundary . An easy function

  17. Labeling of Patient Specimens

    Science.gov (United States)

    2011-01-26

    noted during the event that the actu.al number of near miss incidmts reported monthly was low due to laboratory personnel performing rounds each...specimens never leaves label and if moved it is labeled), All orders in system and all near misses and errors reported to patient safety Purchase/Install...Meeting 14 Aug 09, 1400 in lab break room thru out Develop TICK sheet to track near misses .JDI Ms. Clark Clinics will provide toPS 1st working day of

  18. Instrumented Charpy : analysis of the instrumentation and the effect of different metallurgical conditions of an ultra-high strength steel on the dynamic fracture toughness

    OpenAIRE

    2009-01-01

    Resumo: O ensaio Charpy clássico, utilizado desde o início do século XX, permite determinar a energia global de fratura, propriedade mecânica conhecida como tenacidade. Esta energia tem caráter qualitativo e na aplicação no projeto estrutural, reduz-se a comparações entre as curvas de transição dúctil-frágil dos materiais. Com o surgimento da instrumentação do ensaio Charpy clássico é possível determinar a propriedade mecânica, tenacidade à fratura dinâmica, KId, muito utilizada na Mecânica d...

  19. Post-deformation examination of specimens subjected to SCC testing

    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); Busby, Jeremy T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Leonard, Keith J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-01

    This report details the results of post-radiation and post-deformation characterizations performed during FY 2015–FY 2016 on a subset of specimens that had previously been irradiated at high displacement per atom (dpa) damage doses. The specimens, made of commercial austenitic stainless steels and alloys, were subjected to stress-corrosion cracking tests (constant extension rate testing and crack growth testing) at the University of Michigan under conditions typical of nuclear power plants. After testing, the specimens were returned to Oak Ridge National Laboratory (ORNL) for further analysis and evaluation.

  20. a Study of Stress Relaxation Rate in Un-Irradiated and Neutron-Irradiated Stainless Steel

    Science.gov (United States)

    Ghauri, I. M.; Afzal, Naveed; Zyrek, N. A.

    Stress relaxation rate in un-irradiated and neutron-irradiated 303 stainless steel was investigated at room temperature. The specimens were exposed to 100 mC, Ra-Be neutron source of continuous energy 2-12 MeV for a period ranging from 4 to 16 days. The tensile deformation of the specimens was carried out using a Universal Testing Machine at 300 K. During the deformation, straining was frequently interrupted by arresting the cross head to observe stress relaxation at fixed load. Stress relaxation rate, s, was found to be stress dependent i.e. it increased with increasing stress levels σ0 both in un-irradiated and irradiated specimens, however the rate was lower in irradiated specimens than those of un-irradiated ones. A further decrease in s was observed with increase in exposure time. The experiential decrease in the relaxation rate in irradiated specimens is ascribed to strong interaction of glide dislocations with radiation induced defects. The activation energy for the movement of dislocations was found to be higher in irradiated specimens as compared with the un-irradiated ones.

  1. Effect of low temperatures on charpy impact toughness of austempered ductile irons

    Science.gov (United States)

    Riabov, Mikhail V.; Lerner, Yury S.; Fahmy, Mohammed F.

    2002-10-01

    Impact properties of standard American Society for Testing Materials (ASTM) grades of austempered ductile iron (ADI) were evaluated at subzero temperatures in unnotched and V-notched conditions and compared with ferritic and pearlitic grades of ductile irons (DIs). It was determined that there is a decrease in impact toughness for all ADI grades when there is a decrease in content of retained austenite and a decrease in test temperature, from room temperature (RT) to -60 °C. However, the difference in impact toughness values was not so noticeable for low retained austenite containing grade 5 ADI at both room and subzero temperatures as it was for ADI grade 1. Furthermore, the difference in impact toughness values of V-notched specimens of ADI grades 1 and 5 tested at -40 °C was minimal. The impact behaviors of ADI grade 5 and ferritic DI were found to be more stable than those of ADI grades 1, 2, 3, and 4 and pearlitic DI when the testing temperature was decreased. The impact toughness of ferritic DI was higher than that of ADI grades 1 and 2 at both -40 °C and -60 °C. The impact properties of ADI grades 4 and 5 were found to be higher than that of pearlitic DI at both -40 °C and -60 °C. The scanning electron microscopy (SEM) study of fracture surfaces revealed mixed ductile and quasicleavage rupture morphology types in all ADI samples tested at both -40 °C and -60 °C. With decreasing content of retained austenite and ductility, the number of quasicleavage facets increased from ADI grade 1-5. It was also found that fracture morphology of ADI did not experience significant changes when the testing temperature decreased. Evaluation of the bending angle was used to support impact-testing data. Designers and users of ADI castings may use the data developed in this research as a reference.

  2. ATF Neutron Irradiation Program Irradiation Vehicle Design Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Geringer, J. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Cetiner, N. O. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; McDuffee, J. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division

    2016-03-01

    The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post irradiation examination and characterization of irradiated materials and the shipment of irradiated materials to Japan. This report discusses the conceptual design, the development and irradiation of the test vehicles.

  3. Type specimen studies in Pleurotus

    NARCIS (Netherlands)

    Petersen, Ronald H.; Krisai-Greilhuber, Irmgard

    1999-01-01

    An epitype specimen is designated for Pleurotus cornucopiae. Morphological examination of Mexican material and the type specimen of P. opuntiae showed that the distribution of this species includes North Africa and the highlands of Mexico. The type specimen of Lentinus (Pleurotus) eugrammus reveals

  4. Study of irradiation creep of vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Strain, R.V.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    Thin-wall tubing was produced from the 832665 (500 kg) heat of V-4 wt.% Cr-4 wt.% Ti to study its irradiation creep behavior. The specimens, in the form of pressurized capsules, were irradiated in Advanced Test Reactor and High Flux Isotope Reactor experiments (ATR-A1 and HFIR RB-12J, respectively). The ATR-A1 irradiation has been completed and specimens from it will soon be available for postirradiation examination. The RB-12J irradiation is not yet complete.

  5. Views of TAGSI on effects of neutron irradiation on ductile tearing in ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Knott, J.F. [School of Metallurgy and Materials, University of Birmingham, Birmingham B15 2TT (United Kingdom); Lidbury, D.P.G. [Serco Technical and Assurance Services, Walton House, 404 Faraday Street, Birchwood Park, Warrington WA3 6GA (United Kingdom)], E-mail: david.lidbury@serco.com

    2009-07-15

    The paper reviews information pertaining to effects of neutron irradiation on 'upper-shelf' Charpy impact behaviour and on elastic/plastic fracture mechanics characterising parameters, again for 'upper shelf' conditions, in which the initiation and early growth of a crack involve ductile tearing. The hardening and associated reduction in strain-hardening capacity induced by irradiation gives rise to a decrease in Charpy upper shelf energy. Effects on J-based parameters are more complicated. The material resistance parameters tend to increase for low dose, but decrease at high dose, when the decrease in crack-tip ductility outweighs the effect of hardening. High doses can produce 'fast shear' fracture, which propagates rapidly and is therefore more likely to induce brittle cleavage fracture. The situation is exacerbated if the irradiation also promotes inter-granular segregation and fracture, hence reducing the local brittle fracture stress. For the levels of irradiation experienced by the types of UK civil reactors in operation, no fracture instability is expected to arise as a result of ductile fracture mechanisms alone.

  6. Current understanding of the effects of enviromental and irradiation variables on RPV embrittlement

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G.R.; Lucas, G.E.; Wirth, B.; Liu, C.L. [Univ. of California, Santa Barbara, CA (United States)

    1997-02-01

    Radiation enhanced diffusion at RPV operating temperatures around 290{degrees}C leads to the formation of various ultrafine scale hardening phases, including copper-rich and copper-catalyzed manganese-nickel rich precipitates. In addition, defect cluster or cluster-solute complexes, manifesting a range of thermal stability, develop under irradiation. These features contribute directly to hardening which in turn is related to embrittlement, manifested as shifts in Charpy V-notch transition temperature. Models based on the thermodynamics, kinetics and micromechanics of the embrittlement processes have been developed; these are broadly consistent with experiment and rationalize the highly synergistic effects of most important irradiation (temperature, flux, fluence) and metallurgical (copper, nickel, manganese, phosphorous and heat treatment) variables on both irradiation hardening and recovery during post-irradiation annealing. A number of open questions remain which can be addressed with a hierarchy of new theoretical and experimental tools.

  7. Neutron irradiation effects on the ductile-brittle transition of ferritic/martensitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Alexander, D.J. [Oak Ridge National Lab., TN (United States)

    1997-08-01

    Ferritic/martensitic steels such as the conventional 9Cr-1MoVNb (Fe-9Cr-1Mo-0.25V-0.06Nb-0.1C) and 12Cr-1MoVW (Fe-12Cr-1Mo-0.25V-0.5W-0.5Ni-0.2C) steels have been considered potential structural materials for future fusion power plants. The major obstacle to their use is embrittlement caused by neutron irradiation. Observations on this irradiation embrittlement is reviewed. Below 425-450{degrees}C, neutron irradiation hardens the steels. Hardening reduces ductility, but the major effect is an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy, as measured by a Charpy impact test. After irradiation, DBTT values can increase to well above room temperature, thus increasing the chances of brittle rather than ductile fracture.

  8. [Histological findings in an irradiated choroidal melanoma].

    Science.gov (United States)

    Koinzer, S; Hasselbach, H; Bräsen, J H; Leuschner, I; Roider, J

    2011-06-01

    Histological findings of choroidal melanomas after proton beam irradiation have been reported for complicated cases after enucleation. We present specimens of a tumor after transretinal probe excision. One year after irradiation, the biopsy was examined histologically. The specimens showed pigmented, spindle-shaped cells staining positively for Melan-A and HMB-45. Ki-67 showed low proliferation. Caspase-3 staining was normal. The melanoma still contained vital and even single proliferating cells, but regressed afterwards without additional therapy.

  9. Status of Post Irradiation Examination of FCAB and FCAT Irradiation Capsules

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD)

    2016-09-29

    A series of irradiation programs are ongoing to address the need for determining the radiation tolerance of FeCrAl alloys. These irradiation programs, deemed the FCAT and FCAB irradiation programs, use the High Flux Isotope Reactor (HFIR) to irradiate second generation wrought FeCrAl alloys and early-generation powder-metallurgy (PM) oxide dispersion-strengthened (ODS) FeCrAl alloys. Irradiations have been or are being performed at temperatures of 200°C, 330°C, and 550°C from doses of 1.8 dpa up to 16 dpa. Preliminary post-irradiation examination (PIE) on low dose (<2 dpa) irradiation capsules of tensile specimens has been performed. Analysis of co-irradiated SiC thermometry have shown reasonable matching between the nominal irradiation temperatures and the target irradiation temperatures. Room temperature tensile tests have shown typical radiation-induced hardening and embrittlement at irradiations of 200°C and 330°C, but a propensity for softening when irradiated to 550°C for the wrought alloys. The PM-ODS FeCrAl specimens showed less hardening compared to the wrought alloys. Future PIE includes high temperature tensile tests on the low dose irradiation capsules as well as the determination of reference fracture toughness transition temperature, To, in alloys irradiated to 7 dpa and higher.

  10. Status of Post Irradiation Examination of FCAB and FCAT Irradiation Capsules

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD)

    2016-09-29

    A series of irradiation programs are ongoing to address the need for determining the radiation tolerance of FeCrAl alloys. These irradiation programs, deemed the FCAT and FCAB irradiation programs, use the High Flux Isotope Reactor (HFIR) to irradiate second generation wrought FeCrAl alloys and early generation powder-metallurgy (PM) oxide dispersion strengthened (ODS) FeCrAl alloys. Irradiations have been or are being performed at temperatures of 200°C, 330°C, and 550°C from doses of 1.8 dpa up to 16 dpa. Preliminary post-irradiation examination (PIE) on low dose (<2 dpa) irradiation capsules of tensile specimens has been performed. Analysis of co-irradiated SiC thermometry have shown reasonable matching between the nominal irradiation temperatures and the target irradiation temperatures. Room temperature tensile tests have shown typical radiation-induced hardening and embrittlement at irradiations of 200°C and 330°C but a propensity for softening when irradiated to 550°C for the wrought alloys. The PM-ODS FeCrAl specimens showed less hardening compared to the wrought alloys. Future PIE includes high temperature tensile tests on the low dose irradiation capsules as well as the determination of reference fracture toughness transition temperature, To, in alloys irradiated to 7 dpa and higher.

  11. Rows of Dislocation Loops in Aluminium Irradiated by Aluminium Ions

    DEFF Research Database (Denmark)

    Henriksen, L.; Johansen, A.; Koch, J.

    1967-01-01

    Single-crystal aluminium specimens, irradiated with 50-keV aluminium ions, contain dislocation loops that are arranged in regular rows along <110 > directions. ©1967 The American Institute of Physics......Single-crystal aluminium specimens, irradiated with 50-keV aluminium ions, contain dislocation loops that are arranged in regular rows along directions. ©1967 The American Institute of Physics...

  12. Microstructure and mechanical properties of neutron irradiated OFHC-copper before and after post-irradiation annealing

    DEFF Research Database (Denmark)

    Singh, B.N.; Edwards, D.J.; Toft, P.

    2001-01-01

    Tensile specimens of OFHC-copper were irradiated with fission neutrons in the DR-3 reactor at Risø National Laboratory at 100 deg. C to different displacement dose levels in the range of 0.01 to 0.3 dpa (NRT). Some of the specimens were tensile tested inthe as-irradiated condition at 100 deg. C...

  13. Mechanical response of proton beam irradiated nitinol

    Energy Technology Data Exchange (ETDEWEB)

    Afzal, Naveed [Centre for Advanced Studies in Physics, GC University, Lahore (Pakistan); Ghauri, I.M., E-mail: ijaz.phys@gmail.co [Centre for Advanced Studies in Physics, GC University, Lahore (Pakistan); Mubarik, F.E.; Amin, F. [Centre for Advanced Studies in Physics, GC University, Lahore (Pakistan)

    2011-01-01

    The present investigation deals with the study of mechanical behavior of proton beam irradiated nitinol at room temperature. The specimens in austenitic phase were irradiated over periods of 15, 30, 45 and 60 min at room temperature using 2 MeV proton beam obtained from Pelletron accelerator. The stress-strain curves of both unirradiated and irradiated specimens were obtained using a universal testing machine at room temperature. The results of the experiment show that an intermediate rhombohedral (R) phase has been introduced between austenite and martensite phase, which resulted in the suppression of direct transformation from austenite to martensite (A-M). Stresses required to start R-phase ({sigma}{sub RS}) and martensitic phase ({sigma}{sub MS}) were observed to decrease with increase in exposure time. The hardness tests of samples before and after irradiation were also carried out using Vickers hardness tester. The comparison reveals that the hardness is higher in irradiated specimens than that of the unirradiated one. The increase in hardness is quite sharp in specimens irradiated for 15 min, which then increases linearly as the exposure time is increased up to 60 min. The generation of R-phase, variations in the transformation stresses {sigma}{sub RS} and {sigma}{sub MS} and increase in hardness of irradiated nitinol may be attributed to lattice disorder and associated changes in crystal structure induced by proton beam irradiation.

  14. Visual interface for the automation of the instrumented pendulum of Charpy tests used in the surveillance program of reactors vessel of nuclear power plants; Interfase visual para la automatizacion del pendulo instrumentado de pruebas Charpy utilizado en el programa de vigilancia de la vasija de reactores de centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Rojas S, A.S.; Sainz M, E.; Ruiz E, J.A. [ININ, Carretera Mexico-Toluca Km.36.5, Mpio. de Ocoyoacac, Estado de Mexico (Mexico)]. E-mail: asrs@nuclear.inin.mx; esm@nuclear.inin.mx; jare@nuclear.inin.mx

    2004-07-01

    Inside the Programs of Surveillance of the nuclear power stations periodic information is required on the state that keep the materials with those that builds the vessel of the reactor. This information is obtained through some samples or test tubes that are introduced inside the core of the reactor and it is observed if its physical characteristics remain after having been subjected to the radiation changes and temperature. The rehearsal with the instrumented Charpy pendulum offers information on the behavior of fracture dynamics of a material. In the National Institute of Nuclear Research (ININ) it has an instrumented Charpy pendulum. The operation of this instrument is manual, having inconveniences to carry out rehearsals with radioactive material, handling of high and low temperatures, to fulfill the normative ones for the realization of the rehearsals, etc. In this work the development of a computational program is presented (virtual instrument), for the automation of the instrumented pendulum. The system has modules like: Card of data acquisition, signal processing, positioning system, tempered system, pneumatic system, compute programs like it is the visual interface for the operation of the instrumented Charpy pendulum and the acquisition of impact signals. This system shows that given the characteristics of the nuclear industry with radioactive environments, the virtual instrumentation and the automation of processes can contribute to diminish the risks to the personnel occupationally exposed. (Author)

  15. 2003 Dead Bald Eagle Specimen

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — The specimen report states the Bald Eagle was found along the side of the I-95 by a motorist who contacted Santee National Wildlife Refuge. The Bald Eagle was taken...

  16. Manufacturing of Plutonium Tensile Specimens

    Energy Technology Data Exchange (ETDEWEB)

    Knapp, Cameron M [Los Alamos National Laboratory

    2012-08-01

    Details workflow conducted to manufacture high density alpha Plutonium tensile specimens to support Los Alamos National Laboratory's science campaigns. Introduces topics including the metallurgical challenge of Plutonium and the use of high performance super-computing to drive design. Addresses the utilization of Abaqus finite element analysis, programmable computer numerical controlled (CNC) machining, as well as glove box ergonomics and safety in order to design a process that will yield high quality Plutonium tensile specimens.

  17. Reprocessing technology development for irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, H.; Sakamoto, N. [Oarai Research Establishment, Ibaraki-ken (Japan); Tatenuma, K. [KAKEN Co., Ibaraki-ken (Japan)] [and others

    1995-09-01

    At present, beryllium is under consideration as a main candidate material for neutron multiplier and plasma facing material in a fusion reactor. Therefore, it is necessary to develop the beryllium reprocessing technology for effective resource use. And, we have proposed reprocessing technology development on irradiated beryllium used in a fusion reactor. The preliminary reprocessing tests were performed using un-irradiated and irradiated beryllium. At first, we performed beryllium separation tests using un-irradiated beryllium specimens. Un-irradiated beryllium with beryllium oxide which is a main impurity and some other impurities were heat-treated under chlorine gas flow diluted with Ar gas. As the results high purity beryllium chloride was obtained in high yield. And it appeared that beryllium oxide and some other impurities were removed as the unreactive matter, and the other chloride impurities were separated by the difference of sublimation temperature on beryllium chloride. Next, we performed some kinds of beryllium purification tests from beryllium chloride. And, metallic beryllium could be recovered from beryllium chloride by the reduction with dry process. In addition, as the results of separation and purification tests using irradiated beryllium specimens, it appeared that separation efficiency of Co-60 from beryllium was above 96%. It is considered that about 4% Co-60 was carried from irradiated beryllium specimen in the form of cobalt chloride. And removal efficiency of tritium from irradiated beryllium was above 95%.

  18. NSUF Irradiated Materials Library

    Energy Technology Data Exchange (ETDEWEB)

    Cole, James Irvin [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The Nuclear Science User Facilities has been in the process of establishing an innovative Irradiated Materials Library concept for maximizing the value of previous and on-going materials and nuclear fuels irradiation test campaigns, including utilization of real-world components retrieved from current and decommissioned reactors. When the ATR national scientific user facility was established in 2007 one of the goals of the program was to establish a library of irradiated samples for users to access and conduct research through competitively reviewed proposal process. As part of the initial effort, staff at the user facility identified legacy materials from previous programs that are still being stored in laboratories and hot-cell facilities at the INL. In addition other materials of interest were identified that are being stored outside the INL that the current owners have volunteered to enter into the library. Finally, over the course of the last several years, the ATR NSUF has irradiated more than 3500 specimens as part of NSUF competitively awarded research projects. The Logistics of managing this large inventory of highly radioactive poses unique challenges. This document will describe materials in the library, outline the policy for accessing these materials and put forth a strategy for making new additions to the library as well as establishing guidelines for minimum pedigree needed to be included in the library to limit the amount of material stored indefinitely without identified value.

  19. Irradiation damage

    Energy Technology Data Exchange (ETDEWEB)

    Howe, L.M

    2000-07-01

    There is considerable interest in irradiation effects in intermetallic compounds from both the applied and fundamental aspects. Initially, this interest was associated mainly with nuclear reactor programs but it now extends to the fields of ion-beam modification of metals, behaviour of amorphous materials, ion-beam processing of electronic materials, and ion-beam simulations of various kinds. The field of irradiation damage in intermetallic compounds is rapidly expanding, and no attempt will be made in this chapter to cover all of the various aspects. Instead, attention will be focused on some specific areas and, hopefully, through these, some insight will be given into the physical processes involved, the present state of our knowledge, and the challenge of obtaining more comprehensive understanding in the future. The specific areas that will be covered are: point defects in intermetallic compounds; irradiation-enhanced ordering and irradiation-induced disordering of ordered alloys; irradiation-induced amorphization.

  20. Definition of the minimum longitude of insert in the rebuilding of Charpy test tubes for surveillance and life extension of vessels in Mexico; Definicion de la longitud minima de inserto en la reconstitucion de probetas Charpy para vigilancia y extension de vida de vasijas en Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Romero C, J.; Hernandez C, R.; Rocamontes A, M., E-mail: jesus.romero@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    In the National Institute of Nuclear Research (Mexico) a welding system for the rebuilding of Charpy test tubes has been developed, automated, qualified and used for the surveillance of the mechanical properties (mainly embrittlement) of the vessel. This system uses the halves of the rehearsed Charpy test tubes of the surveillance capsules extracted of the reactors, to obtain, of a rehearsed test tube, two reconstituted test tubes. This rebuilding process is used so much in the surveillance program like in the potential extension of the operation license of the vessel. To the halves of Charpy test tubes that have been removed the deformed part by machine are called -insert- and in a very general way the rebuilding consists in weld with the welding process -Stud Welding- two metallic implants in the ends of the insert, to obtain a reconstituted test tube. The main characteristic of this welding are the achieved small dimensions, so much of the areas welded as of the areas affected by the heat. The applicable normative settles down that the minim longitude of the insert for the welding process by Stud Welding it should be of 18 mm, however according to the same normative this longitude can diminish if is demonstrated analytic or experimentally that the central volume of 1 cm{sup 3} in the insert is not affected. In this work the measurement of the temperature profiles to different distances of the welding interface is presented, defining an equation for the maximum temperatures reached in function of the distance, on the other hand the real longitude affected in the test tube by means of metallography is determined and this way the minimum longitude of the insert for this developed rebuilding system was determined. (Author)

  1. Modeling and Testing Miniature Torsion Specimens for SiC Joining Development Studies for Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Henager, Charles H.; Nguyen, Ba Nghiep; Kurtz, Richard J.; Roosendaal, Timothy J.; Borlaug, Brennan A.; Ferraris, Monica; Ventrella, Andrea; Katoh, Yutai

    2015-08-19

    The international fusion community has designed a miniature torsion specimen for neutron irradiation studies of joined SiC and SiC/SiC composite materials. Miniature torsion joints based on this specimen design were fabricated using displacement reactions between Si and TiC to produce Ti3SiC2 + SiC joints with CVD-SiC and tested in torsion-shear prior to and after neutron irradiation. However, many of these miniature torsion specimens fail out-of-plane within the CVD-SiC specimen body, which makes it problematic to assign a shear strength value to the joints and makes it difficult to compare unirradiated and irradiated joint strengths to determine the effects of the irradiation. Finite element elastic damage and elastic-plastic damage models of miniature torsion joints are developed that indicate shear fracture is likely to occur within the body of the joined sample and cause out-of-plane failures for miniature torsion specimens when a certain modulus and strength ratio between the joint material and the joined material exists. The model results are compared and discussed with regard to unirradiated and irradiated joint test data for a variety of joint materials. The unirradiated data includes Ti3SiC2 + SiC/CVD-SiC joints with tailored joint moduli, and includes steel/epoxy and CVD-SiC/epoxy joints. The implications for joint data based on this sample design are discussed.

  2. Microstructural change on electron irradiated oxide dispersion strengthened ferritic steels

    Science.gov (United States)

    Kinoshita, H.; Akasaka, N.; Takahashi, H.; Shibahara, I.; Onose, S.

    1992-09-01

    Oxide dispersion strengthened (ODS) ferritic steels were irradiated in a high voltage electron microscope (HVEM) to study their response to irradiation. Fe-13Cr with 0.25 wt% Y2O3 as dispersed particles and containing additions of either 0.45% Nb, 0.45% V and 0.67% Zr were irradiated at 673 and 723 K up to 15 dpa. The Y2O3 particles in all specimens were stable under these irradiation conditions. During irradiation, two types of dislocations were formed but observable voids were not formed. Furthermore, plate-like and granular-like precipitates formed in both the irradiated and nonirradiated regions.

  3. Microstructure property analysis of HFIR-irradiated reduced-activation ferritic/martensitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, H. E-mail: tanigawa@popsvr.tokai.jaeri.go.jp; Hashimoto, N.; Sakasegawa, H.; Klueh, R.L.; Sokolov, M.A.; Shiba, K.; Jitsukawa, S.; Kohyama, A

    2004-08-01

    The effects of irradiation on the Charpy impact properties of reduced-activation ferritic/martensitic steels were investigated on a microstructural basis. It was previously reported that the ductile-brittle transition temperature (DBTT) of F82H-IEA and its heat treatment variant increased by about 130 K after irradiation at 573 K up to 5 dpa. Moreover, the shifts in ORNL9Cr-2WVTa and JLF-1 steels were much smaller, and the differences could not be interpreted as an effect of irradiation hardening. The precipitation behavior of the irradiated steels was examined by weight analysis and X-ray diffraction analysis on extraction residues, and SEM/EDS analysis was performed on extraction replica samples and fracture surfaces. These analyses suggested that the difference in the extent of DBTT shift could be explained by (1) smaller irradiation hardening at low test temperatures caused by irradiation-induced lath structure recovery (in JLF-1), and (2) the fracture stress increase caused by the irradiation-induced over-solution of Ta (in ORNL9Cr-2WVTa)

  4. Food irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Webb, T.

    1986-01-01

    The proposed use of gamma radiation from cobalt 60 and cesium 137 for food irradiation in the United Kingdom is discussed, with particular reference to the possible dangers and disadvantages to the safety and wholesomeness of the food.

  5. Irradiation creep relaxation of void swelling-driven stresses

    Energy Technology Data Exchange (ETDEWEB)

    Hall, M.M., E-mail: hallmm63@comcast.net [MacRay Consulting, 1366 Hillsdale Drive, Monroeville, PA (United States)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Irradiation void swelling can cause distortion of reactor core components. Black-Right-Pointing-Pointer Constrained swelling can drive stresses beyond acceptable levels. Black-Right-Pointing-Pointer Compressive stresses decrease irradiation swelling rates. Black-Right-Pointing-Pointer Irradiation creep relaxes swelling-driven stresses and core restraint forces. Black-Right-Pointing-Pointer Swelling-driven creep stresses are consistent with predictions of a proposed model. - Abstract: Swelling-driven-creep test specimens are used to measure the compressive stresses that develop due to constraint of irradiation void swelling. These specimens use a previously non-irradiated 20% CW Type 316 stainless steel holder to axially restrain two Type 304 stainless steel tubular specimens that were previously irradiated in the US Experimental Breeder Reactor (EBR-II) at 490 Degree-Sign C. One specimen was previously irradiated to fluence levels in the void nucleation regime (9 dpa) and the other in the quasi-steady void growth regime (28 dpa). A lift-off compliance measurement technique was used post-irradiation to determine compressive stresses developed during reirradiation of the two specimen assemblies in Row 7 of EBR-II at temperatures of 547 Degree-Sign C and 504 Degree-Sign C, respectively, to additional damage levels each of about 5 dpa. Results obtained on the higher fluence swelling-driven-creep specimen show that compressive stress due to constraint of swelling retards void swelling to a degree that is consistent with active load uniaxial compression specimens that were irradiated as part of a previously reported multiaxial in-reactor creep experiment. Swelling results obtained on the lower fluence swelling-driven creep specimen show a much larger effect of compressive stress in reducing swelling, demonstrating that the larger effect of stress on swelling is on void nucleation as compared to void growth. Test results are

  6. [Food irradiation].

    Science.gov (United States)

    Migdał, W

    1995-01-01

    A worldwide standard on food irradiation was adopted in 1983 by Codex Alimentarius Commission of the Joint Food Standard Programme of the Food and Agriculture Organization (FAO) of the United Nations and the World Health Organization (WHO). As a result, 41 countries have approved the use of irradiation for treating one or more food items and the number is increasing. Generally, irradiation is used to: food loses, food spoilage, disinfestation, safety and hygiene. The number of countries which use irradiation for processing food for commercial purposes has been increasing steadily from 19 in 1987 to 33 today. In the frames of the national programme on the application of irradiation for food preservation and hygienization an experimental plant for electron beam processing has been established in Institute of Nuclear Chemistry and Technology. The plant is equipped with a small research accelerator Pilot (19MeV, 1 kW) and an industrial unit Elektronika (10MeV, 10 kW). On the basis of the research there were performed at different scientific institutions in Poland, health authorities have issued permission for irradiation for: spices, garlic, onions, mushrooms, potatoes, dry mushrooms and vegetables.

  7. Evaluation of Ion Irradiation Behavior of ODS Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jin Sung; Kim, Min Chul; Hong, Jun Hwa; Han, Chang Hee; Chang, Young Mun; Bae, Chang Soo; Bae, Yoon Young; Chang, Moon Hee

    2006-08-15

    FM steel (Grade 92) and ODS alloy(MA956) specimens were ion irradiated with 122 MeV Ne ions. Irradiation temperatures were about 450 and 550 .deg. C and the peak dose was 1, 5, and 10 dpa. Cross-sectional TEM samples were prepared by the electrolytic Ni-plating after pre-treatment of the irradiated specimens. Irradiation cavities in FM steel and ODS alloy specimens were not much different in size; about 20 nm in diameter in both specimens irradiated at around 450 .deg. C. However, the size distribution of cavities in FM steel specimens was broader than that in ODS alloy specimen, indicating that the cavity growth probably via coalescence). It was noticeable that the location and the preferential growth of the cavities in FM steel specimens: cavities on the PAGB (prior austenite grain boundary) was significantly larger than those within the grains. This could be an important issue for the mechanical properties, especially high temperature creep, fracture toughness, and so on. The dependency of the dose threshold and swelling on the ratio of the inert gas concentration/dpa was analysed for the various irradiation source, including He, Ne, Fe/He, and fast neutron, and the empirical correlation was established.

  8. Irradiation creep of dispersion strengthened copper alloy

    Energy Technology Data Exchange (ETDEWEB)

    Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A. [and others

    1997-04-01

    Dispersion strengthened copper alloys are under consideration as reference materials for the ITER plasma facing components. Irradiation creep is one of the parameters which must be assessed because of its importance for the lifetime prediction of these components. In this study the irradiation creep of a dispersion strengthened copper (DS) alloy has been investigated. The alloy selected for evaluation, MAGT-0.2, which contains 0.2 wt.% Al{sub 2}O{sub 3}, is very similar to the GlidCop{trademark} alloy referred to as Al20. Irradiation creep was investigated using HE pressurized tubes. The tubes were machined from rod stock, then stainless steel caps were brazed onto the end of each tube. The creep specimens were pressurized by use of ultra-pure He and the stainless steel caps subsequently sealed by laser welding. These specimens were irradiated in reactor water in the core position of the SM-2 reactors to a fluence level of 4.5-7.1 x 10{sup 21} n/cm{sup 2} (E>0.1 MeV), which corresponds to {approx}3-5 dpa. The irradiation temperature ranged from 60-90{degrees}C, which yielded calculated hoop stresses from 39-117 MPa. A mechanical micrometer system was used to measure the outer diameter of the specimens before and after irradiation, with an accuracy of {+-}0.001 mm. The irradiation creep was calculated based on the change in the diameter. Comparison of pre- and post-irradiation diameter measurements indicates that irradiation induced creep is indeed observed in this alloy at low temperatures, with a creep rate as high as {approx}2 x 10{sup {minus}9}s{sup {minus}1}. These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys.

  9. Specimen Collection and Submission Manual

    Science.gov (United States)

    2016-06-01

    a validated test algorithm . If there is insufficient specimen volume for testing, there may be delays, and the request may be referred to management...Additional guidance on packing and shipping infectious substances can be found through American Society for Microbiology: http://www.asm.org...images/pdf/Clinical/ pack -ship-7-15-2011.pdf TR-16-161 DISTRIBUTION STATEMENT A: Approved for public release; distribution is unlimited. UNCLASSIFIED

  10. Fungal contaminants in cytopathology specimens

    Directory of Open Access Journals (Sweden)

    Prashant Sharma

    2014-02-01

    Full Text Available A pseudo-epidemic of environmental fungi, most likely by Fusarium spp., leading to inappropriate investigations for disseminated systemic mycosis is described. Subtle diagnostic clues, including the specimens affected, the nature of the host response, and the type of fungal elements noted helped to determine the nature of contaminants. The potential pitfall can be avoided by the knowledge of pertinent disease biology, prompt consultation for infectious diseases, and investigations of the potential environmental sources followed by source control.

  11. Saturation behavior of irradiation hardening in F82H irradiated in the HFIR

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  12. Modeling and testing miniature torsion specimens for SiC joining development studies for fusion

    Energy Technology Data Exchange (ETDEWEB)

    Henager, C.H., E-mail: chuck.henager@pnnl.gov [Pacific Northwest National Laboratory, Richland, WA (United States); Nguyen, B.N.; Kurtz, R.J.; Roosendaal, T.J.; Borlaug, B.A. [Pacific Northwest National Laboratory, Richland, WA (United States); Ferraris, M.; Ventrella, A. [Politecnico di Torino, Torino (Italy); Katoh, Y. [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

    2015-11-15

    The international fusion community has designed a miniature torsion specimen for neutron irradiation studies of joined SiC and SiC/SiC composite materials. Miniature torsion joints based on this specimen design were fabricated using displacement reactions between Si and TiC to produce Ti{sub 3}SiC{sub 2} + SiC joints with SiC and tested in torsion-shear prior to and after neutron irradiation. However, many miniature torsion specimens fail out-of-plane within the SiC specimen body, which makes it problematic to assign a shear strength value to the joints and makes it difficult to compare unirradiated and irradiated strengths to determine irradiation effects. Finite element elastic damage and elastic–plastic damage models of miniature torsion joints are developed that indicate shear fracture is more likely to occur within the body of the joined sample and cause out-of-plane failures for miniature torsion specimens when a certain modulus and strength ratio between the joint material and the joined material exists. The model results are compared and discussed with regard to unirradiated and irradiated test data for a variety of joint materials. The unirradiated data includes Ti{sub 3}SiC{sub 2} + SiC/CVD-SiC joints with tailored joint moduli, and includes steel/epoxy and CVD-SiC/epoxy joints. The implications for joint data based on this sample design are discussed. - Highlights: • Finite element damage models developed and applied to understand miniature torsion specimen. • Damage models correctly predict torsion joint failure locations for wide range of materials. • Tests of strong, stiff ceramic joints will not produce accurate shear strength values. • Miniature torsion specimen has diminished test utility but still valuable.

  13. Ascertaining the micromechanical damage parameters using the small scale test specimens

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, N.N. (HBNI, RSD, BARC, Trombay (India)), e-mail: naveenm@barc.gov.in; Durgaprasad, P.V.; Dutta, B.K. (Reactor Safety Division, Bhabha Atomic Research Centre Trombay (India)); Dey, G.K. (Material Science Division, Bhabha Atomic Research Centre Trombay (India))

    2009-07-01

    Objective of the study is to ascertain the damage parameters and stress strain behaviour of material under irradiated condition. To achieve this goal, following methodology is employed; a) Elastic-plastic and micro-mechanical analysis of small punch test is carried out. From the elastic plastic analysis, friction factor between the ball and specimen is found. From micro mechanical analysis, Gurson damage parameters are calibrated by comparing simulation results with experimental result of unirradiated material; b) load-displacement behaviour of small punch tests are obtained by assuming the damage parameters are unchanged due to irradiation and with approximate shift in the stress strain curve; c) Comparing the above small punch results with experimental load displacement data of irradiated sample, the stress-strain data of irradiated samples is obtained. At the next stage, the fracture properties like J-R curve can be evaluated for standard CT specimens by employing the calibrated micromechanical damage parameters and stress strain data

  14. Irradiation embrittlement investigation of the Shippingport Station neutron shield tank

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T.; Shack, W.J.; Chopra, O.K.

    1989-01-01

    A joint effort between the US Nuclear Regulatory Commission and the US Department of Energy is under way to investigate the low- temperature, low-fluence rate embrittlement of reactor vessel support structures through analysis of the decommissioned Shippingport Station neutron shield tank (NST). The Shippingport NST operated at a temperature of 130/degree/F (55/degree/C) and saw a maximum fluence of approximately 6 /times/ 10/sup 17/ n/cm/sup 2/, E > 1 MeV. To characterize the embrittlement behavior of the NST material, eleven 6-inch diameter discs were removed from the irradiated inner shell of the NST along with the corresponding material from the slightly irradiated outer shell. Standard Charpy V-notch tests on the most highly irradiated NST material indicate a shift in the transition temperature (measured at 15 ft-lbs) of approximately 45/degree/F (25/degree/C) after 9.25 effective full power years of operation. This shift is not as severe as that expected based on surveillance results at the High Flux Isotope Reactor. The resultant transition temperature, however, is close to the NST service temperature. A low toughness at service temperature is also indicated. 10 refs., 8 figs., 1 tab.

  15. Status of lithium-filled specimen subcapsules for the HFIR-MFE-RB10J experiment

    Energy Technology Data Exchange (ETDEWEB)

    Robertson, J.P.; Howell, M.; Lenox, K.E. [Oak Ridge National Lab., TN (United States)

    1998-09-01

    The HFIR-MFE-RB-10J experiment will be irradiated in a Removable Beryllium position in the HFIR for 10 reactor cycles, accumulating approximately 5 dpa in steel. The upper region of the capsule contains two lithium-filled subcapsules containing vanadium specimens. This report describes the techniques developed to achieve a satisfactory lithium fill with a specimen occupancy of 26% in each subcapsule.

  16. Effect of heat treatment and irradiation temperature on impact behavior of irradiated reduced-activation ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Alexander, D.J. [Oak Ridge National Lab., TN (United States)

    1998-03-01

    Charpy tests were conducted on eight normalized-and-tempered reduced-activation ferritic steels irradiated in two different normalized conditions. Irradiation was conducted in the Fast Flux Test Facility at 393 C to {approx}14 dpa on steels with 2.25, 5, 9, and 12% Cr (0.1% C) with varying amounts of W, V, and Ta. The different normalization treatments involved changing the cooling rate after austenitization. The faster cooling rate produced 100% bainite in the 2.25 Cr steels, compared to duplex structures of bainite and polygonal ferrite for the slower cooling rate. For both cooling rates, martensite formed in the 5 and 9% Cr steels, and martensite with {approx}25% {delta}-ferrite formed in the 12% Cr steel. Irradiation caused an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy. The difference in microstructure in the low-chromium steels due to the different heat treatments had little effect on properties. For the high-chromium martensitic steels, only the 5 Cr steel was affected by heat treatment. When the results at 393 C were compared with previous results at 365 C, all but a 5 Cr and a 9 Cr steel showed the expected decrease in the shift in DBTT with increasing temperature.

  17. Investigation on the effects of gamma irradiation on bitumen

    Energy Technology Data Exchange (ETDEWEB)

    Mello, M.S.; Braz, D.; Motta, L.M.G., E-mail: Laura@coc.ufrj.br [Universidade Federal Rio de Janeiro (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Centro de Tecnologia; Leite, L.F.M., E-mail: leniml@petrobras.com.br [Centro de Pesquisa e Desenvolvimento Leopoldo Americo Miguez de Mello (CENPES/RJ), Rio de Janeiro (Brazil)

    2011-07-01

    Brazil has more than 218,000 km of asphalt-paved highways. Bitumen is a generic term for natural or manufactured black or dark-colored solid, semisolid, or viscous cementitious materials that are composed mainly of high molecular weight hydrocarbons (90-95%). Several papers have shown that the irradiation process has changed the mechanical behavior in some polymers. This work aims to analyze the behavior of Brazilian irradiated Bitumen (CAP 50-70). In order to provide a preliminary evaluation, bitumen samples and cylindrical specimens of asphaltic mixture were tested. The bitumen samples were irradiated 0.1 to 300 kGy, and asphaltic mixture specimen was irradiated 5 to 300 kGy. The cylindrical asphaltic mixture specimen of 10.16 cm diameter used in this study was molded using an asphalt-aggregate mixture. The specimens were irradiated in LIN/UFRJ/Brazil using a Gamma cell Co{sup 60} source of gamma irradiation with an applied dose rate of 29.7 Gy/min. After irradiated, the bitumen samples were subjected to penetration test and the asphaltic mixtures were subjected to indirect tensile strength test (diametral compression) for determination of the resilient modulus, according to ASTM method D 4123. The results of these experiments for each dose were compared with the control (nonirradiated). As expected, the penetration results showed that the ratio (irradiated/non-irradiated) decreases with increasing of irradiation dose for bitumen samples and the resilient modulus results showed that the ratio (irradiated/non-irradiated) increases with increasing of irradiation dose for asphaltic mixture. (author)

  18. Characterizing the transition region of an A508 cl3 steel using small specimens by the reference temperature and the weak-link distances

    Energy Technology Data Exchange (ETDEWEB)

    Miranda, C.A.J. [IPEN-CNEN/SP, Sao Paulo, SP (Brazil)

    2001-07-01

    An experimental program was developed to characterize the transition region of an A508 cl3 steel. Some fracture mechanic specimens were tested in the transition region using three geometries with thickness B < 1 T (CT, SENB and pre-cracked Charpy). Fractographic observations were made in the fracture surfaces to identify the weak-links and to measure their distances to the crack-tip. These distances are compared with the theoretical values that come from a deterministic methodology. The results are presented and discussed in terms of the obtained J{sub c} values, the reference temperature values, To, associated with each geometry and test temperature, and the measured r{sub wl} distances and the theoretical ones. (author)

  19. Hydraulically Driven Grips For Hot Tensile Specimens

    Science.gov (United States)

    Bird, R. Keith; Johnson, George W.

    1994-01-01

    Pair of grips for tensile and compressive test specimens operate at temperatures up to 1,500 degrees F. Grips include wedges holding specimen inside furnace, where heated to uniform temperature. Hydraulic pistons drive wedges, causing them to exert clamping force. Hydraulic pistons and hydraulic fluid remain outside furnace, at room temperature. Cooling water flows through parts of grips to reduce heat transferred to external components. Advantages over older devices for gripping specimens in high-temperature tests; no need to drill holes in specimens, maintains constant gripping force on specimens, and heated to same temperature as that of specimen without risk of heating hydraulic fluid and acuator components.

  20. Irradiation creep of various ferritic alloys irradiated at ˜400°C in the PFR and FFTF reactors

    Science.gov (United States)

    Toloczko, M. B.; Garner, F. A.; Eiholzer, C. R.

    1998-10-01

    Irradiation creep of three ferritic alloys at ˜400 ∘C has been studied. Specimens were in the form of pressurized tubes. In a joint US/UK creep study, two identical sets of creep specimens constructed from one heat of HT9 were irradiated in fast reactors, one in the Prototypic Fast Reactor (PFR) and the other in the Fast Flux Test Facility (FFTF). The specimens in PFR were irradiated to a dose of ˜50 dpa, whereas the specimens in FFTF were irradiated to a dose of 165 dpa. The observed swelling and creep behavior were very different in the two reactors. Creep specimens constructed from D57, a developmental alloy ferritic alloy, were also irradiated in PFR to a dose of ˜50 dpa. Creep behavior typical of previous studies on ferritic alloys was observed. Finally, creep specimens constructed from MA957, a Y 2O 3 dispersion-hardened ferritic alloy, were irradiated in FFTF to a dose of ˜110 dpa. This alloy exhibited a large amount of densification, and the creep behavior was different than observed in more conventional ferritic or ferritic-martensitic alloys.

  1. AGC-1 Post Irradiation Examination Status

    Energy Technology Data Exchange (ETDEWEB)

    David Swank

    2011-09-01

    The Next Generation Nuclear Plant (NGNP) Graphite R&D program is currently measuring irradiated material property changes in several grades of nuclear graphite for predicting their behavior and operating performance within the core of new Very High Temperature Reactor (VHTR) designs. The Advanced Graphite Creep (AGC) experiment consisting of six irradiation capsules will generate this irradiated graphite performance data for NGNP reactor operating conditions. All six AGC capsules in the experiment will be irradiated in the Advanced Test Reactor (ATR), disassembled in the Hot Fuel Examination Facility (HFEF), and examined at the INL Research Center (IRC) or Oak Ridge National Laboratory (ORNL). This is the first in a series of status reports on the progress of the AGC experiment. As the first capsule, AGC1 was irradiated from September 2009 to January 2011 to a maximum dose level of 6-7 dpa. The capsule was removed from ATR and transferred to the HFEF in April 2011 where the capsule was disassembled and test specimens extracted from the capsules. The first irradiated samples from AGC1 were shipped to the IRC in July 2011and initial post irradiation examination (PIE) activities were begun on the first 37 samples received. PIE activities continue for the remainder of the AGC1 specimen as they are received at the IRC.

  2. 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)

  3. ATF Neutron Irradiation Program Technical Plan

    Energy Technology Data Exchange (ETDEWEB)

    Geringer, J. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division

    2016-03-01

    The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post irradiation examination and characterization of irradiated materials and the shipment of irradiated materials to JAEA in Japan. This report discusses the technical plan of the experimental study.

  4. ATF Neutron Irradiation Program Technical Plan

    Energy Technology Data Exchange (ETDEWEB)

    Geringer, J. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division

    2016-03-01

    The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post-irradiation examination and characterization of irradiated materials and the shipment of irradiated materials to JAEA in Japan. This report discusses the technical plan of the experimental study.

  5. Measurements and Counts for Notacanthidae Specimens

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Taxonomic data were collected for specimens of deep-sea spiny eels (Notacanthidae) from the Hawaiian Ridge by Bruce C. Mundy. Specimens were collected off the north...

  6. Comparison of microstructural properties and Charpy impact behaviour between different plates of the Eurofer97 steel and effect of isothermal ageing

    Science.gov (United States)

    Stratil, Ludek; Hadraba, Hynek; Bursik, Jiri; Dlouhy, Ivo

    2011-09-01

    The microstructure and fracture properties of the Eurofer97 steel plates of thickness 14 mm and 25 mm were investigated in as-received state and in state after long-term thermal ageing (550 °C/5000 h). Detailed microstructure studies were carried out by means of optical light, electron and quantitative electron microscopy. Mechanical properties were evaluated by means of Charpy impact testing and hardness testing and fracture surfaces were fractographically analysed in macro and microscales. The microstructure of the Eurofer97 consisted of tempered martensite with M 23C 6 and MX precipitates. Microstructure of 14 mm plate was more homogenous and fine grained than 25 mm plate. Due to different microstructure the tDBTT of thicker plate was on +10 °C higher than for 14 mm plate for which reached -60 °C. Slight microstructural changes on the level of subgrain consisting of their partial recrystallization and slight carbide coarsening were observed after applied ageing. The isothermal ageing caused evident shift in tDBTT about +5 °C, which was most likely caused by recrystallization of subgrains.

  7. ESR detection procedure of irradiated papaya containing high water content

    Science.gov (United States)

    Kikuchi, Masahiro; Shimoyama, Yuhei; Ukai, Mitsuko; Kobayashi, Yasuhiko

    2011-05-01

    ESR signals were recorded from irradiated papaya at liquid nitrogen temperature (77 K), and freeze-dried irradiated papaya at room temperature (295 K). Two side peaks from the flesh at the liquid nitrogen temperature indicated a linear dose response for 3-14 days after the γ-irradiation. The line shapes recorded from the freeze-dried specimens were sharper than those at liquid nitrogen temperature.

  8. ESR detection procedure of irradiated papaya containing high water content

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Masahiro, E-mail: kikuchi.masahiro@jaea.go.j [Quantum Beam Science Directorate, Japan Atomic Energy Agency, 1233 Watanuki-machi, Takasaki, Gunma 370-1292 (Japan); Shimoyama, Yuhei [Quantum Beam Science Directorate, Japan Atomic Energy Agency, 1233 Watanuki-machi, Takasaki, Gunma 370-1292 (Japan); Ukai, Mitsuko [Hokkaido University of Education, 1-2 Hachiman-cho, Hakodate, Hokkaido 040-8567 (Japan); Kobayashi, Yasuhiko [Quantum Beam Science Directorate, Japan Atomic Energy Agency, 1233 Watanuki-machi, Takasaki, Gunma 370-1292 (Japan)

    2011-05-15

    ESR signals were recorded from irradiated papaya at liquid nitrogen temperature (77 K), and freeze-dried irradiated papaya at room temperature (295 K). Two side peaks from the flesh at the liquid nitrogen temperature indicated a linear dose response for 3-14 days after the {gamma}-irradiation. The line shapes recorded from the freeze-dried specimens were sharper than those at liquid nitrogen temperature.

  9. Polishing methods for metallic and ceramic transmission electron microscopy specimens: Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Kestel, B.J.

    1986-03-01

    In recent years, the increasing sophistication of transmission electron microscope (TEM) studies of materials has necessitated more exacting methods of specimen preparation. The present report describes improved equipment and techniques for electropolishing and chemically polishing a wide variety of specimens. Many of the specimens used in developing or improving the techniques to be described were irradiated with heavy ions such as nickel or vanadium to study radiation damage. The high cost of these specimens increased the need for reproducible methods of initial preparation postirradiation processing, and final thinning for TEM examination. A technique was also developed to salvage specimens that had previously been thinned but were unusable for various reasons. Jet polishing is, in general, the method of choice for surface polishing, sectioning, and thinning. The older beaker electropolishing method is included in this report because it is inexpensive and simple, and gives some insight into how the more recent methods were developed.

  10. Demonstration of finite element simulations in MOOSE using crystallographic models of irradiation hardening and plastic deformation

    Energy Technology Data Exchange (ETDEWEB)

    Patra, Anirban [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wen, Wei [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martinez Saez, Enrique [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tome, Carlos [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-05-31

    This report describes the implementation of a crystal plasticity framework (VPSC) for irradiation hardening and plastic deformation in the finite element code, MOOSE. Constitutive models for irradiation hardening and the crystal plasticity framework are described in a previous report [1]. Here we describe these models briefly and then describe an algorithm for interfacing VPSC with finite elements. Example applications of tensile deformation of a dog bone specimen and a 3D pre-irradiated bar specimen performed using MOOSE are demonstrated.

  11. Features of structural response of mechanically loaded crystallites to irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Korchuganov, Aleksandr V., E-mail: avkor@ispms.ru [Institute of Strength Physics and Materials Science SB RAS, Tomsk, 634055 (Russian Federation); National Research Tomsk State University, Tomsk, 634050 (Russian Federation)

    2015-10-27

    A molecular dynamics method is employed to investigate the origin and evolution of plastic deformation in elastically deformed iron and vanadium crystallites due to atomic displacement cascades. Elastic stress states of crystallites result from different degrees of specimen deformation. Crystallites are deformed under constant-volume conditions. Atomic displacement cascades with the primary knock-on atom energy up to 50 keV are generated in loaded specimens. It is shown that irradiation may cause not only the Frenkel pair formation but also large-scale structural rearrangements outside the irradiated area, which prove to be similar to rearrangements proceeding by the twinning mechanism in mechanically loaded specimens.

  12. In-situ TEM ion irradiation investigations on U3Si2 at LWR temperatures

    Science.gov (United States)

    Miao, Yinbin; Harp, Jason; Mo, Kun; Bhattacharya, Sumit; Baldo, Peter; Yacout, Abdellatif M.

    2017-02-01

    The radiation-induced amorphization of U3Si2 was investigated by in-situ transmission electron microscopy using 1 MeV Kr ion irradiation. Both arc-melted and sintered U3Si2 specimens were irradiated at room temperature to confirm the similarity in their responses to radiation. The sintered specimens were then irradiated at 350 °C and 550 °C up to 7.2 × 1015 ions/cm2 to examine their amorphization behavior under light water reactor (LWR) conditions. U3Si2 remains crystalline under irradiation at LWR temperatures. Oxidation of the material was observed at high irradiation doses.

  13. Progress report on the design of a varying temperature irradiation experiment for operation in HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A.L. [Oak Ridge National Lab., TN (United States); Muroga, T.

    1997-04-01

    The purpose of this experiment is to determine effects of temperature variation during irradiation on microstructure and mechanical properties of potential fusion reactor structural materials. A varying temperature irradiation experiment is being performed under the framework of the Japan-USA Program of Irradiation Tests for fusion Research (JUPITER) to study the effects of temperature variation on the microstructure and mechanical properties of candidate fusion reactor structural materials. An irradiation capsule has been designed for operation in the High Flux Isotope Reactor at Oak Ridge National Laboratory that will allow four sets of metallurgical test specimens to be irradiated to exposure levels ranging from 5 to 10 dpa. Two sets of specimens will be irradiated at constant temperature of 500{degrees}C and 350{degrees}C. Matching specimen sets will be irradiated to similar exposure levels, with 10% of the exposure to occur at reduced temperatures of 300{degrees}C and 200{degrees}C.

  14. Development of hardened PVF : PMMA polyblend: effect of gamma and electron irradiation

    Indian Academy of Sciences (India)

    R Bajpai; N B Dhagat; R Katare; Pragyesh Agrawal; S C Datt

    2003-06-01

    Specimens of poly(vinyl formal) (PVF) : poly(methyl methacrylate) (PMMA) polyblends with different weight percentage ratios were subjected to gamma irradiation (1 to 50 Mrad) and electron irradiation (1 to 20 Mrad). The effect of irradiation on the strength of the blend specimens was studied by measuring the surface microhardness using a Vickers microhardness tester attached to a Carl Zeiss NU 2 Universal research microscope. Significant changes were observed in the Vickers microhardness number, $H_v$. The $H_v$ values of gamma irradiated specimens are found to be higher than the unirradiated specimens indicating an occurrence of radiational crosslinking. The maximum value of $H_v$ is obtained at the gamma radiation dose of 15 Mrad. In case of electron irradiation the radiational crosslinking is found to take place for the blend specimens having lower wt% content of PMMA (0 and 1 wt%) in PVF matrix. On the other hand degradation of polymeric system is observed for the blends having PMMA content more than 1 wt%. The maximum value of $H_v$ is obtained for all the blend specimens at the electron irradiation dose of 8 Mrad. The degree of crosslinking in polyblends due to gamma irradiation is found to be more than electron irradiation. The scissioning mechanism is found to predominate in the polyblend system in case of electron irradiation.

  15. Clinicopathological study of hysterectomised specimens

    Directory of Open Access Journals (Sweden)

    Saravana A.

    2016-12-01

    Full Text Available Background: Hysterectomy is the commonest major surgical procedure performed in gynecology. It can be done by abdominal or vaginal route and with the help of laparoscopy. Hysterectomy is an effective treatment option for many conditions like fibroid, abnormal uterine bleeding, endometriosis, adenomyosis, uterine prolapse, pelvic inflammatory disease and cancer of reproductive organ when other treatment options are contraindicated or have failed, or if the woman no longer wishes to retain her menstrual and reproductive. The aim and objective of the study was to correlate indications of hysterectomy with histopathological findings in hysterectomised patients. Methods: A retrospective study was carried on 113 hysterectomised cases over a period of one year from June 2015 to May 2016. The data regarding the patient’s age, parity, clinical diagnosis, type of hysterectomy and histopathological diagnosis were reviewed by the records and analyzed. Results: A total of 113 cases of hysterectomies were studied. Hysterectomies were distributed over a wide age ranging from 20 years to 75 years. Most common age group was 41-50 years. Among hystectomies majority were done through vaginal route 86 (76.1% and 26 (23% cases were done through abdominal route. Most common clinical diagnosis was fibroid uterus in 44(38.9% cases. Most of the hysterectomies were done for benign conditions. In final histopathological report most common diagnosis was fibroid uterus in 45(39.8% hysterectomy specimens. It was correlated well with clinical diagnosis. Next most common histopathological diagnosis was Adenomyosis. Conclusions: Histopathological analysis correlated well with preoperative clinical diagnosis in majority of cases. The commonest indication and histopathological finding in our study was fibroid uterus. Next most common histopathological finding was Adenomyosis. Most commonly hysterectomies were done through vaginal route in our study.

  16. [The German Environmental Specimen Bank].

    Science.gov (United States)

    Schröter-Kermani, Christa; Gies, Andreas; Kolossa-Gehring, Marike

    2016-03-01

    The main objective of the German Environmental Specimen Bank (ESB) is the long-term storage of environmental and human samples under stable deep-freeze conditions for future research. The ESB is unique in providing a continuous historical record of environmental and human exposure to chemicals in Germany. ESB was started parallel to the development of the first German Chemicals Legislation in the late 1970s. In 1979, the ESB test operation began. After the Chemicals Law came into force in 1982, the ESB was established as a permanent facility in 1985. With the new European Chemicals Legislation, REACH, in 2007 responsibility for the safety of commercial chemicals and risk assessment was assigned to the industry. Since then, the ESB has become even more important in verifying the self-assessment of the industry, in evaluating the effectiveness of regulations, thus ensuring the protection of humans and the environment against adverse effects caused by exposure to chemicals. These objectives are pursued by the regular monitoring of contaminations and the assessment of temporal trends. Demonstrating the necessity of deriving exposure reduction measures, ESB results serve as key information for policy-makers. Information on preventing exposure to chemicals is available to the general public and to the public health services. The ESB is thus an important monitoring instrument of the Federal Ministry for the Environment, Nature Conservation, Building and Nuclear Safety. The Federal Environment Agency operates the ESB based on its own concepts, heads the scientific data evaluation and transfers results into the environmental policy arena and to the general public.

  17. Nondestructive DNA extraction from museum specimens.

    Science.gov (United States)

    Hofreiter, Michael

    2012-01-01

    Natural history museums around the world hold millions of animal and plant specimens that are potentially amenable to genetic analyses. With more and more populations and species becoming extinct, the importance of these specimens for phylogenetic and phylogeographic analyses is rapidly increasing. However, as most DNA extraction methods damage the specimens, nondestructive extraction methods are useful to balance the demands of molecular biologists, morphologists, and museum curators. Here, I describe a method for nondestructive DNA extraction from bony specimens (i.e., bones and teeth). In this method, the specimens are soaked in extraction buffer, and DNA is then purified from the soaking solution using adsorption to silica. The method reliably yields mitochondrial and often also nuclear DNA. The method has been adapted to DNA extraction from other types of specimens such as arthropods.

  18. Surface, structural and tensile properties of proton beam irradiated zirconium

    Science.gov (United States)

    Rafique, Mohsin; Chae, San; Kim, Yong-Soo

    2016-02-01

    This paper reports the surface, structural and tensile properties of proton beam irradiated pure zirconium (99.8%). The Zr samples were irradiated by 3.5 MeV protons using MC-50 cyclotron accelerator at different doses ranging from 1 × 1013 to 1 × 1016 protons/cm2. Both un-irradiated and irradiated samples were characterized using Field Emission Scanning Electron Microscope (FESEM), X-ray Diffraction (XRD) and Universal Testing Machine (UTM). The average surface roughness of the specimens was determined by using Nanotech WSxM 5.0 develop 7.0 software. The FESEM results revealed the formation of bubbles, cracks and black spots on the samples' surface at different doses whereas the XRD results indicated the presence of residual stresses in the irradiated specimens. Williamson-Hall analysis of the diffraction peaks was carried out to investigate changes in crystallite size and lattice strain in the irradiated specimens. The tensile properties such as the yield stress, ultimate tensile stress and percentage elongation exhibited a decreasing trend after irradiation in general, however, an inconsistent behavior was observed in their dependence on proton dose. The changes in tensile properties of Zr were associated with the production of radiation-induced defects including bubbles, cracks, precipitates and simultaneous recovery by the thermal energy generated with the increase of irradiation dose.

  19. Making Durable Specimens For Electron Microscopy

    Science.gov (United States)

    Doychak, Joseph

    1989-01-01

    Consistent metal-oxide cross sections prepared quickly. New process makes TEM/STEM cross sections of metal/oxide interfaces. After specimen bars oxidized, placed in specially designed mold. Following encapsulation in zinc alloy, 3-mm-diameter specimen bar sliced into disks suitable for further preparation steps. Technique used to prepare 3-mm-diameter specimens of cross sections of oxides of alloys intended for use at temperatures greater than approximately 600 degree C.

  20. RSB: Research Specimen Banking across the Institution

    OpenAIRE

    Pense, Rick; Grose, Tim; Anderson, Lynn; Lee, H

    2001-01-01

    Research Specimen Banking (RSB) system is a component of the translational investigations infrastructure at Moffitt Cancer Center & Research Institute. It was implemented to provide specimen management functions to support basic science cancer research taking place in conjunction with caner clinical trials. RSB handles the receipt and distribution of clinical specimens to the research labs, with identifiers that both mask personal identity and enable linkage of clinical data to correlative re...

  1. Validation of the shear punch-tensile correlation technique using irradiated materials

    Energy Technology Data Exchange (ETDEWEB)

    Hankin, G.L.; Faulkner, R.G. [Loughborough Univ., Leicestershire (United Kingdom); Toloczko, M.B. [Washington State Univ., WA (United States); Hamilton, M.L. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    It was recently demonstrated that tensile data could be successfully related to shear punch data obtained on transmission electron microscopy (TEM) discs for a variety of irradiated alloys exhibiting yield strengths that ranged from 100 to 800 MPa. This implies that the shear punch test might be a viable alternative for obtaining tensile properties using a TEM disk, which is much smaller than even the smallest miniature tensile specimens, especially when irradiated specimens are not available or when they are too radioactive to handle easily. The majority of the earlier tensile-shear punch correlation work was done using a wide variety of unirradiated materials. The current work extends this correlation effort to irradiated materials and demonstrates that the same relationships that related shear punch tests remain valid for irradiated materials. Shear punch tests were performed on two sets of specimens. In the first group, three simple alloys from the {sup 59}Ni isotopic doping series in the solution annealed and cold worked conditions were irradiated at temperatures ranging from 365 to 495 C in the Fast Flux Test Facility. The corresponding tensile data already existed for tensile specimens fabricated from the same raw materials and irradiated side-by-side with the disks. In the second group, three variants of 316 stainless steel were irradiated in FFTF at 5 temperatures between 400 and 730 C to doses ranging from 12.5 to 88 dpa. The specimens were in the form of both TEM and miniature tensile specimens and were irradiated side-by-side.

  2. Atom probe tomography characterizations of high nickel, low copper surveillance RPV welds irradiated to high fluences

    Science.gov (United States)

    Miller, M. K.; Powers, K. A.; Nanstad, R. K.; Efsing, P.

    2013-06-01

    The Ringhals Units 3 and 4 reactors in Sweden are pressurized water reactors (PWRs) designed and supplied by Westinghouse Electric Company, with commercial operation in 1981 and 1983, respectively. The reactor pressure vessels (RPVs) for both reactors were fabricated with ring forgings of SA 508 class 2 steel. Surveillance blocks for both units were fabricated using the same weld wire heat, welding procedures, and base metals used for the RPVs. The primary interest in these weld metals is because they have very high nickel contents, with 1.58 and 1.66 wt.% for Unit 3 and Unit 4, respectively. The nickel content in Unit 4 is the highest reported nickel content for any Westinghouse PWR. Although both welds contain less than 0.10 wt.% copper, the weld metals have exhibited high irradiation-induced Charpy 41-J transition temperature shifts in surveillance testing. The Charpy impact 41-J shifts and corresponding fluences are 192 °C at 5.0 × 1023 n/m2 (>1 MeV) for Unit 3 and 162 °C at 6.0 × 1023 n/m2 (>1 MeV) for Unit 4. These relatively low-copper, high-nickel, radiation-sensitive welds relate to the issue of so-called late-blooming nickel-manganese-silicon phases. Atom probe tomography measurements have revealed ˜2 nm-diameter irradiation-induced precipitates containing manganese, nickel, and silicon, with phosphorus evident in some of the precipitates. However, only a relatively few number of copper atoms are contained within the precipitates. The larger increase in the transition temperature shift in the higher copper weld metal from the Ringhals R3 Unit is associated with copper-enriched regions within the manganese-nickel-silicon-enriched precipitates rather than changes in their size or number density.

  3. Impact behavior of reduced-activation steels irradiated to 24 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Alexander, D.J. [Oak Ridge National Laboratory, TN (United States)

    1996-04-01

    Charpy impact properties of eight reduced-activation Cr-W ferritic steels were determined after irradiation to {approx}21-24 dpa in the Fast Flux Test Facility (FFTF) at 365{degree}C. Chromium concentrations in the eight steels ranged from 2.25 to 12wt% Cr (steels contained {approx}0.1%C). the 2 1/4Cr steels contained variations of tungsten and vanadium, and the steels with 5, 9, and 12% Cr, contained a combination of 2% W and 0.25% V. A 9Cr in FFTF to {approx}6-8 and {approx}15-17 dpa. Irradiation caused an increase in the DBTT and decrease in the USE, but there was little further change in the DBTT from that observed after the 15-17 dpa irradiation, indicating that the shift had essentially saturated with fluence. The results are encouraging because they indicate that the effect of irradiation on toughness can be faorably affected by changing composition and microstructure.

  4. Design of unique pins for irradiation of higher actinides in a fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Basmajian, J.A.; Birney, K.R.; Weber, E.T.; Adair, H.L.; Quinby, T.C.; Raman, S.; Butler, J.K.; Bateman, B.C.; Swanson, K.M.

    1982-03-01

    The actinides produced by transmutation reactions in nuclear reactor fuels are a significant factor in nuclear fuel burnup, transportation and reprocessing. Irradiation testing is a primary source of data of this type. A segmented pin design was developed which provides for incorporation of multiple specimens of actinide oxides for irradiation in the UK's Prototype Fast Reactor (PFR) at Dounreay Scotland. Results from irradiation of these pins will extend the basic neutronic and material irradiation behavior data for key actinide isotopes.

  5. Steam-chemical reactivity for irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Anderl, R.A.; McCarthy, K.A.; Oates, M.A.; Petti, D.A.; Pawelko, R.J.; Smolik, G.R. [Idaho National Engineering and Environmental Lab., Idaho Falls, ID (United States)

    1998-01-01

    This paper reports the results of an experimental investigation to determine the influence of neutron irradiation effects and annealing on the chemical reactivity of beryllium exposed to steam. The work entailed measurements of the H{sub 2} generation rates for unirradiated and irradiated Be and for irradiated Be that had been previously annealed at different temperatures ranging from 450degC to 1200degC. H{sub 2} generation rates were similar for irradiated and unirradiated Be in steam-chemical reactivity experiments at temperatures between 450degC and 600degC. For irradiated Be exposed to steam at 700degC, the chemical reactivity accelerated rapidly and the specimen experienced a temperature excursion. Enhanced chemical reactivity at temperatures between 400degC and 600degC was observed for irradiated Be annealed at temperatures of 700degC and higher. This reactivity enhancement could be accounted for by the increased specific surface area resulting from development of a surface-connected porosity in the irradiated-annealed Be. (author)

  6. Characterisation of the fracture properties in the ductile to brittle transition region of the weld material of a reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Scibetta, M. [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Ferreno, D., E-mail: ferrenod@unican.es [University of Cantabria, ETS Ingenieros de Caminos, Av/Los Castros s/n, 39005 Santander (Spain); Gorrochategui, I. [Centro Tecnologico de Componentes (CTC), Parque Cientifico y Tecnologico de Cantabria, Isabel Torres No 1, 39011 Santander (Spain); Nuclenor, SA, C/Hernan Cortes 26, 39003 Santander (Spain); Lacalle, R. [University of Cantabria, ETS Ingenieros de Caminos, Av/Los Castros s/n, 39005 Santander (Spain); Walle, E. van [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Martin, J. [Nuclenor, SA, C/Hernan Cortes 26, 39003 Santander (Spain); Gutierrez-Solana, F. [University of Cantabria, ETS Ingenieros de Caminos, Av/Los Castros s/n, 39005 Santander (Spain)

    2011-04-15

    This work presents the results of the fracture characterisation of the weld material of a nuclear vessel, currently in service, in the ductile to brittle transition region. The tests consisted of Charpy impact and tensile tests, performed in the framework of the surveillance programme of the plant. Moreover, in the context of this research, K{sub Jc} fracture toughness tests on pre-cracked Charpy V notch specimens (evaluated according to the Master Curve methodology) together with some mini-tensile tests, were performed; non-irradiated and several irradiated material conditions were characterised. The analysis of the experimental results revealed some inconsistencies concerning the material embrittlement as measured through Charpy and K{sub Jc} fracture tests: in order to obtain an adequate understanding of the results, an extended experimental scope well beyond the regulatory framework was developed, including Charpy tests and K{sub Jc} fracture tests, both performed on reconstituted specimens. Moreover, Charpy specimens irradiated in the high flux BR2 material test reactor were tested with the same purpose. With this extensive experimental programme, a coherent and comprehensive description of the irradiation behaviour of the weld material in the transition region was achieved. Furthermore it revealed better material properties in comparison with the initial expectations based on the information obtained in the framework of the surveillance programme.

  7. Apparent fracture toughness of acrylic bone cement: effect of test specimen configuration and sterilization method.

    Science.gov (United States)

    Lewis, G

    1999-01-01

    The plane strain fracture toughness of Palacos R bone cement was determined using linear elastic fracture mechanics (LEFM) principles and three different test specimen configurations: single edge notched three-point (SENB), rectangular compact tension (RCT), and chevron notched short rod (CNSR). Another aspect of the study was an investigation of the effect of three methods used to sterilize the powder constituents of the cement-none, gamma irradiation and ethylene oxide--on the fracture toughness of the fully polymerized material. A detailed justification is provided for using LEFM. The fracture toughness results obtained using the CNSR specimens were, on average, 14 and 16% higher than those obtained using the SENB and RCT types, respectively. These differences are accounted for in terms of differences in four aspects of these specimen configuration (namely, residual stress effects, loading rate, material inhomogeneity, and the nature of the test). For a given specimen configuration, gamma irradiation produced a statistically significant decrease in fracture toughness which, it is suggested, is due to the concomitant depreciation in molecular weight. For a given cement type, there is no statistically significant difference in fracture toughness results obtained using SENB and RCT specimens. It is thus suggested that either of these configurations can be used to determine the fracture toughness of acrylic bone cement.

  8. A Study on Mechanical behavior of Tensile Specimen Fabricated by Laser Cutting

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Y. G.; Kim, G. S.; Baik, S. J.; Baek, S. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The mechanical testing data are required for the assessment of dry storage of the spent nuclear fuel. Laser cutting system could be useful tools for material processing such as cutting in radioactive environment due to non-contact nature, ease in handling and the laser cutting process is most advantageous, offering the narrow kerf width and heat affected zone by using small beam spot diameter. The feasibility of the laser cutting system was demonstrated for the fabrication of various types of the unirradiated cladding with and without oxide layer on the specimens. In the present study, the dimensional measurement and tensile test were conducted to investigate the mechanical behavior of the axial tensile test specimens depending on the material processing methods in a hot cell at IMEF (Irradiated Materials Examination Facility) of KAERI. Laser cutting system was used to fabricate the tensile test specimens, and the mechanical behavior was investigated using the dimensional measurement and tensile test. It was shown that the laser beam machining could be a useful tool to fabricate the specimens and this technique will be developed for the fabrication of various types of irradiated specimens in a hotcell.

  9. 16 CFR Figure 3 to Part 1610 - Specimen Holder Supported in Specimen Rack

    Science.gov (United States)

    2010-01-01

    ... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Specimen Holder Supported in Specimen Rack 3 Figure 3 to Part 1610 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION FLAMMABLE FABRICS ACT... Holder Supported in Specimen Rack ER25MR08.002...

  10. Mode-II-Fracture Specimen And Holder

    Science.gov (United States)

    Buzzard, Robert J.; Ghosn, Louis; Succop, George

    1991-01-01

    Test specimen and loading frame developed for fatigue and fracture testing of materials under mode-II (sliding-mode) loading. Assembly placed in compression-testing machine. Loads directed oppositely along centerline cause self-similar crack to propagate. Enables consistently accurate alignment of specimens before insertion of specimen/frame assemblies into compression-testing machine. Makes design attractive for testing in hostile environments in which access to machine or furnace limited. Additional feature, with little or no modification, placed horizontally into impact testing machine and subjected to loading at high speeds.

  11. 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.

  12. Microstructure and microhardness of CLAM steel irradiated up to 20.8 dpa in STIP-V

    Science.gov (United States)

    Peng, Lei; Ge, Hongen; Dai, Yong; Huang, Qunying; Ye, Minyou

    2016-01-01

    Specimens of China low activation martensitic (CLAM) steel were irradiated in the fifth experiment of SINQ target irradiation program (STIP-V) up to 20.8 dpa/1564 appm He. Microhardness measurements and transmission electron microscope (TEM) observations have been performed to investigate irradiation induced hardening effects. The results of CLAM steel specimens show similar trend in microhardness and microstructure changes with irradiation dose, compared to F82H/Optimax-A steels irradiated in STIP-I/II. Defects and helium bubbles were observed in all specimens, even at a very low dose of 5.4 dpa. For defects and bubbles, the mean size and number density increased with increasing irradiation dose to 13 dpa, and then the mean size increased and number density decreased with the increasing irradiation dose to 20.8 dpa.

  13. Cracking behavior of thermally aged and irradiated CF-8 cast austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y., E-mail: Yiren_Chen@anl.gov [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Alexandreanu, B.; Chen, W.-Y.; Natesan, K. [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Li, Z.; Yang, Y. [University of Florida, Gainesville, FL 32611 (United States); Rao, A.S. [US Nuclear Regulatory Commission, 11545 Rockville Pike, Rockville, MD 20852 (United States)

    2015-11-15

    To assess the combined effect of thermal aging and neutron irradiation on the cracking behavior of CF-8 cast austenitic stainless steel, crack growth rate (CGR) and fracture toughness J-R curve tests were carried out on compact-tension specimens in high-purity water with low dissolved oxygen. Both unaged and thermally aged specimens were irradiated at ∼320 °C to 0.08 dpa. Thermal aging at 400 °C for 10,000 h apparently had no effect on the corrosion fatigue and stress corrosion cracking behavior in the test environment. The cracking susceptibility of CF-8 was not elevated significantly by neutron irradiation at 0.08 dpa. Transgranular cleavage-like cracking was the main fracture mode during the CGR tests, and a brittle morphology of delta ferrite was often seen on the fracture surfaces at the end of CGR tests. The fracture toughness J-R curve tests showed that both thermal aging and neutron irradiation can induce significant embrittlement. The loss of fracture toughness due to neutron irradiation was more pronounced in the unaged than aged specimens. After neutron irradiation, the fracture toughness values of the unaged and aged specimens were reduced to a similar level. G-phase precipitates were observed in the aged and irradiated specimens with or without prior aging. The similar microstructural changes resulting from thermal aging and irradiation suggest a common microstructural mechanism of inducing embrittlement in CF-8.

  14. Structure of Wet Specimens in Electron Microscopy

    Science.gov (United States)

    Parsons, D. F.

    1974-01-01

    Discussed are past work and recent advances in the use of electron microscopes for viewing structures immersed in gas and liquid. Improved environmental chambers make it possible to examine wet specimens easily. (Author/RH)

  15. Testing Biopsy and Cytology Specimens for Cancer

    Science.gov (United States)

    ... Your Diagnosis Exams and Tests for Cancer Testing Biopsy and Cytology Specimens for Cancer Waiting to hear ... who tell you whether the cells in your biopsy sample are cancer or not. How is cancer ...

  16. Structure of Wet Specimens in Electron Microscopy

    Science.gov (United States)

    Parsons, D. F.

    1974-01-01

    Discussed are past work and recent advances in the use of electron microscopes for viewing structures immersed in gas and liquid. Improved environmental chambers make it possible to examine wet specimens easily. (Author/RH)

  17. Impact of specimen adequacy on the assessment of renal allograft biopsy specimens.

    Science.gov (United States)

    Cimen, S; Geldenhuys, L; Guler, S; Imamoglu, A; Molinari, M

    2016-01-01

    The Banff classification was introduced to achieve uniformity in the assessment of renal allograft biopsies. The primary aim of this study was to evaluate the impact of specimen adequacy on the Banff classification. All renal allograft biopsies obtained between July 2010 and June 2012 for suspicion of acute rejection were included. Pre-biopsy clinical data on suspected diagnosis and time from renal transplantation were provided to a nephropathologist who was blinded to the original pathological report. Second pathological readings were compared with the original to assess agreement stratified by specimen adequacy. Cohen's kappa test and Fisher's exact test were used for statistical analyses. Forty-nine specimens were reviewed. Among these specimens, 81.6% were classified as adequate, 6.12% as minimal, and 12.24% as unsatisfactory. The agreement analysis among the first and second readings revealed a kappa value of 0.97. Full agreement between readings was found in 75% of the adequate specimens, 66.7 and 50% for minimal and unsatisfactory specimens, respectively. There was no agreement between readings in 5% of the adequate specimens and 16.7% of the unsatisfactory specimens. For the entire sample full agreement was found in 71.4%, partial agreement in 20.4% and no agreement in 8.2% of the specimens. Statistical analysis using Fisher's exact test yielded a P value above 0.25 showing that - probably due to small sample size - the results were not statistically significant. Specimen adequacy may be a determinant of a diagnostic agreement in renal allograft specimen assessment. While additional studies including larger case numbers are required to further delineate the impact of specimen adequacy on the reliability of histopathological assessments, specimen quality must be considered during clinical decision making while dealing with biopsy reports based on minimal or unsatisfactory specimens.

  18. Ultrasonic analysis of spherical composite test specimens

    Energy Technology Data Exchange (ETDEWEB)

    Brosey, W.D.

    1984-08-22

    Filament wound spherical test specimens have been examined ultrasonically as part of a program to determine the effectiveness of various nondestructive evaluation techniques for analysis of mechanical characteristics of a composite with enclosed geometry. The Kevlar-epoxy composite specimens contained simulated defect conditions which were located, and the extent of damage determined, using ultrasonic analysis. Effects of transducer frequency and signal parameters have been examined to determine optimum conditions for flaw detection. The data were displayed in rectangular and axonometric projection.

  19. AGC-2 Graphite Pre-irradiation Data Package

    Energy Technology Data Exchange (ETDEWEB)

    David Swank; Joseph Lord; David Rohrbaugh; William Windes

    2010-08-01

    The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterized prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.

  20. Microstructure and mechanical properties of neutron irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, E.; Kawamura, H. [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Terai, T.; Tanaka, S.

    1998-01-01

    Microstructure and mechanical properties of the neutron irradiated beryllium with total fast neutron fluences of 1.3 - 4.3 x 10{sup 21} n/cm{sup 2} (E>1 MeV) at 327 - 616degC were studied. Swelling increased by high irradiation temperature, high fluence, and by the small grain size and high impurity. Obvious decreasing of the fracture stress was observed in the bending test and in small grain specimens which had many helium bubbles on the grain boundary. Decreasing of the fracture stress for small grain specimens was presumably caused by crack propagation on the grain boundaries which weekend by helium bubbles. (author)

  1. The Design and Safety Review for a Creep Capsule With Four Specimens

    Energy Technology Data Exchange (ETDEWEB)

    Cho, M. S.; Seo, C. G.; Choi, M. H.; Choo, K. N.; Sohn, J. M.; Kang, Y. H.; Kim, B. G. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    To review the safety in the irradiation test of a creep capsule with 4 specimens, the design analysis for the incore creep test in the normal and abnormal operation condition was performed. In the normal operation condition, the reactivity effect was estimated and the analysis on the stress and temperature was performed. In the abnormal operation condition, the effect exerted on the capsule body by a leakage or breakage of the stress loading unit was analyzed.

  2. Post Irradiation Examination of a Thermo-Mechanically Improved Version of EUROFER ODS

    Energy Technology Data Exchange (ETDEWEB)

    Lucon, E.; Leenaers, A.; Vandermeulen, W.

    2006-08-15

    EUROFER is a 9Cr-1W-0.2V-0.1Ta reduced activation ferritic/martensitic (RAFM) steel, presently considered within the European Union as the primary candidate structural material in a fusion power plant. Its mechanical strength properties currently prevent its use at temperatures higher than 500-550 degrees Celsius. In an effort to extend the range of operating temperatures to 600-650 degrees Celsius and therefore enhance the efficiency of the machine, a different production route, Oxide Dispersion Strengthening (ODS), is being investigated. The characteristics of different versions of EUROFER ODS have been assessed in recent years, leading to the improvement of the material by a combination of optimized production process and post-thermal treatment. Until recently, the mechanical properties of EUROFER ODS had only been investigated in the unirradiated condition, and no information was available for the irradiation response of the material. However, mechanical samples have been irradiated during 2004-2005 at 300 degrees Celsius in the Belgian Reactor 2 (BR2) in Mol to an accumulated dose of 1.73 dpa; tensile, Charpy impact and fracture toughness tests have been performed in the hot cell laboratories of the Belgian Nuclear Centre (SCK-CEN). Metallographic and microstructural investigations were also performed on the investigated material in both the unirradiated and irradiated condition.

  3. Effect of boron on post irradiation tensile properties of reduced activation ferritic steel (F-82H) irradiated in HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Shiba, Kiyoyuki; Suzuki, Masahide; Hishinuma, Akimichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Dept. of Materials Science and Engineering; Pawel, J.E. [Oak Ridge National Lab., TN (United States). Metals and Ceramics Div.

    1994-12-31

    Reduced activation ferritic/martensitic steel, F-82H (Fe-8Cr-2W-V-Ta), was irradiated in the High Flux Isotope Reactor (HFIR) to doses between 11 and 34 dpa at 400 and 500 C. Post irradiation tensile tests were performed at the nominal irradiation temperature in vacuum. Some specimens included {sup 10}B or natural boron (nB) to estimate the helium effect on tensile properties. Tensile properties including the 0.2% offset yield stress, the ultimate tensile strength, the uniform elongation and the total elongation were measured. The tensile properties were not dependent on helium content in specimens irradiated to 34 dpa, however {sup 10}B-doped specimens with the highest levels of helium showed slightly higher yield strength and less ductility than boron-free specimens. Strength appears to go through a peak, and ductility through a trough at about 11 dpa. The irradiation to more than 21 dpa reduced the strength and increased the elongation to the unirradiated levels. Ferritic steels are one of the candidate alloys for nuclear fusion reactors because of their good thermophysical properties, their superior swelling resistance, and the low corrosion rate in contact with potential breeder and coolant materials.

  4. Irradiation creep of vanadium-base alloys

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Billone, M.C.; Strain, R.V.; Smith, D.L. [Argonne National Lab., IL (United States); Matsui, H. [Tohoku Univ. (Japan)

    1998-03-01

    A study of irradiation creep in vanadium-base alloys is underway with experiments in the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) in the United States. Test specimens are thin-wall sealed tubes with internal pressure loading. The results from the initial ATR irradiation at low temperature (200--300 C) to a neutron damage level of 4.7 dpa show creep rates ranging from {approx}0 to 1.2 {times} 10{sup {minus}5}/dpa/MPa for a 500-kg heat of V-4Cr-4Ti alloy. These rates were generally lower than reported from a previous experiment in BR-10. Because both the attained neutron damage levels and the creep strains were low in the present study, however, these creep rates should be regarded as only preliminary. Substantially more testing is required before a data base on irradiation creep of vanadium alloys can be developed and used with confidence.

  5. Neutrophil myeloperoxidase destruction by ultraviolet irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Hanker, J.; Giammara, B.; Strauss, G.

    1988-01-01

    The peroxidase activity of enriched leukocyte preparations on coverslips was determined cytochemically with a newly developed method. The techniques utilizes diaminobenzidine medium and cupric nitrate intensification and is suitable for analysis with light microscopy, SEM, and TEM. Blood specimens from control individuals were studied with and without in vitro UV irradiation and compared with those from psoriasis patients exposed therapeutically to various types of UV in phototherapy. All UV irradiated samples showed diminished neutrophil myeloperoxidase (MP) activity although that of the principal eosinophil peroxidase was unaffected. The SEMs supported the contention that decreased neutrophil MP activity might be related to UV induced degranulation. It is believed to be possible, eventually, to equate the observed MP degranulation effect after UV irradiation with diminished ability to fight bacterial infections.

  6. Closeout of JOYO-1 Specimen Fabrication Efforts

    Energy Technology Data Exchange (ETDEWEB)

    ME Petrichek; JL Bump; RF Luther

    2005-10-31

    Fabrication was well under way for the JOYO biaxial creep and tensile specimens when the NR Space program was canceled. Tubes of FS-85, ASTAR-811C, and T-111 for biaxial creep specimens had been drawn at True Tube (Paso Robles, CA), while tubes of Mo-47.5 Re were being drawn at Rhenium Alloys (Cleveland, OH). The Mo-47.5 Re tubes are now approximately 95% complete. Their fabrication and the quantities produced will be documented at a later date. End cap material for FS-85, ASTAR-811C, and T-111 had been swaged at Pittsburgh Materials Technology, Inc. (PMTI) (Large, PA) and machined at Vangura (Clairton, PA). Cutting of tubes, pickling, annealing, and laser engraving were in process at PMTI. Several biaxial creep specimen sets of FS-85, ASTAR-811C, and T-111 had already been sent to Pacific Northwest National Laboratory (PNNL) for weld development. In addition, tensile specimens of FS-85, ASTAR-811C, T-111, and Mo-47.5 Re had been machined at Kin-Tech (North Huntington, PA). Actual machining of the other specimen types had not been initiated. Flowcharts 1-3 detail the major processing steps each piece of material has experienced. A more detailed description of processing will be provided in a separate document [B-MT(SRME)-51]. Table 1 lists the in-process materials and finished specimens. Also included are current metallurgical condition of these materials and specimens. The available chemical analyses for these alloys at various points in the process are provided in Table 2.

  7. Experimental determination of creep properties of Beryllium irradiated to relevant fusion power reactor doses

    Science.gov (United States)

    Scibetta, M.; Pellettieri, A.; Sannen, L.

    2007-08-01

    A dead weight machine has been developed to measure creep in irradiated beryllium relevant to fusion power reactors. Due to the external compressive load, the material will creep and the specimen will shrink. However, the specimen also swells due to the combined effect of internal pressure in helium bubbles and creep. One of the major challenges is to unmask swelling and derive intrinsic creep properties. This has been achieved through appropriate pre-annealing experiments. Creep has been measured on irradiated and unirradiated specimens. The temperature and stress dependence is characterized and modeled using the product of an Arrhenius' law for the temperature dependence and a power law for the stress dependence. Irradiation increases the sensitivity to creep but the irradiation effects can be rationalized by taking into account the irradiation-induced porosity. Experimental evidence supports dislocation climb by vacancy absorption to be the most plausible intrinsic creep mechanism.

  8. Effects of the neutronic irradiation on the impact tests. Efectos de la irradiacion neutronica sobre los ensayos de resiliencia

    Energy Technology Data Exchange (ETDEWEB)

    Lapea, J.; Perosanz, F.J.; Hernandez, M.T.

    1993-01-01

    The changes that the Charpy curves suffer when steel is exposed to neutronic fluence are studied. Three steels with different chemical composition were chosen, two of them (JPF and JPJ) being treated at only one neutronic fluence, while the last one (JRQ) was irradiated at three fluences. In this way, it was possible to compare the effect of increasing the neutronic dose, and to study the experimental results as a function of the steel chemical composition. Two characteristic facts have been observed: the displacement of the curve at higher temperatures, and decrease of the upper shelf energy (USE). The mechanical recovery of the materials after two different thermal treatments is also described, and a comparation between the experimental results obtained and the damage prediction formulas given by different regulatory international organisms in the nuclear field is established. Author. 11 refs.

  9. 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

  10. Effect of 16.3 dpa neutron irradiation on fatigue lifetime of the RAFM steel EUROFER97

    Energy Technology Data Exchange (ETDEWEB)

    Materna-Morris, E., E-mail: edeltraud.materna-morris@kit.edu [KIT Karlsruhe Institute of Technology, Campus Nord, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Moeslang, A.; Rolli, R.; Schneider, H.-C. [KIT Karlsruhe Institute of Technology, Campus Nord, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

    2011-10-15

    Low cycle fatigue specimens of the reduced-activation martensitic/ferritic steel EUROFER97 were neutron irradiated at 250 deg. C up to an accumulated dose of 16.3 dpa. After irradiation, the specimens were push-pull fatigue tested under strain-controlled conditions at 250 deg. C to determine the impact of irradiation on lifetime, fracture behavior, and microstructure. The typical cyclic softening of martensitic/ferritic steels was observed. Furthermore, a considerable increase of lifetime after irradiation and subsequent cycling at lower strain amplitudes was remarkable. This behavior was attributed to the homogeneous distribution of stable irradiation-induced dislocation loops and small precipitates acting as barriers for the cyclic motion of dislocations, thereby influencing substantially crack initiation and crack network formation. While in the un-irradiated material push-pull fatigue sweeps the dislocations to the boundaries, a significant fraction of dislocations was fixed at irradiation-induced defects after irradiation and fatigue testing.

  11. Significance of endoscopic biopsy after preoperative irradiation therapy for rectal cancer

    Energy Technology Data Exchange (ETDEWEB)

    Takiguchi, Nobuhiro; Sarashina, Hiromi; Saito, Norio; Nunomura, Masao; Kohda, Keishi; Nakajima, Nobuyuki (Chiba Univ. (Japan). School of Medicine)

    1994-05-01

    To evaluate the utility of endoscopic biopsy before and after preoperative irradiation therapy for rectal cancer, we examined histologically both biopsy specimens and resected materials of forty-three patients. Two pieces of biopsy materials were taken both before and after irradiation therapy (total dose 42.6 Gy) from the marginal wall of the tumor, cavity and transitional mucosa, respectively. In biopsy specimens, according to the degree of degeneration of cancer cells, cases with remarkable changes of nucleus, nucleolus, and cytoplasm due to irradiation were classified into the severely degenerated group. According to the histological examinations of resected materials, twenty-four cases were under Grade 1b (Gr I), and nineteen cases were over Grade 2 (Gr II). The rates of cancer cells found in biopsy materials after irradiation were 91.7% in Gr I and were 47.4% in Gr II, respectively (p<0.01). Among the cases, 54.5% in Gr I and 100% in Gr II belonged to the severely degenerated group (p<0.05). Transitional mucosas were not greatly damaged by irradiation. As a result, the greater the irradiation effect was, the fewer cancer cells were found and the more degenerated cancer cells were found in biopsy specimens. But the rate of severely degenerated cells found in the biopsy specimens of little effect cases was high. So it was thought to be too difficult to predict the histological radiation effect of resected specimens from only biopsy specimens. (author).

  12. An irradiation test of heat-resistant ceramic composite materials. Interim report on post-irradiation examinations of the first preliminary irradiation test: 97M-13A

    Energy Technology Data Exchange (ETDEWEB)

    Baba, Shin-ichi; Takahashi, Tsuneo; Ishihara, Masahiro; Hayashi, Kimio; Sozawa, Shizuo; Saito, Takashi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Suzuki, Yoshio [Nuclear Engineering, Co. Ltd., Osaka (Japan); Saito, Tamotsu; Sekino, Hajime [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    The Japan Atomic Energy Research Institute (JAERI) has been carrying out the research on radiation damage mechanism of heat-resistant ceramic composite materials, as one of the subjects of the innovative basic research on high temperature engineering using the High Temperature Engineering Test Reactor (HTTR). A series of preliminary irradiation tests is being made using the Japan Materials Testing Reactor (JMTR). The present report describes results of post-irradiation examinations (PIE) so far on specimens irradiated in the first capsule, designated 97M-13A, to fast neutron fluences of 1.2-1.8x10{sup 24} m{sup -2} (E>1 MeV) at temperatures of 573, 673 and 843 K. In the PIE, measurements were made on (1) dimensional changes, (2) thermal expansions, (3) X-ray parameters and (4) {gamma}-ray spectra. The results for the carbon/carbon and SiC/SiC composites were similar to those in existing literatures. The temperature monitor effect was observed both for SiC fiber- and particle-reinforced SiC composites as in the case of monolithic SiC. Namely, the curve of the coefficient of thermal expansion (CTE) of these specimens showed a rapid drop above a temperature around the irradiation temperature +100 K in the first ramp (ramp rate: 10 K/min), while in the second ramp the CTE curves were almost the same as those of un-irradiated SiC specimens. (author)

  13. Irradiation conditions for fiber laser bonding of HAp-glass ceramics with bovine cortical bone.

    Science.gov (United States)

    Tadano, Shigeru; Yamada, Satoshi; Kanaoka, Masaru

    2014-01-01

    Orthopedic implants are widely used to repair bones and to replace articulating joint surfaces. It is important to develop an instantaneous technique for the direct bonding of bone and implant materials. The aim of this study was to develop a technique for the laser bonding of bone with an implant material like ceramics. Ceramic specimens (10 mm diameter and 1 mm thickness) were sintered with hydroxyapatite and MgO-Al2O3-SiO2 glass powders mixed in 40:60 wt% proportions. A small hole was bored at the center of a ceramic specimen. The ceramic specimen was positioned onto a bovine bone specimen and a 5 mm diameter area of the ceramic specimen was irradiated using a fiber laser beam (1070-1080 nm wavelength). As a result, the bone and the ceramic specimens bonded strongly under the irradiation conditions of a 400 W laser power and a 1.0 s exposure time. The maximum shear strength was 5.3 ± 2.3 N. A bonding substance that penetrated deeply into the bone specimen was generated around the hole in the ceramic specimen. On using the fiber laser, the ceramic specimen instantaneously bonded to the bone specimen. Further, the irradiation conditions required for the bonding were investigated.

  14. STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE/SILICON CARBIDE JOINTS

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [ORNL; Koyanagi, Takaaki [ORNL; Kiggans, Jim [ORNL; Cetiner, Nesrin [ORNL; McDuffee, Joel [ORNL

    2014-09-01

    Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

  15. Tensile properties of a titanium modified austenitic stainless steel and the weld joints after neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Shiba, K.; Ioka, I.; Jitsukawa, S.; Hamada, A.; Hishinuma, A. [and others

    1996-10-01

    Tensile specimens of a titanium modified austenitic stainless steel and its weldments fabricated with Tungsten Inert Gas (TIG) and Electron Beam (EB) welding techniques were irradiated to a peak dose of 19 dpa and a peak helium level of 250 appm in the temperature range between 200 and 400{degrees}C in spectrally tailored capsules in the Oak Ridge Research Reactor (ORR) and the High Flux Isotope Reactor (HFIR). The He/dpa ratio of about 13 appm/dpa is similar to the typical helium/dpa ratio of a fusion reactor environment. The tensile tests were carried out at the irradiation temperature in vacuum. The irradiation caused an increase in yield stress to levels between 670 and 800 MPa depending on the irradiation temperature. Total elongation was reduced to less than 10%, however the specimens failed in a ductile manner. The results were compared with those of the specimens irradiated using irradiation capsules producing larger amount of He. Although the He/dpa ratio affected the microstructural change, the impact on the post irradiation tensile behavior was rather small for not only base metal specimens but also for the weld joint and the weld metal specimens.

  16. Experimental tests of irradiation-anneal-reirradiation effects on mechanical properties of RPV plate and weld materials

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [Electric Power Research Inst., Charlotte, NC (United States); Hawthorne, J.R. [Materials Engineering Associates, Inc., Lanham, MD (United States); Rochau, G.E. [Sandia National Labs., Albuquerque, NM (United States)

    1999-10-01

    The Charpy-V (C{sub v}) notch ductility and tensile properties of three reactor pressure vessel (RPV) steel materials were determined for the 288 C (550 F) irradiated (I), 288 C (550 F) irradiated + 454 C (850 F) - 168 h postirradiation annealed (IA), and 288 C (550 F) reirradiated (IAR) conditions. Total fluences of the I condition and the IAR condition were, respectively, 3.33 {times} 10{sup 19} n/cm{sup 2} and 4.18 {times} 10{sup 19} n/cm{sup 2}, E > 1 MeV. The irradiation portion of the IAR condition represents an incremental fluence increase of 1.05 {times} 10{sup 19} n/cm{sup 2}, E > 1 MeV, over the I-condition fluence. The annealing treatment produced full C{sub v} upper shelf recovery and full or nearly full recovery in the C{sub v} 41 J (30 ft-lb) transition temperature. The C{sub v} transition temperature increases produced by the reirradiation exposure were 22% to 43% of the increase produced by the first cycle irradiation exposure. A somewhat greater radiation embrittlement sensitivity and a somewhat greater reirradiation embrittlement sensitivity were exhibited by the low nickel content plate than the high nickel content plate. The IAR-condition properties of the surface vs. interior regions of the low nickel content plate are also compared.

  17. The Value of Agricultural Voucher Specimens

    OpenAIRE

    Barkworth,Mary; Wolf, Paul; Kinosian,Sylvia; Dyreson, Curtis; Pearse, Will; Brandt, Ben; Cobb,Neil

    2017-01-01

    Voucher specimens are the ultimate raw data of biodiversity studies because they document the interpretation of the names used in papers and reports resulting from such studies. The value of voucher specimens is increased by making their records web-accessible but they can be further enhanced by linking them to other online resources, particularly if the links are birectional.  In this presentation, we discuss the potential benefits of such links for a group of agricultural significance, the ...

  18. Food irradiation makes progress

    Energy Technology Data Exchange (ETDEWEB)

    Kooij, J. van (Joint FAO/IAEA Div. of Isotope and Radiation Applications of Atomic Energy for Food and Agricultural Development, Vienna (Austria))

    1984-06-01

    In the past fifteen years, food irradiation processing policies and programmes have been developed both by a number of individual countries, and through projects supported by FAO, IAEA and WHO. These aim at achieving general acceptance and practical implementation of food irradiation through rigorous investigations of its wholesomeness, technological and economic feasibility, and efforts to achieve the unimpeded movement of irradiated foods in international trade. Food irradiation processing has many uses.

  19. 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

  20. Food irradiation in China

    Energy Technology Data Exchange (ETDEWEB)

    Wedekind, L.

    1986-08-01

    The paper concerns food irradiation in The People's Republic of China. Its use is envisaged to prolong storage times and to improve the quality of specific foodstuffs. Commercialisation in China, demonstration plants, seasonal shortages and losses, Shanghai irradiation centre, health and safety approval, prospects for wider applications and worldwide use of food irradiation, are all discussed.

  1. Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Chopra, O. K.; Soppet, W. K.; Shack, W. J.; Yang, Y.; Allen, T. R.; Univ. of Wisconsin at Madison

    2010-02-16

    Cracking behavior of stainless steels specimens irradiated in the BOR-60 at about 320 C is studied. The primary objective of this research is to improve the mechanistic understanding of irradiation-assisted stress corrosion cracking (IASCC) of core internal components under conditions relevant to pressurized water reactors. The current report covers several baseline tests in air, a comparison study in high-dissolved-oxygen environment, and TEM characterization of irradiation defect structure. Slow strain rate tensile (SSRT) tests were conducted in air and in high-dissolved-oxygen (DO) water with selected 5- and 10-dpa specimens. The results in high-DO water were compared with those from earlier tests with identical materials irradiated in the Halden reactor to a similar dose. The SSRT tests produced similar results among different materials irradiated in the Halden and BOR-60 reactors. However, the post-irradiation strength for the BOR-60 specimens was consistently lower than that of the corresponding Halden specimens. The elongation of the BOR-60 specimens was also greater than that of their Halden specimens. Intergranular cracking in high-DO water was consistent for most of the tested materials in the Halden and BOR-60 irradiations. Nonetheless, the BOR-60 irradiation was somewhat less effective in stimulating IG fracture among the tested materials. Microstructural characterization was also carried out using transmission electron microscopy on selected BOR-60 specimens irradiated to {approx}25 dpa. No voids were observed in irradiated austenitic stainless steels and cast stainless steels, while a few voids were found in base and grain-boundary-engineered Alloy 690. All the irradiated microstructures were dominated by a high density of Frank loops, which varied in mean size and density for different alloys.

  2. Fission neutron irradiation of copper containing implanted and transmutation produced helium

    DEFF Research Database (Denmark)

    Singh, B.N.; Horsewell, A.; Eldrup, Morten Mostgaard

    1992-01-01

    . The distributions of helium prior to fission neutron irradiation were determined by a combination of transmission electron microscopy (TEM) and positron annihilation techniques (PAT). These specimens, together with pure copper, were then irradiated with fission neutrons in a single capsule in fast flux test...

  3. 16 CFR Figure 6 to Subpart A of... - Dummy Specimen in Specimen Holder

    Science.gov (United States)

    2010-01-01

    ... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Dummy Specimen in Specimen Holder 6 Figure 6 to Subpart A of Part 1209 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION CONSUMER PRODUCT SAFETY ACT REGULATIONS INTERIM SAFETY STANDARD FOR CELLULOSE INSULATION The Standard Pt. 1209, Subpt. A...

  4. 37 CFR 2.56 - Specimens.

    Science.gov (United States)

    2010-07-01

    ... indicate membership in the collective organization. (5) A certification mark specimen must show how a... of a union or other organization performed the work or labor on the goods or services. (c) A... be a digitized image in .jpg or .pdf format....

  5. Measurement of Diameter Changes during Irradiation Testing

    Energy Technology Data Exchange (ETDEWEB)

    Davis, K. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Knudson, D. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Crepeau, J. C. [Univ. of Idaho, Idaho Falls, ID (United States); Solstad, S. [Inst. for Energy Technologoy, Halden (Norway)

    2015-03-01

    New materials are being considered for fuel, cladding, and structures in advanced and existing nuclear reactors. Such materials can experience significant dimensional and physical changes during irradiation. Currently in the US, such changes are measured by repeatedly irradiating a specimen for a specified period of time and then removing it from the reactor for evaluation. The time and labor to remove, examine, and return irradiated samples for each measurement makes this approach very expensive. In addition, such techniques provide limited data and handling may disturb the phenomena of interest. In-pile detection of changes in geometry is sorely needed to understand real-time behavior during irradiation testing of fuels and materials in high flux US Material and Test Reactors (MTRs). This paper presents development results of an advanced Linear Variable Differential Transformer-based test rig capable of detecting real-time changes in diameter of fuel rods or material samples during irradiation in US MTRs. This test rig is being developed at the Idaho National Laboratory and will provide experimenters with a unique capability to measure diameter changes associated with fuel and cladding swelling, pellet-clad interaction, and crud buildup.

  6. Rehydration of forensically important larval Diptera specimens.

    Science.gov (United States)

    Sanford, Michelle R; Pechal, Jennifer L; Tomberlin, Jeffery K

    2011-01-01

    Established procedures for collecting and preserving evidence are essential for all forensic disciplines to be accepted in court and by the forensic community at large. Entomological evidence, such as Diptera larvae, are primarily preserved in ethanol, which can evaporate over time, resulting in the dehydration of specimens. In this study, methods used for rehydrating specimens were compared. The changes in larval specimens with respect to larval length and weight for three forensically important blow fly (Diptera: Calliphoridae) species in North America were quantified. Phormia regina (Meigen), Cochliomyia macellaria (F.), and Chrysomya rufifacies (Macquart) third-instar larvae were collected from various decomposing animals and preserved with three preservation methods (80% ethanol, 70% isopropyl alcohol, and hot-water kill then 80% ethanol). Preservative solutions were allowed to evaporate. Rehydration was attempted with either of the following: 80% ethanol, commercial trisodium phosphate substitute solution, or 0.5% trisodium phosphate solution. All three methods partially restored weight and length of specimens recorded before preservation. Analysis of variance results indicated that effects of preservation, rehydration treatment, and collection animal were different in each species. The interaction between preservative method and rehydration treatment had a significant effect on both P. regina and C. macellaria larval length and weight. In addition, there was a significant interaction effect of collection animal on larval C. macellaria measurements. No significant effect was observed in C. rufifacies larval length or weight among the preservatives or treatments. These methods could be used to establish a standard operating procedure for dealing with dehydrated larval specimens in forensic investigations.

  7. Study of the Effect of Swelling on Irradiation Assisted Stress Corrosion Cracking

    Energy Technology Data Exchange (ETDEWEB)

    Teysseyre, Sebastien Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This report describes the methodology used to study the effect of swelling on the crack growth rate of an irradiation-assisted stress corrosion crack that is propagating in highly irradiated stainless steel 304 material irradiated to 33 dpa in the Experimental Breeder Reactor-II. The material selection, specimens design, experimental apparatus and processes are described. The results of the current test are presented.

  8. Evolution of microstructure after irradiation creep in several austenitic steels irradiated up to 120 dpa at 320 °C

    Energy Technology Data Exchange (ETDEWEB)

    Renault-Laborne, A., E-mail: alexandra.renault@cea.fr [DEN-Service de Recherches Métallurgiques Appliquées, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Garnier, J.; Malaplate, J. [DEN-Service de Recherches Métallurgiques Appliquées, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Gavoille, P. [DEN-Service d' Etudes des Matériaux Irradiés, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Sefta, F. [EDF R& D, MMC, Site des Renardières, F-77818, Morêt-sur-Loing Cedex (France); Tanguy, B. [DEN-Service d' Etudes des Matériaux Irradiés, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France)

    2016-07-15

    Irradiation creep was investigated in different austenitic steels. Pressurized tubes with stresses of 127–220 MPa were irradiated in BOR-60 at 320 °C to 120 dpa. Creep behavior was dependent on both chemical composition and metallurgical state of steels. Different steels irradiated with and without stress were examined by TEM. Without stress, the irradiation produced high densities of dislocation lines and Frank loops and, depending on the type of steels, precipitates. Stress induced an increase of the precipitate mean size and density and, for some grades, an increase of the mean loop size and a decrease of their density. An anisotropy of Frank loop density or size induced by stress was not observed systematically. Dislocation line microstructure seems not to be different between the stressed and unstressed specimens. No cavities were detectable in these specimens. By comparing with the data from this work, the main irradiation creep models are discussed.

  9. Effect of irradiation temperature and strain rate on the mechanical properties of V-4Cr-4Ti irradiated to low doses in fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zinkle, S.J.; Snead, L.L.; Rowcliffe, A.F.; Alexander, D.J.; Gibson, L.T. [Oak Ridge National Lab., TN (United States)

    1998-09-01

    Tensile tests performed on irradiated V-(3-6%)Cr-(3-6%)Ti alloys indicate that pronounced hardening and loss of strain hardening capacity occurs for doses of 0.1--20 dpa at irradiation temperatures below {approximately}330 C. The amount of radiation hardening decreases rapidly for irradiation temperatures above 400 C, with a concomitant increase in strain hardening capacity. Low-dose (0.1--0.5 dpa) irradiation shifts the dynamic strain aging regime to higher temperatures and lower strain rates compared to unirradiated specimens. Very low fracture toughness values were observed in miniature disk compact specimens irradiated at 200--320 C to {approximately}1.5--15 dpa and tested at 200 C.

  10. Metal surface swelling by heavy charged particle irradiation

    CERN Document Server

    Terasawa, M; Liu, L; Tsubakino, H; Niibe, M

    2002-01-01

    Austenitic stainless steel specimens of Type 316SS were irradiated with 200 keV He sup + or N sup + ions, and the irradiated specimen surfaces were observed by atomic force microscopy (AFM). At the ion irradiation in high temperatures the specimens show surface step-up. In case of He sup + , the surface swelling is remarkable and increases linearly with He ion fluence, which indicates the swelling is due to formation of He bubbles. The irradiated surface is sometimes in irregularity, especially at and near grain boundary, remarkable ridging is observed. In case of N sup + , the surface step-up is less remarkable compared with He sup +. The swelling shows a so-called bi-linear behavior, i.e. a threshold of N sup + fluence appears and beyond the threshold the swelling increase is almost linearly presumably due to evolution of voids induced by the N sup + irradiation in high temperature. Denudation of void formation adjacent to grain boundary is recognized.

  11. 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)

  12. Residual voltage of. gamma. -irradiated polyethylene containing antioxidant

    Energy Technology Data Exchange (ETDEWEB)

    Yamanaka, Sanshiro; Fukuda, Tadashi; Sawa, Goro; Ieda, Masayuki

    1987-03-01

    Various cables used for nuclear power plants, especially those used for safety protection system must keep their functions even if a LOCA occurred at the end of 40 year plant life. When cross-linked polyethylene is exposed to radiation in the atmosphere, the tensile strength and elongation begin to decrease by the irradiation above 1.5 x 10/sup 8/ rad. In order to improve the reliability of cross-linked polyethylene, the method of increasing the radiation resistance by mixing additives has been examined. Moreover, it is necessary to diagnose the deterioration of cables by nondestructive method, but such field test is difficult now. For the purpose of detecting the radiation deterioration of polyethylene power cables nondestructively using residual voltage, the authors examined the effect of gamma ray irradiation exerted on the residual voltage of low density polyethylene to which phenol compound (Irganox 1010) or amine compound (Antage F) was added, subsequently to the study on the polyethylene without additives. The specimens, experimental method and results are reported. In the specimens without irradiation, the addition of Antage F increased the residual voltage and current. The residual voltage of the specimens with a large amount of addition showed the same value after irradiation as that in the case without irradiation. (Kako, I.).

  13. Effect of initial oxygen content on the void swelling behavior of fast neutron irradiated copper

    Energy Technology Data Exchange (ETDEWEB)

    Zinkle, S.J. [Oak Ridge National Lab., TN (United States); Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    Density measurements were performed on high purity copper specimens containing {le}10 wt.ppm and {approximately}120 wt.ppm oxygen following irradiation in FFTF MOTA 2B. Significant amounts of swelling were observed in both the oxygen-free and oxygen-doped specimens following irradiation to {approximately}17 dpa at 375 C and {approximately}47 dpa at 430 C. Oxygen doping up to 360 appm (90 wt.ppm) did not significantly affect the void swelling of copper for these irradiation conditions.

  14. Degradation of shear stiffness of Nomex honeycomb sandwich panel in laser irradiation

    Science.gov (United States)

    Wang, Jiawei; Jiang, Houman; Wu, Lixiong; Zhu, Yongxiang; Wei, Chenghua; Ma, Zhiliang; Wang, Lijun

    2017-05-01

    Based on the overhanging beam three-point bending method, the experimental system was set up to measure the variety of shear stiffness of Nomex honeycomb sandwich panel in laser irradiation. The shear stiffness of the specimens under different laser power density was measured. The result shows that the thermal effect during the laser irradiation leads to the degradation of mechanical properties of Nomex honeycomb sandwich panel. High temperature rise rate in the specimen is another main reason for the shear stiffness degeneration. This research provides a reference for the degradation of mechanical properties of composite materials in laser irradiation and proposes a new method for the study of laser interaction with matter.

  15. Positron annihilation and thermally stimulated current of electron beam irradiated polyetheretherketone

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, Shigetaka; Shinyama, Katsuyoshi; Baba, Makoto [Hachinohe Inst. of Tech., Hachinohe, Aomori (Japan); Suzuki, Takenori

    1997-03-01

    Positron lifetime measurements were applied to electron beam irradiated poly(ether-ether-ketone). The lifetime, {tau}{sub 3}, of the ortho-positronium of unirradiated and 5 MGy irradiated specimen became rapidly longer above about 150degC. {tau}{sub 3} of 50 MGy and 100 MGy irradiated specimen was shorter than that of unirradiated one. Thermally stimulated current (TSC) decreased with increasing the dose before voltage application. In the case of voltage application, a TSC peak appeared and the peak value decreased with increased the dose. The correlation between the results of positron annihilation and TSC was investigated. (author)

  16. Improvement of carbon fiber surface properties using electron beam irradiation

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    Carbon fiber-reinforced advance composites have been used for struetural applications, mainly on account of their mechanical properties. The main factor for a good mechanical performance of carbon fiber-reinforced composite is the interfacial interaction between its components, which are carbon fiber and polymeric matrix. The aim of this study is to improve the surface properties of the carbon fiber using ionizing radiation from an electron beam to obtain better adhesion properties in the resultant composite. EB radiation was applied on the carbon fiber itself before preparing test specimens for the mechanical tests. Experimental results showed that EB irradiation improved the tensile strength of carbon fiber samples. The maximum value in tensile strength was reached using doses of about 250kGy. After breakage, the morphology aspect of the tensile specimens prepared with irradiated and non-irradiated carbon fibers were evaluated. SEM micrographs showed modifications on the carbon fiber surface.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-16

    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.

  18. Breast specimen shrinkage following formalin fixation

    Directory of Open Access Journals (Sweden)

    Horn CL

    2014-02-01

    Full Text Available Christopher L Horn, Christopher Naugler Department of Pathology and Laboratory Medicine, University of Calgary, and Calgary Laboratory Services, Calgary, AB, Canada Abstract: Accurate measurement of primary breast tumors and subsequent surgical margin assessment is critical for pathology reporting and resulting patient therapy. Anecdotal observations from pathology laboratory staff indicate possible shrinkage of breast cancer specimens due to the formalin fixation process. As a result, we conducted a prospective study to investigate the possible shrinkage effects of formalin fixation on breast cancer specimens. The results revealed no significant changes in tumor size, but there were significant changes in the distance to all surgical resection margins from the unfixed to fixed state. This shrinkage effect could interfere with the accuracy of determining distance to margin assessment and tumor-free margin assessment. Thus, changes in these measurements due to the formalin fixation process have the potential to alter treatment options for the patient. Keywords: breast margins, formalin, shrinkage, cancer

  19. Is routine histopathology of tonsil specimen necessary?

    Directory of Open Access Journals (Sweden)

    Agida S Adoga

    2011-01-01

    Full Text Available Background: Tonsillar diseases are common in paediatric and adult otolaryngological practice. These diseases require tonsillectomy. Specimens are subjected to histopathology routinely in my institution for fear of infections and tumour without consideration for risk factors. The financial burden is on the patients and waste of histopathologist′s man hour because other specimens are left un-attended. This study aims to find out the necessity of routine histopathology of tonsil specimens. Materials and Methods : A 2 year retrospective review of the histopathological results of two (paediatric and adult groups of 61 patients managed for tonsillar diseases at the ENT UNIT of Jos University Teaching Hospital from July 2005 to June, 2007. Data extracted included biodata, clinical features and histopathological diagnosis. Result : The 61 patients comprise 35 children and 26 adults. The youngest and oldest paediatric patients were 1 year and 3 months and 16 years respectively, a range of 1 year 3 months to 16 years. The youngest and oldest adults were 17 and 50 years with a range of 17-50 years. Groups mean ages were 5.1 and 28.5 years. The gender ratios were 1:2.7 and 1:1.9 respectively. One adult was HIV positive. The histopathological diagnosis were chronic nonspecific tonsillitis in 10(16.6%, follicular tonsillitis in 23(38.3%, chronic suppurative tonsillitis in 10(16.6%, lymphoid hyperplasia in 18(30.0% and lymphoma in 1(1.0% respectively. Conclusion : Histopathologic request for tonsillectomy specimens should be based on certain risk factors with consideration of the cost to patients and to spare the histopathologist′s man hour.

  20. SQA specimen paper 2013, national 5, French

    CERN Document Server

    SQA

    2013-01-01

    Practise for your exam on the offical National 5 specimen paper from the Scottish Qualifications Authority. Plus each book includes additional model papers and extra revision guidance, making them an essential purchase for any student.; Discover how to get your best grade with answers checked by senior examiners.; Prepare for your exams with study skills guidance sections.; Gain vital extra marks and avoid common mistakes with examiner tips

  1. Damage modeling in Small Punch Test specimens

    DEFF Research Database (Denmark)

    Martínez Pañeda, Emilio; Cuesta, I.I.; Peñuelas, I.

    2016-01-01

    Ductile damage modeling within the Small Punch Test (SPT) is extensively investigated. The capabilities ofthe SPT to reliably estimate fracture and damage properties are thoroughly discussed and emphasis isplaced on the use of notched specimens. First, different notch profiles are analyzed...... and constraint conditionsquantified. The role of the notch shape is comprehensively examined from both triaxiality and notchfabrication perspectives. Afterwards, a methodology is presented to extract the micromechanical-basedductile damage parameters from the load-displacement curve of notched SPT samples...

  2. Comparison between the Strength Levels of Baseline Nuclear-Grade Graphite and Graphite Irradiated in AGC-2

    Energy Technology Data Exchange (ETDEWEB)

    Carroll, Mark Christopher [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-07-01

    This report details the initial comparison of mechanical strength properties between the cylindrical nuclear-grade graphite specimens irradiated in the second Advanced Graphite Creep (AGC-2) experiment with the established baseline, or unirradiated, mechanical properties compiled in the Baseline Graphite Characterization program. The overall comparative analysis will describe the development of an appropriate test protocol for irradiated specimens, the execution of the mechanical tests on the AGC-2 sample population, and will further discuss the data in terms of developing an accurate irradiated property distribution in the limited amount of irradiated data by leveraging the considerably larger property datasets being captured in the Baseline Graphite Characterization program. Integrating information on the inherent variability in nuclear-grade graphite with more complete datasets is one of the goals of the VHTR Graphite Materials program. Between “sister” specimens, or specimens with the same geometry machined from the same sub-block of graphite from which the irradiated AGC specimens were extracted, and the Baseline datasets, a comprehensive body of data will exist that can provide both a direct and indirect indication of the full irradiated property distributions that can be expected of irradiated nuclear-grade graphite while in service in a VHTR system. While the most critical data will remain the actual irradiated property measurements, expansion of this data into accurate distributions based on the inherent variability in graphite properties will be a crucial step in qualifying graphite for nuclear use as a structural material in a VHTR environment.

  3. Low dose irradiation creep of pure nickel. [17 or 15 MeV deuterons

    Energy Technology Data Exchange (ETDEWEB)

    Henager, C.H. Jr.

    1984-10-01

    A detailed climb-controlled glide model of low dose irradiation creep has been developed to rationalize irradiation creep data of pure nickel irradiated in a light ion irradiation creep apparatus. Experimental irradiation creep data were obtained to study the effects of initial microstructure and stress on low dose irradiation creep in pure nickel. Pure nickel specimens (99.992% Ni), with three different microstructures, were irradiated with 17 or 15 MeV deuterons at 473 K and stresses ranging from 0.35 to 0.9 of the unirradiated yield stress. Transmission electron microscopy revealed that the microstructure following irradiation to 0.05 dpa consisted of a high density of small dislocation loops, some small voids and network dislocations. The creep model predicted creep rates proportional to the mobile dislocation density and a comparison of experimental irradiation creep rates as a function of homologous stress revealed a dependence on initial microstructure of the magnitude predicted by the measured dislocation densities. The three microstructures that were irradiated consisted of 85% and 25% cold-worked Ni specimens and well-annealed Ni specimens. A weak stress dependence of irradiation creep was observed in 85% cold-worked Ni in agreement with experimental determinations of the stress dependence of irradiation creep by others. The weak stress dependence was shown to be a consequence of the stress independence of the dislocation climb velocity and the weak stress dependence of the barrier removal process. The irradiation creep rate was observed to be proportional to the applied stress. This linear stress dependence was suggested to be due to the stress dependence of the mobile dislocation density. 101 references, 27 figures, 11 tables.

  4. Incidental prostate cancer in radical cystoprostatectomy specimens

    Institute of Scientific and Technical Information of China (English)

    Xiao-Dong Jin; Zhao-Dian Chen; Bo Wang; Song-Liang Cai; Xiao-Lin Yao; Bai-Ye Jin

    2008-01-01

    Aim: To investigate the rates of prostate cancer (Pca) in radical cystoprostatectomy (RCP) specimens for bladder cancer in mainland China. To determine the follow-up outcome of patients with two concurrent cancers and identify whether prostate-specific antigen (PSA) is a useful tool for the detection of Pca prior to surgery. Methods: From January 2002 to January 2007, 264 male patients with bladder cancer underwent RCP at our center. All patients underwent digital rectal examination (DRE) and B ultrasound. Serum PSA levels were tested in 168 patients. None of the patients had any evidence of Pca before RCP. Entire prostates were embedded and sectioned at 5 mm intervals. Results: Incidental Pca was observed in 37 of 264 (14.0%) RCP specimens. Of these, 12 (32.4%) were clinically significant according to an accepted definition. The PSA levels were not significantly different between patients with Pca and those without Pca, nor between patients with significant Pca and those with insignificant Pca. Thirty-four patients with incidental Pca were followed up. During a mean follow-up period of 26 months, two patients with PSA > 4 ng/mL underwent castration. None of the patients died of Pca. Conclusion: The incidence of Pca in RCP specimens in mainland China is lower than that in most developed countries. PSA cannot identify asymptomatic Pca prior to RCP. In line with published reports, incidental Pca does not impact the prognosis of bladder cancer patients undergoing RCP.

  5. 42 CFR 493.1232 - Standard: Specimen identification and integrity.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 5 2010-10-01 2010-10-01 false Standard: Specimen identification and integrity... Nonwaived Testing General Laboratory Systems § 493.1232 Standard: Specimen identification and integrity. The... optimum integrity of a patient's specimen from the time of collection or receipt of the specimen through...

  6. Specimen loading list for the varying temperature experiment

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A.L.; Sitterson, R.G. [Oak Ridge National Lab., TN (United States)

    1998-09-01

    The varying temperature experiment HFIR-RB-13J has been assembled and inserted in the reactor. Approximately 5300 specimens were cleaned, inspected, matched, and loaded into four specimen holders. A listing of each specimen loaded into the steady temperature holder, its position in the capsule, and the identification of the corresponding specimen loaded into the varying temperature holder is presented in this report.

  7. HMSRP Hawaiian Monk Seal Specimen Data (includes physical specimens, collection information, status, storage locations, and laboratory results associated with individual specimens)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set includes physical specimens, paper logs and Freezerworks database of all logged information on specimens collected from Hawaiian monk seals since 1975....

  8. The Next Generation Nuclear Plant Graphite Creep Experiment Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep 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 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 graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years 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 Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will 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 will not be subjected to a compressive load during irradiation. The six stacks will have differing compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of

  9. Effect of proton irradiation on irradiation assisted stress corrosion cracking in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Han Ok; Hwang, Mi Jin; Kim, Sung Woo; Hwang, Seong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Irradiation assisted stress corrosion cracking (IASCC) involves the cracking and failure of materials under irradiation environment in nuclear power plant water environment. The major factors and processes governing an IASCC are suggested by others. The IASCC of the reactor core internals due to the material degradation and the water chemistry change has been reported in high stress stainless steel components, such as fuel elements (Boiling Water Reactors) in the 1960s, a control rod in the 1970s, and a baffle former bolt in recent years of light water reactors (Pressurized Water Reactors). Many irradiated stainless steels that are resistant to inergranular cracking in 288 .deg. C argon are susceptible to IG cracking in the simulated BWR environment at the same temperature. Under the circumstances, a lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate an IASCC in a PWR, but the mechanism in a PWR is not fully understood yet as compared with that in a BWR owing to a lack of data from laboratories and fields. Therefore, it is strongly necessary to review and analyze recent researches of an IASCC in both BWR and PWR for establishing a proactive management technology for the IASCC of core internals in Korean PWRs. The objective of this research to find IASCC behavior of proton irradiated 316 stainless steels in a high-temperature water chemistry environment. The IASCC initiation susceptibility on 1, 3, 5 DPA proton irradiated 316 austenite stainless steel was evaluated in PWR environment. SCC area ratio on the fracture surface was similar regardless of irradiation level. Total crack length on the irradiated surface increases in order of specimen 1, 3, 5 DPA. The total crack length at the side surface is a better measure in evaluating IASCC initiation susceptibility for proton-irradiated samples.

  10. Final report on neutron irradiation at low temperature to investigate plastic instability and at high temperature to study caviation

    DEFF Research Database (Denmark)

    Singh, B.N; Eldrup, Morten Mostgaard; Golubov, D.J.

    2005-01-01

    Effects of neutron irradiation on defect accumulation and physical and mechanical properties of pure iron and F82H and EUROFER 97 ferritic-martensitic steels have been investigated. Tensile specimens were neutron irradiated to a dose level of 0,23 dpa at333 and 573 K. Electrical resistivity...... and tensile properties were measured both in the unirradiated and irradiated condition. Some additional specimens of pure iron were irradiated at 333 K to doses of 10-3, 10-2 and 10-1 dpa and tensile tested at 333 K.To investigate the effect of helium on cavity nucleation and growth, specimens of pure iron...... and EUROFER 97 were implanted with different amounts of helium at 323 K and subsequently neutron irradiated to doses of 10-3, 10-2 and 10-1 dpa at 323 K. Defectmicrostructures were investigated using positron annihilation spectroscopy (PAS) and transmission electron microscopy (TEM). Numerical calculations...

  11. Influence of the irradiation temperature on the surface structure and physical/chemical properties of Ar ion-irradiated bulk metallic glasses

    Energy Technology Data Exchange (ETDEWEB)

    Menéndez, E., E-mail: Enric.MenendezDalmau@fys.kuleuven.be [KU Leuven, Instituut voor Kern-en Stralingsfysica, Celestijnenlaan 200 D, 3001 Leuven (Belgium); Hynowska, A.; Fornell, J.; Suriñach, S. [Departament de Física, Facultat de Ciències, Universitat Autònoma de Barcelona, E-08193 Bellaterra (Spain); Montserrat, J. [Institut de Microelectrònica de Barcelona (IMB-CNM), CSIC, Campus Universitat Autònoma Barcelona, E-08193 Bellaterra (Spain); Temst, K.; Vantomme, A. [KU Leuven, Instituut voor Kern-en Stralingsfysica, Celestijnenlaan 200 D, 3001 Leuven (Belgium); Baró, M.D. [Departament de Física, Facultat de Ciències, Universitat Autònoma de Barcelona, E-08193 Bellaterra (Spain); García-Lecina, E. [Surfaces Division, IK4-CIDETEC, Parque Tecnológico de San Sebastián, E-20009 Donostia (Spain); Pellicer, E., E-mail: Eva.Pellicer@uab.cat [Departament de Física, Facultat de Ciències, Universitat Autònoma de Barcelona, E-08193 Bellaterra (Spain); Sort, J., E-mail: Jordi.Sort@uab.cat [Institució Catalana de Recerca i Estudis Avançats (ICREA) and Departament de Física, Universitat Autònoma de Barcelona, E-08193 Bellaterra (Spain)

    2014-10-15

    Highlights: • Ion irradiation is performed on bulk metallic glasses at 300 K and close to T{sub g}. • Nanocrystallization is observed after high-temperature irradiation. • The mechanical properties are enhanced after the irradiation procedures. • Corrosion resistance is improved after irradiation close to T{sub g}. - Abstract: Surface treatments using multiple Ar ion irradiation processes with a maximum energy and fluence of 200 keV and 1 × 10{sup 16} ions/cm{sup 2}, respectively, have been performed on two different metallic glasses: Zr{sub 55}Cu{sub 28}Al{sub 10}Ni{sub 7} and Ti{sub 40}Zr{sub 10}Cu{sub 38}Pd{sub 12}. Analogous irradiation procedures have been carried out at room temperature (RT) and at T = 620 K (≈0.9 T{sub g}, where T{sub g} denotes the glass transition). The structure, mechanical behavior, wettability and corrosion resistance of the irradiated alloys have been compared with the properties of the as-cast and annealed (T = 620 K) non-irradiated specimens. While ion irradiation at RT does not significantly alter the amorphous structure of the alloys, ion irradiation close to T{sub g} promotes decomposition/nanocrystallization. Consequently, the hardness (H) and reduced Young’s modulus (E{sub r}) decrease after irradiation at RT but they both increase after irradiation at 620 K. While annealing close to T{sub g} increases the hydrophobicity of the samples, irradiation induces virtually no changes in the contact angle when comparing with the as-cast state. Concerning the corrosion resistance, although not much effect is found after irradiation at RT, an improvement is observed after irradiation at 620 K, particularly for the Ti-based alloy. These results are of practical interest in order to engineer appropriate surface treatments based on ion irradiation, aimed at specific functional applications of bulk metallic glasses.

  12. Wildlife specimen collection, preservation, and shipment

    Science.gov (United States)

    White, C. LeAnn; Dusek, Robert J.; Franson, J. Christian; Friend, Milton; Gibbs, Samantha E.J.; Wild, Margaret A.

    2015-01-01

    Specimens are used to provide supporting information leading to the determination of the cause of disease or death in wildlife and for disease monitoring or surveillance. Commonly used specimens for wildlife disease investigations include intact carcasses, tissues from carcasses, euthanized or moribund animals, parasites, ingested food, feces, or environmental samples. Samples from live animals or the environment (e.g., contaminated feed) in the same vicinity as a mortality event also may be helpful. The type of specimen collected is determined by availability of samples and biological objectives. Multiple fresh, intact carcasses from affected species are the most useful in establishing a cause for a mortality event. Submission of entire carcasses allows observation of gross lesions and abnormalities, as well as disease testing of multiple tissues. Samples from live animals may be more appropriate when sick animals cannot be euthanized (e.g., threatened or endangered species) or for research and monitoring projects examining disease or agents circulating in apparently healthy animals or those not exhibiting clinical signs. Samples from live animals may include collections of blood, hair, feathers, feces, or ectoparasites, or samples obtained by swabbing lesions or orifices. Photographs and videos are useful additions for recording field and clinical signs and conveying conditions at the site. Collection of environmental samples (e.g., feces, water, feed, or soil) may be appropriate when animals cannot be captured for sampling or the disease agent may persist in the environment. If lethal collection is considered necessary, biologists should refer to the policies, procedures, and permit requirements of their institution/facility and the agency responsible for species management (U.S. Fish and Wildlife Service or State natural resource agency) prior to use in the field. If threatened or endangered species are found dead, or there is evidence of illegal take, field

  13. Histological evaluation of 400 cholecystectomy specimens

    Directory of Open Access Journals (Sweden)

    H Kumar

    2015-09-01

    Full Text Available Background: A majority of gallbladder specimens show changes associated with chronic cholecystitis; however few harbour a highly lethal carcinoma. This study was conducted to review the significant histopathological findings encountered in gallbladder specimens received in our laboratory.Materials and Methods: Four hundred cholecystectomy specimens were studied over a period of five years (May, 2002 to April, 2007 received at department of pathology, Kasturba Medical College, Mangalore, India. Results: Gallstones and associated diseases were more common in women in the 4th to 5th decade as compared to men with M: F ratio of 1:1.33. Maximum number of patients (28.25% being 41 to 50 years old. Histopathologically, the most common diagnosis was chronic cholecystitis (66.75%, followed by chronic active cholecystitis (20.25%, acute cholecystitis (6%, gangrenous cholecystitis (2.25%,xanthogranulomatous cholecystitis (0.50%, empyema (1%, mucocele (0.25%, choledochal cyst (0.25%, adenocarcinoma gallbladder (1.25% and  normal  gallbladders (1%.Conclusion: All lesions were found more frequently in women except chronic active cholecystitis. Gallstones were present in (80.25% cases, and significantly associated with various lesions (P value 0.009. Pigment stones were most common, followed by cholesterol stones and mixed stones. Adequate  sectioning  is  mandatory  in  all  cases  to  assess  epithelial changes arising from cholelithiasis and chronic cholecystitis as it has been known to progress to malignancy in some cases.

  14. Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

    Energy Technology Data Exchange (ETDEWEB)

    Dethloff, Christian, E-mail: christian.dethloff@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Gaganidze, Ermile [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Svetukhin, Vyacheslav V. [Ulyanovsk State University, Leo Tolstoy Str. 42, 432970 Ulyanovsk (Russian Federation); Aktaa, Jarir [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2012-07-15

    Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different {sup 10}B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

  15. Effects of water and irradiation temperatures on IASCC susceptibility of type 316 stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Tsukada, Takashi E-mail: ttsukada@popsvr.tokai.jaeri.go.jp; Miwa, Yukio; Jitsukawa, Shiro; Shiba, Kiyoyuki; Ouchi, Asao

    2004-08-01

    Effects of water and irradiation temperatures on irradiation-assisted stress corrosion cracking (IASCC) of type 316 stainless steel were investigated. Type 316 stainless steel was irradiated at 333-673 K to a dose level of 16 dpa. Susceptibility to IASCC was evaluated by slow strain rate testing in oxygenated water in the temperature range of 513-573 K. Irradiation at 603 and 673 K caused IASCC in 513 K water, but irradiation below 473 K did not induce IASCC at 513 K. Specimens irradiated at 333 K did not show IASCC susceptibility in 513 K water, but high susceptibility was observed in 573 K water. Effect of irradiation temperature is discussed from the view points of microstructural and microcompositional changes.

  16. Mechanical strength of martensitic 10%-Cr-steel after low-dose irradiation in HFR

    Energy Technology Data Exchange (ETDEWEB)

    Materna-Morris, E. [Inst. fuer Materialforschung 1, Kernforschungszentrum Karlsruhe GmbH (Germany); Romer, O. [Hauptabteilung fuer Versuchstechnik/Heisse Zellen, Kernforschungszentrum Karlsruhe GmbH (Germany)

    1995-12-31

    Within the framework of the SIENA irradiation program (Steel Irradiation in an Enhanced Neutron Arrangement), the materials MANET I (DIN 1.4914) and AISI 316 L favored for the NET (Next European Torus) fusion reactor were investigated. The martensitic 9-12% chromium steels are considered as alternative materials for components of fusion reactors, because of their low He embrittlement and the good swelling behavior. Irradiations of the martensitic tensile specimens were performed in the reactor at 300 C, 400 C and 475 C, respectively with irradiation doses of 5, 10 and 15 dpa attained. Following the post-irradiation tensile tests, considerable hardening of the material was observed at low irradiation and test temperatures. In the microstructure, dislocation loops and He bubbles were found to occur as irradiation induced material changes. The dislocation loops contribute significantly to material embrittlement. (orig.).

  17. Quantitative TEM analysis of precipitation and grain boundary segregation in neutron irradiated EUROFER 97

    Science.gov (United States)

    Dethloff, Christian; Gaganidze, Ermile; Aktaa, Jarir

    2014-11-01

    Characterization of irradiation induced microstructural defects is essential for assessing the applicability of structural steels like the Reduced Activation Ferritic/Martensitic steel EUROFER 97 in upcoming fusion reactors. In this work Transmission Electron Microscopy (TEM) is used to analyze the types and structure of precipitates, and the evolution of their size distributions and densities caused by neutron irradiation to a dose of 32 displacements per atom (dpa) at 330-340 °C in the irradiation experiment ARBOR 1. A significant growth of MX and M23C6 type precipitates is observed after neutron irradiation, while the precipitate density remains unchanged. Hardening caused by MX and M23C6 precipitate growth is assessed by applying the Dispersed Barrier Hardening (DBH) model, and shown to be of minor importance when compared to other irradiation effects like dislocation loop formation. Additionally, grain boundary segregation of chromium induced by neutron irradiation was investigated and detected in irradiated specimens.

  18. Isolation of Klebsiella terrigena from clinical specimens.

    Science.gov (United States)

    Podschun, R; Ullmann, U

    1992-04-01

    In a three-year survey conducted from 1988 to 1990 Klebsiella isolates from human clinical specimens were subjected to additional tests to identify any Klebsiella terrigena strains. Ten strains of Klebsiella terrigena (0.4%) were found among 2355 indole-negative Klebsiella isolates. Most of the isolates were recovered from the respiratory tract. In the API20EC system almost exclusively biotypes no. 1777771 and 1777671 were observed. Serotyping revealed capsule types K2, K5 and K18 in two strains each. In antibiotic susceptibility tests the strains were shown to be comparable in sensitivity to Klebsiella pneumoniae.

  19. Behavior of beryllium pebbles under irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Dalle-Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik; Baldwin, D.L.; Gelles, D.S.; Greenwood, L.R.; Kawamura, H.; Oliver, B.M.

    1998-01-01

    Beryllium pebbles are being considered in fusion reactor blanket designs as neutron multiplier. An example is the European `Helium Cooled Pebble Bed Blanket.` Several forms of beryllium pebbles are commercially available but little is known about these forms in response to fast neutron irradiation. Commercially available beryllium pebbles have been irradiated to approximately 1.3 x 10{sup 22} n/cm{sup 2} (E>1 MeV) at 390degC. Pebbles 1-mm in diameter manufactured by Brush Wellman, USA and by Nippon Gaishi Company, Japan, and 3-mm pebbles manufactured by Brush Wellman were included. All were irradiated in the below-core area of the Experimental Breeder Reactor-II in Idaho Falls, USA, in molybdenum alloy capsules containing helium. Post-irradiation results are presented on density change measurements, tritium release by assay, stepped-temperature anneal, and thermal ramp desorption tests, and helium release by assay and stepped-temperature anneal measurements, for Be pebbles from two manufacturing methods, and with two specimen diameters. The experimental results on density change and tritium and helium release are compared with the predictions of the code ANFIBE. (author)

  20. Response of the canine esophagus to irradiation.

    Science.gov (United States)

    Gillette, S M; Poulson, J M; Deschesne, K M; Chaney, E L; Gillette, E L

    1998-09-01

    One hundred twenty-eight beagle dogs were randomized to receive thoracic irradiation with doses between 0 and 72 Gy in 1.5-Gy fractions over 6 weeks. Dogs were randomized to have either 33, 67 or 100% of their lung volume irradiated. The entire thoracic portion of the esophagus and variable portions of the fundus of the stomach were included in the treatment field at all volumes. Sixteen of the 128 dogs entered in the study developed clinical signs of esophagitis. These 16 dogs received doses between 45 and 72 Gy. Clinical signs of esophagitis/gastritis included dysphagia, anorexia, emesis, excessive salivation and weight loss that required force-feeding of a liquid diet. An ED50 of 67.2 Gy (95% CI 61.45-79.7 Gy) was calculated for the occurrence of clinical signs that required some supportive treatment. Three of the 16 dogs receiving 63 or 72 Gy failed to respond to treatment and were euthanized. Twenty-five other dogs were euthanized prior to 2 years due to other treatment-related complications. Two dogs died of causes not related to treatment. No late esophageal complications were observed in the remaining 98 dogs out to 2 years after irradiation. Esophageal specimens from 79 dogs were available for quantitative histological analysis 2 years after irradiation. Histological analysis showed a decrease in the percentage of glandular tissue with a corresponding increase in lamina propria and muscle.

  1. Microstructural examination of irradiated vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States); Chung, H.M. [Argonne National Lab., IL (United States)

    1997-04-01

    Microstructural examination results are reported for a V-5Cr-5Ti unirradiated control specimens of heat BL-63 following annealing at 1050{degrees}C, and V-4Cr-4Ti heat BL-47 irradiated in three conditions from the DHCE experiment: at 425{degrees}C to 31 dpa and 0.39 appm He/dpa, at 600{degrees}C to 18 dpa and 0.54 appm He/dpa and at 600{degrees}C to 18 dpa and 4.17 appm He/dpa.

  2. Modelling property changes in graphite irradiated at changing irradiation temperature

    CSIR Research Space (South Africa)

    Kok, S

    2011-01-01

    Full Text Available A new method is proposed to predict the irradiation induced property changes in nuclear; graphite, including the effect of a change in irradiation temperature. The currently used method; to account for changes in irradiation temperature, the scaled...

  3. Room temperature fatigue behavior of OFHC copper and CuAl25 specimens of two sizes

    DEFF Research Database (Denmark)

    Singhal, A.; Stubbins, J.F.; Singh, B.N.

    1994-01-01

    requiring an understanding of their fatigue behavior.This paper describes the room temperature fatigue behavior of unirradiated OFHC (oxygen-free high-conductivity) copper and CuAl25 (copper strengthened with a 0.25% atom fraction dispersion of alumina). The response of two fatigue specimen sizes to strain......Copper and its alloys are appealing for application in fusion reactor systems for high heat flux components where high thermal conductivities are critical, for instance, in divertor components. The thermal and mechanical loading of such components will be, at least in part, cyclic in nature, thus...... controlled fatigue loading is examined, and differences in behavior are discussed. Specimens with the smaller size are now being irradiated in several reactors...

  4. Evaluation of irradiation assisted stress corrosion cracking (IASCC) of type 316 stainless steel irradiated in FBR

    Energy Technology Data Exchange (ETDEWEB)

    Tsukada, T. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)); Jitsukawa, S. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)); Shiba, K. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)); Sato, Y. (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan)); Shibahara, I. (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan)); Nakajima, H. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan))

    1993-12-01

    Type 316 stainless steel from the core of the experimental fast breeder reactor (FBR) JOYO was examined by the slow strain rate tensile (SSRT) test in pure, oxygenated-water and air and by the electrochemical potentiokinetic reactivation (EPR) test to evaluate a susceptibility to the irradiation assisted stress corrosion cracking (IASCC) and the radiation-induced segregation (RIS). The solution annealed and 20% cold-worked materials had been irradiated at 425 C to a neutron fluence of 8.3x10[sup 26] n/m[sup 2] (> 0.1 MeV) which is equivalent to 40 displacement per atom (dpa). Intergranular cracking was induced by the SSRT in water at 200 and 300 C, but was not observed on specimen tested in water at 60 C and in air at 300 C. This indicates that irradiation increased a susceptibility to stress corrosion cracking (SCC) in water. After the EPR test, grain boundary etching was observed in addition to grain face etching. This suggests Cr depletion may have occurred both at grain boundary and at defect clusters during the irradiation. The results are compared with the behavior of similar materials irradiated with different neutron spectrum. (orig.)

  5. Specimen preparation for cryogenic coherent X-ray diffraction imaging of biological cells and cellular organelles by using the X-ray free-electron laser at SACLA.

    Science.gov (United States)

    Kobayashi, Amane; Sekiguchi, Yuki; Oroguchi, Tomotaka; Okajima, Koji; Fukuda, Asahi; Oide, Mao; Yamamoto, Masaki; Nakasako, Masayoshi

    2016-07-01

    Coherent X-ray diffraction imaging (CXDI) allows internal structures of biological cells and cellular organelles to be analyzed. CXDI experiments have been conducted at 66 K for frozen-hydrated biological specimens at the SPring-8 Angstrom Compact Free-Electron Laser facility (SACLA). In these cryogenic CXDI experiments using X-ray free-electron laser (XFEL) pulses, specimen particles dispersed on thin membranes of specimen disks are transferred into the vacuum chamber of a diffraction apparatus. Because focused single XFEL pulses destroy specimen particles at the atomic level, diffraction patterns are collected through raster scanning the specimen disks to provide fresh specimen particles in the irradiation area. The efficiency of diffraction data collection in cryogenic experiments depends on the quality of the prepared specimens. Here, detailed procedures for preparing frozen-hydrated biological specimens, particularly thin membranes and devices developed in our laboratory, are reported. In addition, the quality of the frozen-hydrated specimens are evaluated by analyzing the characteristics of the collected diffraction patterns. Based on the experimental results, the internal structures of the frozen-hydrated specimens and the future development for efficient diffraction data collection are discussed.

  6. Specimen preparation for cryogenic coherent X-ray diffraction imaging of biological cells and cellular organelles by using the X-ray free-electron laser at SACLA

    Science.gov (United States)

    Kobayashi, Amane; Sekiguchi, Yuki; Oroguchi, Tomotaka; Okajima, Koji; Fukuda, Asahi; Oide, Mao; Yamamoto, Masaki; Nakasako, Masayoshi

    2016-01-01

    Coherent X-ray diffraction imaging (CXDI) allows internal structures of biological cells and cellular organelles to be analyzed. CXDI experiments have been conducted at 66 K for frozen-hydrated biological specimens at the SPring-8 Angstrom Compact Free-Electron Laser facility (SACLA). In these cryogenic CXDI experiments using X-ray free-electron laser (XFEL) pulses, specimen particles dispersed on thin membranes of specimen disks are transferred into the vacuum chamber of a diffraction apparatus. Because focused single XFEL pulses destroy specimen particles at the atomic level, diffraction patterns are collected through raster scanning the specimen disks to provide fresh specimen particles in the irradiation area. The efficiency of diffraction data collection in cryogenic experiments depends on the quality of the prepared specimens. Here, detailed procedures for preparing frozen-hydrated biological specimens, particularly thin membranes and devices developed in our laboratory, are reported. In addition, the quality of the frozen-hydrated specimens are evaluated by analyzing the characteristics of the collected diffraction patterns. Based on the experimental results, the internal structures of the frozen-hydrated specimens and the future development for efficient diffraction data collection are discussed. PMID:27359147

  7. Effects of helium content of microstructural development in Type 316 stainless steel under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Maziasz, P.J.

    1985-11-01

    This work investigated the sensitivity of microstructural evolution, particularly precipitate development, to increased helium content during thermal aging and during neutron irradiation. Helium (110 at. ppM) was cold preinjected into solution annealed (SA) DO-heat type 316 stainess steel (316) via cyclotron irradiation. These specimens were then exposed side by side with uninjected samples. Continuous helium generation was increased considerably relative to EBR-II irradiation by irradiation in HFIR. Data were obtained from quantitative analytical electron microscopy (AEM) in thin foils and on extraction replicas. 480 refs., 86 figs., 19 tabs.

  8. Irradiation Creep in Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Ubic, Rick; Butt, Darryl; Windes, William

    2014-03-13

    An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

  9. Irradiation creep and swelling of various austenitic alloys irradiated in PFR and FFTF

    Energy Technology Data Exchange (ETDEWEB)

    Garner, F.A.; Toloczko, M.B. [Pacific Northwest National Lab., Richland, WA (United States)] [and others

    1996-10-01

    In order to use data from surrogate neutron spectra for fusion applications, it is necessary to analyze the impact of environmental differences on property development. This is of particular importance in the study of irradiation creep and its interactions with void swelling, especially with respect to the difficulty of separation of creep strains from various non-creep strains. As part of an on-going creep data rescue and analysis effort, the current study focuses on comparative irradiations conducted on identical gas-pressurized tubes produced and constructed in the United States from austenitic steels (20% CW 316 and 20% CW D9), but irradiated in either the Prototype Fast Reactor (PFR) in the United Kingdom or the Fast Flux Test Facility in the United States. In PFR, Demountable Subassemblies (DMSA) serving as heat pipes were used without active temperature control. In FFTF the specimens were irradiated with active ({+-}{degrees}5C) temperature control. Whereas the FFTF irradiations involved a series of successive side-by-side irradiation, measurement and reinsertion of the same series of tubes, the PFR experiment utilized simultaneous irradiation at two axial positions in the heat pipe to achieve different fluences at different flux levels. The smaller size of the DMSA also necessitated a separation of the tubes at a given flux level into two groups (low-stress and high-stress) at slightly different axial positions, where the flux between the two groups varied {le}10%. Of particular interest in this study was the potential impact of the two types of separation on the derivation of creep coefficients.

  10. Effect of LASER Irradiation on the Shear Bond Strength of Zirconia Ceramic Surface to Dentin

    Directory of Open Access Journals (Sweden)

    Sima Shahabi

    2012-09-01

    Full Text Available Background and Aims: Reliable bonding between tooth substrate and zirconia-based ceramic restorations is always of great importance. The laser might be useful for treatment of ceramic surfaces. The aim of the present study was to investigate the effect of laser irradiation on the shear bond strength of zirconia ceramic surface to dentin. Materials and Methods: In this experimental in vitro study, 40 Cercon zirconia ceramic blocks were fabricated. The surface treatment was performed using sandblasting with 50-micrometer Al2O3, CO2 laser, or Nd:YAG laser in each test groups. After that, the specimens were cemented to human dentin with resin cement. The shear bond strength of ceramics to dentin was determined and failure mode of each specimen was analyzed by stereo-microscope and SEM investigations. The data were statistically analyzed by one-way analysis of variance and Tukey multiple comparisons. The surface morphology of one specimen from each group was investigated under SEM. Results: The mean shear bond strength of zirconia ceramic to dentin was 7.79±3.03, 9.85±4.69, 14.92±4.48 MPa for CO2 irradiated, Nd:YAG irradiated, and sandblasted specimens, respectively. Significant differences were noted between CO2 (P=0.001 and Nd:YAG laser (P=0.017 irradiated specimens with sandblasted specimens. No significant differences were observed between two laser methods (P=0.47. The mode of bond failure was predominantly adhesive in test groups (CO2 irradiated specimens: 75%, Nd:YAG irradiated: 66.7%, and sandblasting: 41.7%. Conclusion: Under the limitations of the present study, surface treatment of zirconia ceramics using CO2 and Nd:YAG lasers was not able to produce adequate bond strength with dentin surfaces in comparison to sandblasting technique. Therefore, the use of lasers with the mentioned parameters may not be recommended for the surface treatment of Cercon ceramics.

  11. High Temperature Tensile Properties of Unirradiated and Neutron Irradiated 20 Cr-35 Ni Austenitic Steel

    Energy Technology Data Exchange (ETDEWEB)

    Roy, R.B.; Solly, B.

    1966-12-15

    The tensile properties of an unirradiated and neutron irradiated (at 40 deg C) 20 % Cr, 35 % Ni austenitic steel have been studied at 650 deg C, 750 deg C and 820 deg C. The tensile elongation and mode of fracture (transgranular) of unirradiated specimens tested at room temperature and 650 deg C are almost identical. At 750 deg C and 820 deg C the elongation decreases considerably and a large part of the total elongation is non-uniform. Furthermore, the mode of fracture at these temperatures is intergranular and microscopic evidence suggests that fracture is caused by formation and linkup of grain boundary cavities. YS and UTS decrease monotonically with temperature. Irradiated specimens show a further decrease in ductility and an increase in the tendency to grain boundary cracking. Irradiation has no significant effect on the YS, but the UTS are reduced. The embrittlement of the irradiated specimens is attributed to the presence of He and Li atoms produced during irradiation and the possible mechanisms are discussed. Prolonged annealing of irradiated and unirradiated specimens at 650 deg C appears to have no significant effect on tensile properties.

  12. 2H-SiC Dendritic Nanocrystals In Situ Formation from Amorphous Silicon Carbide under Electron Beam Irradiation

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    Under electron beam irradiation, the in-situ formation of 2H-SiC dentritic nanocrystals from amorphous silicon carbide at room temperature was observed. The homogenous transition mainly occurs at the thin edge and on the surface of specimen where the energy obtained from electron beam irradiation is high enough to cause the amorphous crystallizing into 2H-SiC.

  13. The effect of fibre content, fibre size and alkali treatment to Charpy impact resistance of Oil Palm fibre reinforced composite material

    Science.gov (United States)

    Fitri, Muhamad; Mahzan, Shahruddin

    2016-11-01

    In this research, the effect of fibre content, fibre size and alkali treatment to the impact resistance of the composite material have been investigated, The composite material employs oil palm fibre as the reinforcement material whereas the matrix used for the composite materials are polypropylene. The Oil Palm fibres are prepared for two conditions: alkali treated fibres and untreated fibres. The fibre sizes are varied in three sizes: 5mm, 7mm and 10mm. During the composite material preparation, the fibre contents also have been varied into 3 different percentages: 5%, 7% and 10%. The statistical approach is used to optimise the variation of specimen determined by using Taguchi method. The results were analyzed also by the Taguchi method and shows that the Oil Palm fibre content is significantly affect the impact resistance of the polymer matrix composite. However, the fibre size is moderately affecting the impact resistance, whereas the fibre treatment is insignificant to the impact resistance of the oil palm fibre reinforced polymer matrix composite.

  14. Enzymatic detection of formalin-fixed museum specimens for DNA analysis and enzymatic maceration of formalin-fixed specimens

    DEFF Research Database (Denmark)

    Sørensen, Margrethe; Redsted Rasmussen, Arne; Simonsen, Kim Pilkjær

    2016-01-01

    % ethanol. The method was subsequently tested on wild-living preserved specimens and an archived specimen. The protease enzyme used was SavinaseH 16 L, Type EX from Novozymes A/S. The enzymatic screening test demands only simple laboratory equipment. The method is useful for natural history collections...... in museums where DNA analyses of archived specimens are performed. Wasted time and resources can be avoided through the detection of formalin-fixed specimens because these specimens yield low-quality, damaged DNA. In addition to the screening method, it is shown that formalin-preserved specimens can...

  15. Reheating of zinc-titanate sintered specimens

    Directory of Open Access Journals (Sweden)

    Labus N.

    2015-01-01

    Full Text Available The scope of this work was observing dimensional and heat transfer changes in ZnTiO3 samples during heating in nitrogen and air atmosphere. Interactions of bulk specimens with gaseous surrounding induce microstructure changes during heating. Sintered ZnTiO3 nanopowder samples were submitted to subsequent heating. Dilatation curves and thermogravimetric with simultaneous differential thermal analysis TGA/DTA curves were recorded. Reheating was performed in air and nitrogen atmospheres. Reheated samples obtained at different characteristic temperatures in air were analyzed by X-ray diffraction (XRD. Microstructures obtained by scanning electron microscopy (SEM of reheated sintered samples are presented and compared. Reheating in a different atmosphere induced different microstructures. The goal was indicating possible causes leading to the microstructure changes. [Projekat Ministarstva nauke Republike Srbije, br. OI172057 i br. III45014

  16. Dissolution of bulk specimens of silicon nitride

    Science.gov (United States)

    Davis, W. F.; Merkle, E. J.

    1981-01-01

    An accurate chemical characterization of silicon nitride has become important in connection with current efforts to incorporate components of this material into advanced heat engines. However, there are problems concerning a chemical analysis of bulk silicon nitride. Current analytical methods require the pulverization of bulk specimens. A pulverization procedure making use of grinding media, on the other hand, will introduce contaminants. A description is given of a dissolution procedure which overcomes these difficulties. It has been found that up to at least 0.6 g solid pieces of various samples of hot pressed and reaction bonded silicon nitride can be decomposed in a mixture of 3 mL hydrofluoric acid and 1 mL nitric acid overnight at 150 C in a Parr bomb. High-purity silicon nitride is completely soluble in nitric acid after treatment in the bomb. Following decomposition, silicon and hydrofluoric acid are volatilized and insoluble fluorides are converted to a soluble form.

  17. Evaluation of impacts of stress triaxiality on plastic deformability of RAFM steel using various types of tensile specimen

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Taichiro, E-mail: kato.taichiro@jaea.go.jp [Japan Atomic Energy Agency, 2-166, Obuchi-omotedate, Rokkasho, Aomori 039-3212 (Japan); Ohata, Mitsuru [Osaka University, 2-1, Yamada-Oka, Suita, Osaka 565-0871 (Japan); Nogami, Shuhei [Tohoku University, 6-6-01-2, Aramaki-aza-Aoba, Aoba-ku, Sendai, Miyagi 980-8579 (Japan); Tanigawa, Hiroyasu [Japan Atomic Energy Agency, 2-166, Obuchi-omotedate, Rokkasho, Aomori 039-3212 (Japan)

    2016-11-01

    Highlights: • The fracture ductility is lower as the stress triaxiality is higher. • Voids of the interrupted RB1 specimen were observed along grain boundaries and expanded parallel to the tensile axis. • Voids of interrupted R0.2 specimen were rounded shape than those of RB1. • The fracture surface of specimens were observed the elongated and the equiaxed dimples. • The decrease of plastic deformability of the notched specimen was caused by the process of voids formation and crack growth due to the effect of plastic constraint of the notch. - Abstract: A case study on a fusion blanket design such as DEMO indicated that there could be some sections with high stress triaxiality, a parameter to evaluate the magnitude of plastic constraint, in the case of plasma disruption or coolant loss accident. Therefore, it is necessary to accurately understand the ductility loss limit of structural material in order to conduct the structural design assessment of the irradiated and embrittled fusion reactor blanket. Tensile tests were conducted by using three kinds of tensile specimen shapes to investigate of the plastic deformability of F82H. From the results, the fracture ductility is lower as the stress triaxiality is higher. Voids of the interrupted RB1 specimen were observed along grain boundaries and expanded parallel to the tensile axis. That of interrupted R0.2 specimen was rounded shape compared with those of RB1. The fracture surface of RB1 and R0.2 specimens were observed the elongated dimples and the equiaxed dimples without so much elongation, respectively. It is considered that the decrease of plastic deformability for the notched specimen was caused by the process of voids formation and crack growth due to the effect of plastic constraint of the notch.

  18. Irradiation damage of SiC semiconductor device (I)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ji Yeon; Kim, Weon Ju

    2000-09-01

    This report reviewed the irradiation damage of SiC semiconductor devices and examined a irradiation behavior of SiC single crystal as a pre-examination for evaluation of irradiation behavior of SiC semiconductor devices. The SiC single was crystal irradiated by gamma-beam, N+ ion and electron beam. Annealing examinations of the irradiated specimens also were performed at 500 deg C. N-type 6H-SiC dopped with N+ ion was used and irradiation doses of gamma-beam, N+ion and electron beam were up to 200 Mrad, 1x10{sup 16} N{sup +} ions/cm{sup 2} and 3.6 x 10{sup 17} e/cm{sup 2} and 1.08 x 10{sup 18} e/cm{sup 2} , respectively. Irradiation damages were analyzed by the EPR method. Additionally, properties of SiC, information about commercial SiC single crystals and the list of web sites with related to the SiC device were described in the appendix.

  19. Tensile test of dumbbell-shaped specimen in thickness direction

    Science.gov (United States)

    Iizuka, Takashi

    2016-10-01

    Sheet metal forming is widely used in manufacturing shops, and evaluation of forming limit for sheet metal is important. However, specimen shape influences on the fracture of the sheet metal. As one of methods to decrease these effects, an uniaxial tensile test using specimen dumbbell-shaped in thickness direction had been examined using FEM analysis. In this study, actually specimen dumbbell-shaped in thickness direction was fabricated using a new incremental sheet forging method, and uniaxial tensile test was conducted. Load-stroke diagram, fracture morphologies, stress-strain curves and shape after fracture were investigated, and effects of specimen shape were considered. Elongation was larger as using specimen dumbbell-shaped in the width direction. Stress-strain curves until necking occurred were less influenced by specimen shape. However, yield stress decreased and local elongation increased as using specimen dumbbell-shaped in the width direction. The reasons why these tendencies showed were considered in the view of specimen shapes.

  20. A new specimen management system using RFID technology.

    Science.gov (United States)

    Shim, Hun; Uh, Young; Lee, Seung Hwan; Yoon, Young Ro

    2011-12-01

    The specimen management system with barcode needs to be improved in order to solve inherent problems in work performance. This study describes the application of Radio Frequency Identification (RFID) which is the solution for the problems associated with specimen labeling and management. A new specimen management system and architecture with RFID technology for clinical laboratory was designed. The suggested system was tested in various conditions such as durability to temperature and aspect of effective utilization of new work flow under a virtual hospital clinical laboratory environment. This system demonstrates its potential application in clinical laboratories for improving work flow and specimen management. The suggested specimen management system with RFID technology has advantages in comparison to the traditional specimen management system with barcode in the aspect of mass specimen processing, robust durability of temperature, humidity changes, and effective specimen tracking.

  1. Effect of irradiation damage on hydrothermal corrosion of SiC

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, Sosuke, E-mail: kondo@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011 (Japan); Lee, Moonhee; Hinoki, Tatsuya [Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011 (Japan); Hyodo, Yoshihiro; Kano, Fumihisa [Power and Industrial Systems Research and Development Center, Toshiba Corporation, Yokohama, Kanagawa 235-8523 (Japan)

    2015-09-15

    The hydrothermal corrosion behavior (320 °C, 20 MPa, 168 h) of high-purity chemical-vapor-deposited (CVD) SiC pre-irradiated with 5.1-MeV Si ions at 400 and 800 °C and 0.1–2.6 dpa was studied in order to clarify the effects of irradiation damage on SiC corrosion. Regardless of the pre-irradiation conditions, selective corrosion was observed at the grain boundaries and stacking faults even at the unirradiated regions. In contrast to the complete loss of the irradiated regions observed in the specimens irradiated at 400 °C during the autoclave test, a number of large grains survived in the case of the specimens irradiated at 800 °C. The corrosion rates at the irradiated regions increased with increasing irradiation fluence, with a significant dependence in the lower dpa regime similar to that observed in the point-defect swelling. SiO{sub 2} formation was not detected in any case. Cross-sectional scanning transmission electron microscopy (TEM) and electron energy loss spectroscopy (EELS) analyses of the surfaces of the surviving grains revealed oxygen diffusion to a depth of 3.0 nm from the surface. A significant reduction of the oxygen diffusion barrier at the surface was implicated as one of the key mechanisms of the acceleration of the ion-irradiated SiC corrosion rates.

  2. A non-destructive DNA sampling technique for herbarium specimens.

    Science.gov (United States)

    Shepherd, Lara D

    2017-01-01

    Herbarium specimens are an important source of DNA for plant research but current sampling methods require the removal of material for DNA extraction. This is undesirable for irreplaceable specimens such as rare species or type material. Here I present the first non-destructive sampling method for extracting DNA from herbarium specimens. DNA was successfully retrieved from robust leaves and/or stems of herbarium specimens up to 73 years old.

  3. Drone Transport of Microbes in Blood and Sputum Laboratory Specimens.

    Science.gov (United States)

    Amukele, Timothy K; Street, Jeff; Carroll, Karen; Miller, Heather; Zhang, Sean X

    2016-10-01

    Unmanned aerial vehicles (UAVs) could potentially be used to transport microbiological specimens. To examine the impact of UAVs on microbiological specimens, blood and sputum culture specimens were seeded with usual pathogens and flown in a UAV for 30 ± 2 min. Times to recovery, colony counts, morphologies, and matrix-assisted laser desorption ionization-time of flight mass spectrometry (MALDI-TOF MS)-based identifications of the flown and stationary specimens were similar for all microbes studied.

  4. The type specimen of Anoura geoffroyi lasiopyga (Chiroptera: Phyllostomidae)

    Science.gov (United States)

    Arroyo-Cabrales, Joaquin; Gardner, A.L.

    2003-01-01

    In 1868, Wilhelm Peters described Glossonycteris lasiopyga, based on a specimen provided by Henri de Saussure and collected in Mexico. The type specimen was presumed to be among those housed in the collections of the Zoologisches Museum of the Humboldt Universitat in Berlin, Germany. Our study of one of Saussure?s specimens from Mexico, discovered in the collections of the Museum d?Histoire Naturelle, Geneva, Switzerland, demonstrates that it and not one of the Berlin specimens is the holotype.

  5. Tensile Hoop Behavior of Irradiated Zircaloy-4 Nuclear Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jaramillo, Roger A [ORNL; Hendrich, WILLIAM R [ORNL; Packan, Nicolas H [ORNL

    2007-03-01

    A method for evaluating the room temperature ductility behavior of irradiated Zircaloy-4 nuclear fuel cladding has been developed and applied to evaluate tensile hoop strength of material irradiated to different levels. The test utilizes a polyurethane plug fitted within a tubular cladding specimen. A cylindrical punch is used to compress the plug axially, which generates a radial displacement that acts upon the inner diameter of the specimen. Position sensors track the radial displacement of the specimen outer diameter as the compression proceeds. These measurements coupled with ram force data provide a load-displacement characterization of the cladding response to internal pressurization. The development of this simple, cost-effective, highly reproducible test for evaluating tensile hoop strain as a function of internal pressure for irradiated specimens represents a significant advance in the mechanical characterization of irradiated cladding. In this project, nuclear fuel rod assemblies using Zircaloy-4 cladding and two types of mixed uranium-plutonium oxide (MOX) fuel pellets were irradiated to varying levels of burnup. Fuel pellets were manufactured with and without thermally induced gallium removal (TIGR) processing. Fuel pellets manufactured by both methods were contained in fuel rod assemblies and irradiated to burnup levels of 9, 21, 30, 40, and 50 GWd/MT. These levels of fuel burnup correspond to fast (E > 1 MeV) fluences of 0.27, 0.68, 0.98, 1.4 and 1.7 1021 neutrons/cm2, respectively. Following irradiation, fuel rod assemblies were disassembled; fuel pellets were removed from the cladding; and the inner diameter of cladding was cleaned to remove residue materials. Tensile hoop strength of this cladding material was tested using the newly developed method. Unirradiated Zircaloy-4 cladding was also tested. With the goal of determining the effect of the two fuel types and different neutron fluences on clad ductility, tensile hoop strength tests were

  6. IR and UV irradiations on ion bombarded polycrystalline silver

    Energy Technology Data Exchange (ETDEWEB)

    Latif, Anwar, E-mail: anwarlatif@uet.edu.p [Department of Physics, University of Engineering and Technology, Lahore 54890 (Pakistan); Khaleeq-ur-Rahman, M.; Bhatti, K.A.; Rafique, M.S.; Rizvi, Z.H. [Department of Physics, University of Engineering and Technology, Lahore 54890 (Pakistan)

    2010-10-15

    Ion bombarded polycrystalline fine polished silver surfaces are exposed to Nd:YAG (1064 nm, 10 mJ, 12 ns) and KrF excimer (248 nm, 57 mJ, 20 ns) lasers to examine structural and morphological changes employing X-ray diffractometry and optical microscopy, respectively. Irradiation causes considerable changes in grain sizes. Hydrodynamic sputtering is found to be dominant in heat affected zones (HAZs). Craters with irregular boundary and non-uniform thermal conduction are resulted on laser ablated surfaces of ion bombarded specimens. No disturbance takes place in the d-spacing of the planes of irradiated samples.

  7. Fixture For Hot Stress Tests Of Thin Specimens

    Science.gov (United States)

    Gates, Thomas S.

    1993-01-01

    Fixture designed to hold and heat thin, rectangular-cross-section specimen of composite material during hot lengthwise-stress test. Suitable for testing same specimen in either tension or compression. Clamps lightly onto specimen, providing both heat via thermal conduction and lateral support needed to prevent buckling during compression test.

  8. 49 CFR 219.205 - Specimen collection and handling.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Specimen collection and handling. 219.205 Section... § 219.205 Specimen collection and handling. (a) General. Urine and blood specimens must be obtained, marked, preserved, handled, and made available to FRA consistent with the requirements of this subpart...

  9. The CAS Bio-specimen Centers in Sound Progress

    Institute of Scientific and Technical Information of China (English)

    LI Liangqian; QIAO Gexia; YAO Yijian

    2010-01-01

    @@ Bio-specimen centers, including herbaria and zoological museums, are the most integrated places for the storage of specimens, which are real samples and the most important vouchers for taxonomic and biodiversity studies.The information carried by the specimens is of substantial reference for research on species distribution, history, status, phylogeny and evolution, etc.

  10. Adaptation of Museum Specimens for Use in Anatomical Teaching Aids

    Science.gov (United States)

    Harris, P. F.; And Others

    1977-01-01

    Color transparencies are prepared of a re-colored anatomical specimen after placing labels temporarily in position to indicate specific structures. The specimen is also radiographed to show skeletal and soft tissue structures. Cross-reference among the specimen, photographs, and radiographs is supplemented by examination and self-assessment…

  11. Reliability of 46,XX results on miscarriage specimens: a review of 1,222 first-trimester miscarriage specimens

    National Research Council Canada - National Science Library

    Lathi, Ruth B; Gustin, Stephanie L F; Keller, Jennifer; Maisenbacher, Melissa K; Sigurjonsson, Styrmir; Tao, Rosina; Demko, Zach

    2014-01-01

    To examine the rate of maternal contamination in miscarriage specimens. Retrospective review of 1,222 miscarriage specimens submitted for chromosome testing with detection of maternal cell contamination (MCC...

  12. Measurement of the dynamic fracture toughness with notched PMMA specimen under impact loading

    OpenAIRE

    2009-01-01

    International audience; In the present study three-point-bend impact experiments were conducted using an instrumented Charpy pendulum with a laser displacement measurement to better understand the correlation between impact velocity and the dynamic effects observed on the load-time curves. The experiments were performed at impact velocities ranging from 1 to 4 m/s. The aim of this work is to measure the dynamic fracture toughness at high impact velocities where the classical method is limited...

  13. New suoid specimens from Gebel Zelten, Libya

    Directory of Open Access Journals (Sweden)

    Pickford, M.

    2006-12-01

    Full Text Available A restricted collection of suoids from Gebel Zelten was made in the 1990’s by the Spanish-Libyan Palaeontology Expedition. Dr Dolores Soria filmed the specimens with a video camera and took measurements of the teeth with vernier calipers. This paper uses the images from the video, which, even though somewhat limited in terms of picture quality, are of interest because they represent the first known snout of the gigantic suid Megalochoerus khinzikebirus. The images reveal that it is basically an enlarged version of Libycochoerus massai, but with relatively small premolars. The sanithere specimens from the site were photographed with an Olympus 1.4 megapixel digital camera, and the image quality is better than from the video camera. These specimens throw light on the degree of sexual dimorphism exhibited by sanitheres, a feature that was previously inferred from isolated teeth, but which can now be confirmed on the basis of the two mandible fragments from Gebel Zelten. This paper is dedicated to the memory of Dr Soria. This paper takes into account a few undescribed suid post-cranial bones from Gebel Zelten housed in the Natural History Museum, London, collected during the 1960’s by R. Savage.Una limitada colección de suoideos procedentes de Gebel Zelten fue hecha a finales de los años 1990 por una expedición paleontológica internacional, con participación española y libia. La Dra. Dolores Soria filmó los ejemplares con una cámara de vídeo y tomó las medidas de los dientes con calibre. En este trabajo se utilizan las imágenes filmadas, que, aunque algo limitadas en términos de calidad fotográfica, son interesantes porque representan las primeras conocidas del rostro del suido gigante Megalochoerus khinzikebirus. Las imágenes revelan que básicamente es una versión agrandada de Libycochoerus massai, pero con premolares relativamente más pequeños. Los ejemplares de saniterios fueron fotografiados con una cámara digital Olympus

  14. AGC-2 Irradiation Report

    Energy Technology Data Exchange (ETDEWEB)

    Rohrbaugh, David Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Windes, William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-06-01

    The Next Generation Nuclear Plant (NGNP) will be a helium-cooled, very high temperature reactor (VHTR) with a large graphite core. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphites have been developed and are considered suitable candidates for the new NGNP reactor design. To support the design and licensing of NGNP core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade with a specific emphasis on data related to the life limiting effects of irradiation creep on key physical properties of the NGNP candidate graphites. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change. In this case, the graphite physically changes i.e., first shrinking and then expanding with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the

  15. Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2009-05-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep 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 graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years 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 at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any

  16. Apparatus and method for magnetically processing a specimen

    Science.gov (United States)

    Ludtka, Gerard M; Ludtka, Gail M; Wilgen, John B; Kisner, Roger A; Jaramillo, Roger A

    2013-09-03

    An apparatus for magnetically processing a specimen that couples high field strength magnetic fields with the magnetocaloric effect includes a high field strength magnet capable of generating a magnetic field of at least 1 Tesla and a magnetocaloric insert disposed within a bore of the high field strength magnet. A method for magnetically processing a specimen includes positioning a specimen adjacent to a magnetocaloric insert within a bore of a magnet and applying a high field strength magnetic field of at least 1 Tesla to the specimen and to the magnetocaloric insert. The temperature of the specimen changes during the application of the high field strength magnetic field due to the magnetocaloric effect.

  17. Apparatus and method for magnetically processing a specimen

    Energy Technology Data Exchange (ETDEWEB)

    Ludtka, Gerard M; Ludtka, Gail M; Wilgen, John B; Kisner, Roger A; Jaramillo, Roger A

    2013-09-03

    An apparatus for magnetically processing a specimen that couples high field strength magnetic fields with the magnetocaloric effect includes a high field strength magnet capable of generating a magnetic field of at least 1 Tesla and a magnetocaloric insert disposed within a bore of the high field strength magnet. A method for magnetically processing a specimen includes positioning a specimen adjacent to a magnetocaloric insert within a bore of a magnet and applying a high field strength magnetic field of at least 1 Tesla to the specimen and to the magnetocaloric insert. The temperature of the specimen changes during the application of the high field strength magnetic field due to the magnetocaloric effect.

  18. Carbon loss during irradiation of T4 bacteriophages and E. coli bacteria in electron microscopes

    Energy Technology Data Exchange (ETDEWEB)

    Dubochet, J.

    1975-08-01

    The loss of /sup 14/C due to electron irradiation has been measured on labeled T4 bacteriophages and E. coli bacteria under conditions relevant for practical electron microscopy for fixed and scanning beam exposure. During irradiation, the remaining material became less and less sensitive to further carbon loss. Surface migration of molecular fragments and adsorbed molecules is involved in the process of beam damage. In CTEM under normal working conditions, the parameters on which the carbon loss depends cannot all be controlled. There is no perceptible carbon loss when the irradiation is made at liquid helium temperature. The material surrounding the biological object, the way in which the electron dose is given to the specimen, and the vacuum conditions in the specimen chamber influence the process. We conclude that secondary reactions of molecular fragments formed after inelastic scattering events are of importance and depend on the layer of molecules adsorbed onto the surface of the specimen. (auth)

  19. Correlation between locally deformed structure and oxide film properties in austenitic stainless steel irradiated with neutrons

    Science.gov (United States)

    Chimi, Yasuhiro; Kitsunai, Yuji; Kasahara, Shigeki; Chatani, Kazuhiro; Koshiishi, Masato; Nishiyama, Yutaka

    2016-07-01

    To elucidate the mechanism of irradiation-assisted stress corrosion cracking (IASCC) in high-temperature water for neutron-irradiated austenitic stainless steels (SSs), the locally deformed structures, the oxide films formed on the deformed areas, and their correlation were investigated. Tensile specimens made of irradiated 316L SSs were strained 0.1%-2% at room temperature or at 563 K, and the surface structures and crystal misorientation among grains were evaluated. The strained specimens were immersed in high-temperature water, and the microstructures of the oxide films on the locally deformed areas were observed. The appearance of visible step structures on the specimens' surface depended on the neutron dose and the applied strain. The surface oxides were observed to be prone to increase in thickness around grain boundaries (GBs) with increasing neutron dose and increasing local strain at the GBs. No penetrative oxidation was observed along GBs or along surface steps.

  20. Irradiation spectrum and ionization-induced diffusion effects in ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Zinkle, S.J. [Oak Ridge National Lab., TN (United States)

    1997-08-01

    There are two main components to the irradiation spectrum which need to be considered in radiation effects studies on nonmetals, namely the primary knock-on atom energy spectrum and ionizing radiation. The published low-temperature studies on Al{sub 2}O{sub 3} and MgO suggest that the defect production is nearly independent of the average primary knock-on atom energy, in sharp contrast to the situation for metals. On the other hand, ionizing radiation has been shown to exert a pronounced influence on the microstructural evolution of both semiconductors and insulators under certain conditions. Recent work on the microstructure of ion-irradiated ceramics is summarized, which provides evidence for significant ionization-induced diffusion. Polycrystalline samples of MgO, Al{sub 2}O{sub 3}, and MgAl{sub 2}O{sub 4} were irradiated with various ions ranging from 1 MeV H{sup +} to 4 MeV Zr{sup +} ions at temperatures between 25 and 650{degrees}C. Cross-section transmission electron microscopy was used to investigate the depth-dependent microstructural of the irradiated specimens. Dislocation loop nucleation was effectively suppressed in specimens irradiated with light ions, whereas the growth rate of dislocation loops was enhanced. The sensitivity to irradiation spectrum is attributed to ionization-induced diffusion. The interstitial migration energies in MgAl{sub 2}O{sub 4} and Al{sub 2}O{sub 3} are estimated to be {le}0.4 eV and {le}0.8 eV, respectively for irradiation conditions where ionization-induced diffusion effects are expected to be negligible.

  1. A general mixed mode fracture mechanics test specimen: The DCB-specimen loaded with uneven bending moments

    DEFF Research Database (Denmark)

    Sørensen, Bent F.; Jørgensen, K.; Jacobsen, T.K.;

    2004-01-01

    A mixed mode specimen is proposed for fracture mechanics characterisation of adhesive joints, laminates and multilayers. The specimen is a double cantilever beam specimen loaded with uneven bending moments at the two free beams. By varying the ratiobetween the two applied moments, the full mode...

  2. Irradiation Defects in Silicon Crystal

    Institute of Scientific and Technical Information of China (English)

    2003-01-01

    The application of irradiation in silicon crystal is introduced.The defects caused by irradiation are reviewed and some major ways of studying defects in irradiated silicon are summarized.Furthermore the problems in the investigation of irradiated silicon are discussed as well as its properties.

  3. Food irradiation; Napromieniowanie zywnosci

    Energy Technology Data Exchange (ETDEWEB)

    Migdal, W. [Instytut Chemii i Techniki Jadrowej, Doswiadczalna Stacja Radiacyjnego Utrwalania Plodow Rolnych, Warsaw (Poland)

    1995-12-31

    A worldwide standard on food irradiation was adopted in 1983 by codex Alimentarius Commission of the Joint Food Standard Programme of the Food and Agriculture Organization (FAO) of the United Nations and The World Health Organization (WHO). As a result, 41 countries have approved the use of irradiation for treating one or more food items and the number is increasing. Generally, irradiation is used to: food loses, food spoilage, disinfestation, safety and hygiene. The number of countries which use irradiation for processing food for commercial purposes has been increasing steadily from 19 in 1987 to 33 today. In the frames of the national programme on the application of irradiation for food preservation and hygienization an experimental plant for electron beam processing has been established in Inst. of Nuclear Chemistry and Technology. The plant is equipped with a small research accelerator Pilot (19 MeV, 1 kW) and industrial unit Electronika (10 MeV, 10 kW). On the basis of the research there were performed at different scientific institutions in Poland, health authorities have issued permissions for irradiation for; spices, garlic, onions, mushrooms, potatoes, dry mushrooms and vegetables. (author) 14 refs, 3 tabs

  4. Total lymphoid irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Sutherland, D.E.; Ferguson, R.M.; Simmons, R.L.; Kim, T.H.; Slavin, S.; Najarian, J.S.

    1983-05-01

    Total lymphoid irradiation by itself can produce sufficient immunosuppression to prolong the survival of a variety of organ allografts in experimental animals. The degree of prolongation is dose-dependent and is limited by the toxicity that occurs with higher doses. Total lymphoid irradiation is more effective before transplantation than after, but when used after transplantation can be combined with pharmacologic immunosuppression to achieve a positive effect. In some animal models, total lymphoid irradiation induces an environment in which fully allogeneic bone marrow will engraft and induce permanent chimerism in the recipients who are then tolerant to organ allografts from the donor strain. If total lymphoid irradiation is ever to have clinical applicability on a large scale, it would seem that it would have to be under circumstances in which tolerance can be induced. However, in some animal models graft-versus-host disease occurs following bone marrow transplantation, and methods to obviate its occurrence probably will be needed if this approach is to be applied clinically. In recent years, patient and graft survival rates in renal allograft recipients treated with conventional immunosuppression have improved considerably, and thus the impetus to utilize total lymphoid irradiation for its immunosuppressive effect alone is less compelling. The future of total lymphoid irradiation probably lies in devising protocols in which maintenance immunosuppression can be eliminated, or nearly eliminated, altogether. Such protocols are effective in rodents. Whether they can be applied to clinical transplantation remains to be seen.

  5. Effect of irradiation on the dental pulp tissues in streptozotocin-induced diabetic rats

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Ho Duk; Hwang, Eui Hwan; Lee, Sang Rae [Kyunghee University College of Medicine, Seoul (Korea, Republic of)

    2005-03-15

    To observe the histological changes in the pulp tissues of mandibular molars in streptozotocin-induced diabetic rats after irradiation. The male Sprague-Dawley rats weighing approximately 250 gm were divided into four groups : control, diabetes, irradiation, and diabetes-irradiation groups. Diabetes mellitus was induced in the rats by injecting streptozotocin. Rats in control and irradiation groups were injected with citrate buffer only. After 5 days, the head and neck region of the rats in irradiation and diabetes-irradiation groups were irradiated with a single absorbed dose of 10 Gy. All the rats were sacrificed at 3, 7, 14, 21, and 28 days after irradiation. The specimen including the mandibular molars were sectioned and observed using a histopathological method. In the diabetes group, capillary dilatation was observed. However, there was no obvious morphologic alteration of the odontoblasts. In the irradiation group, generalized necrosis of the dental pulp tissues was observed. Vacuolation of the odontoblasts and dilatation of the capillaries were noted in the early experimental phases. In the diabetes-irradiation group, generalized degeneration of the dental pulp tissues was observed. Vacuolation of the dental pulp cells and the odontoblasts was noted in the late experimental phases. This experiment suggest that dilatation of the capillaries in the dental pulp tissue is induced by diabetic state, and generalized degeneration of the dental pulp tissues is induced by irradiation of the diabetic group.

  6. The effect of oxygen on void stability in ion-irradiated steel

    Science.gov (United States)

    Seitzman, Larry E.; Dodd, R. Arthur; Kulcinski, Gerald L.

    1990-07-01

    The effect of oxygen on void stability in an Fe-17Ni-13Cr austenitic ternary alloy has been investigated using 15 MeV nickel-ion irradiation at elevated temperatures and preimplantation of 6 MeV oxygen at room temperature. The nickel irradiation was performed over a temperature range of 550 °C to 650 °C. Utilizing transverse specimen preparation techniques, the irradiated steel was examined by transmission electron microscopy (TEM). As little as 10 appm preimplanted oxygen caused a significant increase in the void number density when the steel was irradiated at 550 °C. A near-surface void-denuded zone occurs in the irradiated steel, while a region depleted of visible voids also occurs in the steel injected with 300 appm oxygen or greater and irradiated at 550 °C.

  7. The microstructure of neutron irradiated type-348 stainless steel and its relation to creep and hardening

    Science.gov (United States)

    Thomas, L. E.; Beeston, J. M.

    1982-06-01

    Annealed type-348 stainless steel specimens irradiated to 33 to 39 dpa at 350°C were examined by transmission electron microscopy to determine the cause of pronounced irradiation creep and hardening. The irradiation produced very high densities of 1-2 nm diameter helium bubbles, 2-20 nm diameter faulted (Frank) dislocation loops and 10 nm diameter precipitate particles. These defects account for the observed irradiation hardening but do not explain the creep strains. Too few point defects survive as faulted dislocation loops for significant creep by the stress-induced preferential absorption (SIPA) mechanism and there are not enough unfaulted dislocations for creep by climb-induced glide. Also, the irradiation-induced precipitates are face-centred cubic G-phase (a niobium nickel suicide), and cannot cause creep. It is suggested that the irradiation creep occurs by a grain-boundary movement mechanism such as diffusion accomodated grain-boundary sliding.

  8. Comparison of defect cluster accumulation and pattern formation in irradiated copper and nickel

    Energy Technology Data Exchange (ETDEWEB)

    Zinkle, S.J.; Snead, L.L. [Oak Ridge National Lab., TN (United States); Edwards, D.J. [Pacific Northwest Lab., Richland, WA (United States)] [and others

    1995-04-01

    The objective of this study is to compare the contrasting behavior of defect cluster formation in neutron-irradiated copper and nickel specimens. Transmission electron microscopy was used to examine the density and spatial distribution of defect clusters produced in copper and nickel as the result of fission neutron irradiation to damage levels of 0.01 to 0.25 displacements per atom (dpa) at irradiation temperature between 50 and 230{degrees}C. A comparison with published results in the literature indicates that defect cluster wall formation occurs in nickel irradiated at 0.2 to 0.4 T{sub M} in a wide variety of irradiation spectra. Defect cluster wall formation apparently only occurs in copper during low temperature irradiation with electrons and light ions. These results are discussed in terms of the thermal spike model for energetic displacement cascades.

  9. Space environmental durability of spacecrafts materials using ion beam irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. H.; Kim, D. W.; Lee, S. M.; Lee, I. T.; Ok, J. G. [Seoul Nat. Univ., Seoil (Korea, Republic of)

    2006-04-15

    Simulation of space proton effects by ion beam irradiation : due to diverse energy spectrums and fluxes of space protons are distributed in space according to the altitude and location in orbits, hard to simulate simply on the ground. JPL-1991 solar proton event is chosen to simulate the specific proton model. Cyclotrons for radiological treatments are utilized as main facility which can accelerate protons with MeV energy and possible to simulate the fluxes. Specimens are prepared with ITO aluminized polyimide Kapton and VDA Mylar. Mechanical, chemical changes analyses, and visual analysis of crystalline change : for assessment of mechanical properties of irradiated specimens, 50N micro tensile system is used for the ultimate tensile strength and elongation. Additional ESPI equipment can measure the elongation rate, yield strength, and elastic modulus. XPS is used for strength change from the molecular binding energy in crystal. SEM is also used for morphological visula analysis.

  10. Adequacy of urine cytology specimens: an assessment of collection techniques.

    Science.gov (United States)

    Hundley, Andrew F; Maygarden, Susan; Wu, Jennifer M; Visco, Anthony G; Connolly, AnnaMarie

    2007-09-01

    The objective of this study was to determine whether the method of urine collection impacts the adequacy and cell counts of cytology specimens in a low-risk population. Voided, post-cystometrogram (CMG), and bladder irrigant specimens were collected and evaluated for cytologic adequacy and average cell count by a single cytopathologist masked to the source of each sample. Data were analyzed to detect differences in specimen adequacy and cell counts based on method of collection. Both the voided and post-CMG specimens (97.3%, 93.7% respectively) were significantly more likely to be adequate compared to the bladder irrigant specimen (11.7%, p urine dipstick (p = 0.03). No cytologic abnormalities were diagnosed. Whereas both spontaneously voided and post-CMG specimens were consistently adequate for interpretation, spontaneous voided specimens were optimal with regard to maximizing cell count/hpf.

  11. Evaluation of hybrid composite materials in cylindrical specimen geometries

    Science.gov (United States)

    Liber, T.; Daniel, I. M.

    1976-01-01

    Static and fatigue properties of three composite materials and hybrids were examined. The materials investigated were graphite/epoxy, S-glass/epoxy, PRD-49 (Kevlar 49)/epoxy, and hybrids in angle-ply configurations. A new type of edgeless cylindrical specimen was developed. It is a flattened tube with two flat sides connected by curved sections and it is handled much like the standard flat coupon. Special specimen fabrication, tabbing, and tab region reinforcing techniques were developed. Axial modulus, Poisson's ratio, strength, and ultimate strain were obtained under static loading from flattened tube specimens of nine laminate configurations. In the case of graphite/epoxy the tubular specimens appeared to yield somewhat higher strength and ultimate strain values than flat specimens. Tensile fatigue tests were conducted with all nine types of specimens and S-N curves obtained. Specimens surviving 10 million cycles of tensile loading were subsequently tested statically to failure to determine residual properties.

  12. Recording and submitting specimen history data

    Science.gov (United States)

    Bodenstein, Barbara L.; Franson, J. Christian; Friend, Milton; Gibbs, Samantha E.J.; Wild, Margaret A.

    2016-06-14

    SummaryIn wildlife disease investigations, determining the history or background of a problem is the first significant step toward establishing a diagnosis and aiding agencies with management considerations. The diagnostic process and overall investigation is often greatly expedited by a chronological record accompanying specimens submitted for laboratory evaluation. Knowing where and when the outbreak is taking place, what the environmental conditions and species involved are, and clinical signs in sick animals, along with necropsy findings and diagnostic test results are important for understanding the natural history or epizootiology of disease outbreaks. It becomes increasingly difficult to retrospectively obtain all of the pertinent history as time passes. The most helpful information is that which is obtained at the time of the die-off event by perceptive field biologists and other observers. Significant events preceding morbidity and/or mortality also provide valuable information on which to base corrective actions. In this chapter, readers will find information regarding what type of information should be recorded, how it should be recorded and why it is relevant to a disease investigation. A thoughtful approach in providing as much information as possible surrounding the situation including about host species and the biotic and abiotic environment, greatly aids in determining the most likely causative agent(s).

  13. Results of irradiating aluminum and homogeneous alloy YMn2 by 23 MeV γ-quanta in a molecular deuterium atmosphere at 2 kbar pressure

    Science.gov (United States)

    Didyk, A. Yu.; Wisniewski, R.

    2014-03-01

    Specimens of a number of metal were placed successively along the length in a deuterium high-pressure chamber of the "finger type" (DHPC-FT). The specimens were: two aluminum rods, a copper rod, two YMn2 alloy specimens, and stainless steel. The molecular deuterium pressure in the DHPC-FT chamber was 2 kbar. The specimens were irradiated by braking γ-quanta with boundary energy 23 MeV. After irradiation, all specimens were investigated on scanning electron microscopes (SEM) with electron probe X-ray microelement analysis (XMA). Considerable changes in the structure of the surfaces and elemental composition of the measured aluminum, destruction of the homogeneous YMn2 alloy specimen, and the "formation of monocrystalline specimens" of the YMn2 type and structures resembling manganese-based "crystals" were observed. A phenomenological explanation of the observed phenomena and effects based on nuclear reactions is proposed with consideration of certain new approaches, which are examined.

  14. Final report on graphite irradiation test OG-3. [Fast neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Price, R.J.; Beavan, L.A.

    1977-01-01

    The results of dimensional, thermal expansivity, thermal conductivity, Young's modulus, and tensile strength measurements on graphite specimens irradiated in capsule OG-3 are presented. The graphite grades investigated included near-isotropic H-451 (three different preproduction lots), TS-1240, and SO818; needle coke H-327; and European coal tar pitch coke grades P/sub 3/JHA/sub 2/N, P/sub 3/JHAN, and ASI2-500. Data were obtained in the temperature range 823/sup 0/K to 1673/sup 0/K. The peak fast neutron fluence in the experiment was 3 x 10/sup 25/ n/m/sup 3/ (E greater than 29 fJ)/sub HTGR/; the total accumulated fluence exceeded 9 x 10/sup 25/ n/m/sup 2/ on some H-451 specimens and 6 x 10/sup 25/ n/m/sup 2/ on some TS-1240 specimens. Irradiation-induced dimensional changes on H-451 graphite differed slightly from earlier predictions. For an irradiation temperature of about 1225/sup 0/K, axial shrinkage rates at high fluences were somewhat higher than predicted, and the fluence at which radial expansion started (about 9 x 10/sup 25/ n/m/sup 2/ at 1275/sup 0/K) was lower. TS-1240 graphite underwent smaller dimensional changes than H-451 graphite, while limited data on SO818 and ASI2-500 graphites showed similar behavior to H-451. P/sub 3/JHAN and P/sub 3/JHA/sub 2/N graphites displayed anisotropic behavior with rapid axial shrinkage. Comparison of dimensional changes between specimens from three logs of H-451 and of TS-1240 graphites showed no significant log-to-log variations for H-451, and small but significant log-to-log variations for TS-1240. The thermal expansivity of the near-isotropic graphites irradiated at 865-1045/sup 0/K first increased by 5 percent to 10 percent and then decreased. At higher irradiation temperatures the thermal expansivity decreased by up to 50 percent. Changes in thermal conductivity were consistent with previously established curves. Specimens which were successively irradiated at two different temperatures took on the

  15. Mechanical properties of neutron-irradiated nickel-containing martensitic steels: II. Review and analysis of helium-effects studies

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L. [Oak Ridge National Laboratory, Metals and Ceramics Division, Building 4500S, P.O. Box 2008, MS 6151, Oak Ridge, TN 37831-6151 (United States)]. E-mail: kluehrl@ornl.gov; Hashimoto, N. [Oak Ridge National Laboratory, Metals and Ceramics Division, Building 4500S, P.O. Box 2008, MS 6151, Oak Ridge, TN 37831-6151 (United States); Sokolov, M.A. [Oak Ridge National Laboratory, Metals and Ceramics Division, Building 4500S, P.O. Box 2008, MS 6151, Oak Ridge, TN 37831-6151 (United States); Maziasz, P.J. [Oak Ridge National Laboratory, Metals and Ceramics Division, Building 4500S, P.O. Box 2008, MS 6151, Oak Ridge, TN 37831-6151 (United States); Shiba, K. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Tokai, Ibaraki 319-1195 (Japan); Jitsukawa, S. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Tokai, Ibaraki 319-1195 (Japan)

    2006-10-15

    In part I of this helium-effects study on ferritic/martensitic steels, results were presented on tensile and Charpy impact properties of 9Cr-1MoVNb (modified 9Cr-1Mo) and 12Cr-1MoVW (Sandvik HT9) steels and these steels containing 2% Ni after irradiation in the High Flux Isotope Reactor (HFIR) to 10-12 dpa at 300 and 400 deg. C and in the Fast Flux Test Facility (FFTF) to 15 dpa at 393 deg. C. The results indicated that helium caused an increment of hardening above irradiation hardening produced in the absence of helium. In addition to helium-effects studies on ferritic/martensitic steels using nickel doping, studies have also been conducted over the years using boron doping, ion implantation, and spallation neutron sources. In these previous investigations, observations of hardening and embrittlement were made that were attributed to helium. In this paper, the new results and those from previous helium-effects studies are reviewed and analyzed.

  16. Characterization of corrosion layers on irradiated and non-irradiated surfaces in BWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kysela, J.; Balek, V.; Zmitko, M.; Brozova, A.; Burda, J. [Nuclear Research Inst., Rez (Czech Republic); Hoffmann, H.; Ruehle, W. [VGB Essen (Germany); Bezdicka, P. [Institute of Inorganic Chemistry, ASCR, Rez (Czech Republic)

    2002-07-01

    Stress corrosion cracking of low-alloyed steel 22NiMoCr37 is evaluated with the goal to determine crack growth rate in irradiated steel under conditions simulating closely conditions of BWR RPV under operation. For the experiment, in pile BWR experimental loop has been built at Nuclear Research Institute, Rez. During the experiment, specimens are loaded by cyclic and constant load. Crack growth is monitored by means of potential drop measurement and COD. Corrosion layers formed on specimens in reactor water loop exposed to BWR primary water chemistry and radiation were studied. Two sets of specimens were placed in loop channels. One set of specimens was situated in reactor conditions and the second set out of reactor, other parameters like water chemistry (e.g. concentration of hydrogen, oxygen and conductivity), temperature and flow rate were identical. By means of this an effect of radiation could be studied. The differences in chemical composition, structure and microstructure of corrosion products were characterized by SEM and X-ray powder diffractometry. The differences in microstructure of corrosion layer formed under different conditions were observed. (authors)

  17. Tritium and helium retention and release from irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Anderl, R.A.; Longhurst, G.R.; Oates, M.A.; Pawelko, R.J. [Idaho National Engineering and Environmental Lab., Idaho Falls, ID (United States)

    1998-01-01

    This paper reports the results of an experimental effort to anneal irradiated beryllium specimens and characterize them for steam-chemical reactivity experiments. Fully-dense, consolidated powder metallurgy Be cylinders, irradiated in the EBR-II to a fast neutron (>0.1 MeV) fluence of {approx}6 x 10{sup 22} n/cm{sup 2}, were annealed at temperatures from 450degC to 1200degC. The releases of tritium and helium were measured during the heat-up phase and during the high-temperature anneals. These experiments revealed that, at 600degC and below, there was insignificant gas release. Tritium release at 700degC exhibited a delayed increase in the release rate, while the specimen was at 700degC. For anneal temperatures of 800degC and higher, tritium and helium release was concurrent and the release behavior was characterized by gas-burst peaks. Essentially all of the tritium and helium was released at temperatures of 1000degC and higher, whereas about 1/10 of the tritium was released during the anneals at 700degC and 800degC. Measurements were made to determine the bulk density, porosity and specific surface area for each specimen before and after annealing. These measurements indicated that annealing caused the irradiated Be to swell, by as much as 14% at 700degC and 56% at 1200degC. Kr gas adsorption measurements for samples annealed at 1000degC and 1200degC determined specific surface areas between 0.04 m{sup 2}/g and 0.1 m{sup 2}/g for these annealed specimens. The tritium and helium gas release measurements and the specific surface area measurements indicated that annealing of irradiated Be caused a porosity network to evolve and become surface-connected to relieve internal gas pressure. (author)

  18. Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures: Summary report

    Energy Technology Data Exchange (ETDEWEB)

    Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F. [Oak Ridge National Lab., TN (United States)] [and others

    1996-04-01

    Disk compact specimens of candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at nominal irradiation temperatures of either 90 250{degrees}C. These specimens have been tested over a temperature range from 20 to 250{degrees}C to determine J-integral values and tearing moduli. The results show that irradiation at these temperatures reduces the fracture toughness of austenic stainless steels, but the toughness remains quite high. The toughness decreases as the temperature increases. Irradiation at 250{degrees}C is more damaging that at 90{degrees}C, causing larger decreases in the fracture toughness. The ferritic-martensitic steels HT-9 and F82H show significantly greater reductions in fracture toughness that the austenitic stainless steels.

  19. Application of the Master Curve approach for the irradiation embrittlement evaluation of pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, H.W.; Boehmert, J. [Forschungszentrum Rossendorf e.V., Inst. fuer Sicherheitsforschung, Dresden (Germany)

    2003-09-01

    The master curve (MC) approach and the associated reference temperature, T{sub 0}, as defined in the test standard ASTM E1921, is rapidly moving from the research laboratory to application in integrity assessment of components and structures. T{sub 0} is the index temperature for the universal MC, which considers the toughness behaviour of a specific material. ''The Structural Integrity Assessment Procedures for European Industry'' (SINTAP) contain a MC extension for analysing the fracture behaviour of inhomogeneous ferritic steels. This paper presents the application of the MC approach to the T{sub 0} determination of different types of Russian WWER-type reactor pressure vessel (RPV) steels. In addition the SINTAP-MC approach was applied to determine an alternative reference temperature, T{sub R}. The influence of different microstructures and compositions within one type of RPV steel and the effect of irradiation with fast neutrons on T{sub 0} are experimentally evaluated. In general the MC based T{sub 0} is about 72 K below the Charpy V-notch transition temperature related to an impact energy of 48 J. The paper demonstrates the application of MC based T{sub 0} and T{sub R} as an alternative reference temperature for neutron embrittled RPV steels used in the RPV integrity assessment. (orig.)

  20. Irradiation and food processing.

    Science.gov (United States)

    Sigurbjörnsson, B; Loaharanu, P

    1989-01-01

    After more than four decades of research and development, food irradiation has been demonstrated to be safe, effective and versatile as a process of food preservation, decontamination or disinfection. Its various applications cover: inhibition of sprouting of root crops; insect disinfestation of stored products, fresh and dried food; shelf-life extension of fresh fruits, vegetables, meat and fish; destruction of parasites and pathogenic micro-organisms in food of animal origin; decontamination of spices and food ingredients, etc. Such applications provide consumers with the increase in variety, volume and value of food. Although regulations on food irradiation in different countries are largely unharmonized, national authorities have shown increasing recognition and acceptance of this technology based on the Codex Standard for Irradiated Foods and its associated Code of Practice. Harmonization of national legislations represents an important prerequisite to international trade in irradiated food. Consumers at large are still not aware of the safety and benefits that food irradiation has to offer. Thus, national and international organizations, food industry, trade associations and consumer unions have important roles to play in introducing this technology based on its scientific values. Public acceptance of food irradiation may be slow at the beginning, but should increase at a faster rate in the foreseeable future when consumers are well informed of the safety and benefits of this technology in comparison with existing ones. Commercial applications of food irradiation has already started in 18 countries at present. The volume of food or ingredients treated on a commercial scale varies from country to country ranging from several tons of spices to hundreds of thousands of tons of grains per annum. With the increasing interest of national authorities and the food industry in applying the process, it is anticipated that some 25 countries will use some 55 commercial

  1. Characterization of BOR-60 Irradiated 14YWT-NFA1 Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Saleh, Tarik A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aydogan, Eda [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Quintana, Matthew Estevan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Romero, Tobias J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-15

    Tubes of FCRD 14YWT-NFA1 Alloy were placed in the BOR-60 reactor and irradiated under a fast flux neutron environment to two conditions: 7 dpa at 360-370 °C and 6 dpa at 385-430 °C. Small sections of the tube were cut and sent to UC Berkeley for nanohardness testing and focused ion beam (FIB) milling of TEM specimens. FIB specimens were sent back to LANL for final FIB milling and TEM imaging. Hardness data and TEM images are presented in this report. This is the first fast reactor neutron irradiated information on the 14YWT-NFA1 alloy.

  2. Status of ATR-A1 irradiation experiment on vanadium alloys and low-activation steels

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

    The ATR-A1 irradiation experiment in the Advanced Test Reactor (ATR) was a collaborative U.S./Japan effort to study at low temperatures the effects of neutron damage on vanadium alloys. The experiment also contained a limited quantity of low-activation ferritic steel specimens from Japan as part of the collaboration agreement. The irradiation was completed on May 5, 1996, as planned, after achieving an estimated neutron damage of 4.7 dpa in vanadium. The capsule has since been kept in the ATR water canal for the required radioactivity cool-down. Planning is underway for disassembly of the capsule and test specimen retrieval.

  3. Helium release from neutron-irradiated Li{sub 2}O single crystals

    Energy Technology Data Exchange (ETDEWEB)

    Yamaki, Daiju; Tanifuji, Takaaki; Noda, Kenji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Helium release behavior in post-irradiation heating tests was investigated for Li{sub 2}O single crystals which had been irradiated with thermal neutrons in JRR-4 and JRR-2, and fast neutrons in FFTF. It is clarified that the helium release curves from JRR-4 and JRR-2 specimens consists of only one broad peak. From the dependence of the peak temperatures on the neutron fluence and the crystal diameter, and the comparison with the results obtained for sintered pellets, it is considered that the helium generated in the specimen is released through the process of bulk diffusion with trapping by irradiation defects such as some defect clusters. For the helium release from FFTF specimens, two broad peaks were observed in the release curves. It is considered to suggest that two different diffusion paths exist for helium migration in the specimen, that is, bulk diffusion and diffusion through the micro-crack due to the heavy irradiation. In addition, helium bubble formation after irradiation due to the high temperature over 800K is suggested. (J.P.N.)

  4. 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.

  5. Preliminary report on the irradiation conditions of the HFIR JP-23 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Ermi, A.M. [Westinghouse Hanford Company, Richland, WA (United States); Gelles, D.S. [Pacific Northwest Laboratory, Richland, WA (United States)

    1995-04-01

    The objective of this effort was to irradiate a series of alloys over the temperature range 300 to 600{degrees}C to approximately 10 dpa in the High Flux Isotope Reactor (HFIR). The alloys covered a wide range of materials and treatments. The Japanese specimen matrix consisted of ferritic steels, vanadium alloys, copper alloys, molybdenum alloys, and titanium-aluminum compounds. The US specimen matrix consisted of vanadium alloys, 316 stainless steels, and isotopically tailored ferritic and austenitic alloys.

  6. Sub ablative Er: YAG laser irradiation on surface roughness of eroded dental enamel.

    Science.gov (United States)

    Curylofo-Zotti, Fabiana Almeida; Lepri, Taísa Penazzo; Colucci, Vivian; Turssi, Cecília Pedroso; Corona, Silmara Aparecida Milori

    2015-11-01

    This study evaluated the effects of Er:YAG laser irradiation applied at varying pulse repetition rate on the surface roughness of eroded enamel. Bovine enamel slabs (n = 10) were embedded in polyester resin, ground, and polished. To erosive challenges, specimens were immersed two times per day in 20mL of concentrated orange juice (pH = 3.84) under agitation, during a two-day period. Specimens were randomly assigned to irradiation with the Er:YAG laser (focused mode, pulse energy of 60 mJ and energy density of 3.79 J/cm(2) ) operating at 1, 2, 3, or 4 Hz. The control group was left nonirradiated. Surface roughness measurements were recorded post erosion-like formation and further erosive episodes by a profilometer and observed through atomic force microscopy (AFM). Analysis of variance revealed that the control group showed the lowest surface roughness, while laser-irradiated substrates did not differ from each other following post erosion-like lesion formation. According to analysis of covariance, at further erosive episodes, the control group demonstrated lower surface roughness (P > 0.05), than any of the irradiated groups (P dental enamel eroded. The AFM images showed that the specimens irradiated by the Er:YAG laser at 1 Hz presented a less rough surface than those irradiated at 2, 3, and 4 Hz. © 2015 Wiley Periodicals, Inc.

  7. The Birmingham Irradiation Facility

    Science.gov (United States)

    Dervan, P.; French, R.; Hodgson, P.; Marin-Reyes, H.; Wilson, J.

    2013-12-01

    At the end of 2012 the proton irradiation facility at the CERN PS [1] will shut down for two years. With this in mind, we have been developing a new ATLAS scanning facility at the University of Birmingham Medical Physics cyclotron. With proton beams of energy approximately 30 MeV, fluences corresponding to those of the upgraded Large Hadron Collider (HL-LHC) can be reached conveniently. The facility can be used to irradiate silicon sensors, optical components and mechanical structures (e.g. carbon fibre sandwiches) for the LHC upgrade programme. Irradiations of silicon sensors can be carried out in a temperature controlled cold box that can be scanned through the beam. The facility is described in detail along with the first tests carried out with mini (1×1 cm2) silicon sensors.

  8. The Birmingham Irradiation Facility

    CERN Document Server

    Dervan, P; Hodgson, P; Marin-Reyes, H; Wilson, J

    2013-01-01

    At the end of 2012 the proton irradiation facility at the CERN PS [1] will shut down for two years. With this in mind, we have been developing a new ATLAS scanning facility at the University of Birmingham Medical Physics cyclotron. With proton beams of energy approximately 30 MeV, fluences corresponding to those of the upgraded Large Hadron Collider (HL-LHC) can be reached conveniently. The facility can be used to irradiate silicon sensors, optical components and mechanical structures (e.g. carbon fibre sandwiches) for the LHC upgrade programme. Irradiations of silicon sensors can be carried out in a temperature controlled cold box that can be scanned through the beam. The facility is described in detail along with the first tests carried out with mini (1 x 1 cm^2 ) silicon sensors.

  9. Irradiation of food

    Energy Technology Data Exchange (ETDEWEB)

    MacGregor, J.; Stanbrook, I.; Shersby, M.

    1989-07-12

    The House of Commons was asked to support the Government's intention to allow the use of the irradiation of foodstuffs under conditions that will fully safeguard the interests of the consumer. The Government, it was stated, regards this process as a useful additional way to ensure food safety. The effect of the radiation in killing bacteria will enhance safety standards in poultry meat, in some shell-fish and in herbs and spices. The problem of informing the public when the food has been irradiated, especially as there is no test to detect the irradiation, was raised. The subject was debated for an hour and a half and is reported verbatim. The main point raised was over whether the method gave safer food as not all bacteria were killed in the process. The motion was carried. (U.K.).

  10. Fracture mechanics characterisation of medium-size adhesive joint specimens

    DEFF Research Database (Denmark)

    Sørensen, Bent F.; Jacobsen, T.K.

    2004-01-01

    Medium-size specimens (adhesive layer were tested in four point bending to determine their load carrying capacity. Specimens having different thickness were tested. Except for onespecimen, the cracking occurred as cracking...... along the adhesive layer; initially cracking occurred along the adhesive/laminate interface, but after some crack extension the cracking took place inside the laminate (for one specimen the later part of thecracking occurred unstably along the adhesive/ laminate interface). Crack bridging by fibres...

  11. Direct observation of unstained biological specimens in water by the frequency transmission electric-field method using SEM.

    Directory of Open Access Journals (Sweden)

    Toshihiko Ogura

    Full Text Available Scanning electron microscopy (SEM is a powerful tool for the direct visualization of biological specimens at nanometre-scale resolution. However, images of unstained specimens in water using an atmospheric holder exhibit very poor contrast and heavy radiation damage. Here, we present a new form of microscopy, the frequency transmission electric-field (FTE method using SEM, that offers low radiation damage and high-contrast observation of unstained biological samples in water. The wet biological specimens are enclosed in two silicon nitride (SiN films. The metal-coated SiN film is irradiated using a focused modulation electron beam (EB at a low-accelerating voltage. A measurement terminal under the sample holder detects the electric-field frequency signal, which contains structural information relating to the biological specimens. Our results in very little radiation damage to the sample, and the observation image is similar to the transmission image, depending on the sample volume. Our developed method can easily be utilized for the observation of various biological specimens in water.

  12. Design Analysis of the Mixed Mode Bending Sandwich Specimen

    DEFF Research Database (Denmark)

    Quispitupa, Amilcar; Berggreen, Christian; Carlsson, Leif A.

    2010-01-01

    A design analysis of the mixed mode bending (MMB) sandwich specimen for face–core interface fracture characterization is presented. An analysis of the competing failure modes in the foam cored sandwich specimens is performed in order to achieve face–core debond fracture prior to other failure modes....... The analysis facilitates selection of the appropriate geometry for the MMB sandwich specimen to promote debond failure. An experimental study is performed using MMB sandwich specimens with a H100 PVC foam core and E-glass–polyester faces. The results reveal that debond propagation is successfully achieved...... for the chosen geometries and mixed mode loading conditions....

  13. A Debonded Sandwich Specimen Under Mixed Mode Bending (MMB)

    DEFF Research Database (Denmark)

    Quispitupa, Amilcar; Berggreen, Christian; Carlsson, Leif A.

    2008-01-01

    Face/core interface crack propagation in sandwich specimens is analyzed. A thorough analysis of the typical failure modes in sandwich composites was performed in order to design the MMB specimen to promote face/core debond fracture. Displacement, compliance and energy release rate expressions...... for the MMB specimen were derived from a superposition analysis. An experimental verification of the methodology proposed was performed using MMB sandwich specimens with H100 PVC foam core and E-glass/polyester non-crimp quadro-axial [0/45/90/-45]s DBLT-850 faces. Different mixed mode loadings were applied...

  14. A Debonded Sandwich Specimen Under Mixed Mode Bending (MMB)

    DEFF Research Database (Denmark)

    Quispitupa, Amilcar; Berggreen, Christian; Carlsson, Leif A.

    2008-01-01

    Face/core interface crack propagation in sandwich specimens is analyzed. A thorough analysis of the typical failure modes in sandwich composites was performed in order to design the MMB specimen to promote face/core debond fracture. Displacement, compliance and energy release rate expressions...... for the MMB specimen were derived from a superposition analysis. An experimental verification of the methodology proposed was performed using MMB sandwich specimens with H100 PVC foam core and E-glass/polyester non-crimp quadro-axial [0/45/90/-45]s DBLT-850 faces. Different mixed mode loadings were applied...

  15. High-frequency ultrasonic imaging of thickly sliced specimens

    Science.gov (United States)

    Miyasaka, Chiaki; Tittmann, Bernhard R.; Chandraratna, Premindra A. N.

    2003-07-01

    It has been reported that a mechanical scanning reflection acoustic microscope (hereinafter called simply "SAM"), using high frequency ultrasonic tone-burst waves, can form a horizontal cross-sectional image (i.e., c-scan image) showing a highly resolved cellular structure of biological tissue. However, the tissue prepared for the SAM has been mostly a thinly sectioned specimen. In this study, the SAM images of specimens thickly sectioned from the tissue were analyzed. Optical and scanning acoustic microscopies were used to evaluate tissues of human small intestine and esophagus. For preparing thin specimens, the tissue was embedded in paraffin, and substantially sectioned at 5-10μm by the microtome. For optical microscopy, the tissue was stained with hematoxylin and eosin, and affixed onto glass substrates. For scanning acoustic microscopy, two types of specimens were prepared: thinly sectioned specimens affixed on the glass substrate, wherein the specimens were deparaffinized in xylene, but not stained, and thickely sectioned specimens. Images of the thick specimens obtained with frequency at 200 MHz revealed cellular structures. The morphology was very similar to that seen in the thinly sectioned specimens with optical and scanning acoustic microscopy. In addition, scanning electron microscopy was used to compare the images of biological tissue. An acoustic lens with frequency at 200 MHz permitted the imaging of surface and/or subsurface of microstructures in the thick sections of small intestine and esophagus.

  16. Irradiation creep of 3C-SiC and microstructural understanding of the underlying mechanisms

    Science.gov (United States)

    Kondo, Sosuke; Koyanagi, Takaaki; Hinoki, Tatsuya

    2014-05-01

    Irradiation-induced creep in high-purity silicon carbide was studied by an ion-irradiation method under various irradiation conditions. The tensioned surfaces of bent thin specimens were irradiated with 5.1 MeV Si2+ ions up to 3 dpa at 280-1200 °C, which is referred to as a single-ion experiment. Additional He+ ions were irradiated simultaneously in the dual-ion experiment to study the effects of transmuted helium on irradiation creep. Irradiation creep was observed above 400 °C in the single-ion case, where a linear relationship between irradiation creep and swelling (C/S) was observed at 400-800 °C for all stress levels (150, 225, and 300 MPa). The proportional constant of the C/S relationship was strongly dependent on temperature and stress. A rapid reduction in creep strain was observed above 1000 °C. On the basis of the microstructural analysis, anisotropic distribution of self-interstitial atom (SIA) clusters was suspected to be the primary creep mechanism. Some interesting results were obtained from re-irradiation under stress after the irradiation without stress. The creep strain was significantly retarded by pre-irradiation to even 0.01 dpa at 400 and 600 °C. This implies that the loop orientation was determined very early in the irradiation regime. For the dual-ion cases, irradiation creep was absent or very limited at all irradiation temperatures studied (400-800 °C). Microstructural analysis indicated that helium inhibited the stable growth of SIA clusters and prevented them from exhibiting anisotropic distribution.

  17. Defects annihilation behavior of neutron-irradiated SiC ceramics densified by liquid-phase-assisted method after post-irradiation annealing

    Directory of Open Access Journals (Sweden)

    Mohd Idzat Idris

    2016-12-01

    Full Text Available Numerous studies on the recovery behavior of neutron-irradiated high-purity SiC have shown that most of the defects present in it are annihilated by post-irradiation annealing, if the neutron fluence is less than 1×1026 n/m2 (>0.1MeV and the irradiation is performed at temperatures lower than 973K. However, the recovery behavior of SiC fabricated by the nanoinfiltrated and transient eutectic phase (NITE process is not well understood. In this study, the effects of secondary phases on the irradiation-related swelling and recovery behavior of monolithic NITE-SiC after post-irradiation annealing were studied. The NITE-SiC specimens were irradiated in the BR2 reactor at fluences of up to 2.0–2.5×1024 n/m2 (E>0.1MeV at 333–363K. This resulted in the specimens swelling up ∼1.3%, which is 0.1% higher than the increase seen in concurrently irradiated high-purity SiC. The recovery behaviors of the specimens after post-irradiation thermal annealing were examined using a precision dilatometer; the specimens were heated at temperatures of up to 1673K using a step-heating method. The recovery curves were analyzed using a first-order model, and the rate constants for each annealing step were obtained to determine the activation energy for volume recovery. The NITE-A specimen (containing 12 wt% sintering additives recovered completely after annealing at ∼1573K; however, it shrank because of the volatilization of the oxide phases at 1673K. The NITE-B specimen (containing 18wt% sintering additives did not recover fully, since the secondary phase (YAG was crystallized during the annealing process. The recovery mechanism of NITE-A SiC was based on the recombination of the C and Si Frenkel pairs, which were very closely sited or only slightly separated at temperatures lower than 1223K, as well as the recombination of the slightly separated C Frenkel pairs and the migration of C and Si interstitials at temperatures of 1223–1573K. That is to say, the

  18. Irradiation of Polyimide and Neutron Poison Materials by Using a HANARO Capsule

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Cho, M. S.; Shin, Y. T.; Kim, B. G.; Seo, C. G.; Kim, Y. K. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    A material capsule system has been developed for an irradiation test of non-fissile materials in HANARO (High flux Advanced Neutron Application ReactOr).This capsule system has been actively utilized for the various material irradiation tests requested by users from research institutes, universities, and the industries. The capsules were mainly designed for an irradiation of the RPV (Reactor Pressure Vessel) and reactor core materials, and Zr-based alloys of parts of nuclear fuel assembly. Recently, irradiation tests of neutron poison materials and Polyimide were requested by Westinghouse Electric Company (WEC) and Hanyang University, respectively. As a candidate material of control rod of AP1000 reactor, Ag and Ag-In-Cd alloys were requested to be irradiated in HANARO by WEC. Polyimide has been studied as a shielding material against thermal and fast neutrons. The irradiation of these new materials which might affect the safety of a reactor was carried out for the first time in HANARO. As a preliminary test, small amount of these materials were determined to be inserted in a KNF (Korea Nuclear Fuel) irradiation capsule of 07M-13N. Due to the new materials, the irradiation test of the 07M-13N capsule was examined and approved by the 'HANARO Safety Review Committee'. The 07M-13N capsule was safely 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 of these new materials were irradiated up to a maximum fast neutron fluence of 1.13x1021(n/cm{sup 2}) (E>1.0MeV) and the dpa of the irradiated specimens were evaluated as 1.87.

  19. Convoluted dislocation loops induced by helium irradiation in reduced-activation martensitic steel and their impact on mechanical properties

    Energy Technology Data Exchange (ETDEWEB)

    Luo, Fengfeng [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory, School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Yao, Z. [Department of Mechanical and Materials Engineering, Queen' s University, Kingston, ON, Canada K7L 3N6 (Canada); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory, School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Suo, Jinping [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Wen, Yongming [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory, School of Physics and Technology, Wuhan University, Wuhan 430072 (China)

    2014-06-01

    Helium irradiation induced dislocation loops in reduced-activation martensitic steels were investigated using transmission electron microscopy. The specimens were irradiated with 100 keV helium ions to 0.8 dpa at 350 °C. Unexpectedly, very large dislocation loops were found, significantly larger than that induced by other types of irradiations under the same dose. Moreover, the large loops were convoluted and formed interesting flower-like shape. The large loops were determined as interstitial type. Loops with the Burgers vectors of b=〈100〉 were only observed. Furthermore, irradiation induced hardening caused by these large loops was observed using the nano-indentation technique.

  20. Status of ITER task T213 collaborative irradiation screening experiment on Cu/SS joints in the Russian Federation SM-2-reactor

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, D.J. [Pacific Northwest National Lab., Richland, WA (United States); Fabritsiev, S.A. [D.V. Efremov Inst., St. Petersburg (Russian Federation); Pokrovsky, A.S. [SRIAR, Dimitrovgrad (Russian Federation); Zinkle, S.J. [Oak Ridge National Lab., TN (United States)] [and others

    1996-04-01

    Specimen fabrication is underway for an irradiation screening experiment planned to start in January 1996 in the SM-2 reactor in Dimitrovgrad, Russia. The purpose of the experiment is to evaluate the effects of neutron irradiation at ITER-relevant temperatures on the bond integrity performance of Cu/SS and Be/Cu joints, as well as to further investigate the base metal properties of irradiated copper alloys. Specimens from each of the four ITER parties (U.S., EU, japan, and RF) will be irradiated to a dose of {approx}0.2 dpa at two different temperatures, 150 and 300{degrees}C. The specimens will consist of Cu/SS and Be/Cu joints in several different geometries, as well as a large number of specimens from the base materials. Fracture toughness data on base metal and Cu/SS bonded specimens will be obtained from specimens supplied by the U.S. Due to lack of material, the Be/Cu specimens supplied by the U.S will only be irradiated as TEM disks.

  1. Mechanical-property changes of polymeric and composite materials after low-temperature proton irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Snead, C.L. Jr.; Czajkowski, C.J.; Skaritka, J. [Brookhaven National Lab., Upton, NY (United States). Dept. of Advanced Technology; Morena, J. [Ace Inc., Stuart, FL (United States)

    1999-02-01

    The mechanical properties of polymeric and composite materials are known to be sensitive to ionizing radiation. Most of the existing data, however, is the result of near-room-temperature irradiations, most commonly with {sup 60}Co gamma irradiation. For use of these materials in applications such as for magnetic fusion magnets, where operation will be at cryogenic temperatures in sometimes severe radiation fields, knowledge of the materials` radiation response to low-temperature irradiations is required. This paper reports the results of mechanical-property-change measurements made at 4.2K on a number of potential magnet materials following 200-MeV-proton irradiation at temperatures below 20K. Standard three-point bend tests were performed at 4.2K for short-beam shear determinations in the laminate materials and for shear strength in the remainder of the specimens. Specimens were warmed to room temperature for one week prior to the mechanical testing in order to emulate the expected the expected mechanical state of the material assuming room-temperature cycling in the expected magnet applications. Data are presented in the form of yield stresses before and after irradiations with percentages of change. There were five specimens per test dose for each material. Data are presented for exposures ranging from nominally 10{sup 7} to 10{sup 9} rad. Results of the mechanical tests range from complete delamination and distortion of the specimens at 10{sup 9} rad to an increase in the yield stress of 63% after 10{sup 9} rad. The latter specimen did, however, evidence significant embrittlement. The phenomenon of irradiation-induced strengthening due to enhanced cross linking in undercured polymers was observed in some cases.

  2. Solar Irradiance Variability

    CERN Document Server

    Solanki, Sami K

    2012-01-01

    The Sun has long been considered a constant star, to the extent that its total irradiance was termed the solar constant. It required radiometers in space to detect the small variations in solar irradiance on timescales of the solar rotation and the solar cycle. A part of the difficulty is that there are no other constant natural daytime sources to which the Sun's brightness can be compared. The discovery of solar irradiance variability rekindled a long-running discussion on how strongly the Sun affects our climate. A non-negligible influence is suggested by correlation studies between solar variability and climate indicators. The mechanism for solar irradiance variations that fits the observations best is that magnetic features at the solar surface, i.e. sunspots, faculae and the magnetic network, are responsible for almost all variations (although on short timescales convection and p-mode oscillations also contribute). In spite of significant progress important questions are still open. Thus there is a debat...

  3. Cellular Response to Irradiation

    Institute of Scientific and Technical Information of China (English)

    LIU Bo; YAN Shi-Wei

    2011-01-01

    To explore the nonlinear activities of the cellular signaling system composed of one transcriptional arm and one protein-interaction arm, we use an irradiation-response module to study the dynamics of stochastic interactions.It is shown that the oscillatory behavior could be described in a unified way when the radiation-derived signal and noise are incorporated.

  4. Wholesomeness of irradiated food

    Energy Technology Data Exchange (ETDEWEB)

    Raica, Nicholas; McDowell, Marion E.; Darby, William J.

    1963-01-15

    The wholesomeness of irradiated foods was evaluated in mice, rats, dogs, and monkeys over a 2-year period, or 4 generations. Data are presented on the effects of a diet containing radiation-processed foods on growth, reproduction, hematology, histopathology, carcinogenicity, and life span. (86 references) (C.H.)

  5. Disassembly of the fusion-1 capsule after irradiation in the BOR-60 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H. [Argonne National Lab., IL (United States); Kazakov, V.A.; Chakin, V.P. [and others

    1997-04-01

    A U.S./Russia (RF) collaborative irradiation experiment, Fusion-1, was completed in June 1996 after reaching a peak exposure of {approx}17 dpa in the BOR-60 fast reactor at the Research Institute of Atomic Reactors (RIAR) in Russia. The specimens were vanadium alloys, mainly of recent heats from both countries. In this reporting period, the capsule was disassembled at the RIAR hot cells and all test specimens were successfully retrieved. For the disassembly, an innovative method of using a heated diffusion oil to melt and separate the lithium bond from the test specimens was adopted. This method proved highly successful.

  6. Post-irradiation experiments on physical thermal and microstructural properties of neutron-irradiated ceramics. 2

    Energy Technology Data Exchange (ETDEWEB)

    Yano, Toyohiko [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors

    1999-03-01

    Succeeding to the report on the post-irradiation experiments conducted in the previous year, this is a summary report on the post-irradiation experiments of physical, thermal and microstructural properties of neutron-irradiated various ceramics, which are expected to be applied to the in-core materials of an Advanced Fast Breeder Reactor in near future. Four candidate ceramics, Al{sub 2}O{sub 3}, AlN, SiC and Si{sub 3}N{sub 4} were fast-neutron-irradiated up to a fluence of 3.9x10{sup 26} n/m{sup 2}, different irradiation conditions from the previous report specimens, in the CMIR-4 rig in the JOYO experimental fast reactor in JNC. The following observations were performed: (1) Microstructural observation by means of transmission electron microscopy, (2) Measurement of swelling, (3) Measurement of thermal diffusivity by a laser-flash method, (4) Recovery of swelling by isochronal annealing, and (5) Recovery of thermal diffusivity by isochronal annealing. Obtained main results are summarized as follows. Macroscopic length changes by neutron irradiation of Al{sub 2}O{sub 3} and AlN were measured to be 1.8-2.0% and these of SiC and Si{sub 3}N{sub 4} to be 0.2-0.4%, respectively. Thermal diffusivities of all irradiated materials degraded to 0.03-0.05 cm{sup 2}/s, irrespective of materials which had large difference before irradiation. Microstructural observation of irradiated materials by TEM revealed that Al{sub 2}O{sub 3} contained high-density loops, microvoids in grains, and microcracking along grain boundaries, AlN contained high-density loops and microcracking along grain boundaries, SiC contained high-density loops, and Si{sub 3}N{sub 4} contained loops lying on the planes parallel to the c-axis, respectively. Macroscopic length of Al{sub 2}O{sub 3} and AlN started to recover at around 800deg or 1100degC, respectively, irrespective of irradiation temperature, and reduced quickly. Macroscopic length of SiC recovered gradually from near the irradiation temperature

  7. Failed PCR of Ganoderma type specimens affects nomenclature.

    Science.gov (United States)

    Paterson, R R M; Lima, N

    2015-06-01

    The nomenclature of Ganoderma used as a Chinese medicine is debated. A group of researchers could not amplify the DNA of type specimens and concluded the DNA was degraded irreparably. New topotypes were used as the type specimens which was premature. The use of internal amplification controls is recommended to determine if other factors were involved as alternative explanations.

  8. 16 CFR Figure 7 to Subpart A of... - Specimen Tray

    Science.gov (United States)

    2010-01-01

    ... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Specimen Tray 7 Figure 7 to Subpart A of Part 1209 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION CONSUMER PRODUCT SAFETY ACT... to Subpart A of Part 1209—Specimen Tray EC03OC91.037 ...

  9. 46 CFR 57.06-4 - Production testing specimen requirements.

    Science.gov (United States)

    2010-10-01

    ... plates three-fourths inch or less in thickness one reduced section tensile specimen and two free-bend... saw into as many portions of the thickness as necessary, as shown in Figure 57.06-4(f)(1)(ii) each of..., the specimen may be cut with a thin saw into as many portions of the thickness as necessary as shown...

  10. A cylindrical specimen holder for electron cryo-tomography

    Energy Technology Data Exchange (ETDEWEB)

    Palmer, Colin M., E-mail: cpalmer@mrc-lmb.cam.ac.uk; Löwe, Jan, E-mail: jyl@mrc-lmb.cam.ac.uk

    2014-02-01

    The use of slab-like flat specimens for electron cryo-tomography restricts the range of viewing angles that can be used. This leads to the “missing wedge” problem, which causes artefacts and anisotropic resolution in reconstructed tomograms. Cylindrical specimens provide a way to eliminate the problem, since they allow imaging from a full range of viewing angles around the tilt axis. Such specimens have been used before for tomography of radiation-insensitive samples at room temperature, but never for frozen-hydrated specimens. Here, we demonstrate the use of thin-walled carbon tubes as specimen holders, allowing the preparation of cylindrical frozen-hydrated samples of ribosomes, liposomes and whole bacterial cells. Images acquired from these cylinders have equal quality at all viewing angles, and the accessible tilt range is restricted only by the physical limits of the microscope. Tomographic reconstructions of these specimens demonstrate that the effects of the missing wedge are substantially reduced, and could be completely eliminated if a full tilt range was used. The overall quality of these tomograms is still lower than that obtained by existing methods, but improvements are likely in future. - Highlights: • The missing wedge is a serious problem for electron cryo-tomography. • Cylindrical specimens allow the missing wedge to be eliminated. • Carbon nanopipettes can be used as cylindrical holders for tomography of frozen-hydrated specimens. • Cryo-tomography of cylindrical biological samples demonstrates a reduction of deleterious effects associated with the missing wedge.

  11. Preparation of Articular Cartilage Specimens for Scanning Electron Microscopy.

    Science.gov (United States)

    Stupina, T A

    2016-08-01

    We developed and adapted a technology for preparation of articular cartilage specimens for scanning electron microscopy. The method includes prefixation processing, fixation, washing, and dehydration of articular cartilage specimens with subsequent treatment in camphene and air-drying. The technological result consists in prevention of deformation of the articular cartilage structures. The method is simpler and cheaper than the known technologies.

  12. The whereabouts of pre-nineteenth century bird specimens

    NARCIS (Netherlands)

    Steinheimer, F.D.

    2005-01-01

    The paper lists the whereabouts of surviving pre-nineteenth century bird collections containing altogether about 1500-3000 specimens. They are found in more than 50 institutions world-wide, with Berlin, Leiden, Paris, Stockholm, Tring and Vienna museums each holding more than 200 bird specimens from

  13. On a specimen of Lumbricus terrestris, L. with bifurcated tail

    NARCIS (Netherlands)

    Horst, R.

    1886-01-01

    In the last number of the »Annals and Magazine of Nat. History” (Dec. 1885), I find a notice of Prof. Jeffrey Bell about two Lumbrici with bifid hinder ends, one specimen belonging to L. terrestris, the other to L. foetidus; moreover he mentions a specimen, presenting a similar remarquable arrangeme

  14. The whereabouts of pre-nineteenth century bird specimens

    NARCIS (Netherlands)

    Steinheimer, F.D.

    2005-01-01

    The paper lists the whereabouts of surviving pre-nineteenth century bird collections containing altogether about 1500-3000 specimens. They are found in more than 50 institutions world-wide, with Berlin, Leiden, Paris, Stockholm, Tring and Vienna museums each holding more than 200 bird specimens from

  15. 40 CFR 792.51 - Specimen and data storage facilities.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 31 2010-07-01 2010-07-01 true Specimen and data storage facilities. 792.51 Section 792.51 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) TOXIC SUBSTANCES CONTROL ACT (CONTINUED) GOOD LABORATORY PRACTICE STANDARDS Facilities § 792.51 Specimen and data...

  16. 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.

  17. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  18. A Review of Graphite Irradiation Creep Data from the "OC-Series" of Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Davies, Mark A. [MARAD Co. Ltd., Washington, DC (United States); Burchell, Timothy D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2012-09-01

    The OC-Series graphite irradiation creep experiments were conducted in the early 1970s in the Oak Ridge Research Reactor (ORR) at ORNL. The OC Series consisted of 5 experiments, Capsules 1, 3 and 5 were irradiated at 900°C and Capsules 2 and 4 were irradiated at 600°C. Each capsule contained four columns of specimens, two loaded in compression and two un-loaded. The loaded columns had specimens of different diameter to generate two stress levels, 13.8 MPa and 20.7 MPa. Some of the data from these experiments were presented in extended abstracts at a Carbon Conference (Kennedy et al, 1977: Kennedy and Eatherly, 1979). The data presented some challenges to the accepted approach to irradiation induced creep in graphite adopted in the UK, specifically lateral creep strain behaviour and the effect of irradiation induced creep strain on material properties, e.g. CTE and Poisson’s Ratio. A recent review of irradiation induced creep (Davies & Bradford, 2004) included an anlaysis of the available OC-series data (Mobasheran, 1990) and led to a request to ORNL for an examination of the original OC-Series dataset. An initial search of the ORNL archive revealed additional data from the OC-Series experiment including previously unknown irradiation annealing experiments. This report presents a re-analysis of the available data from the OC-Series archive.

  19. A Review of Graphite Irradiation Creep Data from the "OC-Series" of Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Davies, Mark A. [MARAD Co. Ltd., Washington, DC (United States); Burchell, Timothy D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2012-09-01

    The OC-Series graphite irradiation creep experiments were conducted in the early 1970s in the Oak Ridge Research Reactor (ORR) at ORNL. The OC Series consisted of 5 experiments, Capsules 1, 3 and 5 were irradiated at 900°C and Capsules 2 and 4 were irradiated at 600°C. Each capsule contained four columns of specimens, two loaded in compression and two un-loaded. The loaded columns had specimens of different diameter to generate two stress levels, 13.8 MPa and 20.7 MPa. Some of the data from these experiments were presented in extended abstracts at a Carbon Conference (Kennedy et al, 1977: Kennedy and Eatherly, 1979). The data presented some challenges to the accepted approach to irradiation induced creep in graphite adopted in the UK, specifically lateral creep strain behaviour and the effect of irradiation induced creep strain on material properties, e.g. CTE and Poisson’s Ratio. A recent review of irradiation induced creep (Davies & Bradford, 2004) included an anlaysis of the available OC-series data (Mobasheran, 1990) and led to a request to ORNL for an examination of the original OC-Series dataset. An initial search of the ORNL archive revealed additional data from the OC-Series experiment including previously unknown irradiation annealing experiments. This report presents a re-analysis of the available data from the OC-Series archive.

  20. Embrittlement of molybdenum-rhenium welds under low and high temperature neutron irradiation

    Science.gov (United States)

    Krajnikov, A. V.; Morito, F.; Danylenko, M. I.

    2014-01-01

    The effect of low- and high-temperature neutron irradiation on the tensile strength, microhardness, and fracture mode has been studied for a series of Mo-Re welds with various Re concentrations. Radiation-induced hardening and concurrent ductility reduction are the key after-effects of neutron exposure. Low-temperature irradiation usually leads to a very hard embrittlement. The hardening effect is rather limited and unstable because of the lack of ductility. Irradiated specimens fail by brittle intergranular or transgranular fracture. The damaging effect of neutrons is less pronounced after high-temperature irradiation. The hardening of the matrix is rather high, but irradiated specimens still keep residual plasticity. High-temperature irradiation intensifies homogeneous nucleation of Re-rich phases, and this effect equalises the difference in mechanical properties between the different weld zones. A characteristic ductility loss exposure temperature was found to separate the temperature fields of absolutely brittle and relatively ductile behaviour. It usually varies between 850 K and 1000 K depending on the alloy composition and irradiation conditions.

  1. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  2. Effect of hydrogen sulfide emissions on cement mortar specimens

    Energy Technology Data Exchange (ETDEWEB)

    Idriss, A. F. [Alberta Environment, Science and Technology Branch, Edmonton, AB (Canada); Negi, S. C.; Jofriet, J. C.; Haywoard, G. L. [Guelph Univ., Guelph, ON (Canada)

    2001-07-01

    Six different cement mortar specimens used in animal buildings, where they were exposed to hydrogen sulfide generated from anaerobic fermentation of manure during a period of one year, were investigated. Primary interest was on comparing the corrosion resistance of different cement mortar specimens under long term exposure to hydrogen sulfide. The impressed voltage technique was used to test the specimens in the laboratory. Results revealed that test specimens made with eight per cent silica fume cement replacement performed best and similar Portland cement mortar specimens with a water-cement ratio of 0.55 (PC55) the poorest. All other treatments, (Portland cement with a water to cement ratio of 045, Portland cement Type 50, Portland cement with fibre mesh and Portland cement Type 10 coated with linseed oil) all with water-cement ratios of 0.45, were less effective in preventing corrosion than silica fume replacement.

  3. Micro-structure and micro-hardness of ODS steels after ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Liu, C., E-mail: liuchxin@eng.hokudai.ac.jp [Graduate School of Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Yu, C.; Hashimoto, N.; Ohnuki, S. [Graduate School of Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Ando, M.; Shiba, K.; Jitsukawa, S. [Tokai Laboratory, JAEA, Tokai, Ibaraki 319-1195 (Japan)

    2011-10-01

    The radiation-hardening of oxide dispersion strengthened (ODS) alloys was examined using ion irradiation and nano-indentation. In this work, three ODS steels were irradiated in the TIARA facility at JAEA with 10.5 MeV Fe{sup 3+} ions up to a dose of 20 dpa at 250 and 380 deg. C. Micro-hardness measurements were carried out on the ion-irradiated specimens with ultra-low load indention. Micro-structures were investigated by transmission electron microscopy (TEM) to examine the contribution of various types of defects to the radiation-hardening. All three steels showed increases in the hardness after the ion irradiation, and F82H-ODS showed the lowest radiation-hardening, which suggests that F82H-ODS has the better radiation resistance. Small amounts of particle dissolution was also confirmed in all of the steels after the irradiation.

  4. Micro-structure and micro-hardness of ODS steels after ion irradiation

    Science.gov (United States)

    Liu, C.; Yu, C.; Hashimoto, N.; Ohnuki, S.; Ando, M.; Shiba, K.; Jitsukawa, S.

    2011-10-01

    The radiation-hardening of oxide dispersion strengthened (ODS) alloys was examined using ion irradiation and nano-indentation. In this work, three ODS steels were irradiated in the TIARA facility at JAEA with 10.5 MeV Fe 3+ ions up to a dose of 20 dpa at 250 and 380 °C. Micro-hardness measurements were carried out on the ion-irradiated specimens with ultra-low load indention. Micro-structures were investigated by transmission electron microscopy (TEM) to examine the contribution of various types of defects to the radiation-hardening. All three steels showed increases in the hardness after the ion irradiation, and F82H-ODS showed the lowest radiation-hardening, which suggests that F82H-ODS has the better radiation resistance. Small amounts of particle dissolution was also confirmed in all of the steels after the irradiation.

  5. Dosimetric characteristics of ultraviolet and x-ray-irradiated KBr:Eu{sup 2+} thermoluminescence crystals

    Energy Technology Data Exchange (ETDEWEB)

    Melendrez, R.; Perez-Salas, R. [Programa de Posgrado en Fisica de Materiales, Centro de Investigacion, Cientifica y de Educacion Superior de Ensenada, Apartado Postal 2681, Ensenada, Baja California, 22800 (Mexico); Aceves, R.; Piters, T.M.; Barboza-Flores, M. [Centro de Investigacion en Fisica, Universidad de Sonora, Apartado Postal 5-088, Hermosillo, Sonora, 83190 (Mexico)

    1996-08-01

    Thermoluminescence (TL) characteristics of KBr:Eu{sup 2+} (150 ppm) previously exposed to ultraviolet (UV) light (200{endash}300 nm) and x-ray radiation at room temperature have been determined. The TL glow curve of UV-irradiated samples is composed of six peaks located at 337, 384, 402, 435, 475, and 510 K. The TL glow curves of x-irradiated samples show mainly a TL peak around 384 K. The TL intensities of UV-irradiated (402 and 510 K glow peaks) and x-irradiated specimens present a linear dependence as a function of radiation dose as well as fading stability 300 s after irradiation. These results further enhance the possibilities of using europium-doped materials in nonionizing (actinic region) and ionizing radiation detection and dosimetry applications. {copyright} {ital 1996 American Institute of Physics.}

  6. Post irradiation test report of irradiated DUPIC simulated fuel

    Energy Technology Data Exchange (ETDEWEB)

    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 {gamma}-scanning, profilometry, density, hardness, observation the microstructure and fission product distribution by optical microscope and electron probe microanalyser (EPMA)

  7. Effect of irradiation on expression of clusterin in the rast salivary glands

    Energy Technology Data Exchange (ETDEWEB)

    O, Gyu Myeong; Choi, Yong Suk; Hwang, Eui Hwan; Lee, Sang Rae [Kyung Hee University, Seoul (Korea, Republic of)

    2006-03-15

    To investigate clusterin expression in the acini and ductal cells of rat submandibular glands after Co-60 gamma irradiation. The male Sprague-Dawley rats weighing approximately 250 gm were divided into control and experimental groups. The experimental group was irradiated with a single absorbed dose of 2, 5, 10, and 15 Gy on the head and neck region. All the rats were sacrificed at 1, 3, 7, 14, 21, and 28 days after irradiation. The specimens including the submandibular gland were sectioned and observed using a immunohistochemical method. In the 2 Gy group, clusterin expression was similar to that of the control group at 1 day after irradiation and it was observed in the striated ductal cells at 3 days after irradiation. In the 5 Gy group, clusterin expression was observed in the striated ductal cells at 1 day after irradiation and gradually increased in the 10 and 15 Gy groups. In the 15 Gy group, clusterin expression was prominent in the striated ductal cells at 1 day after irradiation. but it gradually decreased with the experimental period. The destruction of the striated ductal cells was observed in the 2 Gy group at 21 days after irradiation and in the 5, 10, and 15 Gy group at 7 days after irradiation. The destruction of the acinar cells was observed in the 2 Gy group at 28 days after irradiation and in the 5, 10, and 15 Gy groups at 14 days after irradiation. Clusterin expression was induced by low doses of irradiation and it appeared to be involved in the regulation of cellular response to irradiation.

  8. Irradiation behavior of Ti-stabilized 316L type steel

    Science.gov (United States)

    Rodchenkov, B. S.; Kalinin, G. M.; Strebkov, Yu. S.; Shamardin, V. K.; Prokhorov, V. I.; Bulanova, T. M.

    2009-04-01

    Type 316L austenitic steels are widely used for the in-vessel internal structures of fission reactors (core, core support, etc.) and for experimental irradiation facilities. The modifications of 316L Type steel (316L, 316L(N), US 316, J 316, JPCA, etc.) have been considered as structural material for International Thermonuclear Experimental Reactor (ITER). The results of investigation the irradiation behaviour of Ti-stabilized 316 L type steel (0.04 C-15 Cr-11 Ni-2.5 Mo-0.5 Ti) are presented in this work. The specimens cut out from 316L-Ti steel forging were irradiated in the SM-2 reactor up to a dose ˜4 and 10 dpa at 265 ± 15 °C. The tensile properties, fracture toughness and changes in resistance to intergranular stress corrosion cracking (IGSCC) have been investigated after irradiation. The results for Ti-stabilized 316L steel were compared with those for 316L(N)-IG steel irradiated at the same condition.

  9. Irradiation behavior of Ti-stabilized 316L type steel

    Energy Technology Data Exchange (ETDEWEB)

    Rodchenkov, B.S. [Research and Development Institute of Power Engineering (RDIPE), P.O. Box 788, 101000 Moscow (Russian Federation)], E-mail: rodchen@nikiet.ru; Kalinin, G.M.; Strebkov, Yu.S. [Research and Development Institute of Power Engineering (RDIPE), P.O. Box 788, 101000 Moscow (Russian Federation); Shamardin, V.K.; Prokhorov, V.I.; Bulanova, T.M. [State Scientific Center ' Research Institute of Atomic Reactors' , Dimitrovgrad-10, 433510 Ulyanovsk Region (Russian Federation)

    2009-04-30

    Type 316L austenitic steels are widely used for the in-vessel internal structures of fission reactors (core, core support, etc.) and for experimental irradiation facilities. The modifications of 316L Type steel (316L, 316L(N), US 316, J 316, JPCA, etc.) have been considered as structural material for International Thermonuclear Experimental Reactor (ITER). The results of investigation the irradiation behaviour of Ti-stabilized 316 L type steel (0.04 C-15 Cr-11 Ni-2.5 Mo-0.5 Ti) are presented in this work. The specimens cut out from 316L-Ti steel forging were irradiated in the SM-2 reactor up to a dose {approx}4 and 10 dpa at 265 {+-} 15 deg. C. The tensile properties, fracture toughness and changes in resistance to intergranular stress corrosion cracking (IGSCC) have been investigated after irradiation. The results for Ti-stabilized 316L steel were compared with those for 316L(N)-IG steel irradiated at the same condition.

  10. Post Irradiation Mechanical Behaviour of Three EUROFER Joints

    Energy Technology Data Exchange (ETDEWEB)

    Lucon, E.; Leenaers, A.; Vandermeulen, W.

    2006-08-15

    The post-irradiation mechanical properties of three EUROFER joints (two diffusion joints and one TIG weld) have been characterized after irradiation to 1.8 dpa at 300 degrees Celsius in the BR-2 reactor. Tensile, KLST impact and fracture toughness tests have been performed. Based on the results obtained and on the comparison with data from EUROFER base material irradiated under similar conditions, the post-irradiation mechanical behaviour of both diffusion joints (laboratory and mock-up) appears similar to that of the base material. The properties of the TIG joint are affected by the lack of a post-weld heat treatment, which causes the material from the upper part of the weld to be significantly worse than that of the lower region. Thus, specimens from the upper layer exhibit extremely pronounced hardening and embrittlement caused by irradiation. The samples extracted from the lower layer show much better resistance to neutron exposure, although their measured properties do not match those of the diffusion joints. The results presented demonstrate that diffusion joining can be a very promising technique.

  11. Status of FeCrAl ODS Irradiations in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD)

    2016-08-19

    FeCrAl oxide-dispersion strengthened (ODS) alloys are an attractive sub-set alloy class of the more global FeCrAl material class for nuclear applications due to their high temperature steam oxidation resistance and hypothesized enhanced radiation tolerance. A need currently exists to determine the radiation tolerance of these newly developed alloys. To address this need, a preliminary study was conducted using the High Flux Isotope Reactor (HFIR) to irradiate an early generation FeCrAl ODS alloy, 125YF. Preliminary post-irradiation examination (PIE) on these irradiated specimens have shown good radiation tolerance at elevated temperatures (≥330°C) but possible radiation-induced hardening and embrittlement at irradiations of 200°C to a damage level of 1.9 displacement per atom (dpa). Building on this experience, a new series of irradiations are currently being conceptualized. This irradiation series called the FCAD irradiation program will irradiate the latest generation FeCrAl ODS and FeCr ODS alloys to significantly higher doses. These experiments will provide the necessary information to determine the mechanical performance of irradiated FeCrAl ODS alloys at light water reactor and fast reactor conditions.

  12. Impact property of low-activation vanadium alloy after laser welding and heavy neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Nagasaka, Takuya, E-mail: nagasaka@nifs.ac.jp [National Institute for Fusion Science, Toki, Gifu (Japan); The Graduate University for Advanced Studies, Toki, Gifu (Japan); Muroga, Takeo [National Institute for Fusion Science, Toki, Gifu (Japan); The Graduate University for Advanced Studies, Toki, Gifu (Japan); Watanabe, Hideo [Research Institute for Applied Mechanics, Kyushu University, Kasuga (Japan); Miyazawa, Takeshi [The Graduate University for Advanced Studies, Toki, Gifu (Japan); Yamazaki, Masanori [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Oarai, Ibaraki (Japan); Shinozaki, Kenji [Department of Mechanical System Engineering, Graduate School of Engineering, Hiroshima University, Higashi Hiroshima (Japan)

    2013-11-15

    Weld specimens of the reference low activation vanadium alloy, NIFS-HEAT-2, were irradiated up to a neutron fluence of 1.5 × 10{sup 25} n m{sup −2} (E > 0.1 MeV) (1.2 dpa) at 670 K and 1.3 × 10{sup 26} n m{sup −2} (5.3 dpa) at 720 K in the JOYO reactor in Japan. The base metal exhibited superior irradiation resistance with the ductile-to-brittle transition temperature (DBTT) much lower than room temperature (RT) for both irradiation conditions. The weld metal kept the DBTT below RT after the 1.2 dpa irradiation; however, it showed enhanced irradiation embrittlement with much higher DBTT than RT after the 5.3 dpa irradiation. The high DBTT for the weld metal was effectively recovered by a post-irradiation annealing at 873 K for 1 h. Mechanisms of the irradiation embrittlement and its recovery are discussed, based on characterization of the radiation defects and irradiation-induced precipitation.

  13. Status of FeCrAl ODS Irradiations in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD)

    2016-08-19

    FeCrAl oxide-dispersion strengthened (ODS) alloys are an attractive sub-set alloy class of the more global FeCrAl material class for nuclear applications due to their high-temperature steam oxidation resistance and hypothesized enhanced radiation tolerance. A need currently exists to determine the radiation tolerance of these newly developed alloys. To address this need, a preliminary study was conducted using the High Flux Isotope Reactor (HFIR) to irradiate an early generation FeCrAl ODS alloy, 125YF. Preliminary post-irradiation examination (PIE) on these irradiated specimens have shown good radiation tolerance at elevated temperatures (≥330°C) but possible radiation-induced hardening and embrittlement at irradiations of 200°C to a damage level of 1.9 displacement per atom (dpa). Building on this experience, a new series of irradiations are currently being conceptualized. This irradiation series called the FCAD irradiation program will irradiate the latest generation FeCrAl ODS and FeCr ODS alloys to significantly higher doses. These experiments will provide the necessary information to determine the mechanical performance of irradiated FeCrAl ODS alloys at light water reactor and fast reactor conditions.

  14. Plastic zone size for nanoindentation of irradiated Fesbnd 9%Cr ODS

    Science.gov (United States)

    Dolph, Corey K.; da Silva, Douglas J.; Swenson, Matthew J.; Wharry, Janelle P.

    2016-12-01

    The objective of this study is to determine irradiation effects on the nanoindentation plastic zone morphology in a model Fe-9%Cr ODS alloy. Specimens are irradiated to 50 displacements per atom at 400°C with Fe++ self-ions or to 3 dpa at 500°C with neutrons. The as-received specimen is also studied as a control. The nanoindentation plastic zone size is calculated using two approaches: (1) an analytical model based on the expanding spherical cavity analogy, and (2) finite element modeling (FEM). Plastic zones in all specimen conditions extend radially outward from the indenter, ∼4-5 times the tip radius, indicative of fully plastic contact. Non-negligible plastic flow in the radial direction requires the experimentalist to consider the plastic zone morphology when nanoindenting ion-irradiated specimens; a single nanoindent may sample non-uniform irradiation damage, regardless of whether the indent is made top-down or in cross-section. Finally, true stress-strain curves are generated.

  15. Alloy development for irradiation performance. Quarterly progress report for period ending December 31, 1979

    Energy Technology Data Exchange (ETDEWEB)

    Ashdown, B.G. (comp.)

    1980-04-01

    Progress is reported concerning preparation of a materials handbook for fusion, creep-fatigue of first-wall structural materials, test results on miniature compact tension fracture toughness specimens, austenitic stainless steels, Fe-Ni-Cr alloys, iron-base alloys with long-range crystal structure, ferritic steels, irradiation experiments, corrosion testing, and hydrogen permeation studies. (FS)

  16. The effect of irradiation on electrical and electrodynamic properties of nanocarbon-epoxy composites

    Energy Technology Data Exchange (ETDEWEB)

    Matzui, L.; Vovchenko, L.; Lazarenko, O.; Oliynyk, V.; Launetz, V. [Department of Physics, Kyiv National Shevchenko University, Kyiv (Ukraine); Antoni, F.; Muller, D.; Le Normand, F. [Institut d' Electronique du Solide et des Systemes, CNRS, Strasbourg (France)

    2014-12-01

    The aim of this work is to examine the effect of irradiation on the structure, the electrical resistivity and electromagnetic radiation (EMR) shielding (frequency range of f = 25.5-37.5 GHz) of nanocarbon-epoxy composites (CMs). The graphite nanoplatelets (GNPs) and multi-walled carbon nanotubes (MWCNTs) are used as fillers in epoxy polymer matrix (epoxy resin ED20). Nanocarbon filler content is 1-2.2 vol.%. Polymer-nanocarbon CMs were irradiated by different methods: UV irradiation for 51 h, SuK{sub α} X-ray radiation (U = 30 kV, i = 20 mA) for 1 h, electron irradiation (1.8 MeV, the absorbed dose is 1 MGr), ions H{sup +} (1.5 MeV, 5 x 10{sup 16}), C{sup +} (3 MeV, 4 x 10{sup 16}). It was shown, that the different types of radiation provide different influence both on the value and character of the temperature dependence of resistivity. The most significant change of resistivity was observed for the specimen irradiated by UV - resistivity decreases almost by three orders (by 806 times), the temperature dependence of resistivity being weakened. For specimens irradiated by UV and X-ray we also have observed the increase of shielding efficiency as compared with initial CM, while the irradiation of specimen by electrons leads to a decrease of shielding efficiency SE{sub T}. According to Raman investigation results, the irradiation by H{sup +} ions of CM 2 wt.% MWCNTs-ED20 with intensity 1.5 MeV leads to significant destroying of nanotube structure and formation of homogeneous amorphous material. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  17. Irradiated cocoa beans

    Energy Technology Data Exchange (ETDEWEB)

    Ashby, R.; Tesh, J.M.

    1982-11-01

    Groups of 40 male and 40 female CD rats were fed powdered rodent diet containing 25% (w/w) of either non-irradiated, irradiated or fumigated cocoa beans. The diets were supplemented with certain essential dietary constituents designed to satisfy normal nutritional requirements. An additional 40 male and 40 female rats received basal rodent diet alone (ground) and acted as an untreated control. After 70 days of treatment, 15 male and 15 female rats from each group were used to assess reproductive function of the F/sub 0/ animals and growth and development of the F/sub 1/ offspring up to weaning; the remaining animals were killed after 91 days of treatment.

  18. Irradiated brown dwarfs

    CERN Document Server

    Casewell, S L; Lawrie, K A; Maxted, P F L; Dobbie, P D; Napiwotzki, R

    2014-01-01

    We have observed the post common envelope binary WD0137-349 in the near infrared $J$, $H$ and $K$ bands and have determined that the photometry varies on the system period (116 min). The amplitude of the variability increases with increasing wavelength, indicating that the brown dwarf in the system is likely being irradiated by its 16500 K white dwarf companion. The effect of the (primarily) UV irradiation on the brown dwarf atmosphere is unknown, but it is possible that stratospheric hazes are formed. It is also possible that the brown dwarf (an L-T transition object) itself is variable due to patchy cloud cover. Both these scenarios are discussed, and suggestions for further study are made.

  19. Effect of specimen geometry on tensile strength of cortical bone.

    Science.gov (United States)

    Feng, Liang; Jasiuk, Iwona

    2010-11-01

    We investigate the effect of specimen geometry on the ultimate tensile strength of cortical bone measured by a tensile test. This article is motivated by the fact that there is no clear consensus in the literature on a suitable specimen shape for cortical bone testing. We consider three commonly used tensile test specimen shapes: strip, dumbbell with sharp junctions, and dumbbell with rounded junctions. We conduct this study computationally, using a finite element method, and experimentally by testing porcine femurs. Our results show that local stress concentration factors in the specimen lead to reduced values in the measured tensile strength. The higher the stress concentrations are, the lower is the measured strength. We find that the strip specimens are not a good choice due to high stress concentrations. For the same reason, dumbbell specimens with sharp junctions between the grip and gage sections should also be avoided. The dumbbell shaped tensile test specimens with an arc transition and a maximized radius of fillet are a better choice because such geometry lowers stress concentrations.

  20. Flux and composition dependence of irradiation creep of austenitic alloys irradiated in PFR at ˜420°C

    Science.gov (United States)

    Toloczko, M. B.; Garner, F. A.; Standring, J.; Munro, B.; Adaway, S.

    1998-10-01

    Swelling and irradiation creep of five austenitic stainless steel alloys irradiated at ˜420°C in the Prototypic Fast Reactor (PFR) were examined. The specimens were in the form of pressurized creep tubes, constructed in the USA and irradiated in PFR in a joint USA/UK experiment. The alloy compositions varied greatly, with the greatest elemental variation in the nickel content, which ranged from 15% to 40% over the five alloys. For each alloy, at least two identical sets of tubes were constructed. Each tube-set was irradiated at a different neutron flux level. Swelling was observed to vary with both alloy composition and flux. Irradiation creep was examined from the perspective of the overlineB= ɛ¯˙/ overlineσ=B 0+D Ṡ creep model. The values of both creep coefficients, B0 and D, were typical for austenitic stainless steels and were found to be insensitive to flux over the range of fluxes in this experiment. However, the creep coefficients may be mildly sensitive to alloy composition.

  1. Effects of tensile stress on Cu clustering in irradiated Fe–Cu alloy

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, K., E-mail: fujiik@inss.co.jp [Institute of Nuclear Safety System, Inc., Mihama 919-1205 (Japan); Fukuya, K. [Institute of Nuclear Safety System, Inc., Mihama 919-1205 (Japan); Kasada, R.; Kimura, A. [Institute of Advanced Energy, Kyoto University, Uji 611-0011 (Japan); Ohkubo, T. [National Institute for Materials Science, Tsukuba 305-0047 (Japan)

    2015-03-15

    Effects of tensile stress on Cu clustering were explained using atom probe tomography (APT) results of Fe–0.6 wt.%Cu alloy specimens irradiated with 6.4 MeV Fe ions while applying a tensile stress of 60 MPa at room temperature (less than 50 °C) and 290 °C. The hardening under the tensile-stressed irradiation was smaller than that under the stress-free irradiation at both room temperature and 290 °C. APT results showed that well-defined Cu clusters were formed in all specimens even under the room temperature irradiation. The Cu clusters under the tensile-stressed condition were smaller and had higher densities than those under the stress-free condition. The lower Cu content in clusters and more diffuse Cu clustering were obtained for the specimens irradiated under the tensile-stressed condition. The hardening efficiency of Cu clusters was correlated with the Cu content in clusters and the coherency of interface between a cluster and the matrix. Application of tensile stress would control hardening by changing the nature of Cu clusters.

  2. Final Report: Posttest Analysis of Omega II Optical Specimens

    Energy Technology Data Exchange (ETDEWEB)

    Newlander, C D; Fisher, J H

    2007-01-30

    Preliminary posttest analyses have been completed on optical specimens exposed during the Omega II test series conducted on 14 July 2006. The Omega Facility, located at the Laboratory for Laser Energetics (LLE) at the University of Rochester was used to produce X-ray environments through the interaction of intense pulsed laser radiation upon germanium-loaded silica aerogels. The optical specimen testing was supported by GH Systems through experiment design, pre- and post-test analyses, specimen acquisition, and overall technical experience. The test specimens were fabricated and characterized by Surface Optics Corporation (SOC), San Diego, CA and were simple protected gold coatings on silica substrates. Six test specimens were exposed, five filtered with thin beryllium foil filters, and one unfiltered which was exposed directly to the raw environment. The experimental objectives were: (1) demonstrate that tests of optical specimens could be performed at the Omega facility; (2) evaluate the use and survivability of beryllium foil filters as a function of thickness; (3) obtain damage data on optical specimens which ranged from no damage to damage; (4) correlate existing thermal response models with the damage data; (5) evaluate the use of the direct raw environment upon the specimen response and the ability/desirability to conduct sensitive optical specimen tests using the raw environment; and (6) initiate the development of a protocol for performing optical coatings/mirror tests. This report documents the activities performed by GH Systems in evaluating and using the environments provided by LLNL, the PUFFTFT analyses performed using those environments, and the calculated results compared to the observed and measured posttest data.

  3. A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels

    Energy Technology Data Exchange (ETDEWEB)

    Eason, Ernest D. [Modeling and Computing Services, LLC; Odette, George Robert [UCSB; Nanstad, Randy K [ORNL; Yamamoto, Takuya [ORNL

    2007-11-01

    The reactor pressure vessels (RPVs) of commercial nuclear power plants are subject to embrittlement due to exposure to high-energy neutrons from the core, which causes changes in material toughness properties that increase with radiation exposure and are affected by many variables. Irradiation embrittlement of RPV beltline materials is currently evaluated using Regulatory Guide 1.99 Revision 2 (RG1.99/2), which presents methods for estimating the shift in Charpy transition temperature at 30 ft-lb (TTS) and the drop in Charpy upper shelf energy (ΔUSE). The purpose of the work reported here is to improve on the TTS correlation model in RG1.99/2 using the broader database now available and current understanding of embrittlement mechanisms. The USE database and models have not been updated since the publication of NUREG/CR-6551 and, therefore, are not discussed in this report. The revised embrittlement shift model is calibrated and validated on a substantially larger, better-balanced database compared to prior models, including over five times the amount of data used to develop RG1.99/2. It also contains about 27% more data than the most recent update to the surveillance shift database, in 2000. The key areas expanded in the current database relative to the database available in 2000 are low-flux, low-copper, and long-time, high-fluence exposures, all areas that were previously relatively sparse. All old and new surveillance data were reviewed for completeness, duplicates, and discrepancies in cooperation with the American Society for Testing and Materials (ASTM) Subcommittee E10.02 on Radiation Effects in Structural Materials. In the present modeling effort, a 10% random sample of data was reserved from the fitting process, and most aspects of the model were validated with that sample as well as other data not used in calibration. The model is a hybrid, incorporating both physically motivated features and empirical calibration to the U.S. power reactor surveillance

  4. Friction Compensation in the Upsetting of Cylindrical Test Specimens

    DEFF Research Database (Denmark)

    Christiansen, Peter; Martins, P. A. F.; Bay, Niels Oluf

    2016-01-01

    This manuscript presents a combined numerical andexperimental methodology for determining the stress-straincurve of metallic materials from the measurements of forceand displacement obtained in the axial compression of cylindrical test specimens with friction between the specimens and the platens...... model or combined friction models are utilized .Experimental results obtained from cylindrical and Rastegaev test specimens with different lubricants combined with the experimental determination of friction by means of ring compression tests allows compensating the effect of friction...... Appendix is provided for those readers interested in utilizing the associated numerical algorithm for determining the stress straincurves of metallic materials....

  5. Strain rate dependence of the tensile properties of V-(4--5%)Cr-(4--5%)Ti irradiated in EBR-II and HFBR

    Energy Technology Data Exchange (ETDEWEB)

    Zinkle, S.J.; Snead, L.L.; Robertson, J.P.; Rowcliffe, A.F. [Oak Ridge National Lab., TN (United States)

    1998-03-01

    Elevated temperature tensile tests performed on V-(405)Cr-(4-5)Ti indicate that the yield stress increases with increasing strain rate for irradiation and test temperatures near 200 C, and decreases with increasing strain rate for irradiation and test temperatures near 400 C. This observation is in qualitative agreement with the temperature-dependent strain rate effects observed on unirradiated specimens, and implies that some interstitial solute remains free to migrate in irradiated specimens. Additional strain rate data at different temperatures are needed.

  6. Design of a Compact Fatigue Tester for Testing Irradiated Materials

    Energy Technology Data Exchange (ETDEWEB)

    Hartsell, Brian [Fermilab; Campbell, Michael [Fermilab; Fitton, Michael [RAL, Didcot; Hurh, Patrick [Fermilab; Ishida, Taku [KEK, Tsukuba; Nakadaira, Takeshi [KEK, Tsukuba

    2015-06-01

    A compact fatigue testing machine that can be easily inserted into a hot cell for characterization of irradiated materials is beneficial to help determine relative fatigue performance differences between new and irradiated material. Hot cell use has been carefully considered by limiting the size and weight of the machine, simplifying sample loading and test setup for operation via master-slave manipulator, and utilizing an efficient design to minimize maintenance. Funded from a US-Japan collaborative effort, the machine has been specifically designed to help characterize titanium material specimens. These specimens are flat cantilevered beams for initial studies, possibly utilizing samples irradiated at other sources of beam. The option to test spherically shaped samples cut from the T2K vacuum window is also available. The machine is able to test a sample to $10^7$ cycles in under a week, with options to count cycles and sense material failure. The design of this machine will be presented along with current status.

  7. A Comparison of the Irradiation Creep Behavior of Several Graphites

    Energy Technology Data Exchange (ETDEWEB)

    Burchell, Timothy D [ORNL; Windes, Will [Idaho National Laboratory (INL)

    2016-01-01

    Graphite creep strain data from the irradiation creep capsule Advanced Graphite Creep-1 (AGC-1) are reported. This capsule was the first (prototype) of a series of five or six capsules planned as part of the AGC experiment, which was designed to fully characterize the effects of neutron irradiation and the radiation creep behavior of current nuclear graphite. The creep strain data and analysis are reported for the six graphite grades incorporated in the capsule. The AGC-1 capsule was irradiated in the Advanced Test Reactor at Idaho National Laboratory (INL) at approximately 700 C and to a peak dose of 7 dpa (displacements per atom). The specimen s final dose, temperature, and stress conditions have been reported by INL and were used during this analysis. The derived creep coefficients (K) were calculated for each grade and were found to compare well to literature data for the creep coefficient, even under the wide range of AGC-1 specimen temperatures. Comparisons were made between AGC-1 data and historical grade data for creep coefficients.

  8. Recent results on the neutron irradiation of ITER candidate copper alloys irradiated in DR-3 at 250{degrees}C to 0.3 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, D.J. [Pacific Northwest National Lab., Richland, WA (United States); Singh, B.N.; Toft, P.; Eldrup, M.

    1997-04-01

    Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing and bonding thermal treatment with additional specimens re-aged and given a reactor bakeout treatment at 350{degrees}C for 100 h. CuAl-25 was also heat treated to simulate the effects of a bonding thermal cycle on the material. A number of heat treated specimens were neutron irradiated at 250{degrees}C to a dose level of {approximately}0.3 dpa in the DR-3 reactor as Riso. The main effect of the bonding thermal cycle heat treatment was a slight decrease in strength of CuCrZr and CuNiBe alloys. The strength of CuAl-25, on the other hand, remained almost unaltered. The post irradiation tests at 250{degrees}C showed a severe loss of ductility in the case of the CuNiBe alloy. The irradiated CuAl-25 and CuCrZr specimens exhibited a reasonable amount of uniform elongation, with CuCrZr possessing a lower strength.

  9. ARCTOS: a relational database relating specimens, specimen-based science, and archival documentation

    Science.gov (United States)

    Jarrell, Gordon H.; Ramotnik, Cindy A.; McDonald, D.L.

    2010-01-01

    Data are preserved when they are perpetually discoverable, but even in the Information Age, discovery of legacy data appropriate to particular investigations is uncertain. Secure Internet storage is necessary but insufficient. Data can be discovered only when they are adequately described, and visibility increases markedly if the data are related to other data that are receiving usage. Such relationships can be built within (1) the framework of a relational database, or (1) they can be built among separate resources, within the framework of the Internet. Evolving primarily around biological collections, Arctos is a database that does both of these tasks. It includes data structures for a diversity of specimen attributes, essentially all collection-management tasks, plus literature citations, project descriptions, etc. As a centralized collaboration of several university museums, Arctos is an ideal environment for capitalizing on the many relationships that often exist between items in separate collections. Arctos is related to NIH’s DNA-sequence repository (GenBank) with record-to-record reciprocal linkages, and it serves data to several discipline-specific web portals, including the Global Biodiversity Information Network (GBIF). The University of Alaska Museum’s paleontological collection is Arctos’s recent extension beyond the constraints of neontology. With about 1.3 million cataloged items, additional collections are being added each year.

  10. Regulation of food irradiation and detection of irradiated food

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, P.B. [Institute of Geological and Nuclear Sciences, Lower Hutt (New Zealand)

    1998-12-31

    The main international standards for irradiated foods are those produced by the Codex Alimentarius Commission. The international regulatory environment is now favourable towards irradiated foods. Most countries still regulate on a food-by-food, case-by-case basis. However in Asia there is movement towards a Harmonised Regulation for Irradiated Foods. The WHO believes that irradiated foods may be safely irradiated at any dose above 10 kGy. This may lead to the Codex maximum dose being raised or abandoned. If this occurs there are opportunities to produce shelf-stable foods in lightweight packaging that last for years at room temperature. Detection methods for irradiated foods are now available and may assist to reassure consumers that labelling regulations can be enforced. (author)

  11. AGC-2 Irradiation Data Qualification Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Laurence C. Hull

    2012-07-01

    The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The second Advanced Graphite Creep (AGC) experiment (AGC-2) began with Advanced Test Reactor (ATR) Cycle 149A on April 12, 2011, and ended with ATR Cycle 151B on May 5, 2012. The purpose of this report is to qualify AGC-2 irradiation monitoring data following INL Management and Control Procedure 2691, Data Qualification. Data that are Qualified meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Data that do not meet the requirements are Failed. Some data may not quite meet the requirements, but may still provide some useable information. These data are labeled as Trend. No Trend data were identified for the AGC-2 experiment. All thermocouples functioned throughout the AGC-2 experiment. There was one instance where spurious signals or instrument power interruption resulted in a recorded temperature value being well outside physical reality. This value was identified and labeled as Failed data. All other temperature data are Qualified. All helium and argon gas flow data are within expected ranges. Total gas flow was approximately 50 sccm through the capsule. Helium gas flow was briefly increased to 100 sccm during reactor shutdown. All gas flow data are Qualified. At the start of the experiment, moisture in the outflow gas line increased to 200 ppmv then declined to less than 10 ppmv over a period of 5 days. This increase in moisture coincides with the initial heating of the experiment and drying of the system. Moisture slightly exceeded 10 ppmv three other times during the experiment. While these moisture values exceed the 10 ppmv threshold value, the reported measurements are considered accurate and to reflect moisture conditions in the capsule. All moisture data are Qualified. Graphite creep specimens are subjected to one of three loads, 393 lbf

  12. AGC-2 Irradiation Data Qualification Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Laurence C. Hull

    2012-07-01

    The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The second Advanced Graphite Creep (AGC) experiment (AGC-2) began with Advanced Test Reactor (ATR) Cycle 149A on April 12, 2011, and ended with ATR Cycle 151B on May 5, 2012. The purpose of this report is to qualify AGC-2 irradiation monitoring data following INL Management and Control Procedure 2691, Data Qualification. Data that are Qualified meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Data that do not meet the requirements are Failed. Some data may not quite meet the requirements, but may still provide some useable information. These data are labeled as Trend. No Trend data were identified for the AGC-2 experiment. All thermocouples functioned throughout the AGC-2 experiment. There was one instance where spurious signals or instrument power interruption resulted in a recorded temperature value being well outside physical reality. This value was identified and labeled as Failed data. All other temperature data are Qualified. All helium and argon gas flow data are within expected ranges. Total gas flow was approximately 50 sccm through the capsule. Helium gas flow was briefly increased to 100 sccm during reactor shutdown. All gas flow data are Qualified. At the start of the experiment, moisture in the outflow gas line increased to 200 ppmv then declined to less than 10 ppmv over a period of 5 days. This increase in moisture coincides with the initial heating of the experiment and drying of the system. Moisture slightly exceeded 10 ppmv three other times during the experiment. While these moisture values exceed the 10 ppmv threshold value, the reported measurements are considered accurate and to reflect moisture conditions in the capsule. All moisture data are Qualified. Graphite creep specimens are subjected to one of three loads, 393 lbf

  13. 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)

  14. Examination of the fatigue life under combined loading of specimens

    Directory of Open Access Journals (Sweden)

    Fojtík F.

    2008-11-01

    Full Text Available This contribution describes experimental results under combined loading of specimens manufactured from common construction steel 11523. Specimens were gradually loaded by amplitude of the torque, then by combination of torque and tension prestress. The last set of specimens was loaded in combination of torque and inner overpressure. To obtain the required input values the stress-strain analysis of specimens by finite element method in software Ansys was performed within the last experiment. For evaluation of the results the Fuxa's criterion was applied. The performed experiments and their results embody a good agreement with bellow mentioned conjugated strength criterion. The experiments were performed on reconstructed testing machine equipped by pressure chamber.

  15. Simulation Analysis of Standard Metal Specimen Tension Experiment by Fem

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>Some standard metal rod-shaped, plate-shaped or pipe-shaped specimens usually are used to be tensioned to acquire the material properties such as tensional ductility, contractibility ratio on breaking section,

  16. North Mississippi Refuges Complex Dragonfly Vouchered Specimens 2005

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — Report contains a list of dragonflies and photographs of them collected in 2005 from the refuge complex. These were verified by Steve Krotzer and specimens retained...

  17. Description of Specimens in the Marine Mammal Osteology Reference Collection

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The NMFS Alaska Fisheries Science Center National Marine Mammal Laboratory (NMML) Marine Mammal Osteology Collection consists of approximately 2500 specimens (skulls...

  18. Pathologic diagnoses of appendectomy specimens: a 10-year review.

    African Journals Online (AJOL)

    Pathologic diagnoses of appendectomy specimens: a 10-year review. ... Annals of Biomedical Sciences ... Materials and methods: Records of resected appendices with a clinical diagnosis of acute appendicitis submitted to histopathology ...

  19. Images of paraffin monolayer crystals with perfect contrast: minimization of beam-induced specimen motion

    Science.gov (United States)

    Glaeser, R.M.; McMullan, G.; Faruqi, A.R.; Henderson, R.

    2013-01-01

    Quantitative analysis of electron microscope images of organic and biological two-dimensional crystals has previously shown that the absolute contrast reached only a fraction of that expected theoretically from the electron diffraction amplitudes. The accepted explanation for this is that irradiation of the specimen causes beam-induced charging or movement, which in turn causes blurring of the image due to image or specimen movement. In this paper, we used three different approaches to try to overcome this image-blurring problem for monolayer crystals of paraffin. Our first approach was to use an extreme form of spotscan imaging, in which a single image was assembled on film by the successive illumination of up to 50,000 spots each of diameter around 7nm. The second approach was to use the Medipix II detector with its zero-noise readout to assemble a time-sliced series of images of the same area in which each frame from a movie with up to 400 frames had an exposure of only 500 electrons. In the third approach, we simply used a much thicker carbon support film to increase the physical strength and conductivity of the support. Surprisingly, the first two methods involving dose fractionation respectively in space or time produced only partial improvements in contrast whereas the third approach produced many virtually perfect images, in which the absolute contrast predicted from the electron diffraction amplitudes was observed in the images. We conclude that it is possible to obtain consistently almost perfect images of beam-sensitive specimens if they are attached to an appropriately strong and conductive support, but great care is needed in practice and the problem of how best to image ice-embedded biological structures in the absence of a strong, conductive support film requires more work. PMID:21185452

  20. Miniature specimen shear punch test for UHMWPE used in total joint replacements.

    Science.gov (United States)

    Kurt, S M; Jewett, C W; Bergström, J S; Foulds, J R; Edidin, A A

    2002-05-01

    Despite the critical role that shear is hypothesized to play in the damage modes that limit the performance of total hip and knee replacements, the shear behavior of ultra-high molecular weight polyethylene (UHMWPE) remains poorly understood, especially after oxidative degradation or radiation crosslinking. In the present study, we developed the miniature specimen (0.5 mm thickness x 6.4mm diameter) shear punch test to evaluate the shear behavior of UHMWPE used in total joint replacement components. We investigated the shear punch behavior of virgin and crosslinked stock materials, as well as of UHMWPE from tibial implants that were gamma-irradiated in air and shelf aged for up to 8.5 years. Finite element analysis, scanning electron microscopy, and interrupted testing were conducted to aid in the interpretation of the shear punch load-displacement curves. The shear punch load-displacement curves exhibited similar distinctive features. Following toe-in, the load-displacement curves were typically bilinear, and characterized by an initial stiffness, a transition load, a hardening stiffness, and a peak load. The finite element analysis established that the initial stiffness was proportional to the elastic modulus of the UHMWPE, and the transition load of the bilinear curve reflected the development of a plastically deforming zone traversing through the thickness of the sample. Based on our observations, we propose two interpretations of the peak load during the shear punch test: one theory is based on the initiation of crystalline plasticity, the other based on the transition from shear to tension during the tests. Due to the miniature specimen size, the shear punch test offers several potential advantages over bulk test methods, including the capability to directly measure shear behavior, and quite possibly infer ultimate uniaxial behavior as well, from shelf aged and retrieved UHMWPE components. Thus, the shear punch test represents an effective and complementary

  1. Hardening of ODS ferritic steels under irradiation with high-energy heavy ions

    Science.gov (United States)

    Ding, Z. N.; Zhang, C. H.; Yang, Y. T.; Song, Y.; Kimura, A.; Jang, J.

    2017-09-01

    Influence of the nanoscale oxide particles on mechanical properties and irradiation resistance of oxide-dispersion-strengthened (ODS) ferritic steels is of critical importance for the use of the material in fuel cladding or blanket components in advanced nuclear reactors. In the present work, impact of structures of oxide dispersoids on the irradiation hardening of ODS ferritic steels was studied. Specimens of three high-Cr ODS ferritic steels containing oxide dispersoids with different number density and average size were irradiated with high-energy Ni ions at about -50 °C. The energy of the incident Ni ions was varied from 12.73 MeV to 357.86 MeV by using an energy degrader at the terminal so that a plateau of atomic displacement damage (∼0.8 dpa) was produced from the near surface to a depth of 24 μm in the specimens. A nanoindentor (in constant stiffness mode with a diamond Berkovich indenter) and a Vickers micro-hardness tester were used to measure the hardeness of the specimens. The Nix-Gao model taking account of the indentation size effect (ISE) was used to fit the hardness data. It is observed that the soft substrate effect (SSE) can be diminished substantially in the irradiated specimens due to the thick damaged regions produced by the Ni ions. A linear correlation between the nano-hardeness and the micro-hardness was found. It is observed that a higher number density of oxide dispersoids with a smaller average diameter corresponds to an increased resistance to irradiation hardening, which can be ascribed to the increased sink strength of oxides/matrix interfaces to point defects. The rate equation approach and the conventional hardening model were used to analyze the influence of defect clusters on irradiation hardening in ODS ferritic steels. The numerical estimates show that the hardening caused by the interstitial type dislocation loops follows a similar trend with the experiment data.

  2. Stress Analysis of a Secondary-Bending Specimen

    Science.gov (United States)

    1993-11-01

    Control Office Ansett Airlines of Australia, Library 0 Qantas Airways Limited Hawker de Havilland Aust Pty Ltd, Victoria, Library Hawker de Havilland...MELBOURNE, VICTORIA Technical Note 58 STRESS ANALYSIS OF A SECONDARY-BENDING SPECIMEN 0 by R.L. EVANS M. HELLER Approved for public release C) COMMONWEALTH...AND TECHNOLOGY ORGANISATION AERONAUTICAL RESEARCH LABORATORY Technical Note 58 0 STRESS ANALYSIS OF A SECONDARY-BENDING SPECIMEN by R.L. EVANS 0 M

  3. Innovation for reducing blood culture contamination: initial specimen diversion technique.

    Science.gov (United States)

    Patton, Richard G; Schmitt, Timothy

    2010-12-01

    We hypothesized that diversion of the first milliliter of venipuncture blood-the initial specimen diversion technique (ISDT)-would eliminate incompletely sterilized fragments of skin from the culture specimen and significantly reduce our blood culture contamination rate (R). We studied our hypothesis prospectively beginning with our control culture (C) definition: one venipuncture with two sequentially obtained specimens, 10 ml each, the first specimen (M1) for aerobic and the second (M2) for anaerobic media. The test ISDT culture (D) was identical, with the exception that each was preceded by diverting a 1-ml sample (DS) from the same venipuncture. During the first of two sequential 9-month periods, we captured D versus C data (n=3,733), where DMXR and CMXR are R for D and C specimens. Our hypothesis predicted DS would divert soiled skin fragments from DM1, and therefore, CM1R would be significantly greater than DM1R. This was confirmed by CM1R (30/1,061 [2.8%]) less DM1R (37/2,672 [1.4%]; P=0.005), which equals 1.4%. For the second 9-month follow-up period, data were compiled for all cultures (n=4,143), where ADMXR is R for all (A) diversion specimens, enabling comparison to test ISDT. Our hypothesis predicted no significant differences for test ISDT versus all ISDT. This was confirmed by DM1R (37/2,672 [1.4%]) versus ADM1R (42/4,143 [1.0%]; P=0.17) and DM2R (21/2,672 [0.80%]) versus ADM2R (39/4,143 [0.94%]; P=0.50). We conclude that our hypothesis is valid: venipuncture needles soil blood culture specimens with unsterilized skin fragments and increase R, and ISDT significantly reduces R from venipuncture-obtained blood culture specimens.

  4. Progress Report on Alloy 617 Notched Specimen Testing

    Energy Technology Data Exchange (ETDEWEB)

    McMurtrey, Michael David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, Richard Neil [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lillo, Thomas Martin [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    Creep behavior of Alloy 617 has been extensively characterized to support the development of a draft Code Case to qualify Alloy 617 in Section III division 5 of the ASME Boiler and Pressure Vessel Code. This will allow use of Alloy 617 in construction of nuclear reactor components at elevated temperatures and longer periods of time (up to 950°C and 100,000 hours). Prior to actual use, additional concerns not considered in the ASME code need to be addressed. Code Cases are based largely on uniaxial testing of smooth gage specimens. In service conditions, components will generally be under multi axial loading. There is also the concern of the behavior at discontinuities, such as threaded components. To address the concerns of multi axial creep behavior and at geometric discontinuities, notched specimens have been designed to create conditions representative of the states that service components experience. Two general notch geometries have been used for these series of tests: U notch and V notch specimens. The notches produce a tri axial stress state, though not uniform across the specimen. Characterization of the creep behavior of the U notch specimens and the creep rupture behavior of the V notch specimens provides a good approximation of the behavior expected of actual components. Preliminary testing and analysis have been completed and are reported in this document. This includes results from V notch specimens tested at 900°C and 800°C. Failure occurred in the smooth gage section of the specimen rather than at the root of the notch, though some damage was present at the root of the notch, where initial stress was highest. This indicates notch strengthening behavior in this material at these temperatures.

  5. Specimen Sample Preservation for Cell and Tissue Cultures

    Science.gov (United States)

    Meeker, Gabrielle; Ronzana, Karolyn; Schibner, Karen; Evans, Robert

    1996-01-01

    The era of the International Space Station with its longer duration missions will pose unique challenges to microgravity life sciences research. The Space Station Biological Research Project (SSBRP) is responsible for addressing these challenges and defining the science requirements necessary to conduct life science research on-board the International Space Station. Space Station will support a wide range of cell and tissue culture experiments for durations of 1 to 30 days. Space Shuttle flights to bring experimental samples back to Earth for analyses will only occur every 90 days. Therefore, samples may have to be retained for periods up to 60 days. This presents a new challenge in fresh specimen sample storage for cell biology. Fresh specimen samples are defined as samples that are preserved by means other than fixation and cryopreservation. The challenge of long-term storage of fresh specimen samples includes the need to suspend or inhibit proliferation and metabolism pending return to Earth-based laboratories. With this challenge being unique to space research, there have not been any ground based studies performed to address this issue. It was decided hy SSBRP that experiment support studies to address the following issues were needed: Fixative Solution Management; Media Storage Conditions; Fresh Specimen Sample Storage of Mammalian Cell/Tissue Cultures; Fresh Specimen Sample Storage of Plant Cell/Tissue Cultures; Fresh Specimen Sample Storage of Aquatic Cell/Tissue Cultures; and Fresh Specimen Sample Storage of Microbial Cell/Tissue Cultures. The objective of these studies was to derive a set of conditions and recommendations that can be used in a long duration microgravity environment such as Space Station that will permit extended storage of cell and tissue culture specimens in a state consistent with zero or minimal growth, while at the same time maintaining their stability and viability.

  6. The effect of bonding and bakeout thermal cycles on the properties of copper alloys irradiated at 100 C

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, D.J. [Pacific Northwest National Lab., Richland, WA (United States); Singh, B.N.; Toft, P.; Eldrup, M. [Risoe National Lab., Roskilde (Denmark)

    1998-03-01

    This report describes the final irradiation experiment in a series of screening experiments aimed at investigating the effects of bonding and bakeout thermal cycles on irradiated copper alloys. Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing and bonding thermal treatment. The post-irradiation tests at 100 C revealed the greatest loss of ductility occurred in the CuCrZr alloys, irrespective of the pre-irradiation heat treatment, with the uniform elongation dropping to levels of less than 1.5%. The yield and ultimate strengths for all of the individual heat treated samples increased substantially after irradiation. The same trend was observed for the CuNiBe alloys, which overall exhibited a factor of 3 higher uniform elongation after irradiation with almost double the strength. In both alloys irradiation-induced precipitation lead to a large increase in the strength of the solution annealed specimens with a noticeable decrease in uniform elongation. The Al25 alloy also experienced an increase in the overall strength of the alloy after irradiation, accompanied by approximately a 50% decrease in the uniform and total elongation. The additional bakeout treatments given to the CuCrZr and CuNiBe before irradiation served to increase the strength, but in terms of the ductility no improvement or degradation resulted from the additional thermal exposure. The results of this experiment confirm that the al25 possesses the most resistant microstructure to thermal and irradiation-induced changes, while the competing effects of ballistic dissolution and reprecipitation lead to important changes in the two precipitation strengthened alloys. This study and others have repeatedly shown that these materials can only be used if the very low uniform elongation (1% or less) can be accounted for in the design since pre-irradiation thermal processing cannot mitigate the irradiation embrittlement.

  7. The International Environmental Specimen Banks--let's get visible.

    Science.gov (United States)

    Küster, Anette; Becker, Paul R; Kucklick, John R; Pugh, Rebecca S; Koschorreck, Jan

    2015-02-01

    Environmental specimen banks (ESBs) are facilities that archive samples from the environment for future research and monitoring purposes. In addition, the long-term preservation of representative specimens is an important complement to environmental research and monitoring. Today, environmental specimen banking is experiencing a renaissance due to an increase in regulatory interest in ESB biota standards and trend data. The International Environmental Specimen Bank Group (IESB) promotes the worldwide development of techniques and strategies of environmental specimen banking and the international cooperation and collaboration among national ESBs. In order to provide a current and comprehensive overview on international environmental specimen banking activities, a questionnaire was sent to the national ESBs and asked for detailed information on the respective ESBs. The results show the rich diversity of national sampling programs, including more detailed information on archived samples, sampling strategies, and studies that have already been performed in the respective countries. All ESBs completing the survey expressed a strong interest in cooperating with other ESBs on a collaborative project. The collected information of national ESBs is intended to be made publicly available.

  8. ROLE OF SCALE FACTOR DURING TENSILE TESTING OF SMALL SPECIMENS

    Energy Technology Data Exchange (ETDEWEB)

    Gussev, Maxim N [ORNL; Busby, Jeremy T [ORNL; Field, Kevin G [ORNL; Sokolov, Mikhail A [ORNL; Gray, Mr. Sean [University of Michigan

    2014-01-01

    The influence of scale factor (tensile specimen geometry and dimensions) on mechanical test results was investigated for different widely used types of small specimens (SS-1, SS-2, SS-3, and SS-J3) and a set of materials. It was found that the effect of scale factor on the accurate determination of yield stress, ultimate tensile stress, and uniform elongation values was weak; however, clear systematic differences were observed and should be accounted for during interpretation of results. In contrast, total elongation values were strongly sensitive to variations in specimen geometry. Modern experimental methods like digital image correlation allow the impact of scale factor to be reduced. Using these techniques, it was shown that true stress true strain curves describing strain-hardening behavior were very close for different specimen types. The limits of miniaturization are discussed, and an ultra-miniature specimen concept was suggested and evaluated. This type of specimen, as expected, may be suitable for SEM and TEM in situ testing.

  9. Post irradiation plastic properties of F82H derived from the instrumented tensile tests

    Energy Technology Data Exchange (ETDEWEB)

    Taguchi, T. [Neutron Science Research Center, Japan Atomic Energy, Research Institute, Tokai-Mura, Ibaraki-Ken 319-1195 (Japan)]. E-mail: taguchi@popsvr.tokai.jaeri.go.jp; Jitsukawa, S. [Department of Materials Science, Japan Atomic Energy, Research Institute, Tokai-Mura, Ibaraki-Ken 319-1195 (Japan); Sato, M. [KKS, JFE, Kawasaki-Ku, Kawasaki-Shi, Kanagawa-Ken 210-0855 (Japan); Matsukawa, S. [KKS, JFE, Kawasaki-Ku, Kawasaki-Shi, Kanagawa-Ken 210-0855 (Japan); Wakai, E. [Department of Materials Science, Japan Atomic Energy, Research Institute, Tokai-Mura, Ibaraki-Ken 319-1195 (Japan); Shiba, K. [Department of Materials Science, Japan Atomic Energy, Research Institute, Tokai-Mura, Ibaraki-Ken 319-1195 (Japan)

    2004-12-01

    F82H (Fe-8Cr-2W) and its variant doped with 2%Ni were irradiated up to 20 dpa at 300 deg. C in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Post irradiation tensile testing was performed at room temperature. During testing, the images of the specimens including the necked region were continuously recorded. Tests on cold worked material were also carried out for comparison. From the load-displacement curves and the strain distributions obtains from the images, flow stress levels and strain hardening behavior was evaluated. A preliminary constitutive equation for the plastic deformation of irradiated F82H is presented. The results suggest that the irradiation mainly causes defect-induced hardening while it did not strongly affect strain hardening at the same flow stress level for F82H irradiated at 300 deg. C. The strain hardening of Ni doped specimens was, however, strongly affected by irradiation. Results provide basics to determine allowable stress levels at temperatures below 400 deg. C.

  10. Damage structure of austenitic stainless steel 316LN irradiated at low temperature in HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Hashimoto, N.; Robertson, J.P.; Grossbeck, M.L.; Rowcliffe, A.F. [Oak Ridge National Lab., TN (United States); Wakai, E. [Japan Atomic Energy Research Inst. (Japan)

    1998-03-01

    TEM disk specimens of austenitic stainless steel 316LN irradiated to damage levels of about 3 dpa at irradiation temperatures of either about 90 C or 250 C have been investigated by using transmission electron microscopy. The irradiation at 90 C and 250 C induced a dislocation loop density of 3.5 {times} 10{sup 22} m{sup {minus}3} and 6.5 {times} 10{sup 22} m{sup {minus}3}, a black dot density of 2.2 {times} 10{sup 23} m{sup {minus}3} and 1.6 {times} 10{sup 23} m{sup {minus}3}, respectively, in the steels, and a high density (<1 {times} 10{sup 22} m{sup {minus}3}) of precipitates in matrix. Cavities could be observed in the specimens after the irradiation. It is suggested that the dislocation loops, the black dots, and the precipitates cause irradiation hardening, an increase in the yield strength and a decrease in the uniform elongation, in the 316LN steel irradiated at low temperature.

  11. Irradiation dose and temperature dependence of fracture toughness in high dose HT9 steel from the fuel duct of FFTF

    Energy Technology Data Exchange (ETDEWEB)

    Byun, Thak Sang; Toloczko, Mychailo B.; Saleh, Tarik A.; Maloy, Stuart A.

    2013-01-14

    To expand the knowledge base for fast reactor core materials, fracture toughness has been evaluated for high dose HT9 steel using miniature disk compact tension (DCT) specimens. The HT9 steel DCT specimens were machined from the ACO-3 fuel duct of the Fast Flux Test Facility (FFTF), which achieved high doses in the range of 3–148 dpa at 378–504 C. The static fracture resistance (J-R) tests have been performed in a servohydraulic testing machine in vacuum at selected temperatures including room temperature, 200 C, and each irradiation temperature. Brittle fracture with a low toughness less than 50 MPa pm occurred in room temperature tests when irradiation temperature was below 400 C, while ductile fracture with stable crack growth was observed when irradiation temperature was higher. No fracture toughness less than 100 MPa pm was measured when the irradiation temperature was above 430 C. It was shown that the influence of irradiation temperature was dominant in fracture toughness while the irradiation dose has only limited influence over the wide dose range 3–148 dpa. A slow decrease of fracture toughness with test temperature above room temperature was observed for the nonirradiated and high temperature (>430 *C) irradiation cases, which indicates that the ductile–brittle transition temperatures (DBTTs) in those conditions are lower than room temperature. A comparison with the collection of existing data confirmed the dominance of irradiation temperature in the fracture toughness of HT9 steels.

  12. Neutron irradiation of V-Cr-Ti alloys in the BOR-60 fast reactor: Description of the fusion-1 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Rowcliffe, A.F. [Oak Ridge National Laboratory, TN (United States); Tsai, H.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    The FUSION-1 irradiation capsule was inserted in Row 5 of the BOR-60 fast reactor in June 1995. The capsule contains a collaborative RF/U.S. experiment to investigate the irradiation performance of V-Cr-Ti alloys in the temperature range 310 to 350{degrees}C. This report describes the capsule layout, specimen fabrication history, and the detailed test matrix for the U.S. specimens. A description of the operating history and neutronics will be presented in the next semiannual report.

  13. Effects of helium pre-implantation on the microstructure and mechanical properties of irradiated 316 stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B.; Tedeski, G.R.; Lucas, G.E.; Odette, G.R. [Univ. of California, Santa Barbara, CA (United States). Dept. of Chemical and Nuclear Engineering; Stoller, R.E. [Oak Ridge National Lab., TN (United States); Hamilton, M.L. [Pacific Northwest Lab., Richland, WA (United States)

    1994-11-01

    Transmission electron microscopy (TEM) specimens of a First Core heat of 316 stainless steel, in both the solution annealed and 20% cold worked condition, were irradiated to 46 dpa at 420 C, to 49 dpa at 520 C, and to 34 dpa at 600 C in FFTF/MOTA. Prior to irradiation, about half of the specimens were pre-implanted with approximately 100 appm of helium, and of these, several of the solution annealed and pre-implanted specimens were aged at 800 C for 2 hr. Post-irradiation density measurements showed little difference in density between the unimplanted alloys and their helium implanted counterparts. Microstructural observations on specimens irradiated at 420 C and 520 C showed relatively minor differences in defect distributions between the unimplanted and the helium implanted materials; in all cases the defect sizes and number densities were consistent with data in the literature. Where possible, irradiation hardening of the alloys was experimentally evaluated by microhardness and shear punch; experimentally obtained values were compared to values calculated using a computer model based on barrier hardening and the microstructural data. All methods indicated relatively small effects of helium implantation, and both measured and calculated values were in agreement with the range of values reported in the literature.

  14. Craniospinal irradiation techniques

    Energy Technology Data Exchange (ETDEWEB)

    Scarlatescu, Ioana, E-mail: scarlatescuioana@gmail.com; Avram, Calin N. [Faculty of Physics, West University of Timisoara, Bd. V. Parvan 4, 300223 Timisoara (Romania); Virag, Vasile [County Hospital “Gavril Curteanu” - Oradea (Romania)

    2015-12-07

    In this paper we present one treatment plan for irradiation cases which involve a complex technique with multiple beams, using the 3D conformational technique. As the main purpose of radiotherapy is to administrate a precise dose into the tumor volume and protect as much as possible all the healthy tissues around it, for a case diagnosed with a primitive neuro ectoderm tumor, we have developed a new treatment plan, by controlling one of the two adjacent fields used at spinal field, in a way that avoids the fields superposition. Therefore, the risk of overdose is reduced by eliminating the field divergence.

  15. Microwave applications to rock specimen drying in laboratory

    Science.gov (United States)

    Park, Jihwan; Park, Hyeong-Dong

    2014-05-01

    Microwave heating is the process in which electromagnetic wave with 300 MHz - 300 GHz heats dielectric material. Although in the beginning microwave was mainly used in food industry to cook or heat the food, it soon became clear that microwave had a large potential for other applications. It was thus introduced in geological fields of investigation like mineral processing, oil sand and oil shale extraction, soil remediation, waste treatment. However, the drying techniques using microwave was rarely treated in geology field. According to the ISRM suggested methods, experimental rock specimens in laboratory test were dried in 105°C oven for a period of at least 24 hours. In this method, hot air transmits heats to material by means of thermal conduction, and the heat was transferred from the surface to the inside of the rock specimens. The thermal gradient and moisture gradient can deteriorate the specimens, and energy can be wasted in bulk heating the specimens. The aim of our study was to compare physical property, microstructural property, and energy efficiency between microwave drying method and conventional oven drying method, and to suggest new method for rock drying. Granite, basalt, and sandstone were selected as specimens and were made in cylinder shape with 54 mm diameter. To compare two different methods, one set of saturated specimens were dried in 105°C conventional oven and the other set of saturated specimens were dried in microwave oven. After dried, the specimens were cooled and saturated in 20°C water 48 hours. The saturation-drying were repeated 50 cycles, and the physical property and microstructural property were measured every 10 cycles. Absorption and elastic wave velocity were measured to investigate the change of physical property, and microscope image and X-ray computed tomography image were obtained to investigate the change of microstructural property of rock specimens. The electricity consumption of conventional oven and microwave oven

  16. HER2 testing on core needle biopsy specimens from primary breast cancers: interobserver reproducibility and concordance with surgically resected specimens

    Directory of Open Access Journals (Sweden)

    Yamamoto Sohei

    2010-10-01

    Full Text Available Abstract Background Accurate evaluation of human epidermal growth factor receptor type-2 (HER2 status based on core needle biopsy (CNB specimens is mandatory for identification of patients with primary breast cancer who will benefit from primary systemic therapy with trastuzumab. The aim of the present study was to validate the application of HER2 testing with CNB specimens from primary breast cancers in terms of interobserver reproducibility and comparison with surgically resected specimens. Methods A total of 100 pairs of archival formalin-fixed paraffin-embedded CNB and surgically resected specimens of invasive breast carcinomas were cut into sections. All 100 paired sections were subjected to HER2 testing by immunohistochemistry (IHC and 27 paired sections were subjected to that by fluorescence in situ hybridization (FISH, the results being evaluated by three and two observers, respectively. Interobserver agreement levels in terms of judgment and the concordance of consensus scores between CNB samples and the corresponding surgically resected specimens were estimated as the percentage agreement and κ statistic. Results In CNB specimens, the percentage interobserver agreement of HER2 scoring by IHC was 76% (κ = 0.71 for 3 × 3 categories (0-1+ versus 2+ versus 3+ and 90% (κ = 0.80 for 2 × 2 categories (0-2+ versus 3+. These levels were close to the corresponding ones for the surgically resected specimens: 80% (κ = 0.77 for 3 × 3 categories and 92% (κ = 0.88 for 2 × 2 categories. Concordance of consensus for HER2 scores determined by IHC between CNB and the corresponding surgical specimens was 87% (κ = 0.77 for 3 × 3 categories, and 94% (κ = 0.83 for 2 × 2 categories. Among the 13 tumors showing discordance in the mean IHC scores between the CNB and surgical specimens, the results of consensus for FISH results were concordant in 11. The rate of successful FISH analysis and the FISH positivity rate in cases with a HER2 IHC score of

  17. Design study of water chemistry control system for IASCC irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Yuichiro; Ide, Hiroshi; Nabeya, Hideaki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tsukada, Takashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-02-01

    In relation to the aging of Light Water Reactor (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, and the irradiation research on the IASCC is now under schedule. With the progress of the irradiation research on reactor materials, well-controlled environment conditions during irradiation testing are required. Especially for irradiation testing of IASCC studies, water chemistry control is essential in addition to the control of neutron fluence and irradiation temperature. According to these requirements, at the Japan Atomic Energy Research Institute (JAERI), an irradiation testing facility that simulates in-core environment of Boiling Water Reactor (BWR) has been designed to be installed in the Japan Materials Testing Reactor (JMTR). This facility is composed of the Saturated Temperature Capsules (SATCAP) that are installed into the JMTR's core to irradiate material specimens, the Water Control Unit that is able to supply high-temperature and high-pressure chemical controlled water to SATCAP, and other components. This report describes the design study of water chemistry control system of the Water Control Unit. The design work has been performed in the fiscal year 1999. (author)

  18. Gamma irradiation effects on cyanate ester/epoxy insulation materials for superconducting magnets

    Energy Technology Data Exchange (ETDEWEB)

    Li, Jingwen [Key Laboratory of Cryogenics, Technical Institute of Physics and Chemistry, Chinese Academy of Sciences, Beijing 100190 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Wu, Zhixiong, E-mail: zxwu@mail.ipc.ac.cn [Key Laboratory of Cryogenics, Technical Institute of Physics and Chemistry, Chinese Academy of Sciences, Beijing 100190 (China); Huang, Chuanjun [Key Laboratory of Cryogenics, Technical Institute of Physics and Chemistry, Chinese Academy of Sciences, Beijing 100190 (China); Li, Laifeng, E-mail: laifengli@mail.ipc.ac.cn [Key Laboratory of Cryogenics, Technical Institute of Physics and Chemistry, Chinese Academy of Sciences, Beijing 100190 (China)

    2014-12-15

    Highlights: • Irradiation resistance of glass fiber reinforced cyanate ester/epoxy composite was investigated. • The cyanate ester/epoxy resin system has a low viscosity and long pot life. • The T{sub g} of the matrix resin decreased slightly with the increase of irradiation dose. • The ILSS of GFRP composite increased slightly when exposed to 10 MGy of gamma irradiation. - Abstract: Cyanate ester/epoxy resin was used as a cryogenic-grade polymer matrix and glass fiber reinforced polymer (GFRP) composite was manufactured. The processing properties of matrix resin in terms of the isothermal viscosity at 45 °C were investigated. The specimens were exposed with gamma irradiation of 1 MGy, 5 MGy and 10 MGy, respectively. The effect of gamma irradiation on thermal properties and structure of cyanate ester/epoxy matrix was investigated. The interlaminar shear strength (ILSS) of the composites before and after irradiation were investigated at room temperature, 77 K and 4.2 K. Results showed that cyanate ester/epoxy system had a low viscosity and a long pot life at 45 °C. The glass transition temperature of the matrix resin decreased with the increasing irradiation dose. Moreover, the ILSS of GFRP composite slightly increases after irradiation and toughening mechanism was also discussed.

  19. Irradiated stars with convective envelopes

    CERN Document Server

    Lucy, L B

    2016-01-01

    The structure of low-mass stars irradiated by a close companion is considered. Irradiation modifies the surface boundary conditions and thereby also the adiabatic constants of their outer convection zones. This then changes the models' radii and luminosities. For short-period M dwarf binaries with components of similar mass, the radius inflation due to their mutual irradiation is found to be < 0.4%. This is an order of magnitude too small to explain the anomalous radii found for such binaries. Although stronger irradiation of an M dwarf results in a monotonically increasing radius, a saturation effect limits the inflation to < 5%.

  20. International Developments of Food Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Loaharanu, P. [Head, Food Preservation Section, Joint FAO/IAEA Division of Nuclear Techniques in Food and Agriculture, Wagramerstr. 5, A-1400, Vienna (Austria)

    1997-12-31

    Food irradiation is increasingly accepted and applied in many countries in the past decade. Through its use, food losses and food-borne diseases can be reduced significantly, and wider trade in many food items can be facilitated. The past five decades have witnessed a positive evolution on food irradiation according to the following: 1940`s: discovery of principles of food irradiation; 1950`s: initiation of research in advanced countries; 1960`s: research and development were intensified in some advanced and developing countries; 1970`s: proof of wholesomeness of irradiated foods; 1980`s: establishment of national regulations; 1990`s: commercialization and international trade. (Author)