Development of Reconstitution Technology for Surveillance Specimens

International Nuclear Information System (INIS)

Yasushi Atago; Shunichi Hatano; Eiichiro Otsuka

2002-01-01

The Japan Power Engineering and Inspection Corporation (JAPEIC) has been carrying out the project titled 'Nuclear Power Plant Integrated Management Technology (PLIM)' consigned by Japanese Ministry of Economy, Trade and Industry (METI) since 1996FY as a 10-years project. As one of the project themes, development of reconstitution technology for reactor pressure vessel (RPV/RV) surveillance specimens, which are installed in RPVs to monitor the neutron irradiation embrittlement on RPV/RV materials, is now on being carried out to deal with the long-term operation of nuclear power plants. The target of this theme is to establish the technical standard for applicability of reconstituted surveillance specimens including the reconstitution of the Charpy specimens and Compact Tension (CT) specimens. With the Charpy specimen reconstitution, application of 10 mm length inserts is used, which enables the conversion of tests from the LT-direction to the TL-direction. This paper presents the basic data from Charpy and CT specimens of RPV materials using the surveillance specimens obtained for un-irradiated materials including the following. 1) Reconstitution Technology of Charpy Specimens. a) The interaction between plastic zone and Heat Affected Zone (HAZ). b) The effects of the possible deviations from the standard specimens for the reconstituted specimens. 2) Reconstitution Technology of CT specimens. a) The correlation between fracture toughness and plastic zone width. Because the project is now in progress, this paper describes the outline of the results obtained as of the end of 2000 FY. (authors)

Charpy impact behavior of manganese-stabilized martensitic steels

International Nuclear Information System (INIS)

Hu, W.L.; Gelles, D.S.

1986-05-01

Tests were conducted to evaluate the irradiation-induced shift in ductile-to-brittle transition behavior of two manganese stabilized martensitic steels. Miniature Charpy specimens were fabricated from two heats of steel similar in composition to HT-9 but with 0.1% C and Mn contents ranging from 3.3 to 6.6.%. The 3.3% Mn steel showed a transition temperature similar to that of HT-9 in both the unirradiated condition and in specimens irradiated to 11.3 dpa. The steel containing 6.6% Mn exhibited a higher transition temperature after irradiation than the steel containing 3.3% Mn. The upper shelf energy (USE) after irradiation for the manganese stabilized alloys was much higher than for HT-9. 6 refs., 3 figs., 2 tabs

Charpy impact test results on five materials and NIST verification specimens using instrumented 2-mm and 8-mm strikers

International Nuclear Information System (INIS)

Nanstad, R.K.; Sokolov, M.A.

1995-01-01

The Heavy-Section Steel Irradiation Program at Oak Ridge National Laboratory is involved in two cooperative projects, with international participants, both of which involve Charpy V-notch impact tests with instrumented strikers of 2mm and 8mm radii. Two heats of A 533 grade B class I pressure vessel steel and a low upper-shelf (LUS) submerged-arc (SA) weld were tested on the same Charpy machine, while one heat of a Russian Cr-Mo-V forging steel and a high upper-shelf (HUS) SA weld were tested on two different machines. The number of replicate tests at any one temperature ranged from 2 to 46 specimens. Prior to testing with each striker, verification specimens at the low, high, and super high energy levels from the National Institute of Standards and Technology (NIST) were tested. In the two series of verification tests, the tests with the 2mm striker met the requirements at the low and high energy levels but not at the super high energy. For one plate, the 2mm striker showed somewhat higher average absorbed energies than those for the 8-mm striker at all three test temperatures. For the second plate and the LUS weld, however, the 2mm striker showed somewhat lower energies at both test temperatures. For the Russian forging steel and the HUS weld, tests were conducted over a range of temperatures with tests at one laboratory using the 8mm striker and tests at a second laboratory using the 2mm striker. Lateral expansion was measured for all specimens and the results are compared with the absorbed energy results. The overall results showed generally good agreement (within one standard deviation) in energy measurements by the two strikers. Load-time traces from the instrumented strikers were also compared and used to estimate shear fracture percentage. Four different formulas from the European Structural Integrity Society draft standard for instrumented Charpy test are compared and a new formula is proposed for estimation of percent shear from the force-time trace

Development of reconstitution technique of irradiated specimens. 3. Report for FY 1995 and FY 1996 on JAERI-IHI cooperated research program (joint research)

Energy Technology Data Exchange (ETDEWEB)

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

1998-10-01

The cooperated research between Japan Atomic Energy Research Institute and Ishikawajima-Harima Heavy Industries Co., Ltd. on the development of reconstitution technique of irradiated reactor pressure vessel surveillance specimens has been performed from FY 1993. In FY 1993-1994, the method of surface activated joining (SAJ) was applied to reconstitution of Charpy impact specimens. Some verification tests using unirradiated reactor pressure vessel plate materials have shown that SAJ is feasible for a reconstitution technique, in particular, owing to low joining temperature. The present paper reports the results of the cooperated research performed in FY 1995-1996. To improve the quality of the SAJ, the configuration of the end tab surface to be joined with the insert material was modified. The torque measured during joining was also introduced in joining parameters. A nondestructive inspection, temperature measurements in the specimens during joining were performed. The effect of joining on Charpy impact properties was discussed. For practical application of the technique to irradiated specimens, we confirmed that the impact specimens with joining interface gave rise to no failure at the joining position during impact test after neutron irradiation. (author)

The ARBOR irradiation project

International Nuclear Information System (INIS)

Petersen, C.; Shamardin, V.; Fedoseev, A.; Shimansky, G.; Efimov, V.; Rensman, J.

2002-01-01

The irradiation project 'ARBOR', for 'Associated Reactor Irradiation in BOR 60', includes 150 mini-tensile/low cycle fatigue specimens and 150 mini-Charpy (KLST) specimens of nine different RAFM steels. Specimens began irradiation on 22 November 2000 in an specially designed irradiation rig in BOR 60, in a fast neutron flux (>0.1 MeV) of 1.8x10 15 n/cm 2 s and with direct sodium cooling at a temperature less than 340 deg. C. Tensile, low cycle fatigue and Charpy specimens of the following materials are included: EUROFER 97, F82H mod., OPTIFER IVc, EUROFER 97 with different boron contents, ODS-EUROFER 97, as well as EUROFER 97 electron-beam welded and reference bulk material, from NRG, Petten

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

Application of computer techniques to charpy impact testing of irradiated pressure vessel steels

International Nuclear Information System (INIS)

Landow, M.P.; Fromm, E.O.; Perrin, J.S.

1982-01-01

A Rockwell AIM 65 microcomputer has been modified to control a remote Charpy V-notch impact test machine. It controls not only handling and testing of the specimen but also transference and storage of instrumented Charpy test data. A system of electrical solenoid activated pneumatic cylinders and switches provides the interface between the computer and the test apparatus. A command language has been designated that allows the operator to command checkout, test procedure, and data storage via the computer. Automatic compliance with ASTM test procedures is built into the program

The ARBOR irradiation project

Energy Technology Data Exchange (ETDEWEB)

Petersen, C. E-mail: claus.petersen@imf.fzk.de; Shamardin, V.; Fedoseev, A.; Shimansky, G.; Efimov, V.; Rensman, J

2002-12-01

The irradiation project 'ARBOR', for 'Associated Reactor Irradiation in BOR 60', includes 150 mini-tensile/low cycle fatigue specimens and 150 mini-Charpy (KLST) specimens of nine different RAFM steels. Specimens began irradiation on 22 November 2000 in an specially designed irradiation rig in BOR 60, in a fast neutron flux (>0.1 MeV) of 1.8x10{sup 15} n/cm{sup 2} s and with direct sodium cooling at a temperature less than 340 deg. C. Tensile, low cycle fatigue and Charpy specimens of the following materials are included: EUROFER 97, F82H mod., OPTIFER IVc, EUROFER 97 with different boron contents, ODS-EUROFER 97, as well as EUROFER 97 electron-beam welded and reference bulk material, from NRG, Petten.

Development of reconstitution method for surveillance specimens using surface activated joining

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Terumi; Kaihara, Shoichiro; Yoshida, Kazuo; Sato, Akira [Ishikawajima-Harima Heavy Industries Co. Ltd., Tokyo (Japan); Onizawa, Kunio; Nishiyama, Yutaka; Fukaya, Kiyoshi; Suzuki, Masahide

1996-03-01

Evaluation of embrittlement of reactor vessel steel due to irradiation requires surveillance tests. However, many surveillance specimens are necessary for nuclear plants life extension. Therefore, a specimen reconstitution technique has become important to provide the many specimens for continued surveillance. A surface activated joining (SAJ) method has been developed to join various materials together at low temperatures with little deformation, and is useful to bond irradiated specimens. To assess the validity of this method, Charpy impact tests were carried out, and the characteristics caused by heating during joining were measured. The test results showed the Charpy impact values were almost the same as base materials, and surface activated joining reduced heat affected zone to less than 2 mm. (author).

Specimen size effects in Charpy impact testing

International Nuclear Information System (INIS)

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

1989-01-01

Full-size , half-size, and third-size specimens from several different steels have been tested as part of an ongoing alloy development program. The smaller specimens permit more specimens to be made from small trail heats and are much more efficient for irradiation experiments. The results of several comparisons between the different specimen sizes have shown that the smaller specimens show qualitatively similar behavior to large specimens, although the upper-shelf energy level and ductile-to-ductile transition temperature are reduced. The upper-shelf energy levels from different specimen sizes can be compared by using a simple volume normalization method. The effect of specimen size and geometry on the ductile-to-ductile transition temperature is more difficult to predict, although the available data suggest a simple shift in the transition temperature due to specimen size changes.The relatively shallower notch used in smaller specimens alters the deformation pattern, and permits yielding to spread back to the notched surface as well as through to the back. This reduces the constraint and the peak stresses, and thus the initiation of cleavage is more difficult. A better understanding of the stress and strain distributions is needed. 19 refs., 3 figs., 3 tabs

Evaluation and uncertainty estimates of Charpy-impact data

International Nuclear Information System (INIS)

Stallman, F.W.

1982-01-01

Shifts in transition temperature and upper-shelf energy from Charpy tests are used to determine the extent of radiation embrittlement in steels. In order to determine these parameters reliably and to obtain uncertainty estimates, curve fitting procedures need to be used. The hyperbolic tangent or similar models have been proposed to fit the temperature-impact-energy curve. These models are not based on the actual fracture mechanics and are indeed poorly suited in many applications. The results may be falsified by forcing an inflexible curve through too many data points. The nonlinearity of the fit poses additional problems. In this paper, a simple linear fit is proposed. By eliminating data which are irrelevant for the determination of a given parameter, better reliability and accuracy can be achieved. Additional input parameters like fluence and irradiation temperature can be included. This is important if there is a large variation of fluence and temperature in different test specimens. The method has been tested with Charpy specimens from the NRC-HSST experiments

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)

Application of subsize specimens in nuclear plant life extension

International Nuclear Information System (INIS)

Rosinski, S.T.; Kumar, A.S.; Cannon, N.S.; Hamilton, M.L.

1993-01-01

The US Department of Energy is sponsoring a research effort through Sandia National Laboratories and the University of Missouri-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. The correlation methodology is based on the partitioning of the USE into crack initiation and crack propagation energies. To accomplish this partition, both precracked and notched-only specimens will be used. Whereas the USE of notched-only specimens is the sum of both crack initiation and crack propagation energies, the USE of precracked specimens reflects only the crack propagation component. The difference in the USE of the two types of specimens represents a measure of the crack initiation energy. Normalizing the values of the crack initiation energy to the fracture volume of the sample produces similar values for the full-, half-, and third-size specimens. In addition, the ratios of the USE and the crack propagation energy are also in agreement for full-, half-, and third-size specimens. These two observations will be used to predict the USE of full-size specimens based on subsize USE data. This paper provides details of the program and presents 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

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.

Comparison between instrumented precracked Charpy and compact specimen tests of carbon steels

International Nuclear Information System (INIS)

Nanstad, R.K.

1980-01-01

The General Atomic Company High Temperature Gas-Cooled Reactor (HTGR) is housed within a prestressed concrete reactor vessel (PCRV). Various carbon steel structural members serve as closures at penetrations in the vessel. A program of testing and evaluation is underway to determine the need for reference fracture toughness (K/sub IR/) and indexing procedures for these materials as described in Appendix G to Section III, ASME Code for light water reactor steels. The materials of interest are carbon steel forgings (SA508, Class 1) and plates (SA537, Classes 1 and 2) as well as weldments of these steels. The fracture toughness behavior is characterized with instrumented precracked Charpy V-votch specimens (PCVN) - slow-bend and dynamic - and compact specimens (10-mm and 25-mm thicknesses) using both linear elastic (ASTM E399) and elastic-plastic (equivalent Energy and J-Integral) analytical procedures. For the dynamic PCVN tests, force-time traces are analyzed according to the procedures of the Pressure Vessel Research Council (PVRC)/Metal Properties Council (MPC). Testing and analytical procedures are discussed and PCVN results are compared to those obtained with compact specimens

Instrumentation of a Charpy-pendulum. Additional data obtained from it and its application to nuclear reactor pressure vessels surveillance programs; Instrumentacion de un pendulo Charpy. Datos adicionales obtenidos a partir de la misma y su aplicacion a programas de vigilancia de centrales nucleares

Energy Technology Data Exchange (ETDEWEB)

Chomik, Enrique P; Dhers, Horacio; Iorio, Antonio F [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Materiales; Ciriani, Dario F [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Combustibles Nucleares

1999-07-01

Charpy test gives information about a material dynamic fracture behavior. In a plain Charpy test, this information is the absorbed energy during fracture of the specimen, lateral deformation and the percentage of ductile fracture of the specimen. These parameters can then be used for the determination of the material response to a dynamic applied load, and are used at present to determine the brittle-ductile transition temperature of a material. However, there is a lot of additional information that can be obtained from a Charpy test, which is vital for the case of surveillance programs of nuclear power plants, where it is necessary to get the most available information from the specimens to be tested, because each one of them was irradiated for many years under temperature and neutronic flux conditions similar to that of the internal surface of the reactor pressure vessel, which converts these specimens in unique and very expensive ones. This additional information can be obtained from the curve that determines the evolution of the applied force to the specimen throughout the time involved in its fracture. It was possible to instrument a Charpy pendulum at a fraction of the cost necessary to buy an instrumentation package like the ones available in the market, and since the instrumentation equipment obtained is easy to transport. It has the additional advantage that can be used to instrument any other pendulum replacing only the hammer of the pendulum with a instrumented one for that pendulum. (author)

Use of the strength ratio for pre-cracked Charpy specimens for the measuring, of the dynamic toughness of steels

International Nuclear Information System (INIS)

Pereira, L.C.; Darwish, F.A.I.

1981-01-01

The specimen strength ratio (R sub(sb)) determined for precraked Charpy specimens fractured in dynamic bending was correlated with plane strain fracture toughness (K sub(Id)) obtained through valid measurements of the J-integral at the moment of fracture initiation in various microstructures of the AISI 4140 steel. The results indicate a linear relationship between K sub(Id) and R sub(sb) for the microstructures considered in this work. The range of validity of this linear correlation is presented and discussed in terms of the ASTM E399 specimen size criterion. (Author) [pt

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.

Change of Charpy impact fracture behavior of precracked ferritic specimens due to thermal aging in sodium

International Nuclear Information System (INIS)

Hu, W.L.

1985-12-01

A series of tests were conducted to evaluate the effect of sodium on the impact fracture behavior of precracked Charpy specimens made of HT-9 weldment. One set of samples was precracked prior to sodium aging and the other set was precracked after aging in sodium. Both set of specimens exhibited the same DBTT. Samples precracked prior to sodium exposure, however, showed a 40% reduction in the upper shelf energy (USE) as compared to the set precracked after aging. The results suggest that the fracture toughness of the material may be reduced if an existing crack was soaked in sodium at elevated temperature for a period of time

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

International Nuclear Information System (INIS)

Toivonen, A.; Moilanen, P.; Pyykkoenen, M.; Taehtinen, S.; Rintamaa, R.; Saario, T.

1998-01-01

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.)

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.)

Instrumentation of a Charpy-pendulum. Additional data obtained from it and its application to nuclear reactor pressure vessels surveillance programs

International Nuclear Information System (INIS)

Chomik, Enrique P.; Dhers, Horacio; Iorio, Antonio F.; Ciriani, Dario F.

1999-01-01

Charpy test gives information about a material dynamic fracture behavior. In a plain Charpy test, this information is the absorbed energy during fracture of the specimen, lateral deformation and the percentage of ductile fracture of the specimen. These parameters can then be used for the determination of the material response to a dynamic applied load, and are used at present to determine the brittle-ductile transition temperature of a material. However, there is a lot of additional information that can be obtained from a Charpy test, which is vital for the case of surveillance programs of nuclear power plants, where it is necessary to get the most available information from the specimens to be tested, because each one of them was irradiated for many years under temperature and neutronic flux conditions similar to that of the internal surface of the reactor pressure vessel, which converts these specimens in unique and very expensive ones. This additional information can be obtained from the curve that determines the evolution of the applied force to the specimen throughout the time involved in its fracture. It was possible to instrument a Charpy pendulum at a fraction of the cost necessary to buy an instrumentation package like the ones available in the market, and since the instrumentation equipment obtained is easy to transport. It has the additional advantage that can be used to instrument any other pendulum replacing only the hammer of the pendulum with a instrumented one for that pendulum. (author)

Charpy impact test results of ferritic alloys from the HFIR[High Flux Isotope Reactor]-MFE-RB2 test

International Nuclear Information System (INIS)

Hu, W.L.; Gelles, D.S.

1987-03-01

Miniature Charpy specimens of HT-9 in base metal, weld metal and heat affected zone (HAZ) metal conditions, and 9Cr-1Mo in base metal and weld metal conditions have been tested following irradiation in HFIR-MFE-RB2 at 55 0 C to ≅10 dpa. All specimen conditions have degraded properties (both DBTT and USE) in comparison with specimens irradiated to lower dose. 9Cr-Mo degraded more than HT-9 and weld metal performed worse than base metal which performed worse than HAZ material. Property degradation was approximately linear as a function of dose, indicating that degradation response had not saturated by 10 dpa

Conclusions regarding fracture mechanics testing and evaluation of small specimens - As evidenced by the finnish contribution to the IAEA CRP3 programme

Energy Technology Data Exchange (ETDEWEB)

Wallin, K; Valo, M; Rintamaa, R; Torronen, K [Technical Research Centre of Finland, Espoo (Finland); Ahlstrand, R [Imatran Voima Oy (IVO), Helsinki (Finland)

1994-12-31

An extensive mechanical property evaluation has been carried out on various specimens (a Japanese steel plate (JRQ), a French forging material (FFA) and a Japanese forging material (JFL)) in the as-received and irradiated conditions. The mechanical properties measured at different temperatures include Charpy-V notch and instrumented pre-cracked Charpy data and static and dynamic elastic-plastic fracture toughness based on the J-integral, with various specimen size and geometry. Test analysis lead to conclusions regarding the use of small specimen fracture mechanical tests for investigating irradiation effects: CVN{sub pc} and RCT type specimens are suitable for determining the materials fracture toughness even in the ductile/brittle transition region provided the elastic-plastic parameter K{sub JC} is applied together with a statistical size correction. These two specimen types yield equivalent results for the fracture toughness transition shift. Charpy-V appears not to be suitable for estimating the static fracture toughness transition shift. 8 refs., 11 figs.

Prediction of the brittle fracture toughness value of a RPV steel from the analysis of a limited set of Charpy results

International Nuclear Information System (INIS)

Forget, P.; Marini, B.; Verdiere, N.

2001-01-01

Our objective is to establish a method to be able to determine fracture toughness of a reactor pressure vessel (RPV) by using the small number of Charpy specimens used in the reactor surveillance program. Previous studies have shown that it is possible to determine fracture toughness from Charpy tests. Another point is to determine if statistical effects are compatible with a restricted number of specimens, this paper deals with this point and presents a methodology that is applicable to the case of irradiated materials from the surveillance program. Several conclusions can be drawn from this study: -) When determining failure parameters, we gain most accuracy by increasing the number of samples from 3 to about 6; -) it is possible to evaluate brittle fracture toughness using local approach, either by using Beremin or Renevey model; -) The effect of using a small number of Charpy specimens to determine fracture toughness in brittle fracture is evaluated. The error in the evaluation of fracture toughness is much smaller than the experimental dispersion itself. (A.C.)

Charpy V, an application in Mat lab

International Nuclear Information System (INIS)

Castillo M, J.A.; Torres V, M.

2003-01-01

The obtained results with the system Charpy V 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)

Irradiation Effects at 160-240 deg C in Some Swedish Pressure Vessel Steels

Energy Technology Data Exchange (ETDEWEB)

Grounes, M [AB Atomenergi, Nykoeping (Sweden); Myers, H P [Chalmers Institute of Technology, Goeteborg (Sweden); Hannerz, N E [Motala Verkstads AB, Motala (Sweden)

1967-09-15

Tensile specimens, Charpy impact specimens and miniature impact specimens of six steels in different conditions were irradiated to 2.8 x 10{sup 18} and 5.6 x 10{sup 18} n/cm{sup 2} (> 1 MeV) at 160-240 deg C. The steels investigated were SIS 142103, 2103/R3, NO 345, Fortiweld, Fortiweld HS and OK 54 P. There is no correlation between the increase in transition temperature and initial transition temperature. However, changes in strength and ductility can be correlated to the initial yield strength. The increases in transition temperature due to strain aging and irradiation are approximately additive. The irradiation-induced changes in 2103/R3 and Fortiweld HS steels do not vary with position in the thickness of the plate. Different tempering treatments in Fortiweld HS steel do not change the extent of irradiation effects. Normal Charpy V-notch impact specimens and miniature specimens give the same irradiation-induced increase in transition temperature.

Irradiation Effects at 160-240 deg C in Some Swedish Pressure Vessel Steels

International Nuclear Information System (INIS)

Grounes, M.; Myers, H.P.; Hannerz, N.E.

1967-09-01

Tensile specimens, Charpy impact specimens and miniature impact specimens of six steels in different conditions were irradiated to 2.8 x 10 18 and 5.6 x 10 18 n/cm 2 (> 1 MeV) at 160-240 deg C. The steels investigated were SIS 142103, 2103/R3, NO 345, Fortiweld, Fortiweld HS and OK 54 P. There is no correlation between the increase in transition temperature and initial transition temperature. However, changes in strength and ductility can be correlated to the initial yield strength. The increases in transition temperature due to strain aging and irradiation are approximately additive. The irradiation-induced changes in 2103/R3 and Fortiweld HS steels do not vary with position in the thickness of the plate. Different tempering treatments in Fortiweld HS steel do not change the extent of irradiation effects. Normal Charpy V-notch impact specimens and miniature specimens give the same irradiation-induced increase in transition temperature

A new method for improving the reliability of fracture toughness surveillance of nuclear pressure vessel by neutron irradiated embrittlement

International Nuclear Information System (INIS)

Zhang Xinping; Shi Yaowu

1992-01-01

In order to obtain more information from neutron irradiated sample specimens and raise the reliability of fracture toughness surveillance test, it has more important significance to repeatedly exploit the broken Charpy-size specimen which had been tested in surveillance test. In this work, on the renewing design and utilization for Charpy-size specimens, 9 data of fracture toughness can be gained from one pre-cracked side-grooved Charpy-size specimen while at the preset usually only 1 to 3 data of fracture toughness can be obtained from one Chharpy-size specimen. Thus, it is found that the new method would obviously improve the reliability of fracture toughness surveillance test and evaluation. Some factors which affect the reasonable design of pre-cracked deep side-groove Charpy-size compound specimen have been discussed

Testing of irradiated and annealed 15H2MFA materials

International Nuclear Information System (INIS)

Gillemot, F.; Uri, G.

1994-01-01

A set of surveillance samples made from 15H2MFA material has been studied in the laboratory of AEKI. Miniature notched tensile specimens were cut from some remnants of irradiated and broke surveillance charpy remnants. The Absorbed Specific Fracture Energy (ASFE) was measured on the specimens. A cutting machine and testing technique were elaborated for the measurements. The second part of the Charpy remnants was annealed at 460 deg. C and 490 deg. C for 6-8 hours. The specimens were tested similarity and the results were compared. (author). 5 refs, 9 figs

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.

Computer simulation of plastic deformation in the Charpy V-notch impact test

International Nuclear Information System (INIS)

Norris, D.M. Jr.; Quinones, D.F.; Moran, B.

1978-01-01

Calculations describe the dynamic stress and strain states in the standard Charpy specimen from impact to the start of cracking. We model A533 Grade B Class 1 nuclear-pressure-vessel steel at 100 0 C with an elastic-plastic constitutive law. Large deformation and rotation of the material are accounted for. The specimen velocity field during the impact transient is presented and how the early wave effects cause separation of the specimen from the striker is shown. The calculations show why correlations between Charpy fracture energy and fracture toughness have been largely unsuccessful and suggest methods to improve these correlations using the same specimen geometry

Use of miniaturized compact tension specimens for fracture toughness measurements in the upper shelf regime. Electrabel/Tractebel-SCK-CEN Convention 2004 Task 1.1.4/2

International Nuclear Information System (INIS)

Lucon, E.; Scibetta, M.; Chaouadi, R.; Walle, E. van

2005-04-01

In the nuclear field, the importance of direct fracture toughness measurements on RPV materials has been nowadays widely recognized, as opposed to Charpy-based estimations. However, sample dimensions have to be kept small in order to optimize the use of available material (often in the form of previously broken Charpy specimens) or, in the case of new irradiations, make effective use of the limited space available inside irradiation facilities. One of the most appealing geometries for fracture toughness measurements is the miniature Compact Tension specimen, MC(T), which has the following dimensions: B = 4.15 mm, W = 8.3 mm, cross section 10 x 10 mm 2 . Four MC(T) specimens can be machined out of a broken half Charpy, and in the case of irradiation ten MC(T) samples occupy approximately the same volume as a full-size Charpy specimen. The MC(T) geometry was already successfully applied and qualified for fracture toughness assessments in the ductile-to-brittle transition regime, using the Master Curve method (ASTM E1921-03). A further, comprehensive investigation is presented in this report, aimed at assessing the applicability of MC(T) specimens to measure fracture toughness in fully ductile (upper shelf) conditions. In this study, 18 1TC(T) and 20 MC(T) specimens have been tested at different temperatures from three RPV steels and one low-alloy C-Mn steel. The results obtained clearly show that MC(T) samples exhibit lower fracture toughness properties, both in terms of initiation of ductile tearing (according to various test standards) and resistance to ductile crack propagation (J-R curve). The reduction of tearing resistance might be attributed to work hardening prevailing over loss of constraint in the uncracked ligament in a side-grooved specimen, or to the inadequacy of J-integral to represent ductile crack extension in very small specimens. Both arguments will have to be verified with further investigations. (author)

Irradiation effects in strain aged pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

Grounes, M; Myers, H P

1962-02-15

Tensile specimens, Charpy-V notch and subsize impact specimens of an aluminium killed carbon manganese steel, have been irradiated at 160 - 190 deg C in the reactor G1. The total neutron dose received was 2.4 x 10{sup 18} n/cm{sup 2} (> 1 MeV). Specimens were prepared from normalized plate and from strain aged material from the same plate. It was found that the changes in brittle ductile transition temperature due to neutron irradiation and those due to strain ageing must be considered additive.

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.

Charpy impact test results of ferritic alloys at a fluence of 6 x 1022n/cm2

International Nuclear Information System (INIS)

Hu, W.L.

1985-01-01

Charpy impact tests on specimens in the AD-2 reconstitution experiment were completed. One hundred ten specimens made of HT-9 base metal, 9Cr-1Mo base metal and 9Cr-1Mo weldment at various heat treatment conditions were tested in temperature range from -73 0 C to 260 0 C. The specimens were irradiated from 390 0 C to 550 0 C and the fluence of the specimens reached 6 x 10 22 n/cm 2 . This is the first time that the transition behavior of ferritic alloys at high fluence was obtained. This is also the first time that comprehensive results on the irradiated 9Cr-1Mo weldment are available. The test results show a small additional shift in transition temperature for HT-9 base metal irradiated at 390 0 C and 450 0 C as the fluence was raised to 6 x 10 22 n/cm 2 . At higher irradiation temperatures, however, the shift in transition temperature is less conclusive. Further reduction in USE was observed at higher fluence for all the irradiation temperatures. There is no apparent fluence effect for 9Cr-1Mo base metal at all the irradiation temperatures studied. Contrary to the previous finding on HT-9 base metal and weldment, the 9Cr-1Mo weldment shows a higher transition temperature ( + 60 0 C) and a higher USE ( + 100%) as compared to the 9Cr-1MO base metal for the same irradiation conditions. 6 references, 7 figures, 7 tables

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.

The modelling of irradiation embrittlement in submerged-arc welds

International Nuclear Information System (INIS)

Bolton, C.J.; Buswell, J.T.; Jones, R.B.; Moskovic, R.; Priest, R.H.

1996-01-01

Until very recently, the irradiation embrittlement behavior of submerged-arc welds has been interpreted in terms of two mechanisms, namely a matrix damage component and an additional component due to the irradiation-enhanced production of copper-rich precipitates. However, some of the weld specimens from a recent accelerated re-irradiation experiment have shown high Charpy shifts which exceeded the values expected from the measured shift in yield stress. Microstructural examination has revealed the occurrence of intergranular fracture (IGF) in these specimens, accompanied by grain boundary segregation of phosphorus. Theoretical models were developed to predict the parametric dependence of irradiation-enhanced phosphorus segregation on experimental variables. Using these parametric forms, along with the concept of a critical level of segregation for the onset of IGF instead of cleavage, a three mechanism trend curve has been developed. The form of this trend curve, taking into account IGF as well as matrix and copper embrittlement, is thus mechanistically based. The constants in the equation, however, are obtained by a statistical fit to the actual Charpy shift database

Small specimen test technology of fracture toughness in structural material F82H steel for fusion nuclear reactors

International Nuclear Information System (INIS)

Wakai, Eiichi; Ohtsuka, Hideo; Jitsukawa, Shiro; Matsukawa, Shingo; Ando, Masami

2006-03-01

Small specimen test technology (SSTT) has been developed to investigate mechanical properties of nuclear materials. SSTT has been driven by limited availability of effective irradiation volumes in test reactors and accelerator-based neutron and charged particle sources, and it is very useful for the reduction of waste materials produced in nuclear engineering. In this study new bend test machines have been developed to obtain fracture behaviors of F82H steel for very small bend specimens of pre-cracked t/2-1/3CVN (Charpy V-notch) with 20 mm-length and DFMB (deformation and fracture mini bend specimen) with 9 mm-length and disk compact tension of 0.18DCT type, and fracture behaviors were examined to evaluate DBTT (ductile-brittle transition temperature) at temperature from -180 to 25degC. The effect of specimen size on DBTT of F82H steel was also examined by using Charpy type specimens such as 1/2t-CVN, 1/3CVN and t/2-1/3CVN. In this paper, it also provides the information of the specimens irradiated at 250degC and 350degC to about 2 dpa in the capsule of 04M-67A and 04M-68A of JMTR experiments. (author)

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

Irradiation effects on fracture toughness of two high-copper submerged-arc welds, HSSI Series 5

International Nuclear Information System (INIS)

Nanstad, R.K.; Haggag, F.M.; McCabe, D.E.; Iskander, S.K.; Bowman, K.O.; Menke, B.H.

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 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 x 10 19 neutrons/cm 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 √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 min parameter which affects the curve shape

Evaluation of irradiated coating material specimens

International Nuclear Information System (INIS)

Lee, Yong Jin; Nam, Seok Woo; Cho, Lee Moon

2007-12-01

Evaluation result of irradiated coating material specimens - Coating material specimens radiated Gamma Energy(Co 60) in air condition. - Evaluation conditions was above 1 X 10 4 Gy/hr, and radiated TID 2.0 X 10 6 Gy. - The radiated coating material specimens, No Checking, Cracking, Flaking, Delamination, Peeling and Blistering. - Coating system at the Kori no. 1 and APR 1400 Nuclear power plant, evaluation of irradiated coating materials is in accordance with owner's requirement(2.0 X 10 6 Gy)

Modeling and simulation of Charpy impact test of maraging steel 300 using Abaqus

Science.gov (United States)

Madhusudhan, D.; Chand, Suresh; Ganesh, S.; Saibhargavi, U.

2018-03-01

This work emphasizes the modeling and simulation of Charpy impact test to evaluate fracture energy at different pendulum velocities of armor maraging steel 300 using ABAQUS. To evaluate the fracture energy, V-notch specimen is fractured using the Johnson and Cook Damage model. The Charpy impact tests are of great importance related to fracture properties of steels. The objective of this work is to present absorbed energy variation at pendulum velocities of 5 m/sec, 6 m/sec, 7 m/sec and 9 m/sec in addition to stress distribution at v-notch. Finite Element Method of modeling for three dimensional specimens is used for simulation in commercial software of ABAQUS.

Investigation of Ductile-to-Brittle Transition of RPV Materials by using the Pre-cracked Charpy Impact Data

International Nuclear Information System (INIS)

Lee, Ho Jin; Lee, Bong Sang; Hong, Jun Hwa

2005-01-01

Much recent work in the field of elastic-plastic fracture mechanics has been directed to developing a mechanics-based relationship between the onset of cleavage fracture in structural components and that of Charpy V-notch specimens. The assessing processes of the cracks located in the reactor pressure vessel (RPV) is described in the ASME code Sec. III, App. G and Sec. XI, App. A. The RTNDT obtained from the impact test using standard Charpy V-notch (CVN) specimens is used as a reference temperature to assess the integrity of RPV materials. The initial RTNDT, for the Linde 80 weld, was determined by the 67.8J Charpy impact energy instead of drop weight test. Generally, Linde 80 weld has low upper-shelf energy. The initial RTNDT obtained from the Charpy impact energy curve has been considered overly conservative. Recently, master curve method has been investigated to assess the integrity of RPV materials directly. The initial RTT0 obtained from the master curve method is considered more realistic than the initial RTNDT obtained from impact test for low upper-shelf fracture toughness RPV materials. In this research, the correlation of transition regions between the master curves and the Charpy impact energy curves was investigated using the dynamic fracture toughness curve and the impact energy curve obtained from the impact test of pre-cracked Charpy (PCC) specimens. For the low toughness RPV material the ductile-to-brittle transition corresponding to the static master curve was anticipated using the invested correlation

Fractography evaluation of impact and tensile specimens from the HFBR [High Flux Beam Reactor

International Nuclear Information System (INIS)

Czajkowski, C.J.

1989-10-01

The Materials Technology Group of the Department of Nuclear Energy (DNE) at Brookhaven National Laboratory (BNL) has performed a fractographic examination of neutron irradiated and unirradiated tensile and Charpy ''V'' notch specimens. The evaluation was carried out using a scanning electron microscope (SEM) to evaluate the fracture mode. Photomicrographs were then evaluated to determine the extent of ductility present on the fracture surfaces of the unirradiated specimens. Ductility area measurements ranged from 4.6--9.5% on typical photomicrographs examined. 12 figs

Re-utilization by '' Stud Welding'' of capsules charpy-V belonged to surveillance programs

International Nuclear Information System (INIS)

Lapena, J.; Perosanz, F. J.; Gachuz, M.

1998-01-01

The perspectives of nuclear plants life extension that are approximating to their end of design life compels to make new surveillance programs. The re-utilization of specimens belonging to surveillance capsules already tested in these new surveillance programs seems be a solution worldwide accepted. The two possible re-utilization processes of this irradiated material are: Subsized specimens and Reconstitution. While the first alternative (Subsized specimens) outlines serious problems for apply the results, the reconstitution eliminates this problem, since the resulting specimens after of the reconstruction procedure would be of the same dimensions that the original. The reconstruction process involves welds, and therefore it has associated the specific problems of this type of joints. Furthermore, by be tried to material irradiated with certain degree of internal damage, that is the variable to evaluate, requires that the heat contribution to the piece not originate local thermal treatments that alter its mechanical qualities. In this work has been followed the evolution by the variables of the weld process and their influence on the quality by the union from metallographic al point of view as well as mechanical for a weld procedure by Stud Welding. The principal objective is to optimize said parameters to assure a good mechanical continuity, without detriment of the microstructural characteristics of the original material. To verify this last have been accomplished with metallographical tests, temperature profile, hardness and will be carried out also Charpy tests. (Author)

Empirical correlation between mechanical and physical parameters of irradiated pressure vessel steels

International Nuclear Information System (INIS)

Tipping, P.; Solt, G.; Waeber, W.

1991-02-01

Neutron irradiation embrittlement of nuclear reactor pressure vessel (PV) steels is one of the best known ageing factors of nuclear power plants. If the safety limits set by the regulators for the PV steel are not satisfied any more, and other measures are too expensive for the economics of the plant, this embrittlement could lead to the closure of the plant. Despite this, the fundamental mechanisms of neutron embrittlement are not yet fully understood, and usually only empirical mathematical models exist to asses neutron fluence effects on embrittlement, as given by the Charpy test for example. In this report, results of a systematic study of a French forging (1.2 MD 07 B), irradiated to several fluences will be reported. Mechanical property measurements (Charpy tensile and Vickers microhardness), and physical property measurements (small angle neutron scattering - SANS), have been done on specimens having the same irradiation or irradiation-annealing-reirradiation treatment histories. Empirical correlations have been established between the temperature shift and the decrease in the upper shelf energy as measured on Charpy specimens and tensile stresses and hardness increases on the one hand, and the size of the copper-rich precipitates formed by the irradiation on the other hand. The effect of copper (as an impurity element) in enhancing the degradation of mechanical properties has been demonstrated; the SANS measurements have shown that the size and amount of precipitates are important. The correlations represent the first step in an effort to develop a description of neutron irradiation induced embrittlement which is based on physical models. (author) 6 figs., 27 refs

Applicability of the fracture toughness master curve to irradiated reactor pressure vessel steels

International Nuclear Information System (INIS)

Sokolov, M.A.; McCabe, D.E.; Alexander, D.J.; Nanstad, R.K.

1997-01-01

The current methodology for determination of fracture toughness of irradiated reactor pressure vessel (RPV) steels is based on the upward temperature shift of the American Society of Mechanical Engineers (ASME) K Ic curve from either measurement of Charpy impact surveillance specimens or predictive calculations based on a database of Charpy impact tests from RPV surveillance programs. Currently, the provisions for determination of the upward temperature shift of the curve due to irradiation are based on the Charpy V-notch (CVN) 41-J shift, and the shape of the fracture toughness curve is assumed to not change as a consequence or irradiation. The ASME curve is a function of test temperature (T) normalized to a reference nit-ductility temperature, RT NDT , namely, T-RT NDT . That curve was constructed as the lower boundary to the available K Ic database and, therefore, does not consider probability matters. Moreover, to achieve valid fracture toughness data in the temperature range where the rate of fracture toughness increase with temperature is rapidly increasing, very large test specimens were needed to maintain plain-strain, linear-elastic conditions. Such large specimens are impractical for fracture toughness testing of each RPV steel, but the evolution of elastic-plastic fracture mechanics has led to the use of relatively small test specimens to achieve acceptable cleavage fracture toughness measurements, K Jc , in the transition temperature range. Accompanying this evolution is the employment of the Weibull distribution function to model the scatter of fracture toughness values in the transition range. Thus, a probabilistic-based bound for a given data population can be made. Further, it has been demonstrated by Wallin that the probabilistic-based estimates of median fracture toughness of ferritic steels tend to form transition curves of the same shape, the so-called ''master curve'', normalized to one common specimen size, namely the 1T [i.e., 1.0-in

LPTR irradiation of LLL vanadium tensile specimens and LLL Nb--1Zr tensile specimens

International Nuclear Information System (INIS)

MacLean, S.C.; Rowe, C.L.

1977-01-01

The LPTR irradiation of 14 LLL vanadium tensile specimens and 14 LLL Nb-1Zr tensile specimens is described. Sample packaging, the irradiation schedule and neutron fluences for three energy ranges are given

Development of a reconstitution system of Charpy probes for the surveillance of vessels in nucleo electric plants

International Nuclear Information System (INIS)

Romero C, J.; Hernandez, R.; Fernandez, F.; Gonzalez M, A.

2007-01-01

This work describes the development of a welding system, for the rebuilding of halves of Charpy test tubes, the rebuilding consists on welding two implants in those ends of these halves of test tubes, in these welding the main requirement is not to alter the mechanical properties in a minimum volume of 1 cm 3 , the rebuilding is medullary in the surveillance programs of the reactor vessel. In these programs, the mechanical state of the vessel is evaluated, for it there are surveillance capsules with a Charpy witness test tubes series, subjected to a neutron flow similar or bigger to that of the vessel. The objective is to evaluate in advance on the vessel fragilization grade its life design. However the number of capsules with the witness test tubes it is only for the plant design life and at the moment the nucleo electric, negotiates an extension of life of these, until for 20 more years, of there the importance of this material witness's that stores the information of the damage accumulated by the neutron flow. This material requires to be taken advantage it after being rehearsed and the normative one settles down as obligatory to qualify the rebuilding process with all the requirements settled down in the ASTM Designation: E 1253-99 'Standard Guide for Reconstitution of irradiated Charpy-Sized Specimens', to obtain other reconstituted Charpy test tubes that are again introduced in the reactor. When being reconstituted the halves of the original test tubes it is obtained double reconstituted Charpy test tubes. Half of the test tubes they are used in the surveillance program of the vessel, with the surpluses test tubes, it can determine the fracture toughness, property of the material used in the extension methodology of life of vessel. (Author)

The flow effect in the irradiation embrittlement in pressure vessel steels of nuclear power plants

International Nuclear Information System (INIS)

Kempf, Rodolfo A.; Cativa Tolosa, Sebastian; Fortis, Ana M.

2009-01-01

This paper deals with the advances in the study of the mechanical behavior of the Reactor Pressure Vessel steels under accelerate irradiations. The objective is to study the effect of lead factors on the interpretation of the mechanisms that induced the embrittlement of the RPV, like those of the reactors Atucha II and CAREM. It is described a device designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. It is presented also an automatic digital image processing technique for partitioning Charpy fracture surface into regions with a clear physical meaning and appropriate for the work in hot cells. The aim is to obtain the fracture behavior of irradiated specimens with different lead factors in the range of high fluencies and to know the dependence with the composition of the alloy and with the diffusion of other alloy elements. (author)

Computer simulation of the Charpy V-notch toughness test

International Nuclear Information System (INIS)

Norris, D.M. Jr.

1977-01-01

The dynamic Charpy V-notch test was simulated on a computer. The calculational models (for A-533 Grade B class 1 steel) used both a rounded and a flat-tipped striker. The notch stress/strain state was found to be independent of the three-point loading type and was most strongly correlated with notch-opening displacement. The dynamic stress/strain state at the time of fracture initiation was obtained by comparing the calculated deformed shape with that obtained in interrupted Charpy V-notch tests where cracking had started. The calculation was also compared with stress/strain states calculated in other geometries at failure. The distribution and partition of specimen energy was calculated and adiabatic heating and strain rate are discussed

Effect of specimen size on the upper shelf energy of ferritic steels

International Nuclear Information System (INIS)

Kumar, A.S.

1990-01-01

A methodology is proposed that can be used to predict the upper shelf energy (USE) of ferritic steels based on subsize specimen data. The proposed methodology utilizes the partitioning of the USE into energies required for crack initiation and crack propagation. Notched-only Charpy specimens are used in conjunction with precracked specimens to separate the two components. An unirradiated ferritic steel, HT-9, was used to demonstrate the validity of the methodology. Unlike previous correlations that were limited in their applicability to either highly ductile or brittle material, the proposed methodology is expected to be applicable over a wide range of ductility and to be particularly useful for materials that harden significantly during irradiation

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

Temper embrittlement, irradiation induced phosphorus segregation and implications for post-irradiation annealing of reactor pressure vessels

International Nuclear Information System (INIS)

McElroy, R.J.; English, C.A.; Foreman, A.J.; Gage, G.; Hyde, J.M.; Ray, P.H.N.; Vatter, I.A.

1999-01-01

Three steels designated JPB, JPC and JPG from the IAEA Phase 3 Programme containing two copper and phosphorus levels were pre- and post-irradiation Charpy and hardness tested in the as-received (AR), 1200 C/0.5h heat treated (HT) and heat treated and 450 C/2000h aged (HTA) conditions. The HT condition was designed to simulate coarse grained heat-affected zones (HAZ's) and showed a marked sensitivity to thermal ageing in all three alloys. Embrittlement after thermal ageing was greater in the higher phosphorus alloys JPB and JPG. Charpy shifts due to thermal ageing of between 118 and 209 C were observed and accompanied by pronounced intergranular fracture, due to phosphorus segregation. The irradiation embrittlement response was complex. The low copper alloys, JPC and JPB, in the HT and HTA condition exhibited significant irradiation induced Charpy shift but very low or even negative hardness changes indicating non-hardening embrittlement. The higher copper alloy, JPG, also exhibited irradiation hardening in line with its copper content. Fractographic and microchemical studies indicated irradiation induced phosphorus segregation and a transition from cleavage to intergranular failure at grain boundary phosphorus concentrations above a critical level. The enhanced grain boundary phosphorus level increased with dose in agreement with a kinetic segregation model developed at Harwell. The relevance of the thermal ageing studies to RPV Annealing for Plant-Life Extension was identified early in the program. It is of concern that annealing of RPV's has been performed, or is proposed, at temperatures in the range 425--475 C for periods of about 1 week (168h). Much attention has been given to the use of in-situ hardness measurements and machining miniature Charpy and tensile specimens from belt-line plate and weld materials. However, HAZ's, often containing higher phosphorus levels than the present materials, have largely been ignored. A post-irradiation annealing (PIA

Anomalous fracture toughness of irradiated Cr-MoV - Reactor pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

Ahistrand, R [Imatran Voima Oy (IVO), Helsinki (Finland)

1994-12-31

The base metal Crack Opening Displacement (COD) specimens of the irradiation-induced embrittlement surveillance programme in Loviisa 1 revealed an anomalous behaviour of K{sub JC} compared to the Charpy-V results and to expected results according to standards: about 20% of the COD specimens showed an exceptionally low fracture toughness. Abnormal test specimens were analyzed through fractography, metallography and repeated tests using reconstitution technique: the anomalous behaviour appears to be caused by incorrect pre-fatigue cracking of base metal COD specimens. 7 refs., 9 figs.

Computer aided instrumented Charpy test applied dynamic fracture toughness evaluation system

International Nuclear Information System (INIS)

Kobayashi, Toshiro; Niinomi, Mitsuo

1986-01-01

Micro computer aided data treatment system and personal computer aided data analysis system were applied to the traditional instrumented Charpy impact test system. The analysis of Charpy absorbed energy (E i , E p , E t ) and load (P y , P m ), and the evaluation of dynamic toughness through whole fracture process, i.e. J Id , J R curve and T mat was examined using newly developed computer aided instrumented Charpy impact test system. E i , E p , E t , P y and P m were effectively analyzed using moving average method and printed out automatically by micro computer aided data treatment system. J Id , J R curve and T mat could be measured by stop block test method. Then, J Id , J R curve and T mat were effectively estimated using compliance changing rate method and key curve method on the load-load point displacement curve of single fatigue cracked specimen by personal computer aided data analysis system. (author)

Relationships between Charpy impact shelf energies and upper shelf Ksub(IC) values for reactor pressure vessel steels

International Nuclear Information System (INIS)

Witt, F.J.

1983-01-01

Charpy shelf data and lower bound estimates of Ksub(IC) shelf data for the same steels and test temperatures are given. Included are some typical reactor pressure vessel steels as well as some less tough or degraded steels. The data were evaluated with shelf estimates of Ksub(IC) up to and exceeding 550 MPa√m. It is shown that the high shelf fracture toughness representative of tough reactor pressure vessel steels may be obtained from a knowledge of the Charpy shelf energies. The toughness transition may be obtained either by testing small fracture toughness specimens or by Charpy energy indexing. (U.K.)

Irradiation and annealing behavior of 15Kh2MFA reactor pressure vessel steel

International Nuclear Information System (INIS)

Popp, K.; Bergmann, U.; Bergner, F.; Hampe, E.; Leonhardt, W.D.; Schuetzler, H.P.; Viehrig, H.W.

1992-01-01

This work deals with the mechanical properties of RPV steels used WWER-440. The materials under investigation were a forging (base metal 15Kh2MFA) and the corresponding weld. Charpy V-notch specimens and tensile test specimens were irradiated in the WWER-2 Rheinsberg at about 270 C up to the two neutron fluence levels of 4 x 10 18 and 5 x 10 19 n/cm 2 (E>1MeV). Post-irradiation annealing heat treatments were performed, among others a 475 C/152 h treatment of technical interest. (orig.)

Technique of manufacturing specimen of irradiated fuel rods

International Nuclear Information System (INIS)

Min, Duck Seok; Seo, Hang Seok; Min, Duck Kee; Koo, Dae Seo; Lee, Eun Pyo; Yang, Song Yeol

1999-04-01

Technique of manufacturing specimen of irradiated fuel rods to perform efficient PIE is developed by analyzing the relation between requiring time of manufacturing specimen and manufacturing method in irradiated fuel rods. It takes within an hour to grind 1 mm of specimen thickness under 150 rpm in speed of grinding, 600 g gravity in force using no.120, no.240, no.320 of grinding paper. In case of no.400 of grinding paper, it takes more an hour to grind the same thickness as above. It takes up to a quarter to grind 80-130 μm in specimen thickness using no.400 of grinding paper. When grinding time goes beyond 15 minutes, the grinding thickness of specimen does not exist. The polishing of specimen with 150 Rpms in speed of grinding machine, 600 g gravity in force, 10 minutes in polishing time using diamond paste 15 μm on polishing cloths amounts to 50 μm in specimen thickness. In case of diamond paste 9 μm on polishing cloth, the polishing of specimen amounts to 20 μm. The polishing thickness of specimen with 15 minutes in polishing time using 6 μm, 3 μm, 1 μm, 1/4 μm does not exist. Technique of manufacturing specimen of irradiated fuel rods will have application to the destructive examination of PIE. (author). 6 refs., 1 tab., 10 figs

Comparisons of irradiation-induced shifts in fracture toughness, crack arrest toughness, and Charpy impact energy in high-copper welds

International Nuclear Information System (INIS)

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

1991-01-01

The Heavy-Section Steel Irradiation (HSSI) Program is examining relative shifts and changes in shape of fracture and crack-arrest toughness versus temperature behavior for two high-copper welds. Fracture toughness 100-MPa√m temperature shifts are greater than Charpy 41-J shifts for both welds. Mean curve fits to the fracture toughness data provide mixed results regarding curve shape changes, but curves constructed as lower boundaries indicate lower slopes. Preliminary crack-arrest toughness results indicate that shifts of lower-bound curves are approximately the same as CVN 41-J shifts with no shape changes

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.

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

International Nuclear Information System (INIS)

Lucon, E.

2008-01-01

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.

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.

Use of miniature and standard specimens to evaluate effects of irradiation temperature on pressure vessel steels

International Nuclear Information System (INIS)

Haggag, F.M.; Nanstad, R.K.; Byrne, S.T.

1991-01-01

The effects of neutron irradiation on the steel reactor vessel for the modular high-temperature gas-cooled reactor (MHTGR) are being investigated, primarily because the operating temperatures are low [121 to 210 degrees C (250--410 degrees F)] compared to those for commercial light-water reactors (LWRs) [∼288 degrees C (550 degrees F)]. The need for design data on the reference temperature shift necessitated the irradiation at different temperatures of A 533 grade B class 1 plate. A 508 class 3 forging, and welds used for the vessel shell, vessel closure head, the vessel flange. This paper presents results from the first four irradiation capsules of this program. The four capsules were irradiated in the University of Buffalo Reactor to an effective fast fluence of 1 x10 18 neutron/cm 2 [0.68 x 10 18 neutron/cm 2 (>1 MeV)] at temperatures of 288, 204, 163, and 121 degrees C (550, 400, 325, and 250 degrees F), respectively. The yield and ultimate strengths of both steel plate materials of the MHTGR Program increased with decreasing irradiation temperature. Similarly, the 41-J Charpy V-notch (CVN) transition temperature shift increased with decreasing irradiation temperature (in agreement with the increase in yield strength). The miniature tensile and automated ball indentation (ABI) test results (yield strength and flow properties) were in good agreement with those from standard tensile specimens. The miniature tensile and ABI test results were also used in a model that utilizes the changes in yield strength to estimate the CVN ductile-to-brittle transition temperature shift due to irradiation. The model predictions were compared with CVN test results obtained here and in earlier work. 5 refs., 11 figs., 6 tabs

Effects of irradiation on strength and toughness of commercial LWR vessel cladding

International Nuclear Information System (INIS)

Haggag, F.M.; Corwin, W.R.; Alexander, D.J.; Nanstad, R.K.

1987-01-01

The potential for stainless steel cladding to improve the fracture behavior of an operating nuclear reactor pressure vessel, particularly during certain overcooling transients, may depend greatly on the properties of the irradiated cladding. Therefore, weld overlay cladding irradiated at temperatures and to fluences relevant to power reactor operation was examined. The cladding was applied to a pressure vessel steel plate by the three-wire series-arc commercial method. Cladding was applied in three layers to provide adequate thickness for the fabrication of test specimens. The three-wire series-arc procedure, developed by Combustion Engineering, Inc., Chattanooga, Tennessee, produced a highly controlled weld chemistry, microstructure, and fracture properties in all three layers of the weld. Charpy V-notch and tensile specimens were irradiated at 288 0 C to fluence levels of 2 and 5 x 10 19 neutrons/cm 2 (>1 MeV). Postirradiation testing results show that, in the test temperature range from -125 to 288 0 C, the yield strength increased by 8 to 30%, ductility insignificantly increased, while there was almost no change in ultimate tensile strength. All cladding exhibited ductile-to-brittle transition behavior during Charpy impact testing, due to the dominance of delta-ferrite failures at low temperatures. On the upper shelf, energy was reduced, due to irradiation exposure, 15 and 20%, while the lateral expansion was reduced 43 and 41%, at 2 and 5 x 10 19 neutrons/cm 2 (>1 MeV), respectively. In addition, radiation damage resulted in 13 and 28 0 C shifts of the Charpy impact transition temperature at the 41-J level for the low and high fluences, respectively

Test methodology and technology of fracture toughness for small size specimens

Energy Technology Data Exchange (ETDEWEB)

Wakai, E.; Takada, F.; Ishii, T.; Ando, M. [Japan Atomic Energy Agency, Naga-gun, Ibaraki-ken (Japan); Matsukawa, S. [JNE Techno-Research Co., Kanagawa-ken (Japan)

2007-07-01

Full text of publication follows: Small specimen test technology (SSTT) is required to investigate mechanical properties in the limited availability of effective irradiation volumes in test reactors and accelerator-based neutron and charged particle sources. The test methodology guideline and the manufacture processes for very small size specimens have not been established, and we would have to formulate it. The technology to control exactly the load and displacement is also required in the test technology under the environment of high dose radiation produced from the specimens. The objective of this study is to examine the test technology and methodology of fracture toughness for very small size specimens. A new bend test machine installed in hot cell has been manufactured to obtain fracture toughness and DBTT (ductile - brittle transition temperature) of reduced-activation ferritic/martensitic steels for small bend specimens of t/2-1/3PCCVN (pre-cracked 1/3 size Charpy V-notch) with 20 mm length and DFMB (deformation and fracture mini bend specimen) with 9 mm length. The new machine can be performed at temperatures from -196 deg. C to 400 deg. C under unloading compliance method. Neutron irradiation was also performed at about 250 deg. C to about 2 dpa in JMTR. After the irradiation, fracture toughness and DBTT were examined by using the machine. Checking of displacement measurement between linear gauge of cross head's displacement and DVRT of the specimen displacement was performed exactly. Conditions of pre-crack due to fatigue in the specimen preparation were also examined and it depended on the shape and size of the specimens. Fracture toughness and DBTT of F82H steel for t/2-1/3PCCVN, DFMB and 0.18DCT specimens before irradiation were examined as a function of temperature. DBTT of smaller size specimens of DFMB was lower than that of larger size specimen of t/2-1/3PCCVN and 0.18DCT. The changes of fracture toughness and DBTT due to irradiation were also

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

International Nuclear Information System (INIS)

Iskander, S.K.; Sokolov, M.A.; Nanstad, R.K.

1994-01-01

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 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 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

Mechanical properties of 1950's vintage 304 stainless steel weldment components after low temperature neutron irradiation

International Nuclear Information System (INIS)

Sindelar, R.L.; Caskey, G.R. Jr.; Thomas, J.K.; Hawthorne, J.R.; Hiser, A.L.; Lott, R.A.; Begley, J.A.; Shogan, R.P.

1991-01-01

The reactor vessels of the nuclear production reactors at the Savannah River Site (SRS) were constructed in the 1950's from Type 304 stainless steel plates welded with Type 308 stainless steel filler using the multipass metal inert gas process. An irradiated mechanical properties database has been developed for the vessel with materials from archival primary coolant system piping irradiated at low temperatures (75 to 150 degrees C) in the State University of New York at Buffalo reactor (UBR) and the High Flux Isotope Reactor (HFIR) to doses of 0.065 to 2.1 dpa. Fracture toughness, tensile, and Charpy-V impact properties of the weldment components (base, weld, and weld heat-affected-zone (HAZ)) have been measured at temperatures of 25 degrees C and 125 degrees C in the L-C and C-L orientations for materials in both the irradiated and unirradiated conditions for companion specimens. Fracture toughness and tensile properties of specimens cut from an SRS reactor vessel sidewall with doses of 0.1 and 0.5 dpa were also measured at temperatures of 25 and 125 degrees C. The irradiated materials exhibit hardening with loss of work hardenability and a reduction in toughness relative to the unirradiated materials. The HFIR-irradiated materials show an increase in yield strength between about 20% and 190% with a concomitant tensile strength increase between about 15% to 30%. The elastic-plastic fracture toughness parameters and Charpy-V energy absorption both decrease and show only a slight sensitivity to dose. The irradiation-induced decrease in the elastic-plastic fracture toughness (J def at 1 mm crack extension) is between 20% to 65%; the range of J 1C values are 72.8 to 366 kJ/m 2 for the irradiated materials. Similarly, Charpy V-notch results show a 40% to 60% decrease in impact energies

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

Determination of the toughness of a low alloy steel from the Charpy V-notch impact testing

International Nuclear Information System (INIS)

Rossoll, A.

1998-12-01

Charpy V-notch (CVN) impact testing is widely used to characterize the resistance of a material to brittle fracture, by measuring the energy consumed by a specimen during impact. Notably materials undergoing a ductile-to-brittle transition, e.g. ferritic steels, are quality controlled by means of CVN testing, and their ductile-to-brittle transition temperature can be determined. Charpy testing is also widely used in the toughness assessment of large forged components, e.g. pressure vessels for pressurised water reactors (PWR). However, currently no satisfactory link between the Charpy impact energy CVN and the fracture toughness KIc exists. This study aims to establish a non-empirical relationship between the Charpy V-notch energy CVN, and the fracture toughness KIc, on the lower shelf of fracture toughness and the onset of the ductile-to-brittle transition of a A508 Cl.3 steel. The methodology employed is based on the so-called 'local approach'. Brittle cleavage fracture is modelled in terms of the Beremin (1983) model based on 'weakest link' statistics, whereas ductile crack advance preceding cleavage in the transition region is accounted for with the GTN model (Gurson, 1977; Tvergaard, 1982; Tvergaard and Needleman, 1984). Mechanical testing at different strain rates allowed for the establishment of the constitutive equations of the material in an elastic-viscoplastic formulation. Fracture tests on different specimen geometries provided the large data set necessary for statistical evaluation. All specimen types have been modelled with finite element analysis. However, the dynamic nature of the Charpy test requires special consideration. The origin of these dynamic effects was studied, as well as their implications on interpretation of experimental results and on modeling. After a proper modeling procedure had been defined, the local approach was employed for studying fracture. It is found that the fracture toughness can be predicted from the Charpy impact test

Experimental study associated to irradiation of FBR structural material, (4)

International Nuclear Information System (INIS)

1976-01-01

The study presents one of the bases to evaluate the results of the post-irradiation tests to conduct the thermal control tests related to the second JMTR irradiation (70M-61P) of the demestic austenitic stainless steels for the structural material of the FBR performed by Power Reactor and Nuclear Fuel Development Corporation. The thermal control specimens were given the temperature history which simulated that of the irradiation temperature in vacuum by the electrical furnance, and then the tensile, fatigue and Charpy impact tests were performed. The changes of the material properties caused by the thermal history were investigated. (auth.)

Crack-arrest tests on two irradiated high-copper welds

International Nuclear Information System (INIS)

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

1994-03-01

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

Post-irradiation mechanical tests on F82H EB and TIG welds

International Nuclear Information System (INIS)

Rensman, J.; Osch, E.V. van; Horsten, M.G.; D'Hulst, D.S.

2000-01-01

The irradiation behaviour of electron beam (EB) and tungsten inert gas (TIG) welded joints of the reduced-activation martensitic steel IEA heat F82H-mod. was investigated by neutron irradiation experiments in the high flux reactor (HFR) in Petten. Mechanical test specimens, such as tensile specimens and KLST-type Charpy impact specimens, were neutron irradiated up to a dose level of 2-3 dpa at a temperature of 300 deg. C in the HFR reactor in Petten. The tensile results for TIG and EB welds are as expected with practically no strain hardening capacity left. Considering impact properties, there is a large variation in impact properties for the TIG weld. The irradiation tends to shift the DBTT of particularly the EB welds to very high values, some cases even above +250 deg. C. PWHT of EB-welded material gives a significant improvement of the DBTT and USE compared to the as-welded condition

Fractographic and microstructural aspects of fracture toughness testing in irradiated 304 stainless steel

International Nuclear Information System (INIS)

Cullen, W.H.; Hiser, A.L.; Hawthorne, J.R.; Abramczyk, G.A.; Caskey, G.R.

1987-01-01

Fracture toughness and Charpy impact test results on 304 stainless steel baseplate, weld and heat-affected zone (HAZ) tested at 25 0 C and 125 0 C are correlated with the microstructural and fractographic features observed in these materials. Specimens were collected from several sections of 12.7 mm (0.5 in.) wall thickness piping removed from a process system, and were characterized by different material chemistries and thermomechanical histories. As a result, mechanical properties vary over a considerable range from one pipe section to another. The presence of delta ferrite in some of the samples caused significant degradations in the toughness properties for certain crack orientations. Decreases in Charpy impact energies occur in the same material for different crack orientations. Materials irradiated showed 40% decreases in Charpy impact energy, but little change in fracture morphology. An increase in the test temperature resulted in an expected increase in Charpy energies for all materials. Fractographic features did not change appreciably with respect to the 100 0 C increase in test temperature. In unirradiated specimens, a test temperature increase caused lower J/sub Ic/ and J-R curve values with tearing modules values increased. The latter is due to the large decreases in tensile strength with increasing test temperature. The weld metals tend to have the highest tearing resistance, while the HAZ's tend to have the lowest. 30 figs., 3 tabs

Instrumented impact testing machine with reduced specimen oscillation effects

International Nuclear Information System (INIS)

Rintamaa, R.; Rahka, K.; Wallin, K.

1984-07-01

Owing to small and inexpensive specimens the Charpy impact test is widely used in quality control and alloy development. Limitations in power reactor survellance capsules it is also widely used for safety analysis purposes. Instrumenting the tup and computerizing data acquisition, makes dynamic fracture mechanics data measurement possible and convenient. However, the dynamic effects (inertia forces, specimen oscillations) in the impact test cause inaccuracies in the recorded load-time diagram and hence diminish the reliability of the calculated dynamic fracture mechanics parameters. To decrease inaccuracies a new pendulum type of instrumented impact test apparatus has been developed and constructed in the Metals Laboratory of the Technical Research Centre of Finland. This tester is based on a new principle involving inverted test geometry. The purpose of the geometry inversion is to reduce inertia load and specimen oscillation effects. Further, the new impact tester has some other novel features: e.g. the available initia impact energy is about double compared to the conventional standard (300 J) impact tester allowing the use of larger (10 x 20 x 110 mm) bend specimens than normal Charpy specimens. Also, the rotation asix in the three point bending is nearly stationary making COD-measurements possible. An experimental test series is described in which the inertia effects and specimen oscillations are compared in the conventional and new impact tester utilizing Charpy V-notch specimens. Comparison of the two test geometries is also made with the aid of an analytical model using finite element method (FEM) analysis. (author)

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

International Nuclear Information System (INIS)

Gelles, D.S.; Hamilton, M.L.; Hankin, G.L.

1998-01-01

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

Miniature tensile test specimens for fusion reactor irradiation studies

International Nuclear Information System (INIS)

Klueh, R.L.

1985-01-01

Three miniature sheet-type tensile specimens and a miniature rod-type specimen are being used to determine irradiated tensile properties for alloy development for fusion reactors. The tensile properties of type 316 stainless steel were determined with these different specimens, and the results were compared. Reasonably good agreement was observed. However, there were differences that led to recommendations on which specimens are preferred. 4 references, 9 figures, 6 tables

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

International Nuclear Information System (INIS)

Hofman, A.

1999-01-01

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

Belgian Contribution to the IAEA CRP-IV Programme on Assuring Structural Integrity of Reactor Pressure Vessel

International Nuclear Information System (INIS)

Van Walle, E.; Chaouadi, R.; Scibetta, M.; Puzzolante, J.L.; Fabry, A.; Van de Velde, J.

1997-10-01

This report contains the actual status of the Belgian contribution to the IAEA CRP-IV program. Besides Charpy-V impact tests on as-received CRP-IV JRQ-specimens, fracture toughness tests were performed on two geometries: PCCV-specimens and CRB-specimens. The Charpy-V impact results correspond very well with the as-received CRP-III results. The fracture toughness data are also very consistent with identical tests recently performed on remaining as-received CRP-III material. Irradiated broken Charpy-V samples were reconstituted and tested in PCCV-mode. This was done in order to investigate the evolution of the ASME-curve versus the evolution of the mastercurve with irradiation. Initial results were reported. A new CHIVAS-irradiation in the CALLISTO-loop of the BR-2-reactor to support this investigation, is under preparation

Proton irradiation effects on tensile and bend-fatigue properties of welded F82H specimens

Energy Technology Data Exchange (ETDEWEB)

Saito, S., E-mail: saito.shigeru@jaea.go.j [JAEA Tokai, J-PARC Center, 2-4 Shirakata-shirane, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Kikuchi, K.; Hamaguchi, D. [JAEA Tokai, J-PARC Center, 2-4 Shirakata-shirane, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Usami, K.; Ishikawa, A.; Nishino, Y.; Endo, S. [JAEA Tokai, Department of Hot Laboratories, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Kawai, M. [KEK, Tsukuba-shi, Ibaraki-ken 305-0801 (Japan); Dai, Y. [PSI, Spallation Source Division, 5232 Villigen PSI (Switzerland)

2010-03-15

In several institutes, research and development for an accelerator-driven transmutation system (ADS) have been progressed. Ferritic/martensitic (FM) steels are the candidate materials for the beam window of ADS. To evaluate of the mechanical properties of the irradiated materials, the post irradiation examination (PIE) work of the SINQ (Swiss spallation neutron source) target irradiation program (STIP) specimens was carried out at JAEA. In present study, the results of PIE on FM steel F82H and its welded joint have been reported. The present irradiation conditions of the specimens were as follows: proton energy was 580 MeV. Irradiation temperatures were ranged from 130 to 380 deg. C, and displacement damage level was ranged from 5.7 to 11.8 dpa. The results of tensile tests performed at 22 deg. C indicated that the irradiation hardening occurred with increasing the displacement damage up to 10.1 dpa at 320 deg. C irradiation. At higher dose (11.8 dpa) and higher temperature (380 deg. C), irradiation hardening was observed, but degradation of ductility was relaxed in F82H welded joint. In present study, all specimens kept its ductility after irradiation and fractured in ductile manner. The results on bend-fatigue tests showed that the fatigue life (N{sub f}) of F82H base metal irradiated up to 6.3 dpa was almost the same with that of unirradiated specimens. The N{sub f} of the specimens irradiated up to 9.1 dpa was smaller than that of unirradiated specimens. Though the number of specimen was limited, the N{sub f} of F82H EB (15 mm) and EB (3.3 mm) welded joints seemed to increase after irradiation and the fracture surfaces of the specimens showed transgranular morphology. While F82H TIG welded specimens were not fractured by 10{sup 7} cycles.

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).

Effects of irradiation on initiation and crack-arrest toughness of two high-copper welds and on stainless steel cladding

International Nuclear Information System (INIS)

Nanstad, R.K.; Iskander, S.K.; Haggag, F.M.

1990-01-01

The objective of the study on the high-copper welds is to determine the effect of neutron irradiation on the shift and shape of the ASME K Ic and K Ia toughness curves. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plate. Compact specimens fabricated from these welds were irradiated at a nominal temperature of 288 degree C to fluences from 1.5 to 1.9 x 10 19 neutrons/cm 2 (>1 MeV). The fracture toughness test results show that the irradiation-induced shifts at 100 MPa/m were greater than the Charpy 41-J shifts by about 11 and 18 degree C. Mean curve fits indicate mixed results regarding curve shape changes, but curves constructed as lower boundaries to the data do indicate curves of lower slopes. A preliminary evaluation of the crack-arrest 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), compared to those of the ASME K Ia curve did not appear to have been altered by the irradiation. Three-wire stainless steel weld overlay cladding was irradiated at 288 degree C to fluences of 2 and 5 x 10 19 neutrons/cm 2 (>1 MeV). Charpy 41-J temperature shifts of 13 and 28 degree C were observed, respectively. For the lower fluence only, 12.7-mm thick compact specimens showed decreases in both J Ic and the tearing modulus. Comparison of the fracture toughness results with typical plate and a low upper-shelf weld reveals that the irradiated stainless steel cladding possesses low ductile initiation fracture toughness comparable to the low upper-shelf weld. 8 refs., 12 figs., 2 tabs

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

International Nuclear Information System (INIS)

Groot, P.; Franconi, E.

1994-01-01

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Ductile crack initiation in the Charpy V-notch test

International Nuclear Information System (INIS)

Server, W.L.; Norris, D.M. Jr.; Prado, M.E.

1978-01-01

Initiation and growth of a crack in the Charpy V-notch test was investigated by performing both static and impact controlled deflection tests. Test specimens were deformed to various deflections, heat-tinted to mark crack extension and broken apart at low temperature to allow extension measurements. Measurement of the crack extension provided an estimate of crack initiation as defined by different criteria. Crack initiation starts well before maximum load, and is dependent on the definition of ''initiation''. Using a definition of first micro-initiation away from the ductile blunting, computer model predictions agreed favorably with the experimental results

46 CFR 54.05-5 - Toughness test specimens.

Science.gov (United States)

2010-10-01

... shown in Figure 4 of the specification. Special attention is drawn to the fact that the Charpy Keyhole....090-inch. In preparing weld specimens for dropweight testing, weld reinforcement shall be ground flush...

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.

Irradiation effects on fracture toughness of two high-copper submerged-arc welds, HSSI series 5

International Nuclear Information System (INIS)

Nanstad, R.K.; Haggag, F.M.; McCabe, D.E.; Iskander, S.K.; Bowman, K.O.; Menke, B.H.

1992-10-01

The Fifth Irradiation Series in the Heavy-Section Steel irradiation (HSSI) Program was aimed at obtaining a statistically significant fracture toughness data base on two weldments with high-copper contents to determine the shift and shape of the K lc curve as a consequence of irradiation. The program included irradiated Charpy V-notch impact, tensile, and drop-weight specimens in addition to compact fracture toughness specimens. Compact specimens with thicknesses of 25.4, 50.8, and 101.6 mm [1T C(T), 2T C(T), and 4T C(T), respectively] were irradiated. Additionally, unirradiated 6T C(T) and 8T C(T) specimens with the same K lc measuring capacity as the irradiated specimens were tested. The materials for this irradiation series were two weldments fabricated from special heats of weld wire with copper added to the melt. One lot of Linde 0124 flux was used for all the welds. Copper levels for the two welds are 0.23 and 0.31 wt %, while the nickel contents for both welds are 0.60 wt %. Twelve capsules of specimens were irradiated in the pool-side facility of the Oak Ridge Research Reactor at a nominal temperature of 288 degree C and an average fluence of about 1.5 x 10 19 neutrons/cm 2 (> 1 MeV). This volume, Appendices E and F, contains the load-displacement curves and photographs of the fracture toughness specimens from the 72W weld (0.23 wt % Cu) and the 73 W weld (0.31 wt % Cu), respectively

Use of forces from instrumented Charpy V-notch testing to determine crack-arrest toughness

International Nuclear Information System (INIS)

Iskander, S.K.; Nanstad, R.K.; Sokolov, M.A.; McCabe, D.E.; Hutton, J.T.

1996-06-01

The objective of this investigation is an estimation of the crack-arrest toughness, particularly of irradiated materials, from voltage versus time output of an instrumented setup during a test on a Charpy V-notch (CVN) specimen. This voltage versus time trace (which can be converted to force versus displacement) displays events during fracture of the specimen. Various stages of the fracture process can be identified on the trace, including an arrest point indicating arrest of brittle fracture. The force at arrest, F a , versus test temperature, T, relationship is examined to explore possible relationships to other experimental measures of crack-arrest toughness such as the drop-weight nil-ductility temperature (NDT), or crack-arrest toughness, K a . For a wide range of weld and plate materials, the temperature at which F a = 2.45 kN correlates with NDT with a standard deviation, sigma, of about 11 K. Excluding the so-called low upper-shelf energy (USE) welds from the analysis resulted in F a = 4.12 kN and σ = 6.6 K. The estimates of the correlation of the temperature for F a = 7.4 kN with the temperature at 100-MPa√m level for a mean American Society of Mechanical Engineers (ASME) type K Ia curve through crack-arrest toughness values show that prediction of conservative values of K a are possible

An automated tensile machine for small specimens heavily neutron irradiated in FFTF/MOTA

International Nuclear Information System (INIS)

Kohyama, Akira; Sato, Shinji; Hamada, Kenichi

1993-01-01

The objective of this work is to develop a fully automated tensile machine for post-irradiation examination (PIE) of Fast Flux Test Facility (FFTF)/Materials Open Test Assembly (MOTA) irradiated miniature tension specimens. The anticipated merit of the automated tensile machine is to reduce damage to specimens during specimen handling for PIE and to reduce exposure to radioactive specimens. This machine is designed for testing at elevated temperatures, up to 873 K, in a vacuum or in an inert gas environment. Twelve specimen assemblies are placed in the vacuum chamber that can be tested successively in a fully automated manner. A unique automated tensile machine for the PIE of FFTF/MOTA irradiated specimens, the Monbusho Automated Tensile Machine (MATRON) consists of a test frame with controlling units and an automated specimen-loading apparatus. The qualification of the test frame has been completed, and the results have satisfied the machine specifications. The capabilities of producing creep and relaxation data have been demonstrated for Cu, Al, 316SS, and ferritic steels. The specimen holders for the three-point bending test and the small bulge test (small punch test; SP test) were also designed and produced

Progress report on irradiation experiment on small size specimens in high temperature flux module

Energy Technology Data Exchange (ETDEWEB)

Ramesh, M.; Jacquet, P.; Chaouadi, R.

2011-02-15

This report describes the progress made in IFREC/DEMO Research and Development Program during the year 2010 at SCK/CEN. This task is part of demonstrating the possibility to irradiate small specimens in the HFTM modules that will be used in DEMO. Different small specimens of three candidate materials of DEMO fusion reactor will be irradiated with the objective of validating the specimen geometry and size to reliably characterize the mechanical properties of unirradiated and in future of irradiated materials.

Investigation of irradiation embrittlement and annealing behaviour of JRQ pressure vessel steel by instrumented impact tests

Energy Technology Data Exchange (ETDEWEB)

Valo, M; Rintamaa, R; Nevalainen, M; Wallin, K; Torronen, K [Technical Research Centre of Finland, Espoo (Finland); Tipping, P [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1994-12-31

Seven series of A533-B type pressure vessel steel specimens irradiated as well as irradiated - annealed - re-irradiated to different fast neutron fluences (up to 5.10{sup 19}/cm{sup 2}) have been tested with a new type of instrumented impact test machine. The radiation embrittlement and the effect of the intermediate annealing was assessed by using the ductile and cleavage fracture initiation toughness. Although the ductile fracture initiation toughness exhibited scatter, the transition temperature shift corresponding to the dynamic cleavage fracture initiation agreed well with the 41 J Charpy-V shift. The results indicate that annealing is beneficial in restoring mechanical properties in an irradiated nuclear pressure vessel steel. (authors). 8 refs., 11 figs., 1 tab.

The impact of irradiation induced specimen charging on microanalysis in a scanning electron microscope

International Nuclear Information System (INIS)

Stevens-Kalceff, M.A.

2003-01-01

Full text: It is necessary to assess and characterize the perturbing influences of experimental probes on the specimens under investigation. The significant influence of electron beam irradiation on poorly conducting materials has been assessed by a combination of specialized analytical scanning electron and scanning probe microscopy techniques including Cathodoluminescence Microanalysis and Kelvin Probe Microscopy. These techniques enable the defect structure and the residual charging of materials to be characterized at high spatial resolution. Cathodoluminescence is the non-incandescent emission of light resulting from the electron irradiation. CL microscopy and spectroscopy in a Scanning Electron Microscope (SEM) enables high spatial resolution and high sensitivity detection of defects in poorly conducting materials. Local variations in the distribution of defects can be non-destructively characterized with high spatial (lateral and depth) resolution by adjusting electron beam parameters to select the specimen micro-volume of interest. Kelvin Probe Microscopy (KPM) is a Scanning Probe Microscopy technique in which long-range Coulomb forces between a conductive atomic force probe and the specimen enable the surface potential to be characterized with high spatial resolution. A combination of Kelvin Probe Microscopy (KPM) and Cathodoluminescence (CL) microanalysis has been used to characterize ultra pure silicon dioxide exposed to electron irradiation in a Scanning Electron Microscope. Silicon dioxide is an excellent model specimen with which to investigate charging induced effects. It is a very poor electrical conductor, homogeneous and electron irradiation produces easily identifiable surface modification which enables irradiated regions to be easily and unambiguously located. A conductive grounded coating is typically applied to poorly conducting specimens prior to investigation in an SEM to prevent deflection of the electron beam and surface charging, however

Application of Instrumented Charpy Method in Characterisation of Materials

Directory of Open Access Journals (Sweden)

Željko Alar

2015-07-01

Full Text Available Testing of absorbed impact energy according to the Charpy method is carried out to determine the behaviour of a material under the impact load. Instrumented Charpy method allows getting the force displacement curve through the entire test, That curve can be related to force-displacement curve which is obtained by the static tensile test. The purpose of this study was to compare the results of forces obtained by the static tensile test with the forces obtained by the instrumented Charpy method. Experimental part of the work contains testing of the mechanical properties of S275J0 steel by the static tensile test and Impact test on instrumented Charpy pendulum.

Postirradiation notch ductility tests of ESR alloy HT-9 and modified 9Cr-1Mo alloy from UBR reactor experiments

International Nuclear Information System (INIS)

Hawthorne, J.R.

1984-01-01

During this period, irradiation exposures at 300 0 C and 150 0 C to approx. 8 x 10 19 n/cm 2 , E > 0.1 MeV, were completed for the Alloy HT-9 plate and the modified Alloy 9Cr-1Mo plates, respectively. Postirradiation tests of Charpy-V (C/sub v/) specimens were completed for both alloys; other specimen types included in the reactor assemblies were fatigue precracked Charpy-V (PCC/sub v/), half-size Charpy-V, and in the case of the modified 9Cr-1Mo, 2.54 mm thick compact tension specimens

Development of the plant life management technology for RPV steels [ - Current status of surveillance test specimen reconstitution program -

International Nuclear Information System (INIS)

Kazunobu, Sakamoto; Eliichiro, Otsuka; Yoshiaki, Oka; Kosei, Taguchi; Michiyoshi, Yamamoto

2001-01-01

In order to develop the reconstitution technology to standardize surveillance test specimen reconstitution practices to deal with the extended service life of reactor pressure vessels, the Japan Power Engineering and Inspection Corporation (JAPEIC) has been carried out the project entrusted by the Ministry of International Trade and Industry (MITI). We focus on a correlation between the reduction of absorbed energy and the interaction of the heat affected zone (HAZ) and the plastic zone, to establish applicable reconstitution conditions for Charpy specimens. The relationship between the plastic zone width and the absorbed energy has been obtained by estimating the plastic zone width from the hardness distribution of the Charpy specimens. Impact tests of reconstituted specimens with 10 mm-length insert using the surface activated joining method were performed and the test results were compared to those obtained by the standard specimens. By comparing the length of insert material to the sum of HAZ width and plastic zone width, it is clear that the interaction causes the reduction of the absorbed energy. Hence, the applicable conditions of reconstituted Charpy specimens could be assessed by comparing the insert length to the sum of HAZ width and plastic zone width. Moreover the effects of the possible deviations from the standard shape and size specimens for the reconstituted specimens were studied. (authors)

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

International Nuclear Information System (INIS)

Sokolov, Mikhail A; Lucon, Enrico

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 10 11 n/cm 2 /s (>1 MeV) to fluences from 0.5 to 3.4 10 19 n/cm 2 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 > 10 13 n/cm 2 /s and subsequent testing by SCK-CEN. The BR2 irradiations were conducted at about 2 and 4 x 10 13 n/cm 2 /s with irradiation temperature between 295 °C and 300 °C (water temperature), and to fluences between 6 and 10 x 10 19 n/cm 2 . The irradiation-induced shifts of the Master Curve reference temperatures, ΔT 0 , 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, ΔT 0 , 25 °C to 53 °C lower than those from the larger compact specimens, meaning that the irradiation-induced reference temperature shifts, ΔT 0 , were larger than the initial measurements, resulting in much improved agreement between the measured and predicted fracture toughness shifts.

Application of Instrumented Charpy Method in Characterisation of Materials

OpenAIRE

Alar, Željko; Mandić, Davor; Dugorepec, Andrija; Sakoman, Matija

2015-01-01

Testing of absorbed impact energy according to the Charpy method is carried out to determine the behaviour of a material under the impact load. Instrumented Charpy method allows getting the force displacement curve through the entire test, That curve can be related to force-displacement curve which is obtained by the static tensile test. The purpose of this study was to compare the results of forces obtained by the static tensile test with the forces obtained by the instrumented Charpy method...

Statistical analyses of fracture toughness results for two irradiated high-copper welds

International Nuclear Information System (INIS)

Nanstad, R.K.; McCabe, D.E.; Haggag, F.M.; Bowman, K.O.; Downing, D.J.

1990-01-01

The objectives of the Heavy-Section Steel Irradiation Program Fifth Irradiation Series were to determine the effects of neutron irradiation on the transition temperature shift and the shape of the K Ic curve described in Sect. 6 of the ASME Boiler and Pressure Vessel Code. Two submerged-arc welds with copper contents of 0.23 and 0.31% were commercially fabricated in 215-mm-thick plates. Charpy V-notch (CVN) impact, tensile, drop-weight, and compact specimens up to 203.2 mm thick [1T, 2T, 4T, 6T, and 8T C(T)] were tested to provide a large data base for unirradiated material. Similar specimens with compacts up to 4T were irradiated at about 288 degrees C to a mean fluence of about 1.5 x 10 19 neutrons/cm 2 (>1 MeV) in the Oak Ridge Research Reactor. Both linear-elastic and elastic-plastic fracture mechanics methods were used to analyze all cleavage fracture results and local cleavage instabilities (pop-ins). Evaluation of the results showed that the cleavage fracture toughness values determined at initial pop-ins fall within the same scatter band as the values from failed specimens; thus, they were included in the data base for analysis (all data are designated K Jc )

Newly developed non-destructive testing method for evaluation of irradiation brittleness of structural materials using ultrasonic

International Nuclear Information System (INIS)

Ishii, Toshimitsu; Ooka, Norikazu; Kato, Yoshiaki; Saito, Junichi; Hoshiya, Taiji; Shibata, Saburo; Kobayashi, Hideo

1999-01-01

Surveillance testing is important to evaluate neutron irradiation embrittlement of reactor pressure vessel material for long life operation. An alternative test method for evaluating the irradiation embrittlement of the pressure vessel material will have to be proposed to support the limited number of surveillance test specimens in order to manage the plant life to be extended. In this study, ultrasonic testing for irradiated A533B-1 steel and weld metal was applied to examine material degradation nondestructively. With increasing the shift of Charpy 41 J transition temperature, ultrasonic velocity decreased and attenuation coefficient of ultrasonic wave increased. Especially, the difference of ultrasonic velocity for 5 MHz shear wave between as-received and irradiated material is corresponding to the shift of transition temperature showing material degradation. (author)

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}.

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

International Nuclear Information System (INIS)

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

2005-03-01

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

Instrumented Impact Testing: Influence of Machine Variables and Specimen Position

International Nuclear Information System (INIS)

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

2008-01-01

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.

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.

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.

Charpy trend-curve development based on PWR surveillance date

International Nuclear Information System (INIS)

Guthrie, G.L.

1983-01-01

The formulas given in the text provide a useful method of predicting the irradiation induced increase in the 41 joule Charpy transition temperature for plate and weld material. The standard deviations for the least squares fits are 26.4 0 F for the weld equation and 15.6 0 F for the plat relationship. The current method of derivation produces an unbiased estimate of the fluence exponent, resulting in increased reliability for fluence extrapolations. The method given for error estimation provides a relatively rigorous procedure for calculating uncertainties and takes proper account of the effects of uncertainties in the independent variables in any given application of the formulas. 11 refs., 4 figs., 4 tabs

Accelerated irradiation test of Gundremmingen reactor vessel trepan material

Energy Technology Data Exchange (ETDEWEB)

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

1992-08-01

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

Accelerated irradiation test of gundremmingen reactor vessel trepan material

International Nuclear Information System (INIS)

Hawthorne, J.R.

1992-08-01

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

Development of Mini-Compact Tension Test Method for Determining 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)

2017-05-01

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. Validation of the mini compact tension specimen (mini-CT) geometry has been performed on previously well characterized Midland beltline Linde 80 (WF-70) weld in the unirradiated condition. It was shown that the fracture toughness transition temperature, To, measured by these Mini-CT specimens is almost the same as To value that was derived from various larger fracture toughness specimens. Moreover, an International collaborative program has been established to extend the assessment and validation efforts to irradiated Linde 80 weld metal. The program is underway and involves the Oak Ridge National Laboratory (ORNL), Central Research Institute for Electrical Power Industry (CRIEPI), and Electric Power Research Institute (EPRI). The irradiated Mini-CT specimens from broken halves of previously tested Charpy

Measurement of the yield and tensile strengths of neutron-irradiated and post-irradiation recovered vessel steels with notched specimens

International Nuclear Information System (INIS)

Valiente, A.

1996-01-01

Tensile circumferentially notched bars are examined as test specimens for measuring the yield and tensile strengths of nuclear pressure vessel steels under several conditions of irradiation and temperature that a vessel can experience during its service life, including recovery post-irradiation treatment. For all the vessel steels, notch geometries and conditions explored, it has been found that notched specimens fail by plastic collapse, and simple formulae have been derived that allow the yield and tensile strengths to be determined from the yielding and plastic collapse load of a notched specimen. Values measured in this way show good agreement with those measured by the standard tensile test method. (orig.)

ASK Procedure for Instrumented Pre-cracked Charpy-Type Tests

International Nuclear Information System (INIS)

Varga, T.; Njo, D.H.; Prantl, G.

1981-01-01

The essential technical content of the ASK procedure originated from development work in Switzerland since 1963, and practical experiences gained since 1972. The remainder of the content and the format of the procedure are based on the ASTM E 24.03.03. (Tentative Draft Copy) 'Proposed Method for Pre-cracked Charpy Impact and Slow-Bend Testing of Metallic Materials' by C. E. Harbower, 1973. Two different velocities, 5 m/s and 0.1 m/s were used with a Schnadt-type machine of rigid construction. The stiffness of the machine proved to be very suitable for instrumented testing. The instrumented Schnadt-Type machine was equipped with strain gauges both on the top of the pendulum and on the chisel. A static force calibration was followed by energy calibration, comparing potential energy losses with the area under the force-deflection curve. Deflection was measured using a high frequency eddy current method on the pendulum, and for slow testing by means of an inductive gauge on the chisel. Charpy-Type specimens of 1.0 mm max notch depth and 0.12 mm max notch radius were pre-cracked using a resonant fatigue testing machine, or an eccentric drive machine. Crack propagation rate da/dN was measured using 'Russenberger' measuring gauges. In addition a new technique for the detection of dynamic crack initiation, developed at the Institute of Research and Technology (TVFA) in Vienna is discussed and some results presented

Metallographic analysis of irradiated RERTR-3 fuel test specimens

International Nuclear Information System (INIS)

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

2000-01-01

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

Development for advanced materials and testing techniques

Energy Technology Data Exchange (ETDEWEB)

Hishinuma, Akimichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

Recent studies using a JMTR and research reactors of JRR-2 and JRR-3 are briefly summarized. Small specimen testing techniques (SSTT) required for an effective use of irradiation volume and also irradiated specimens have been developed focussing on tensile test, fatigue test, Charpy test and small punch test. By using the small specimens of 0.1 - several mm in size, similar values of tensile and fatigue properties to those by standard size specimens can be taken, although the ductile-brittle transition temperature (DBTT) depends strongly on Charpy specimen size. As for advanced material development, R and D about low activation ferritic steels have been done to investigate irradiation response. The low activation ferritic steel, so-called F82H jointly-developed by JAERI and NKK for fusion, has been confirmed to have good irradiation resistance within a limited dose and now selected as a standard material in the fusion material community. It is also found that TiAi intermetallic compounds, which never been considered for nuclear application in the past, have an excellent irradiation resistance under an irradiation condition. Such knowledge can bring about a large expectation for developing advanced nuclear materials. (author)

Studies performed on neutron-irradiated copper-doped iron specimens by means of neutron small-angle scattering

International Nuclear Information System (INIS)

Naraghi, M.

1978-01-01

By means of neutron small-angle scattering precipitation arising from heat-treatment and reactor irradiation in copper-alloyed iron specimens were studied. Copper content varried between 0 and 1.5%, irradiation temperature between 310 and 563K. The specimens had been cooled from the melt partly fast, partly slowely. By taking account of magnetic scattering and by investigating the azimuthal dependence of the total scattering it became possible to distinguish between copper precipitations and vacancy agglomerates. The most obvious effect in the slowly cooled specimens after irradiation with 2-10 19 fast neutrons per cm 2 at a temperature of 563 K is the existence of copper agglomerates with diameters of the order of magnitude of 5nm. Precipitation already occurs to a much lesser extent by the influence of temperature alone. Fast cooling from the melt or low irradiation temperature reduce precipitation during reactor irradiation. Moreover, there are indications on the formation of vacancy accumulations and dislocation rings, the latter especially in the fast cooled specimens. (orig.) [de

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

International Nuclear Information System (INIS)

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

1990-12-01

The objective of this study was to determine the effect of neutron irradiation on the shift and shape of the lower-bound curve to crack-arrest data. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plate. Crack-arrest specimens fabricated from these welds were irradiated at a nominal temperature of 288 degree C to an average fluence of 1.9 x 10 19 neutrons/cm 2 (>1 MeV). Evaluation of the results shows that the neutron-irradiation-induced crack-arrest toughness temperature shift is about the same as the Charpy V-notch impact temperature shift at the 41-J energy level. The shape of the lower-bound curves (for the range of test temperatures covered) did not seem to have been altered by irradiation compared to those of the ASME K Ia curve. 9 refs., 21 figs., 10 tabs

Dose determination of 600 MeV proton irradiated specimens

International Nuclear Information System (INIS)

Gavillet, D.

1991-01-01

The calculation method for the experimental determination of the atomic production cross section from the γ activity measurements are presented. This method is used for the determination of some isotope production cross sections for 600 MeV proton irradition in MANET steel, copper, tungsten, gold and titanium. The results are compared with some calculation. These values are used to determine the dose of specimens irradiated in the PIREX II facility. The results are discussed in terms of the irradiation parameters. A guide for the use of the production cross section determined in the dosimetry experiment are given. (author) tabs., refs

Characterization by notched and precracked Charpy tests of the in-service degradation of RPV steel fracture toughness

International Nuclear Information System (INIS)

Fabry, A.

1997-01-01

The current engineering and regulatory practice to estimate fracture toughness safety margins for nuclear reactor pressure vessels (RPVs) relies heavily on the CVN impact test. Techniques to estimate in-service toughness degradation directly using a variety of precracked specimens are under development worldwide. Emphasis is on their miniaturization. In the nuclear context, it is essential to address many issues such as representativity of the surveillance programs with respect to the vessel in terms of materials and environment, transferability of test results to the structure (constraint and size effects), lower bound toughness certification, creadibility relative to trends of exising databases. An enhanced RPV surveillance strategy in under development in Belgium. It combines state-of-the-art micromechanical and damage modelling to the evaluation of CVN load-deflection signals, tensile stress-strain curves and slow-bend tests of reconstituted precracked Charpy specimens. A probabilistic micromechanical model has been established for static and dynamic transgranular cleavage initiation fracture toughness in the ductile-brittle transition temperature range. This model allows to project toughness bounds for any steel embrittlement condition from the corresponding CVN and static tensile properties, using a single scaling factor defined by imposing agreement with toughness tests in a single condition. The outstanding finding incorporated by this toughness transfer model is that the microcleavage fracture stress is affected by temperature in the ductile-brittle transition and that this influence is strongly correlated to the flow stress: this explains the shape of the K Ic n K Id temperature curves as well as the actual magnitude of the strain rate and irradiation effects. Furthermore, CVN crack arrest loads and fracture appearance are also taken advantage of in order to estimate K Ia degradation. Finally, the CVN-tensile load-temperature diagram provides substantial

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

International Nuclear Information System (INIS)

Schubert, L.E.; Hamilton, M.L.; Gelles, D.S.

1997-01-01

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 ∼67 dpa and at 370 degrees C to ∼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 ∼67 dpa than after irradiation at 370 degrees C to ∼15 dpa

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.

Thermal analysis on the specimens for low irradiation temperature below 100degC in the HANARO

Energy Technology Data Exchange (ETDEWEB)

Choi, Myoung-Hwan; Kim, Bong-Goo; Lee, Byung-Chul; Kim, Tae-Kyu [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)

2012-03-15

A capsule has been used for an irradiation test of various nuclear materials in the research reactor, HANARO. As a part of the research reactor development project with a plate type fuel, the irradiation tests of beryllium, zircaloy-4 and graphite materials using the capsule will be carried out to obtain the mechanical characteristics at low temperatures below 100degC with 30 MW reactor power. In this study, in order to obtain the preliminary design data of the capsule with various specimens and the temperature of specimens, a thermal analysis is performed by using an ANSYS program. The finite element models for the cross section of the capsule containing the specimen are generated, and the temperatures are evaluated. The analysis results show that most specimens meet the irradiation target temperature. However, some canned graphite specimens have a slightly high temperature, and the gap size has a significant effect on the specimen temperature. Based on those results a detailed design and analysis of the capsule will be completed this year. (author)

Analysis of the Charpy V-notch test for welds

DEFF Research Database (Denmark)

Tvergaard, Viggo; Needleman, A.

2000-01-01

The ductile-brittle transition for a weld is investigated by numerical analyses of Charpy impact specimens. 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. The effect of weld strength undermatch or overmatch is investigated for a comparison material, and analyses are also carried out based on experimentally determined flow...... strength variations in a weldment in a HY100 steel. The predicted work to fracture shows a strong sensitivity to the location of the notch relative to the weld, with the most brittle behavior for a notch close to the narrow heat affected zone. The analyses illustrate the strong dependence of the transition...

Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

International Nuclear Information System (INIS)

Chernobaeva, A.A.; Korolev, Y.N.; Nanstad, R.K.; Nikolaev, Y.A.; Sokolov, M.A.

1998-01-01

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. Even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification of available models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The Working Group agreed that each side would irradiate, anneal, reirradiate (if feasible ), and test two materials of the other. Charpy V-notch (CVN) and tensile specimens were included. Oak Ridge National Laboratory (ORNL) conducted such a program (irradiation and annealing, including static fracture toughness) with two weld metals representative of VVER-440 and VVER-1000 RPVs, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation, annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) Program Plate 02 and Heavy-Section Steel Irradiation (HSSI) Program Weld 73W. The results for each material from each laboratory are compared with those from the other laboratory. The ORNL experiments with the VVER welds included irradiation to about 1 x 10 19 n/cm 2 (>1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 x 10 19 n/cm 2 (>l MeV). In both cases, irradiations were conducted at ∼290 C and annealing treatments were conducted at ∼454 C. The ORNL and RRC

Effects of irradiation on crack-arrest toughness of two high-copper welds

International Nuclear Information System (INIS)

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

1990-01-01

The objective of this study is to determine the effect of neutron irradiation on the shift and shape of the lower-bound curve to crack-arrest data. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plate. Crack-arrest specimens fabricated from these welds were irradiated at a nominal temperature of 288 degree C to an average fluence of 1.9 x 10 19 neutrons/cm 2 (>1 MeV). A preliminary 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), compared to those of the ASME K Ia -curve did not seem to have been altered by irradiation. 10 refs., 9 figs., 7 tabs

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

DEFF Research Database (Denmark)

Tvergaard, Viggo; Needleman, A.

2004-01-01

. The onset of cleavage is taken to occur when the average of the maximum principal stress over a specified volume attains a critical value. The weld analyzed here is overmatched, so that the yield strength for the weld is larger than that of the base material. Analyses are carried out for specimens where...... the notch is cut parallel to the surface of the test piece, as well as more complex geometries where the notched surface of the specimen is rotated relative to the surface of the test piece. It is found that even for a homogeneous material the brittle-ductile transition can be much affected by three...... dimensional effects; for example, curvature of the deformed free surface gives rise to a stress increase that promotes cleavage. Furthermore, for the rotated specimens the weld geometry relative to the notched specimen surface is so complex that only a full 3D analysis is able to account for the interaction...

General views about specimen irradiations in reactors; Considerations generales sur'les irradiations d'echantillons dans les reacteurs

Energy Technology Data Exchange (ETDEWEB)

Seguin, M [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1965-07-01

Specimen irradiation of fissile or non-fissile materials, carried out under circumstances becoming more and more severe and in reactor of increasing flux bas led to an evolution of irradiation rigs. A survey of the problems arising from irradiating under these various circumstances leads to conclude that it is possible to devise one capsule type suitable to every particular case, and that in a wide temperature range. Consequently, once the various irradiation-parameters known, a general method of calculation can be followed so as to determine the various sizes of the parts constituting the capsule. These theoretical calculations might sometimes be corrected through benefits gained from previous irradiations. Similarly, practical experimentation might allow to foresee more handy assembling of the capsule, specimen loading-and unloading being easier at the same time. (author) [French] L'irradiation d'echantillons, fissiles ou non fissiles, dans des conditions imposees de plus en plus strictes et dans des reacteurs a flux de plus en plus eleve, a eu pour consequence une evolution dans la conception des dispositifs d'irradiation. Lorsqu'on examine les problemes souleves par ces differentes irradiations, on en conclut qu'il est possible de concevoir un type de capsule capable de donner satisfaction dans chaque cas particulier, et ce, dans une tres large gamme de temperature. Par consequent, les differents parametres de l'irradiation etant connus, une methode generale de calcul peut etre suivie pour determiner les differentes cotes des pieces constitutives de la capsule. Ces calculs theoriques devront quelquefois etre corriges grace aux enseignements tires d'irradiations precedentes. De meme, l'experience acquise permettra d'envisager un montage plus aise de la capsule, tout en facilitant l'enfournement et le defournement des echantillons.

System for the continuous irradiation of specimens, especially for activation analysis

International Nuclear Information System (INIS)

Dieck, L.E.

1975-01-01

The system is to ensure a continuous irradiation of several specimens, especially for activation analysis. The specimens rotate in a hollow body which is axially movable along a rail. The rotation is effected by a turbine driven by hydraulic or pneumatic power and placed in the hollow body which can be used for example in the rabbit system of a nuclear reactor. The driving medium serves both as conveying medium for the system and drive for the rotating drum and as coolant for the specimens and results in radiation protection. The geometric arrangement and design of both the turbine and the whole system is decribed in detail. (UWI) [de

Heavy-Section Steel Irradiation Program

International Nuclear Information System (INIS)

Corwin, W.R.

1990-08-01

The primary goal of the Heavy-Section Steel Irradiation Program is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior (particularly the fracture toughness properties) of typical pressure-vessel steels as they relate to light-water-reactor pressure-vessel integrity. The program includes direct continuation of irradiation studies previously conducted by the Heavy-Section Steel Technology Program augmented by enhanced examinations of the accompanying microstructural changes. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are examined on a wide range of fracture properties. Detailed statistical analyses of the fracture data on K Ic shift of high-copper welds were performed. Analysis of the first phase of irradiated crack-arrest testing on high-copper welds was completed. Final analysis and publication of the results of the second phase of the irradiation studies on stainless steel weld-overlay cladding were completed. Determinations were made of the variations in chemistry and unirradiated RT NDT of low upper-shelf weld metal from the Midland reactor. Final analyses were performed on the Charpy impact and tensile data from the Second and Third Irradiation series on low upper-shelf welds, and the report on the series was drafted. A detailed survey of existing data on microstructural models and data bases of irradiation damage was performed, and initial development of a reaction-rate-based model was completed. 40 refs., 7 figs., 4 tabs

The Change of the Seebeck Coefficient Due to Neutron Irradiation and Thermal Fatigue of Nuclear Reactor Pressure Vessel Steel and its Application to the Monitoring of Material Degradation

International Nuclear Information System (INIS)

Niffenegger, M.; Reichlin, K.; Kalkhof, D.

2002-05-01

The monitoring of material degradation, that might be caused by neutron irradiation and thermal fatigue, is an important topic in lifetime extension of nuclear power plants. We therefore investigated the application of the Seebeck effect for determining material degradation of common reactor pressure vessel steel. The Seebeck coefficient (SC) of several irradiated Charpy specimens made from Japanese JRQ-steel were measured. The specimens suffered a fluence from 0 up to 4.5 x 10 19 neutrons per cm 2 with energies higher than 1 MeV. The measured changes of the SC within this range were about 500 nV, increasing continuously in the range under investigation. Some indications of saturation appeared at fluencies larger than 4.55 x 10 19 neutrons per cm 2 . We obtained a linear dependency between the SC and the temperature shift ΔT 41 of the Charpy-Energy- Temperature curve which is widely used to characterize material embrittlement. Similar measurements were performed on specimens made from the widely used austenitic steel X6CrNiTi18-10 (according to DIN 1.4541) that were fatigued by applying a cyclic strain amplitude of 0.28%. For this kind of fatigue the observed change of SC was somewhat smaller than for the irradiated specimens. Further investigations were made to quantify the size of the gage volume in which the thermoelectric power is generated. It appeared that the information gathered from a Thermo Electric Power (TEP) measurement is very local. To overcome this problem we propose a novel TEP-method using a Thermoelectric Scanning Microscope (TSM). We finally conclude that the change of the SC has a potential for monitoring of material degradation due to neutron irradiation and thermal fatigue, but it has to be taken into account that several influencing parameters could contribute to the TEP in either an additional or extinguishing manner. A disadvantage of the method is the requirement of a clean surface without any oxide layer. A part of this disadvantage can

Small specimen measurements of dynamic fracture toughness of heavy section steels for nuclear pressure vessel

International Nuclear Information System (INIS)

Tanaka, Y.; Iwadate, T.; Suzuki, K.

1987-01-01

This study presents the dynamic fracture toughness properties (KId) of 12 heats of RPV steels measured using small specimens and analysed based on the current research. The correlation between the KId test and other engineering small specimen tests such as Charpy test and drop weight test are also discussed and a method to predict the KId value is presented. (orig./HP)

Postirradiation fracture toughness tests of ESR alloy HT-9 and modified 9Cr-1Mo alloy from UBR reactor experiments

International Nuclear Information System (INIS)

Hawthorne, J.R.; Reed, J.R.; Sprague, J.A.

1984-01-01

Alloy HT-9 and Modified 9Cr-1Mo are being evaluated for potential applications as first wall materials in magnetic fusion reactors. Objectives of the current research task were to test fatigue-precracked Charpy-V (PCC/sub v/) specimens from representative plates irradiated in the UBR reactor at 149 0 C or 300 0 C, and, to compare the results against postirradiation notch ductility data developed previously for the materials. Both plates represent electroslag refined (ESR) melt processing. PCC/sub v/ specimens of Alloy HT-9 and Modified 9Cr-1Mo alloy were irradiated at 300 0 C and 149 0 C, respectively, to approx.0.8 X 10 20 n/cm 2 , E > 0.1 MeV. During this period, postirradiation tests for fracture toughness were completed and results compared to notch ductility determinations from standard Charpy-V (C/sub v/) specimens irradiated in the same reactor experiments. Fracture surface examinations by SEM are also reported

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.

On the ductile-to-brittle transition behavior of martensitic alloys neutron irradiated to 26 dpa

International Nuclear Information System (INIS)

Hu, W.L.; Gelles, D.S.

1987-01-01

Charpy impact tests were conducted on specimens made of HT-9 and 9Cr-1Mo in various heat treatment conditions which were irradiated in EBR-II to 26 dpa at 390 to 500 0 C. The results are compared with previous results on specimens irradiated to 13 dpa. HT-9 base metal irradiated at low temperatures showed a small additional increase in ductile brittle transition temperature and a decrease in upper shelf energy from 13 to 26 dpa. No fluence effect was observed in 9Cr-1Mo base metal. The 9Cr-1Mo weldment showed degraded DBTT but improved USE response compared to base metal, contrary to previous findings on HT-9. Significant differences were observed in HT-9 base metal between mill annealed material and normalized and tempered material. The highest DBTT for HT-9 alloys was 50 0 C higher than for the worst case in 9Cr-1Mo alloys. Fractography and hardness measurements were also obtained. Significant differences in fracture appearance were observed in different product forms, although no dependence on fluence was observed. Failure was controlled by the preirradiation microstructure

Beryllium irradiation embrittlement test programme. Material and specimen specification, manufacture and qualification

International Nuclear Information System (INIS)

Harries, D.R.; Dalle Donne, M.

1996-06-01

The report presents the specification, manufacture and qualification of the beryllium specimens to be irradiated in the BR2 reactor in Mol to investigate the effect of the neutron irradiation on the embrittlement as a function of temperature and beryllium oxide content. This work was been performed in the framework of the Nuclear Fusion Project of the Forschungszentrum Karlsruhe and is supported by the European Union within the European Fusion Technology Program. (orig.)

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

International Nuclear Information System (INIS)

Zinkle, S.J.

1998-01-01

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 ∼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

Characterization by notched and precracked Charpy tests of the in-service degradation of RPV steel fracture toughness

Energy Technology Data Exchange (ETDEWEB)

Fabry, A.

1997-01-01

The current engineering and regulatory practice to estimate fracture toughness safety margins for nuclear reactor pressure vessels (RPVs) relies heavily on the CVN impact test. Techniques to estimate in-service toughness degradation directly using a variety of precracked specimens are under development worldwide. Emphasis is on their miniaturization. In the nuclear context, it is essential to address many issues such as representativity of the surveillance programs with respect to the vessel in terms of materials and environment, transferability of test results to the structure (constraint and size effects), lower bound toughness certification, creadibility relative to trends of exising databases. An enhanced RPV surveillance strategy in under development in Belgium. It combines state-of-the-art micromechanical and damage modelling to the evaluation of CVN load-deflection signals, tensile stress-strain curves and slow-bend tests of reconstituted precracked Charpy specimens. A probabilistic micromechanical model has been established for static and dynamic transgranular cleavage initiation fracture toughness in the ductile-brittle transition temperature range. This model allows to project toughness bounds for any steel embrittlement condition from the corresponding CVN and static tensile properties, using a single scaling factor defined by imposing agreement with toughness tests in a single condition. The outstanding finding incorporated by this toughness transfer model is that the microcleavage fracture stress is affected by temperature in the ductile-brittle transition and that this influence is strongly correlated to the flow stress: this explains the shape of the K{sub Ic}n K{sub Id} temperature curves as well as the actual magnitude of the strain rate and irradiation effects. Furthermore, CVN crack arrest loads and fracture appearance are also taken advantage of in order to estimate K{sub Ia} degradation. Finally, the CVN-tensile load-temperature diagram

Heavy-section steel irradiation program. Semiannual progress report, September 1993--March 1994

International Nuclear Information System (INIS)

Corwin, W.R.

1995-04-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only component in the primary pressure boundary for which, if it should rupture, the engineering safety systems cannot assure protection from core damage. It is therefore imperative to understand and be able to predict the capabilities and limitations of the integrity inherent in the RPV. In particular, ft is vital to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance that occurs during service. The Heavy-Section Steel (HSS) Irradiation Program has been established; its primary goal is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties of typical pressure-vessel steels, as they relate to light-water RPV integrity. The program includes the direct continuation of irradiation studies previously conducted within the HSS Technology Program augmented by enhanced examinations of the accompanying microstructural changes. During this period, the report on the duplex-type crack-arrest specimen tests from Phase 11 of the K la program was issued, and final preparations for testing the large, irradiated crack-arrest specimens from the Italian Committee for Research and Development of Nuclear Energy and Alternative Energies were completed. Tests on undersize Charpy V-notch (CVN) energy specimens in the irradiated and annealed weld 73W were completed. The results are described in detail in a draft NUREG report. In addition, the ORNL investigation of the embrittlement of the High Flux Isotope RPV indicated that an unusually large ratio of the high-energy gamma-ray flux to fast-neutron flux is most likely responsible for the apparently accelerated embrittlement

Irradiation damage behavior of low alloy steel wrought and weld materials

International Nuclear Information System (INIS)

Stofanak, R.J.; Poskie, T.J.; Li, Y.Y.; Wire, G.L.

1993-01-01

A study was undertaken to evaluate the irradiation damage response of several different types of low alloy steel: vintage type ASTM A302 Grade B (A302B) plates and welds containing different Ni and Cu concentrations, 3.5% Ni steels similar to ASTM A508 Class 4, welds containing about 1% Ni (similar to type 105S), and 3.5% Ni steels with ''superclean'' composition. All materials were irradiated at several different irradiation damage levels ranging from 0.0003 to 0.06 dpa at 232C (450F). Complete Charpy V-notch impact energy transition temperature curves were generated for all materials before and after irradiation to determine transition temperature at 4IJ (30 ft-lb) or 47J (35 ft-lb) and the upper shelf energy. Irradiation damage behavior was measured by shift in Charpy 41J or 47J transition temperature (ΔTT4 41J or ΔTT 47J ) and lowering of upper shelf Charpy energy at a given irradiation damage level. It was found that chemical composition greatly influenced irradiation damage behavior; highest irradiation damage (greatest ΔTT) was found in an A302B type weld containing 1.28% Ni and 0.20% Cu while the least damage was found in 3.5% Ni, 0.05% Cu, superclean wrought materials. Combination of Ni and Cu was found to affect irradiation damage behavior at higher irradiation damage levels in the A302B welds where the 1.28% Ni, 0.20% Cu weld showed more damage than a 0.60% Ni, 0.31% Cu weld. For the 3.5% Ni steels, fabrication influenced irradiation behavior in that a silicon (Si) killed material showed greater irradiation damage than a low silicon material. In general, the 3.5% Ni materials with low copper showed less irradiation damage than the A302B materials

Irradiation of UO2 specimens with molten cores in a pressurized water loop. Test X-2-x

International Nuclear Information System (INIS)

Bain, A.S.

1961-08-01

Two Zircaloy-2 clad specimens containing stoichiometric UO 2 pellets were irradiated in a pressurized water loop for 379 hours at heat ratings sufficient to cause central melting of the UO 2 . There was no appearance of localized overheating or accelerated corrosion of the sheath, but the diametral increases were considerably larger than those observed in loop specimens irradiated at lower heat ratings. The length increases, however, were approximately the same as those measured for specimens at lower ratings. There was a clearly visible demarcation between UO 2 that had been molten and that which had not. The value of ∫ 500 o C Tm kdθ = 74 ± W/cm was essentially the same as that obtained from the short-duration tests in the Hydraulic Rabbit, indicating there is no marked decrease in thermal conductivity of the UO 2 fuel in irradiations up to 379 hours. (author)

Fracture mechanics investigations within the swiss surveillance programme for the pressure vessel of modern nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Ullrich, G; Krompholz, K [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1994-12-31

In the frame of surveillance programmes of Swiss nuclear power plants, irradiation tests have been performed on tensile, impact and wedge opening load specimens as well as on three point bend-type specimens (for J-integral investigations) and pre-cracked Charpy impact specimens (for dynamical stress intensities K{sub ID}). An experimental method (potential drop technique) is used together with a mathematical procedure which allow for the determination of the stress intensity K{sub IC} for small CT-samples instead of large ones: agreement of these both methods is found excellent, and the mapping of both methods to fatigue pre-cracked small specimens (3 PB and Charpy) is possible. The application of the analysis method to dynamical tests is also possible. 15 refs., 9 figs., 1 tab.

Comparative study for the estimation of To shift due to irradiation embrittlement

International Nuclear Information System (INIS)

Lee, Jin Ho; Park, Youn won; Choi, Young Hwan; Kim, Seok Hun; Revka, Volodymyr

2002-01-01

Recently, an approach called the 'Master Curve' method was proposed which has opened a new means to acquire a directly measured material-specific fracture toughness curve. For the entire application of the Master Curve method, several technical issues should be solved. One of them is to utilize existing Charpy impact test data in the evaluation of a fracture transition temperature shift due to irradiation damage. In the U.S. and most Western countries, the Charpy impact test data have been used to estimate the irradiation effects on fracture toughness changes of RPV materials. For the determination of the irradiation shift the indexing energy level of 41 joule is used irrespective of the material yield strength. The Russian Code also requires the Charpy impact test data to determine the extent of radiation embrittlement. Unlike the U.S. Code, however, the Russian approach uses the indexing energy level varying according to the material strength. The objective of this study is to determine a method by which the reference transition temperature shift (ΔT o ) due to irradiation can be estimated. By comparing the irradiation shift estimated according to the U.S. procedure (ΔT 41J ) with that estimated according to the Russian procedure (ΔT F ), it was found that one-to-one relation exists between ΔT o and ΔT F

A perspective on thermal annealing of reactor pressure vessel materials from the viewpoint of experimental results

International Nuclear Information System (INIS)

Iskander, S.K.; Sokolov, M.A.; Nanstad, R.K.

1996-01-01

It is believed that in the next decade or so, several nuclear reactor pressure vessels (RPVs) may exceed the reference temperature limits set by the pressurized thermal shock screening criteria. One of the options to mitigate the effects of irradiation on RPVs is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory to study the annealing response, or ''recovery'' of several irradiated RPV steels. The fracture toughness is one of the important properties used in the evaluation of the integrity of RPVs. Optimally, the fracture toughness is measured directly by fracture toughness specimens, such as compact tension or precracked Charpy specimens, but is often inferred from the results of Charpy V-notch impact specimens. The experimental results are compared to the predictions of models for embrittlement recovery which have been developed by Eason et al. Some of the issues in annealing that still need to be resolved are discussed

Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

Neutron irradiation of silicon carbide (SiC)-based fuel cladding under a high radial heat flux presents a critical challenge for SiC cladding concepts in light water reactors (LWRs). Fission heating in the fuel provides a high heat flux through the cladding, which, combined with the degraded thermal conductivity of SiC under irradiation, results in a large temperature gradient through the thickness of the cladding. The strong temperature dependence of swelling in SiC creates a complex stress profile in SiCbased cladding tubes as a result of differential swelling. The Nuclear Science User Facilities (NSUF) Program within the US Department of Energy Office of Nuclear Energy is supporting research efforts to improve the scientific understanding of the effects of irradiation on SiC cladding tubes. Ultimately, the results of this project will provide experimental validation of multi-physics models for SiC-based fuel cladding during LWR operation. The first objective of this project is to irradiate tube specimens using a previously developed design that allows for irradiation testing of miniature SiC tube specimens subjected to a high radial heat flux. The previous “rabbit” capsule design uses the gamma heating in the core of the High Flux Isotope Reactor (HFIR) to drive a high heat flux through the cladding tube specimens. A compressible aluminum foil allows for a constant thermal contact conductance between the cladding tubes and the rabbit housing despite swelling of the SiC tubes. To allow separation of the effects of irradiation from those due to differential swelling under a high heat flux, a new design was developed under the NSUF program. This design allows for irradiation of similar SiC cladding tube specimens without a high radial heat flux. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of six rabbit capsules to the HFIR. Rabbits of both low and high

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

International Nuclear Information System (INIS)

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

1998-02-01

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

Analysis of the irradiation data for A302B and A533B correlation monitor materials

International Nuclear Information System (INIS)

Wang, J.A.

1996-04-01

The results of Charpy V-notch impact tests for A302B and A533B-1 Correlation Monitor Materials (CMM) listed in the surveillance power reactor data base (PR-EDB) and material test reactor data base (TR-EDB) are analyzed. The shift of the transition temperature at 30 ft-lb (T 30 ) is considered as the primary measure of radiation embrittlement in this report. The hyperbolic tangent fitting model and uncertainty of the fitting parameters for Charpy impact tests are presented in this report. For the surveillance CMM data, the transition temperature shifts at 30 ft-lb (ΔT 30 ) generally follow the predictions provided by Revision 2 of Regulatory Guide 1.99 (R.G. 1.99). Difference in capsule temperatures is a likely explanation for large deviations from R.G. 1.99 predictions. Deviations from the R.G. 1.99 predictions are correlated to similar deviations for the accompanying materials in the same capsules, but large random fluctuations prevent precise quantitative determination. Significant scatter is noted in the surveillance data, some of which may be attributed to variations from one specimen set to another, or inherent in Charpy V-notch testing. The major contributions to the uncertainty of the R.G. 1.99 prediction model, and the overall data scatter are from mechanical test results, chemical analysis, irradiation environments, fluence evaluation, and inhomogeneous material properties. Thus in order to improve the prediction model, control of the above-mentioned error sources needs to be improved. In general the embrittlement behavior of both the A302B and A533B-1 plate materials is similar. There is evidence for a fluence-rate effect in the CMM data irradiated in test reactors; thus its implication on power reactor surveillance programs deserves special attention

Irradiation Embrittlement Monitoring Programs of RPV's in the Slovak Republic NPP's

International Nuclear Information System (INIS)

Kupca, Ludovik

2006-01-01

Four types of surveillance programs were (are) realized in Slovak NPP's: 'Standard Surveillance Specimen Program' (SSSP) was finished in Jaslovske Bohunice V-2 Nuclear Power Plant (NPP) Units 3 and 4, 'Extended Surveillance Specimen Program' (ESSP), was prepared for Jaslovske Bohunice NPP V-2 with aim to validate the SSSP results, For the Mochovce NPP Unit 1 and 2 was prepared completely new surveillance program 'Modern Surveillance Specimen Program' (MSSP), based on the philosophy that the results of MSSP must be available during all NPP service life, For the Bohunice V-1 NPP was finished 'New Surveillance Specimen Program' (NSSP) coordinated by IAEA, which gave arguments for prolongation of service life these units for minimum 20 years, New Advanced Surveillance Specimen Program (ASSP) for Bohunice V-2 NPP (units 3 and 4) and Mochovce NPP (units 1, 2) is approved now. ASSP is dealing with the irradiation embrittlement of heat affected zone (HAZ) and RPV's austenitic cladding, which were not evaluated till this time in surveillance programs. SSSP started in 1979 and was finished in 1990. ESSP program started in 1995 and will be finished in 2007, was prepared with aim of: increasing of neutron fluence measurement accuracy, substantial improvement the irradiation temperature measurement, fixed orientation of samples to the centre of the reactor core, minimum differences of neutron dose for all the Charpy-V notch and COD specimens, the dose rate effect evaluation. In the year 1996 was started the new surveillance specimen program for the Mochovce RPV's unit-1 and 2, based on the fundamental postulate - to provide the irradiation embrittlement monitoring till the end of units operation. The 'New Surveillance Specimen Program' (NSSP) prepared in the year 1999 for the Bohunice V-1 NPP was finished in the year 2004. Main goal of this program was to evaluate the weld material properties degradation due to the irradiation and recovery efficiency by annealing too. The

Fracture Toughness Evaluation of Kori-1 RPV Beltline Weld for a Long-Term Operation

International Nuclear Information System (INIS)

Lee, Bong-Sang; Kim, Min-Chul; Ahn, Sang-Bok; Kim, Byung-Chul; Hong, Jun-Hwa

2007-01-01

Irradiation embrittlement of RPV (reactor pressure vessel) material is the most important aging issue for a long-term operation of nuclear power plants. KORI unit 1, which is the first PWR in Korea, is approaching its initial licensing life of 30 years. In order to operate the reactor for another 10 years and more, it should be demonstrated that the irradiation embrittlement of the reactor will be adequately managed by ensuring that the fracture toughness properties have a certain level of the safety margin. The current regulation requires Charpy V-notch impact data through conventional surveillance tests. It is based on the assumption that Charpy impact test results are well correlated with the fracture toughness properties of many engineering steels. However, Charpy V-notch impact data may not be adequate to estimate the fracture toughness of certain materials, such as Linde 80 welds. During the last decade, a tremendous number of fracture toughness data on many RPV steels have been produced in accordance with the new standard test method, the so-called master curve method. ASTM E1921 represents a revolutionary advance in characterizing fracture toughness of RPV steels, since it permits establishing the ductile to brittle transition portion of the fracture toughness curve with direct measurements on a relatively small number of relatively small specimens, such as pre-cracked Charpy specimens. Actual fracture toughness data from many different RPV steels revealed that the Charpy test estimations are generally conservative with the exception of a few cases. Recent regulation codes in USA permit the master curve fracture toughness methodology in evaluating an irradiation embrittlement of commercial nuclear reactor vessels

Impact behavior of 9-Cr and 12-Cr ferritic steels after low-temperature irradiation

International Nuclear Information System (INIS)

Klueh, R.L.; Vitek, J.M.; Corwin, W.R.; Alexander, D.J.

1987-01-01

Miniature Charpy impact specimens of 9Cr-1MoVNb and 12Cr-1MoVW steels and these steels with 1 and 2% Ni were irradiated in the High-Flux Isotope Reactor (HFIR) at 50 0 C to displacement damage levels of up to 9 dpa. Nickel was added to study the effect of transmutation helium. Irradiation caused an increase in the ductile-brittle transition temperature (DBTT). The 9Cr-1MoVNb steels, with and without nickel, showed a larger shift than the 12Cr-1MoVW steels, with and without nickel. The results indicated that helium also increased the DBTT. The same steels were previously irradiated at higher temperatures. From the present and past tests, the effect of irradiation temperature on the DBTT behavior can be evaluated. For the 9Cr-1MoVNb steel, there is a continuous decrease in the magnitude of the DBTT increase up to an irradiation temperature of about 400 0 C, after which the shift drops rapidly to zero at about 450 0 C. The DBTT of the 12Cr-1MoVW steel shows a maximum increase at an irradiation temperature of about 400 0 C and less of an increase at either higher or lower irradiation temperatures

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.

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...

Radiation embrittlement in pressure vessels of power reactors

International Nuclear Information System (INIS)

Kempf, Rodolfo; Fortis, Ana M.

2007-01-01

It is presented the project to study the effect of lead factors on the mechanical behavior of Reactor Pressure Vessel steels. It is described the facility designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. The objective is to obtain the fracture behavior of irradiated specimens with different lead factors and to know their dependence with the diffusion of alloy elements. (author) [es

Neutron irradiation embrittlement of reactor pressure vessel steel 20 MnMoNi55 weld

International Nuclear Information System (INIS)

Ghoneim, M.M.

1987-05-01

The effect of neutron irradiation on the mechanical and fracture properties of an 'improved' 20 MnMoNi 55 Pressure Vessel Steel (PVS) weld was investigated. In addition to very low residual element content, especially Cu (0.035 wt.%), and relatively higher Ni content (0.9 wt.%), this steel has higher strength (30% more) than the steels used currently in nuclear reactor pressure vessels. The material was irradiated to 3.5x10 19 and 7x10 19 n/cm 2 (E > 1 Mev) at 290 0 C and 2.5x10 19 n/cm 2 (E > 1 MeV) at 160 0 C in FRJ-1 and FRJ-2 research reactors at KFA, Juelich, F.R.G. Test methods used in the evaluation included instrumented impact testing of standard and precracked Charpy specimens, tensile, and fracture toughness testing. Instrumented impact testing provided load and energy vs. time (deflection) data in addition to energy absorption data. The results indicated that the investigated high strength improved steel is more resistant to irradiation induced embrittlement than conventional PVSs. (orig./IHOE)

Methods and devices for small specimen testing at the Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Jitsukawa, Shiro; Kizaki, Minoru; Umino, Akira; Shiba, Kiyoyuki; Hishinuma, Akimichi

1993-01-01

Devices for a punch test on annular notched specimens, small punch (SP) tests, and miniaturized tension tests in hot cells were developed. A micro-manipulator to handle small specimens and an electro-discharge machine (EDM) to extract miniaturized tension specimens and annular notched specimens from transmission electron microscopy (TEM) disks were also fabricated. These devices were designed and made for remote operation in hot cells. Preliminary tests to evaluate the applicability of test methods were carried out. Correlation between SP test results and tensile properties was not strong. Miniaturized tensile results were reasonably similar to the results with larger specimens. The ductile-brittle transition temperature (DBTT) by the punch test on annular notched specimens was higher than that obtained from the SP test. However, materials dependence of the DBTT was different from that measured by standard Charpy V-notch (CVN) tests. This may be due to a specimen size effect

Fracture toughness evaluation in the transition region of reactor pressure vessel steel

International Nuclear Information System (INIS)

Onizawa, K.; Suzuki, M.

1995-01-01

The fracture toughness (K jc and Jc) values at the cleavage fracture initiation in the transition region of a RPV steel were investigated using mainly precracked Charpy specimens. A conventional statistical approach and a fractographic study were applied to analyze the scatter of the fracture toughness values from precracked Charpy specimens. The material used was an ASTM A533B class 1 steel, which was designated as an IAEA correlation monitor material, JRQ. A lower bound transition curve of the fracture toughness for unirradiated condition was determined by the 5% confidence limit from the Weibull and fractographic analyses. The lower bound transition curve after irradiation was evaluated based on the statistics of unirradiated specimens. The results indicated that the shift of the fracture toughness transition curbe were somewhat larger than the Charpy 41J transition temperature. The parameters to determine the lower bound toughness such as the Weibull slope and the amount of ductile crack growth are discussed. The results are also compared with a model based on weakest link theory. (author). 12 refs, 12 figs, 5 tabs

Irradiation temperature measurement of the reactor pressure vessel surveillance specimen in the programmes of radiation degradation monitoring

International Nuclear Information System (INIS)

Kupca, L.; Stanc, S.; Simor, S.

2001-01-01

The information's about the special system of irradiation temperature measurement, used for reactor pressure vessel surveillance specimen, which are placed in reactor thermal shielding canals are presented in the paper. The system was designed and realized in the frame of Extended Surveillance Specimen Programme for NPP V-2 Jaslovske Bohunice and Modern Surveillance Specimen Programme for NPP Mochovce. Base design aspects, technical parameters of realization and results of measurement on the two units in Bohunice and Mochovce NPPs are presented too. (Authors)

Shear compression testing of glass-fibre steel specimens after 4K reactor irradiation: Present status and facility upgrade

International Nuclear Information System (INIS)

Gerstenberg, H.; Kraehling, E.; Katheder, H.

1997-01-01

The shear strengths of various fibre reinforced resins being promising candidate insulators for superconducting coils to be used tinder a strong radiation load, e.g. in future fusion reactors were investigated prior and subsequent to reactor in-core irradiation at liquid helium temperature. A large number of sandwich-like (steel-bonded insulation-steel) specimens representing a widespread variety of materials and preparation techniques was exposed to irradiation doses of up to 5 x 10 7 Gy in form of fast neutrons and γ-radiation. In a systematic study several experimental parameters including irradiation dose, postirradiation storage temperature and measuring temperature were varied before the determination of the ultimate shear strength. The results obtained from the different tested materials are compared. In addition an upgrade of the in-situ test rig installed at the Munich research reactor is presented, which allows combined shear/compression loading of low temperature irradiated specimens and provides a doubling of the testing rate

Analysis of ductile-brittle transition shifts for standard and miniature bending specimens of irradiated steel

International Nuclear Information System (INIS)

Korshunov, M.E.; Korolev, Yu.N.; Krasikov, E.A.; Gabuev, N.N.; Tykhmeev, D.Yu.

1996-01-01

A study is made to reveal if there is a correlation between shifts in temperature curves obtained when testing thin plates and standard specimens on impact bending and fracture toughness. The tests were carried out using steel 25Kh3NM specimens irradiated by 6 x 10 19 cm -2 neutron fluence. A conclusion is made about the possibility to evaluate the degree of radiation-induced embrittlement of reactor steels on the basis of thin plate testing under quasistatic loads [ru

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.

Fracture toughness of A533B Part III - variability of A533B fracture toughness as determined from Charpy data

International Nuclear Information System (INIS)

Druce, S.G.; Eyre, B.L.

1978-08-01

This is the final part of a series of three reports examining the upper shelf fracture toughness of A533B Class 1 pressure vessel steel. Part I (AERE R 8968) critically reviews the current elasto plastic fracture mechanics methodologies employed to characterise toughness following extensive yielding and Part II (AERE R 8969) examines several sources of fracture mechanics data pertinent to A533B Class 1 in the longitudinal (RW) orientation. Part III is a review of the effects of (i) position and orientation within the plate (ii) welding processes and post weld heat treatment and (iii) neutron irradiation as measured by Charpy impact testing. It is concluded that the upper shelf factor energy is dependent on orientation and position and can be reduced by welding, extended post weld heat treatments and neutron irradiation. Neutron irradiation effects are known to be strongly dependent on composition and metallurgical conditions, but an explanation for the variability following extended post weld treatments has yet to be resolved. (author)

FFTF irradiation of fracture mechanics specimens for out-of-core structures

International Nuclear Information System (INIS)

King, D.C.

1978-09-01

The National Program Plan has established data requirements for out-of-core structures for FBRs. Significant FFTF irradiation space with moderate gamma heating levels is required to irradiate relatively large fracture mechanics specimens to total neutron fluences ranging between 5 x 10 21 and 5 x 10 22 n/cm 2 and temperatures which range between 400 0 C (750 0 F) and 650 0 C (1200 0 F). Priority 1 data on stainless steel welds requires a test volume of 7443 cm 3 (454 in 3 ). Priority 2 data on 304 and 316 SS and Inconel 718 materials and Inconel 718 welds requires 2760 cm 3 (168 in 3 ). Priority 3 data on stainless steels, other nickel-base alloys, and ferritics requires 33,118 cm 3 (2021 in 3 ). Priority 4 data at elevated temperatures on stainless steels, other nickel-base alloys and ferritics requires 69,182 cm 3

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

International Nuclear Information System (INIS)

Gelles, D.S.; Shibayama, T.

1998-01-01

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

Reduction of upper shelf energy of highly irradiated RPV steels

Energy Technology Data Exchange (ETDEWEB)

Otaka, M.; Osaki, T. [Japan Nuclear Energy Safety Organization (Japan)

2004-07-01

It is well known that as the embrittlement due to neutron irradiation of reactor pressure vessel (RPV) steels, there is the tendency of the decrease in Charpy absorbed energy at upper shelf region (USE), in addition to the shift of ductile-brittle transition temperature. Concerning to the regulation of the upper shelf region, no method is provided to evaluate integrity for RPV steels with USE of less than 68J in Japanese codes. Under the circumstance, the reduction tendency of USE using simulated Japanese RPV steels, irradiated by fast neutron up to 1 x 10{sup 24} n/m{sup 2}, E>1 MeV in the OECD Halden test reactor, was investigated to establish the basis of the USE prediction after 60 year plant operation for the integrity assessment of the RPVs. This paper describes the results of an atom probe tomography characterization of irradiated steels. A new form of USE prediction equation was developed based on the atom probe tomography characterization and the Charpy impact test results of the irradiated steels. And, the USE prediction equations have been determined through the regression analysis of the test reactor data combined with Japanese surveillance test data. (orig.)

Elastic-plastic analysis of fracture mechanics test specimens. Part 2

International Nuclear Information System (INIS)

Talja, H.; Wallin, K.

1984-12-01

This is second part of the report of the research program 'Comparisons between computational and experimental elastic-plastic results' started at the Technical Research Centre of Finland in 1981. The first part of the research program was reported earlier and contained a two dimensional linear elastic finite element analysis of four specimen geometries (CT, RCT, ASTM-3P and Charpy-V) and testing and elastic-plastic analysis of the specimen (EGF71; 1TCT, material A 542). In this report the second part of the program containing the testing and 2-D elastic-plastic analyses of five specimens is described. The four specimen geometries mentioned above and two different materials (stainless steel AISI 304 and ferrite pressure vessel steel A533B) are considered. The following comparisons are presented in the report: load vs. load displacement curves, J-integral, crack opening displacement (COD), J vs. COD and the size of the plastic zone. The agreement between the computational and experimental results is quite good. Complete agreement can be achieved only with 3-dimensional calculation models. (author)

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).

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

International Nuclear Information System (INIS)

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 3 and H 2 O 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)

In-situ tritium recovery from Li2O irradiated in fast neutron flux - Beatrix-II temperature change specimen

International Nuclear Information System (INIS)

Slagle, O.D.; Hollenberg, G.W.; Kurasawa, T.; Verrall, R.A.

1992-01-01

The Beatrix-II irradiation experiment is an in-situ tritium release experiment to evaluate the stability and tritium release characteristics of Li 2 O under fast neutron irradiation to extended burnups. A thin annular ring specimen capable of temperature changes was irradiated in Phase I of the experiment to a lithium burnup of 5%. The primary emphasis of the test plan was to determine the effect and interrelationship of gas composition and temperature on the tritium inventory with increasing temperature and a series of specific temperature changes were carried out at intervals throughout the experiment to characterize the effect of burnup. Decreasing the amount of hydrogen in the sweep gas resulted in an increase in the tritium inventory in the Li 2 O specimen. The tritium recovery during startup and shutdown was observed to be strongly influenced by the composition of the sweep gas

Analysis of mechanical property data obtained from nuclear pressure vessel surveillance capsules

International Nuclear Information System (INIS)

Perrin, J.S.

1977-01-01

A typical pressure vessel surveillance capsule examination program provides mechanical property data from tensile, Charpy V-notch impact, and, in some cases, fracture mechanics specimens. This data must be analyzed in conjunction with the unirradiated baseline mechanical property data to determine the effect of irradiation on the mechanical properties. In the case of Charpy impact specimens, for example, irradiation typically causes an increase in the transition temperature, and a decrease in the upper shelf energy level. The results of the Charpy impact and other mechanical specimen tests must be evaluated to determine if property changes are occurring in the manner expected when the reactor was put into service. The large amount of data obtained from surveillance capsule examinations in recent years enables one to make fairly good predictions. After the changes in the mechanical properties of specimens from a particular surveillance capsule have been experimentally determined and evaluated, they must be related to the reactor pressure vessel. This requires a knowledge of the neutron fluence of the surveillance capsule, and the ratio of the surveillance capsule fluence to the pressure vessel wall fluence. This ratio is frequently specified by the reactor manufacturer, or can be calculated from a knowledge of the geometry and materials of the reactor components inside the pressure vessel. A knowledge of the exact neutron fluence of the capsule specimens and the capsule to vessel wall neutron fluence ratio is of great importance, since inaccuracies in these numbers cause just as serious a problem as inaccuracies in the mechanical property determinations. A further area causing analysis difficulties is problems encountered in recent capsule programs relating to capsule design, construction, operation, and dismantling. (author)

Effects of thermal aging and neutron irradiation on the mechanical properties of three-wire stainless steel weld overlay cladding

International Nuclear Information System (INIS)

Haggag, F.M.; Nanstad, R.K.

1997-05-01

Thermal aging of three-wire series-arc stainless steel weld overlay cladding at 288 degrees C for 1605 h resulted in an appreciable decrease (16%) in the Charpy V-notch (CVN) upper-shelf energy (USE), but the effect on the 41-J transition temperature shift was very small (3 degrees C). The combined effect of aging and neutron irradiation at 288 degrees C to a fluence of 5 x 10 19 neutrons/cm 2 (> 1 MeV) was a 22% reduction in the USE and a 29 degrees C shift in the 41-J transition temperature. The effect of thermal aging on tensile properties was very small. However, the combined effect of irradiation and aging was an increase in the yield strength (6 to 34% at test temperatures from 288 to -125 degrees C) but no apparent change in ultimate tensile strength or total elongation. Neutron irradiation reduced the initiation fracture toughness (J Ic ) much more than did thermal aging alone. Irradiation slightly decreased the tearing modulus, but no reduction was caused by thermal aging alone. Other results from tensile, CVN, and fracture toughness specimens showed that the effects of thermal aging at 288 or 343 degrees C for 20,000 h each were very small and similar to those at 288 degrees C for 1605 h. The effects of long-term thermal exposure time (50,000 h and greater) at 288 degrees C will be investigated as the specimens become available in 1996 and beyond

Single specimen fracture toughness determination procedure using instrumented impact test

International Nuclear Information System (INIS)

Rintamaa, R.

1993-04-01

In the study a new single specimen test method and testing facility for evaluating dynamic fracture toughness has been developed. The method is based on the application of a new pendulum type instrumented impact tester equipped with and optical crack mouth opening displacement (COD) extensometer. The fracture toughness measurement technique uses the Double Displacement Ratio (DDR) method, which is based on the assumption that the specimen is deformed as two rigid arms that rotate around an apparent centre of rotation. This apparent moves as the crack grows, and the ratio of COD versus specimen displacement changes. As a consequence the onset ductile crack initiation can be detected on the load-displacement curve. Thus, an energy-based fracture toughness can be calculated. In addition the testing apparatus can use specimens with the Double ligament size as compared with the standard Charpy specimen which makes the impact testing more appropriate from the fracture mechanics point of view. The novel features of the testing facility and the feasibility of the new DDR method has been verified by performing an extensive experimental and analytical study. (99 refs., 91 figs., 27 tabs.)

Prediction of fracture toughness K/sub Ic/ of steel from Charpy impact test results

Energy Technology Data Exchange (ETDEWEB)

Iwadate, Tadao; Tanaka, Yasuhiko; Takemata, Hiroyuki; Terashima, Shuhei

1986-08-01

This paper presents a method to predict the fracture toughness K/sub Ic/ and/or K/sub Id/ of steels using their Charpy impact test results and tensile properties. The fracture toughness, Charpy impact and tensile properties of 2 1/4 Cr-1Mo, ASTM A508 Cl.1, A508 Cl.2 A508 Cl.3 and A533 Gr.B Cl.1 steels were measured and analysed on the basis of the excess temperature (test temperature minus FATT) and Rolfe-Novak correlation. The relationship between K/sub Ic//K/sub Ic-us/ and the excess temperature, where K/sub Ic-us/ is the upper-shelf fracture toughness K/sub Ic/ predicted by Rolfe-Novak correlation, discloses that the K/sub Ic/ transition curves of several steels are representable by only one trend curve of K/sub Ic//K/sub Ic-us/ or K/sub Id//K/sub Id-us/ versus excess temperature relation. This curve is denoted as a ''master curve''. By using this curve, the fracture toughness of steel can be predicted using Charpy impact and tensile test results. By taking account of the scattering of both the fracture toughness and Charpy impact test results, the confidence limits of the master curve were also determined. Another approach to develop more general procedure of predicting the fracture toughness K/sub Ic/ is also discussed.

Effect of neutron irradiation on the impact properties of A533B steel

International Nuclear Information System (INIS)

Schubert, L.E.; Kumar, A.S.; Rosinski, S.T.; Hamilton, M.L.

1994-01-01

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, ASTM type A 533 Grade B (A533B) having a low USE (USE 19 n/cm 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

Experimental assessments of notch ductility and tensile strength of stainless steel weldments after 1200C neutron irradiation

International Nuclear Information System (INIS)

Hawthorne, J.R.; Menke, B.H.; Awadalla, N.G.; O'Kula, K.R.

1986-01-01

The Charpy-V (C/sub v/) properties of AISI 300 series stainless steel plate, weld, and weld heat-affected zone (HAZ) materials from commercial production weldments in 406-mm-diameter pipe (12.7-mm wall) were investigated in unirradiated and irradiated conditions. Weld and HAZ tensile properties were also assessed in the two conditions. The plates and weld filler wires represent different steel melts; the welds were produced using the multipass MIG process. Weldment properties in two test orientations were evaluated. Specimens were irradiated in the UBR reactor to 1 x 10 20 n/cm 2 , E >0.1 MeV in a controlled temperature assembly. Specimen tests were performed at 25 0 C and 125 0 C. The radiation-induced reductions in C/sub v/ energy absorption at 25 0 C were about 42 percent for the weld and HAZ materials evaluated. A trend of energy increase with temperature was observed. The concomitant elevation in yield strength was about 53%. In contrast, the increase in tensile strength was only 16%. The postirradiation yield strength of the axial test orientation in the pipe was less than that of the circumferential test orientation. Results for the HAZ indicate that this component may be the weakest link in the weldment from a fracture resistant viewpoint

Embrittlement of a 17Cr ferritic steel irradiated in Phenix

International Nuclear Information System (INIS)

Allegraud, G.; Boutard, J.L.; Boyer, J.M.

1987-01-01

Charpy V and tensile tests have been performed with samples made of 17Cr ferritic steel irradiated in Phenix at temperatures between 390 and 540C up to a maximum dose of 83.3 dpaF. All over the temperature and dose ranges, irradiation leads to an increase of the ductile brittle transition temperature (DBTT). The DBTT and hardening are decreasing functions of the irradiation temperature. Fast neutron flux at 390C hardens the material more than a sole thermal ageing does

Heavy-section steel irradiation program: Embrittlement issues

International Nuclear Information System (INIS)

Corwin, W.R.

1991-01-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents and the potential for major contamination releases. The RPV is one of only two major safety- related components of the plant for which a duplicate or redundant backup system does not exist. In particular, it is vital to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance which occurs during service, since without that radiation damage it is virtually impossible to postulate a realistic scenario which would result in RPV failure. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established by the US Nuclear Regulatory Commission (USNRC) to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water reactor pressure-vessel integrity. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and postirradiation annealing are being examined on a wide range of fracture properties including fracture toughness crack arrest toughness ductile tearing resistance Charpy V-notch impact energy, dropweight nil-ductility temperature and tensile properties. Models based on observations of radiation-induced microstructural changes using the field on microprobe and the high resolution transmission electron microscopy provide improved bases for extrapolating the measured changes in fracture properties to wider ranges of irradiation conditions. The principal materials examined within the HSSI program are high-copper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs

Gamma irradiation test report of simulated grout specimens for gas generation/liquid advection

International Nuclear Information System (INIS)

Hinman, C.A.

1994-01-01

This report presents the results from an irradiation test performed on four specimens of grout that were fabricated from synthetic Double Shell Slurry Feed (DSSF) liquid waste. The objective was to investigate the radiolytic generation of gases and the potential for advective rejection of waste liquids from the grout matrix and to provide experimental information for the validation of the C-Cubed calculated model. It has been demonstrated that a number of gases can be formed within the grout due to radiolytic decomposition of various chemical components that make up the grout. This observation leads to the conjecture that the potential exists for the rejection of a portion of the 60 vol% free liquid from the grout matrix driven by pressurization by these gases. It was found that, for the specimen geometries used in this test series, and for peak radiation dose accumulation rates on the order of 4 to 60 times of the initial rate expected in the grout vaults (300 Rads/hr), no liquid rejection was observed from 2% to 35% of the target exposure expected in the grout vaults (1E+08 Rads). When the irradiation rate exceeded the projected grout vault dose rate by a factor of 200 a small amount of liquid rejection was observed from one of two specimens that had received 20% more than the goal exposure. Because of the differences in the magnitudes of the relative radiation field strengths between this study and an actual grout vault, it is concluded that the potential for liquid rejection by internal gas pressurization from presently configured grout waste forms is very low for the expected conditions

The interpretation of Charpy impact test data using hyper-logistic fitting functions

International Nuclear Information System (INIS)

Helm, J.L.

1996-01-01

The hyperbolic tangent function is used almost exclusively for computer assisted curve fitting of Charpy impact test data. Unfortunately, there is no physical basis to justify the use of this function and it cannot be generalized to test data that exhibits asymmetry. Using simple physical arguments, a semi-empirical model is derived and identified as a special case of the so called hyper-logistic equation. Although one solution of this equation is the hyperbolic tangent, other more physically interpretable solutions are provided. From the mathematics of the family of functions derived from the hyper-logistic equation, several useful generalizations are made such that asymmetric and wavy Charpy data can be physically interpreted

Technology development on analysis program for measuring fracture toughness of irradiated specimens

International Nuclear Information System (INIS)

Shibata, Akira; Takada, Fumiki

2007-03-01

The fracture toughness which represents resistance for brittle or ductile fracture is one of the most important material property concerning linear and non-linear fracture mechanics analyses. In order to respond to needs of collecting data relating to fracture toughness of pressure vessel and austenitic stainless steels, fracture toughness test for irradiated materials has been performed in JMTR hot laboratory. On the other hand, there has been no computer program for analysis of fracture toughness using the test data obtained from the test apparatus installed in the hot cell. Therefore, only load-displacement data have been provided to users to calculate fracture toughness of irradiated materials. Recently, request of analysis of fracture toughness have been increased. Thus a computer program, which calculates the amount of the crack extension, the compliance and the fracture toughness from the data acquired from the test apparatus installed in the hot cell, has been developed. In the program unloading elastic compliance method is applied based on ASTM E1820-01. Through the above development, the request for the fracture toughness analysis can be satisfied and the fracture toughness of irradiated test specimens can be provided to users. (author)

Heavy-section steel irradiation program summary

International Nuclear Information System (INIS)

Corwin, W.R.; Nanstad, R.K.; Iskander, S.K.; Haggag, F.M.

1992-01-01

Since a failure of the RPV carries the potential of major contamination release and severe accident, it is imperative to safe reactor operation to understand and be able to accurately predict failure models of the vessel material. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established with its primary goal to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water RPVs. The program includes the direct continuation of irradiation studies previously conducted within the Heavy-Section Steel Technology Program augmented by enhanced examinations of the accompanying microstructural changes. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and postirradiation annealing are being examined on a wide range of fracture properties including fracture toughness (K Ic and J Ic ), crack-arrest toughness (K Ia ), ductile tearing resistance (dJ/da), Charpy V-notch impact energy, dropweight nil-ductility temperature (NDT), and tensile properties. Models based on observations of radiation-induced microstructural changes using field ion and high-resolution transmission electron microscopy provide a firmer basis for extrapolating the measured changes in fracture properties to wider ranges of irradiation conditions. The principal materials examined within the HSSI Program are highcopper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs. In addition, a limited effort will focus on stainless steel weld overlay cladding, typical of that used on the inner surface of RPVs, since its postirradiation fracture properties have the potential for strongly affecting the extension of small surface flaws during overcooling transients. (orig./GL)

Significance of Charpy and COD tests in the determination of fracture toughness of welds

International Nuclear Information System (INIS)

Caminha Junior, H.M.; Bastian, F.L.

1983-01-01

A comparison is made between the Charpy and crack opening displacement (COD) tests used to acess the fracture toughness of metallic materials. The main problems inherent in these tests are discussed, such as scatter of results and their advantages and limitations. The chief experimental difficulties when they are applied to welds are indicated and the various methods available for calculating the COD from a test graph are described. Comments are made on the use of the Charpy test and the methods of calculating the COD in determing critical defect sizes in welded structures. (Author) [pt

Ductility and failure behaviour of both unirradiated and irradiated zircaloy-4 cladding using plane strain tensile specimens

International Nuclear Information System (INIS)

Carassou, S.; Le Saux, M.; Pizzanelli, J.P.; Rabouille, O.; Averty, X.; Poussard, C.; Cazalis, B.; Desquines, J.; Bernaudat, C.

2010-01-01

In this work, eight PST (Plan Strain Tensile) tests machined from a Zircaloy-4 (Zy-4) cladding irradiated up to 5 annual cycles have been performed at 280, 350 and 480 Celsius degrees. The specimen displacements during the tests were filmed and digitally recorded to allow the use of a Digital Image Correlation (DIC) analysis technique to experimentally determine the local strains on the outer surface of the specimens. The plane strain conditions have been verified and prevail over a wide area between the notches of the specimen, as expected from full 3D FE numerical analysis performed in support of the tests. For the first time, the location of the onset of fracture for this geometry on irradiated material has been experimentally observed: at 280 C.degrees, crack initiates in the vicinity of the notches, in an area where plane strain conditions are not fulfilled, and for a local circumferential strain value of about 5%. At 350 C. degrees and 480 C. degrees, cracks initiate at a location where plane strain conditions prevail, for circumferential strain values respectively close to 10% and greater than 50%. These results have been compared to results obtained previously by similar test on fresh and hydrided material, as well as tests performed as support to the study. At 350 C. degrees, the homogeneous 700 ppm hydrided Zy-4 and the Zy-4 irradiated during 5 annual cycles exhibit similar fracture behaviour, for both fracture hoop strain values (10%) and fracture mode (through-wall slant fracture). For the irradiated material, it has clearly been established that at 350 C. degrees, a brittle fracture occurs at the outer surface in the hydride rim. The crack propagates subsequently toward the inner surface and the notches, where final fracture occurs

The development of fuel pins and material specimens mixed loading irradiation test rig in the experimental fast reactor Joyo. The development of the fuel-material hybrid rig

International Nuclear Information System (INIS)

Oyamatsu, Yasuko; Someya, Hiroyuki

2013-02-01

In the experimental fast reactor Joyo, there were many tests using the irradiation rigs that it was possible to be set irradiation conditions for each compartment independently. In case of no alternative fuel element to irradiate after unloading the irradiated compartments, the irradiation test was restarted with the dummy compartment which the fuel elements was not mounted. If the material specimens are mounted in this space, it is possible to use the irradiation space effectively. For these reasons, the irradiation rig (hybrid rig) is developed that is consolidated with material specimens compartment and fuel elements compartment. Fuel elements and material specimens differ greatly with heat generation, so that the most important issue in developing of hybrid rig is being able to distribute appropriately the coolant flow which satisfies irradiation conditions. The following is described by this report. (1) It was confirmed that the flow distribution of loading the same irradiation rig with the compartment from which a flow demand differs could be satisfied. (2) It was confirmed that temperature setting range of hybrid rig could be equivalent to that of irradiation condition. (3) By standardizing the coolant entrance structure of the compartment lower part, the prospect which can perform easily recombination of the compartment from which a type differs between irradiation rigs was acquired. (author)

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.

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

International Nuclear Information System (INIS)

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 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 s tandard test method for determination of reference temperature, T 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

Feasibility Study of Laser Cutting for Fabrication of Tensile Specimen

International Nuclear Information System (INIS)

Jin, Y. G.; Baik, S. J.; Kim, G. S.; Heo, G. S.; Yoo, B. O.; Ahn, S. B.; Chun, Y. B.

2015-01-01

The specimen fabrication technique was established to machine the specimen from the irradiated materials. The wire cut EDM(electric discharge machine) was modified to fabricate the mechanical testing specimens from irradiated components and fuel claddings. The oxide layer removal system was also developed because the oxide layer on the surface of the irradiated components and claddings interrupted the applying the electric current during the processing. However, zirconium oxide is protective against further corrosion as well as beneficial to mechanical strength for the tensile deformation of the cladding. Thus, it is important to fabricate the irradiated specimens without removal of oxide layer on the surface of the irradiated structural components and claddings. In the present study, laser cutting system was introduced to fabricate the various mechanical testing specimens from the unirradiated fuel cladding and the feasibility of the laser cutting system was studied for the fabrication of various types of irradiated specimens in a hot cell at IMEF (Irradiated Materials Examination Facility) of KAERI. Laser beam machining system was introduced to fabricate the various mechanical testing specimens from the unirradiated fuel cladding and the dimensions were compared for the feasibility of the laser cutting system. The effect of surface oxide layer was also investigated for machining process of the zircaloy-4 fuel cladding and it was found that laser beam machining could be a useful tool to fabricate the specimens with surface oxide layer

Feasibility Study of Laser Cutting for Fabrication of Tensile Specimen

Energy Technology Data Exchange (ETDEWEB)

Jin, Y. G.; Baik, S. J.; Kim, G. S.; Heo, G. S.; Yoo, B. O.; Ahn, S. B.; Chun, Y. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

The specimen fabrication technique was established to machine the specimen from the irradiated materials. The wire cut EDM(electric discharge machine) was modified to fabricate the mechanical testing specimens from irradiated components and fuel claddings. The oxide layer removal system was also developed because the oxide layer on the surface of the irradiated components and claddings interrupted the applying the electric current during the processing. However, zirconium oxide is protective against further corrosion as well as beneficial to mechanical strength for the tensile deformation of the cladding. Thus, it is important to fabricate the irradiated specimens without removal of oxide layer on the surface of the irradiated structural components and claddings. In the present study, laser cutting system was introduced to fabricate the various mechanical testing specimens from the unirradiated fuel cladding and the feasibility of the laser cutting system was studied for the fabrication of various types of irradiated specimens in a hot cell at IMEF (Irradiated Materials Examination Facility) of KAERI. Laser beam machining system was introduced to fabricate the various mechanical testing specimens from the unirradiated fuel cladding and the dimensions were compared for the feasibility of the laser cutting system. The effect of surface oxide layer was also investigated for machining process of the zircaloy-4 fuel cladding and it was found that laser beam machining could be a useful tool to fabricate the specimens with surface oxide layer.

Test and evaluation of pressure vessel materials

International Nuclear Information System (INIS)

Choi, Sun Pil; Hong, Jun Hwa; Nho, Kye Hoe; Han, Dae June; Chi, Se Hwan

1985-01-01

We have prepared a method for analyzing the Charpy impact test data, which is deduced from ''the standard anelastic solid equation''. The theoretical expression for the absorbed energy is in a form of W=Wsub(U)+(Wsub(R)-Wsub(U))/ [1+(ωtau) 2 ] showing the Debye characteristics and where tau is given by the Arrhenius equation; tau=tau 0 exp(ΔH/ksub(B)T). Four measurable parameters, at the present stage, can characterize the dynamic hehavior of cracking (Charpy impact result). They are the upper shelf energy(Wsub(R), the lower shelf energy (Wsub(U)), the activation energy of crack (ΔH, and wtau(0) where w tau(0) are the resonance frequency of the specimen and the jumping pre-exponential factor of propagating crack respectively. However the states of R (relaxed) and U (un-relaxed) should be defined from reasonable physical conditions in the future and it is possible that Wsub(U) is small enough to be taken as zero. The effects of irradiation, alloying elements, and heat treatment on the impact results should be interpreted as changes in the above characteristic parameters. The present method has been applied for weld metal of SA 508-2 irradiated up to a fluence of 4x10 18 n/cm 2 , E>1.0Mev, resulting in about 29% decrease in Wsub(R), negligible change in Wsub(U), 5.6 times increase in ωtau 0 , and no change in ΔH. This seems to indicate that irradiation degrades an average value of YOUNG's modulus so that cracks propagate more easily and it does not effect on breaking the lattice bond. However much more systematic analyses should be necessary for correct judgment. It is concluded that the present method is quite adequate for analyzing the Charpy impact data even though plastic deformation in the specimen was not considered separately so that the method should be applied for various cases in order to evaluate the proper trend of effects of irradiation, alloying elements, and heat treatment on the Charpy impact results. (Author)

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.

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

International Nuclear Information System (INIS)

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

1997-01-01

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 ∼19.5 dpa at 365 degrees C and to ∼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

Comparison of immersion density and improved microstructural characterization methods for measuring swelling in small irradiated disk specimens

International Nuclear Information System (INIS)

Sawai, T.; Suzuki, M.; Hishinuma, A.; Maziasz, P.J.

1992-01-01

The procedure of obtaining microstructural data from reactor-irradiated specimens has been carefully checked for accuracy by comparison of swelling data obtained from transmission electron microscopy (TEM) observations of cavities with density-change data measured using the Precision Densitometer at Oak Ridge National Laboratory (ORNL). Comparison of data measured by both methods on duplicate or, in some cases, on the same specimen has shown some appreciable discrepancies for US/Japan collaborative experiments irradiated in the High Flux Isotope Reactor (HFIR). The contamination spot separation (CSS) method was used in the past to determine the thickness of a TEM foil. Recent work has revealed an appreciable error in this method that can result in an overestimation of the foil thickness. This error causes lower swelling values to be measured by TEM microstructural observation relative to the Precision Densitometer. An improved method is proposed for determining the foil thickness by the CSS method, which includes a correction for the systematic overestimation of foil thickness. (orig.)

Experimental assessments of notch ductility and tensile strength of stainless steel weldments after 1200C neutron irradiation

International Nuclear Information System (INIS)

Hawthorne, J.R.; Menke, B.H.; Awadalla, N.G.; O'Kula, K.R.

1987-01-01

The Charpy-V (C/sub V/) properties of American Iron and Steel Institute (AISI) 300 series stainless steel plate, weld, and weld heat-affected zone (HAZ) materials from commercial production weldments in 406-mm-diameter pipe (12.7-mm wall) were investigated in unirradiated and irradiated conditions. Weld and HAZ tensile properties were also assessed in the two conditions. The plates and weld filler wires represent different steel melts; the welds were produced using the multipass metal inert gas (MIG) process. Weldment properties in two test orientations were evaluated. Specimens were irradiated in a light water cooled and moderated reactor to 1 x 10/sup 20/ n/cm/sup 2/, E > 0.1 MeV, using a controlled temperature assembly. Specimen tests were performed at 25 and 125 0 C. The radiation-induced reductions in C/sub V/ energy absorption at 25 0 C were about 42% for the weld and the HAZ materials evaluated. A trend of energy increase with temperature was observed. The concomitant elevation in yield strength was about 53%. The increase in tensile strength in contrast was only 16%. The postirradiation yield strength of the axial test orientation in the pipe was less than that of the circumferential test orientation. Results for the HAZ indicate that this component may be the weakest link in the weldment from a fracture resistance viewpoint

Post-irradiation annealing of coarse-grained model alloys

Energy Technology Data Exchange (ETDEWEB)

Ray, P H.N.; Wilson, C; McElroy, R J [AEA Reactor Services, Harwell (United Kingdom)

1994-12-31

Thermal ageing and irradiation studies have been carried out on three model alloys (JPC, JPB, JPG) that have identical compositions except for different levels of phosphorus and/or copper. They have been irradiated in three conditions, as-received, heat treated to produce a coarse grained microstructure (similar to heat-affected-zone), and in this condition further aged at 450 C to produce a temper embrittled condition. One of the alloy have been subject to a post-irradiation anneal. The effect of these treatments on mechanical property changes has been characterized by Charpy testing and Vickers hardness measurements; the phosphorus segregation has been studied by a combination of STEM and Auger techniques.

SISTEMA DE AQUISIÇÃO DE DADOS PARA A MÁQUINA DE IMPACTO CHARPY

Directory of Open Access Journals (Sweden)

Jermana Lopes Moraes

2014-08-01

Full Text Available Este trabalho tem o objetivo de implementar e desenvolver um sistema de aquisição de dados para a máquina de impacto Charpy. Assim, é realizado um estudo da máquina de impacto Charpy, do ensaio de impacto e das ferramentas necessárias para desenvolvimento do projeto. Utiliza-se um acelerômetro para determinar a aceleração nos eixos x e y do pêndulo Charpy durante a realização do ensaio. Para leitura e interpretação dos dados enviados pelo acelerômetro utiliza-se a plataforma de hardware Arduino UNO com software específico. Os dados enviados ao Arduino são apresentados em uma interface gráfica desenvolvida no Matlab. Nesta interface é possível inserir os dados iniciais de ensaio Charpy e apresentar ao usuário final os resultados finais de ensaio, como a energia de impacto, a resistência de impacto e a força necessária para romper o corpo de prova. Além disso, é apresentado ao usuário um gráfico da aceleração ao longo da realização do ensaio e o gráfico de força ao longo do tempo. Desta forma, registram-se os dados em um arquivo específico para análise e estudo posterior. A porcentagem de erro entre o valor medidor no mostrador da máquina e o resultado automatizado não ultrapassa o limite de 8 %.

Comparison of different experimental and analytical measures of the thermal annealing response of neutron-irradiated RPV steels

International Nuclear Information System (INIS)

Iskander, S.K.; Sokolov, M.A.; Nanstad, R.K.

1997-01-01

The thermal annealing response of several materials as indicated by Charpy transition temperature (TT) and upper-shelf energy (USE), crack initiation toughness, K Jc , predictive models, and automated-ball indentation (ABI) testing are compared. The materials investigated are representative reactor pressure vessel (RPV) steels (several welds and a plate) that were irradiated for other tasks of the Heavy-Section Steel Irradiation (HSSI) Program and are relatively well characterized in the unirradiated and irradiated conditions. They have been annealed at two temperatures, 343 and 454 C (650 and 850 F) for varying lengths of time. The correlation of the Charpy response and the fracture toughness, ABI, and the response predicted by the annealing model of Eason et al. for these conditions and materials appears to be reasonable. The USE after annealing at the temperature of 454 C appears to recover at a faster rate than the TT, and even over-recovers (i.e., the recovered USE exceeds that of the unirradiated material)

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.

Tensile and impact testing of an HFBR [High Flux Beam Reactor] control rod follower

International Nuclear Information System (INIS)

Czajkowski, C.J.; Schuster, M.H.; Roberts, T.C.; Milian, L.W.

1989-08-01

The Materials Technology Group of the Department of Nuclear Energy (DNE) at Brookhaven National Laboratory (BNL) undertook a program to machine and test specimens from a control rod follower from the High Flux Beam Reactor (HFBR). Tensile and Charpy impact specimens were machined and tested from non-irradiated aluminum alloys in addition to irradiated 6061-T6 from the HFBR. The tensile test results on irradiated material showed a two-fold increase in tensile strength to a maximum of 100.6 ksi. The impact resistance of the irradiated material showed a six-fold decrease in values (3 in-lb average) compared to similar non-irradiated material. Fracture toughness (K I ) specimens were tested on an unirradiated compositionally and dimensionally similar (to HFBR follower) 6061 T-6 material with K max values of 24.8 ± 1.0 Ksi√in (average) being obtained. The report concludes that the specimens produced during the program yielded reproducible and believable results and that proper quality assurance was provided throughout the program. 9 figs., 6 tabs

Elastic-plastic analysis of the SS-3 tensile specimen

International Nuclear Information System (INIS)

Majumdar, S.

1998-01-01

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

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

International Nuclear Information System (INIS)

Odette, G.R.; Donahue, E.; Lucas, G.E.; Sheckherd, J.W.

1996-01-01

The influence of loading rate and constraint on the effective fracture toughness as a function of temperature [K 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 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 normalclose 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 σ*/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 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 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

Irradiation embrittlement of some 15Kh2MFA pressure vessel steels under varying neutron fluence rates

Energy Technology Data Exchange (ETDEWEB)

Valo, M; Bars, B [Technical Research Centre of Finland, Espoo (Finland); Ahlstrand, A [Imatran Voima Oy (IVO), Helsinki (Finland)

1994-12-31

Irradiation sensitivity of two forging materials was measured with Charpy-V and fracture mechanic tests, and with different fluence, fluence rate and irradiation time values. Irradiation sensitivity of the materials was found to be less or equal to the current Russian standard, and appears to be well described by the fluence parameter only. A slight additional effect on embrittlement from a long term low fluence irradiation is noticed, but it stays within the total scatter band of data. 7 refs., 17 figs., 4 tabs.

Heavy-section steel technology program. Volume 1. Quarterly progress report, January-March 1983. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

Pugh, C.E.

1983-09-01

A thermal-strain modification was made to the deformation-plasticity model in the ADINA-ORVIRT fracture-mechanics analysis system in order to be more applicable to combined pressure and thermal loadings. Subcontractors continued studies on crack arrest, cleavage fracture transition, and environmentally assisted crack growth. Charpy testing of state-of-the-art weld specimens in the Fourth HSST Irradiation Series was performed on unirradiated specimens and on a few irradiated specimens for scoping purposes. Finite-flaw capabilities were incorporated into the OCA-II computer code, and parametric studies were carried out to compare fracture predictions with two-dimensional and specific finite flaws. Preparations continued for thermal-shock experiment TSE-7 to be conducted in May. Preparations for the first pressurized-thermal-shock experiment continued.

Fracture toughness measurements with subsize disk compact specimens

International Nuclear Information System (INIS)

Alexander, D.J.

1994-01-01

Special fixtures and test methods have been developed for testing small disk compact specimens (1.25 mm diam by 4.6 mm thick). Specimens of European type 316L austenitic stainless steel were irradiated to damage levels of about 3 dpa at nominal irradiation temperatures of either 90 or 250 C and tested over a temperature range from 20 to 250 C. Results show that irradiation to this dose level at these temperatures reduces the fracture toughness but the toughness remains quite high. The toughness decreases as the test temperature increases. Irradiation at 250 C is more damaging than at 90 C, causing larger decreases in the fracture toughness. The testing shows that it is possible to generate useful fracture toughness data with a small disk compact specimens

Assessment of plastic flow and fracture properties with small specimens test techniques for IFMIF-designed specimens

International Nuclear Information System (INIS)

Spaetig, P.; Campitelli, E.N.; Bonade, R.; Baluc, N.

2005-01-01

The primary mission of the International Fusion Material Irradiation Facility (IFMIF) is to generate a material database to be used for the design of various components, for the licensing and for the assessment of the safe operation of a demonstration fusion reactor. IFMIF is an accelerator-based high-energy neutron source whose irradiation volume is quite limited (0.5 l for the high fluence volume). This requires the use of small specimens to measure the irradiation-induced changes on the physical and mechanical properties of materials. In this paper, we developed finite element models to better analyze the results obtained with two different small specimen test techniques applied to the tempered martensitic steel F82H-mod. First, one model was used to reconstruct the load-deflection curves of small ball punch tests, which are usually used to extract standard tensile parameters. It was shown that a reasonable assessment of the overall plastic flow can be done with small ball punch tests. Second, we investigated the stress field sensitivity at a crack tip to the constitutive behavior, for a crack modeled in plane strain, small-scale yielding and fracture mode I conditions. Based upon a local criterion for cleavage, that appears to be the basis to account for the size and geometry effects on fracture toughness, we showed that the details of the constitutive properties play a key role in modeling the irradiation-induced fracture toughness changes. Consequently, we suggest that much more attention and efforts have to be paid in investigating the post-yield behavior of the irradiated specimens and, in order to reach this goal, we recommend the use of not only tensile specimens but also that of compression ones in the IFMIF irradiation matrices. (author)

Correlation of irradiation-induced transition temperature increases from Cv and KJc/KIc data

International Nuclear Information System (INIS)

Hiser, A.L.

1990-03-01

Reactor pressure vessel (RPV) surveillance capsules contain Charpy-V (C v ) specimens, but many do not contain fracture toughness specimens; accordingly, the radiation-induced shift (increase) in the brittle-to-ductile transition region (ΔT) is based upon the ΔT determined from notch ductility (C v ) tests. Since the ASME K Ic and K IR reference fracture toughness curves are shifted by the ΔT from C v , assurance that this ΔT does not underestimate ΔT associated with the actual irradiated fracture toughness is required to provide confidence that safety margins do not fall below assumed levels. To assess this behavior, comparisons of ΔT's defined by elastic-plastic fracture toughness and C v tests have been made using data from RPV base and weld metals in which irradiations were made under test reactor conditions. Using ''as-measure'' fracture toughness values (K Jc ), average comparisons between ΔT(C v ) and ΔT(K Jc ) are: (a) All data: ΔT(K Jc at sign 100 MPa√ bar m) = ΔT(C v at sign 41 J) +10 degree C; (b) Plates only: ΔT(K Jc at sign 100 MPa√ bar m) = ΔT(C v at sign 41 J) +15 degree C; and (c) Welds only: ΔT(K Jc at sign 100 MPa√ bar m) = ΔT(C v at sign 41 J) -1 degree C. Fluence rate is found to have no significant effect on the relationship between ΔT(C v ) and ΔT(K Jc ). 12 refs., 12 figs., 5 tabs

Monitoring of irradiation effects on the pressure vessel steels of Calder, Chapelcross and Windscale Advanced Gas Cooled Reactor (WAGR) nuclear reactors

International Nuclear Information System (INIS)

Turner, F.

1980-01-01

Tensile, Charpy and bend specimens of plate, forging and weld metal are exposed in the lower and upper zones of the reactors to neutron fluxes covering the range experienced by the vessels. The test conditions are described, the results presented and discussed. (author)

Surveillance extension experience at WWER-440 type reactors

International Nuclear Information System (INIS)

Gillemot, F.; Uri, G.; Oszwald, F.; Trampus, P.

1993-01-01

In WWER-440 reactors, the surveillance specimens are located in accelerated irradiation positions. After 5 years, all specimens are withdrawn and the operational changes are not monitored. At Paks NPP a new surveillance program extension has been settled in order to avoid these original program disadvantages and generate further data for plant lifetime management. This paper includes: research performed to prepare the surveillance extension programme, the evaluation method for the surveillance extension, and first results (Charpy and tensile tests). (authors). 6 refs., 12 figs., 3 tabs

Surveillance extension experience at WWER-440 type reactors

Energy Technology Data Exchange (ETDEWEB)

Gillemot, F; Uri, G [Budapesti Mueszaki Egyetem, Budapest (Hungary); Oszwald, F; Trampus, P

1994-12-31

In WWER-440 reactors, the surveillance specimens are located in accelerated irradiation positions. After 5 years, all specimens are withdrawn and the operational changes are not monitored. At Paks NPP a new surveillance program extension has been settled in order to avoid these original program disadvantages and generate further data for plant lifetime management. This paper includes: research performed to prepare the surveillance extension programme, the evaluation method for the surveillance extension, and first results (Charpy and tensile tests). (authors). 6 refs., 12 figs., 3 tabs.

Irradiation, annealing, and reirradiation research in the ORNL heavy-section steel irradiation program

International Nuclear Information System (INIS)

Nanstad, R.K.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.

1997-01-01

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes experimental results from work performed as part of the Heavy-Section Steel Irradiation (HSSI) Program managed by Oak Ridge National Laboratory (ORNL) for the U.S. Nuclear Regulatory Commission. The HSSI Program focuses on annealing and re-embrittlement response of materials which are representative of those in commercial RPVs and which are considered to be radiation-sensitive. Experimental studies include (1) the annealing of materials in the existing inventory of previously irradiated materials, (2) reirradiation of previously irradiated/annealed materials in a collaborative program with the University of California, Santa Barbara (UCSB), (3) irradiation/annealing/reirradiation of U.S. and Russian materials in a cooperative program with the Russian Research Center-Kurchatov Institute (RRC-KI), (4) the design and fabrication of an irradiation/anneal/reirradiation capsule and facility for operation at the University of Michigan Ford Reactor, (5) the investigation of potential for irradiation-and/or thermal-induced temper embrittlement in heat-affected zones (HAZs) of RPV steels due to phosphorous segregation at grain boundaries, and (6) investigation of the relationship between Charpy impact toughness and fracture toughness under all conditions of irradiation, annealing, and reirradiation

Fundamental study of crack initiation and propagation

International Nuclear Information System (INIS)

Norris, D.M. Jr.; Reaugh, J.E.; Moran, B.; Quinones, D.F.; Wilkins, M.L.

1977-01-01

Objective is to determine the fracture toughness of A533B-1 steel by computer modeling Charpy V-notch tests. A computer model of ductile fracture was developed that predicts fracture initiation. The model contains a set of material-dependent parameters obtained by computer simulations of small specimen tests. The computer calculations give detailed stress and strain histories up to the time of fracture, which are used to determine the model parameter values. The calibrated fracture model, that correctly predicts fracture initiation (and initiation energy) in the Charpy specimen, may then be used to simulate tests of accepted fracture-toughness specimens and hence obtain fracture toughness. The model parameters were calibrated to predict fracture in four different test specimens: two different notched-tension specimens, a simple tension specimen, and a precracked compact-tension specimen. The model was then used in a computer simulation of the Charpy V-notch specimen to initiate and advance a flat fracture. Results were compared with interrupted Charpy tests. Calibration of the model for two additional heat treatments of A533B-1 steel is in progress

Charpy impact test pada kampas rem hybrid komposit phenolic resin matrik dengan penguat serbuk basalt-Alumina-kulit kerang

Directory of Open Access Journals (Sweden)

I N. G. Suma Wijaya

2017-03-01

Full Text Available Abstrak: Penelitian ini adalah difokuskan untuk mengamati ketahanan impact dari material kampas rem kendaraan bermotor. Kampasrem yang diujikan adalah kampas yang terbuat dari material hybrid komposit dengan penguat serbuk basalt – serbuk kulitkerang dan alumina dan pengikat phenolic resin. Material kampas rem hibrid komposit diproses melalui proses sinteringdengan penekanan 2 ton, temperatur 150ºC selama 30 menit. Tujuan penelitian adalah menginvestigasi kekuatan impact daripada bahan hybrid komposit untuk masing – masing variasi terhadap perlakuan impact charpy yang didasarkan pada standarASTM D6110–04. Pengujian dilakukan dengan menganalisa nilai energy yang mampu diserap oleh bahan akibat bebanimpact, selanjutnya patahan impact charpy dianalisa dengan SEM. Diperoleh hasil pengujian charpy impact untuk masing –masing variasi hybrid komposit adalah nilai kekuatan yang tinggi terjadi pada hibrid komposit variasi 2 (HK2 dengan nilai0,000339547 J/mm2, ini disebabkan karena mempunyai ikatan antara metrik dan basalt yang lebih kuat dan sempurnadibandingkan dengan hibrid komposit lainnya. Untuk nilai hibrid komposit variasi 1 (HK1 adalah 0,000304851 J/mm², hibridkomposit variasi 3 (HK3 adalah 0,000334516 J/mm², hibrid komposit variasi 4 (HK4 adalah 0,000325059 J/mm², hibridkomposit variasi 5 (HK5 adalah 0,0003327 J/mm². (2 Dari perbandingan antara kampas rem dipasaran dengan kampas remhibrid komposit maka didapat nilai kekuatannya berbeda pada hibrid komposit variasi 2 (HK2 dengan kampas pembanding(KP yang memiliki nilai kekuatannya lebih besar yaitu, 0,000374867 J/mm².Kata kunci : Kampas rem, Impact charpy, Hibrid komposit, Basalt, Aluminium, Kulit Kerang Abstract: This research is focused to observe the impact resistance of motor vehicles brake lining material. Brake tested are canvasmade of hybrid composite materials with basalt powder reinforced – seashells, alumina powder, and a phenolic resin matrix.Hybrid composite brake

Bright field electron microscopy of biological specimens

International Nuclear Information System (INIS)

Johansen, B.V.

1976-01-01

A preirradiation procedure is described which preserves negatively stained morphological features in bright field electron micrographs to a resolution of about 1.2 nm. Prior to microscopy the pre-irradiation dose (1.6 x 10 -3 C cm -2 ) is given at low electron optical magnification at five different areas on the grid (the centre plus four 'corners'). This pre-irradiation can be measured either with a Faraday cage or through controlled exposure-developing conditions. Uranyl formate stained T2 bacteriophages and stacked disk aggregates of Tobacco Mosaic Virus (TMV) protein served as test objects. A comparative study was performed on specimens using either the pre-irradiation procedure or direct irradiation by the 'minimum beam exposure' technique. Changes in the electron diffraction pattern of the stain-protein complex and the disappearance of certain morphological features in the specimens were both used in order to compare the pre-irradiation method with the direct exposure technique. After identical electron exposures the pre-irradiation approach gave a far better preservation of specimen morphology. Consequently this procedure gives the microscopist more time to select and focus appropriate areas for imaging before deteriorations take place. The investigation also suggested that microscopy should be carried out between 60,000 and 100,000 times magnification. Within this magnification range, it is possible to take advantage of the phase contrast transfer characteristics of the objective lens while the electron load on the object is kept at a moderate level. Using the pre-irradiation procedure special features of the T2 bacteriophage morphology could be established. (author)

Fracture toughness measurements with subsize disk compact specimens

International Nuclear Information System (INIS)

Alexander, D.J.

1992-01-01

Special fixtures and test methods are necessary to facilitate the fracture toughness testing of small disk compact specimens of irradiated candidate materials for first-wall fusion applications. New methods have been developed for both the unloading compliance and potential drop techniques of monitoring crack growth. Provisions have been made to allow the necessary probes and instrumentation to be installed remotely using manipulators for testing of irradiated specimens in a hot cell. Laboratory trials showed that both unloading compliance and potential drop gave useful results. Both techniques gave similar data, and predicted the final crack extension within allowable limits. The results from the small disk compact specimens were similar to results from conventional compact specimen 12.7 mm thick. However, the slopes of the J-R curves from the larger specimens were lower, suggesting that the smaller disk compact specimens may have lost some constraint due to their size. The testing shows that it should be possible to generate useful J-R curve fracture toughness data from the small disk compact specimens

Demonstration of Laser Cutting System for Tube Specimen

Energy Technology Data Exchange (ETDEWEB)

Jin, Y. G.; Kim, G. S.; Heo, G. S.; Baik, S. J.; Kim, H. M.; Ahn, S. B. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The oxide layer removal system was also developed because the oxide layer on the surface of the irradiated fuel cladding and components interrupted the applying the electric current during the processing. However, it was found that the mechanical testing data of the irradiated specimens with removal of oxide layer was less reliable than the specimens with oxide layer . The laser cutting system using Nd:YAG with fiber optic beam delivery has great potential in material processing applications of the irradiated fuel cladding and components due to non-contact process. Thus, the oxide layer doesn't interrupt the fabrication process during the laser cutting system. In the present study, the laser cutting system was designed to fabricate the mechanical testing specimens from the unirradiated fuel cladding with and without oxide. The feasibility of the laser cutting system was demonstrated for the fabrication of various types of unirradiated specimens. The effect of surface oxide layer was also investigated for machining process of the zirlo fuel cladding and it was found that laser beam machining could be a useful tool to fabricate the specimens with surface oxide layer. Based on the feasibility studies and demonstration, the design of the laser cutting machine for fully or partially automatic and remotely operable system will be proposed and made.

Numerical simulation of a Charpy test and correlation of fracture toughness with fracture energy. Vessel steel and duplex stainless steel of the primary loop

International Nuclear Information System (INIS)

Breban, P; Eripret, C.

1995-01-01

The analysis methods used to evaluate the harmlessness of defects in the components of the primary coolant circuit of pressurized water reactor are based on the knowledge of the failure properties of concerned materials. The toughness is used to be measured through tests performed on normalized samples. But in some cases, especially for the vessel steel submitted to irradiation effects or for cast components in duplex stainless steel sensitive to thermal ageing, these measurements are not available on the material aged in operation. Therefore, fracture resistance has been evaluated through Charpy tests. Toughness is thus obtained on the basis of an empirical correlation. To improve these predictions, a modeling of the Charpy test in the framework of the local approach to fracture has been performed, for both materials. For the vessel steel, a complete evaluation of toughness has been achieved on the basis of a bidimensional viscoplastic modeling under large strain assumptions and a post-treatment with a Weibull model (cleavage fracture). The main hypothesis (partition between plain stress and plain strain areas in the bidimensional modeling) was corrected after a three dimensional calculations with the finite element program Code-Aster. The fracture analysis put into evidence that damage considerations like cavity nucleation and growth have to be introduced in the model in order to improve the description of physical phenomena. Two ways of progress have been suggested and are in course of being investigated, one in the framework of local approach to failure, the other with the help of micro-macro relationship. With regard to the duplex steel, the description of a Charpy (U) test allowed to clearly discriminate between crack initiation and propagation phases. A modeling through an equivalent homogenous material with a damage law based on a modified Gurson potential enables to describe quantitatively both phases of fracture. It clearly appears that a reliable

Evaluation of irradiation hardening of proton irradiated stainless steels by nanoindentation

International Nuclear Information System (INIS)

Yabuuchi, Kiyohiro; Kuribayashi, Yutaka; Nogami, Shuhei; Kasada, Ryuta; Hasegawa, Akira

2014-01-01

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 + 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 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 0 . Thus, the nanoindentation technique demonstrated in this study is valuable for measuring irradiation hardening in ion-irradiated materials

Development of a miniaturized bulge test (small punch test) for post-irradiation mechanical property evaluation

International Nuclear Information System (INIS)

Eto, Motokuni; Suzuki, Masahide; Nishiyama, Yutaka; Fukaya, Kiyoshi; Jitsukawa, Shiro; Misawa, Toshihei

1993-01-01

To examine the effectiveness of the small punch test for evaluating strength and toughness of irradiated ferritic steels, detailed procedures are described aiming at standardization of the test. The statistical approach to analysis of the SP energy as a function of temperature for evaluation of DBTT was also reviewed. The method was then applied to neutron-irradiated ferritic steels, which included F-82, F-82H, HT-9, and 2 1/4 Cr-1Mo steel. Fluence and irradiation temperatures ranged from 2 to 12 x 10 23 n/m 2 (E ≥ 1 MeV) and from 573 to 673 K, respectively. Comparison of parameters obtained from the small punch test with the properties measured by the conventional method indicated that: (a) the 0.2% offset stress and the ultimate tensile strength at room temperature can be correlated well with the parameters, P y /(t 0 ) 2 and P max /(t 0 ) 2 , respectively. Here, P y and P max are the loads corresponding to the yield and the maximum, and t 0 is the initial thickness of a specimen; (b) fracture toughness, J IC , can be evaluated using equivalent fracture strain, anti ε qf , and the previously established relationship between these values; and (c) DBTT measured by a Charpy test can be predicted from the results of temperature dependence of SP energy determined from the area under the load-deflection curve using a statistical analysis based on a Weibull distribution

Fracture toughness of irradiated Zr-2.5Nb pressure tube from KAPS-2 evaluated using disk compact tension specimens

International Nuclear Information System (INIS)

Shah, Priti Kotak; Dubey, J.S.; Balakrishnan, K.S.; Shriwastaw, R.S.; Dhotre, M.P.; Bhandekar, A.; Pandit, K.M.; Anantharaman, S.

2013-12-01

The report gives the results of the fracture toughness tests carried out over the range of temperatures on specimens prepared from the irradiated S-07Zr-2.5Nb pressure tube removed from Kakrapar Atomic Power Station-2 (KAPS-2) as a part of materials surveillance programme. The pressure tube had experienced ∼ 8 effective full power years (EFPY) of reactor operation and had hydrogen equivalent (H eq ) content less than 20 ppm along the tube length. The fracture toughness tests have been carried out using 30 mm Disk Compact Tension (DCT) specimens, that were punched out of the irradiated pressure tube. The disk punching was carried out using specially made shielded enclosure and hydraulic press. Fatigue pre-cracking and fracture toughness tests were performed using servo-hydraulic universal testing machine with Direct Current Potential Drop (DCPD) equipment to monitor the crack length. The tests were carried out at different test temperature from ambient to 300℃. The fracture toughness values have been used to estimate the critical pressure for the tube. The fracture properties indicate that such tubes have sufficient toughness to satisfy the Leak-Before-Break (LBB) criterion for in-reactor operation. (author)

Tension test system for irradiated small specimens operated by remote control

International Nuclear Information System (INIS)

Okada, Akira

1993-01-01

A robot-based tension test system has been developed to aid in the mechanical testing of highly radioactive specimens. This system reduces radiation hazards from specimens and allows for the uniform precision of testing results independent of experimenters' skills. The robot system is designed to accommodate a miniaturized tension specimen with a gage section 5.5 by 1.2 mm, with a total length and width of 12.5 and 2.3 mm, respectively, and thickness of about 0.2 mm. The system is composed of a manipulating robot, a vibrational-type specimen feeder, a rotating-type specimen tray, a specimen observation system, a simulated tension text fixture, and a microcomputer for controlling the system. This system accomplishes specimen arrangement in the specimen tray, specimen transportation and loading to the test fixture and testing, and removal of the broken specimen from the fixture. These procedures are performed quickly, safely, and with uniform testing precision by computer control remotely by an unskilled experimenter

Final report on graphite irradiation test OG-2

International Nuclear Information System (INIS)

Price, R.J.; Beavan, L.A.

1975-01-01

Results are presented of dimensional, thermal expansivity, thermal conductivity, Young's modulus, and tensile strength measurements on specimens of nuclear graphites irradiated in capsule OG-2. About half the irradiation space was allocated to H-451 near-isotropic petroleum-coke-based graphite or its subsized prototype grade H-429. Most of these specimens had been previously irradiated. Virgin specimens of another near-isotropic graphite, grade TS-1240, were irradiated. Some previously irradiated specimens of needle-coke-based H-327 graphite and pitch-coke-based P 3 JHAN were also included

Heavy-Section Steel Irradiation Program. Volume 2, No. 1: Semiannual progress report, October 1990--March 1991

International Nuclear Information System (INIS)

Corwin, W.R.

1994-07-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the capabilities and limitations of the integrity inherent in the RPV. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established with its primary goal to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure-vessel steels as they relate to light-water reactor pressure-vessel integrity. The HSSI Program is arranged into nine tasks: (1) program management, (2) K ic curve shift in high-copper welds, (3) K ia curve shift in high-copper welds, (4) irradiation effects on cladding, (5) K ic and K ia curve shifts in low upper-shelf (LUS) weld, (6) irradiation effects in a commercial LUS weld, (7) microstructural analysis of irradiation, (8) in-service aged material evaluations, and (9) correlation monitor materials. During this period, additional analyses on the effects of precleavage stable ductile tearing on the toughness of high-copper welds 72W and 73W demonstrated that the size effects observed in the transition region are not due to substantial differences in ductile tearing behavior. Possible modifications to irradiated duplex crack-arrest specimens were examined to increase the likelihood of their successful testing. Characterization of a second batch of 72W and 73W welds was begun and results of the Charpy V-notch testing is provided. A review of literature on the annealing response of reactor pressure vessel steels was initiated

Predicting crack instability behavior of burst tests from small specimens for irradiated Zr-2.5Nb pressure tubes

International Nuclear Information System (INIS)

Davies, P.H.

1997-01-01

A scaling approach, based on the deformation J-integral at maximum load obtained from small specimens, is proposed for predicting the crack instability behavior of burst tests on irradiated Zr-2.5Nb pressure tubes. An assessment of this approach is carried out by comparison with other toughness criteria such as the modified J-integral and the plastic work dissipation rate approach. The largest discrepancy between the different parameters occurs for materials of intermediate toughness which exhibit the most stable crack growth and tunnelling up to maximum load. A study of one material of intermediate toughness suggests crack-front tunnelling has a significant influence on the results obtained from the 17-mm-wide specimens. It is shown that for a tube of intermediate toughness the different approaches can significantly underpredict the extent of stable crack growth before instability in a burst test even after correcting for tunnelling. The usefulness of a scaling approach in reducing the discrepancy between the small- and large-scale specimen results for this material is demonstrated

Evaluation of Specimen Geometric Effect for Laser Flash Thermal Diffusivity Test

International Nuclear Information System (INIS)

Park, Dae Gyu; Kim, Hee Moon; Song, Woong Sub; Baik, Seung Je; Ryu, Woo Seok; Ahn, Sang Bok; Joo, Young Sun

2012-01-01

KAERI(Korea Atomic Energy Research Institute) is developing a new type of nuclear reactor, the so called 'SMART' (System Integrated Modular Advanced Reactor) reactor. Alloy 690 was selected as the candidate material for the heat exchanger tube of of SMART's steam generator. The SMART R and D is now facing the stage of engineering verification and standard design approval for application of DEMO reactors. Therefore, the material performance under the relevant environment needs to be evaluated. The one of the important material performance issues is thermal conductivity, which the engineering database is necessary for the steam generator design. However, the neutron post irradiation characteristics of alloy 690 are little known. As a result, a PIE (Post Irradiation Examination) of the thermal properties have been plan for a 4 times, so called base line test, 1 st irradiation test, 2 nd and 3 rd irradiation test. But there is some constraint to perform thermal diffusivity test owing to test specimen. Originally thermal diffusivity test are planed using disk shape with 9 mm diameter and 1 mm thick specimen. Due to mismatch of neutron irradiation schedule, thermal diffusivity will be tested by different shape and size specimens at 1 st irradiation test. Therefore, verification of geometric and size effect are necessary for test specimen in order to achieve accurate test results

Recent improvements in size effects correlations for DBTT and upper shelf energy of ferritic steels

International Nuclear Information System (INIS)

Kumar, A.S.; Louden, B.S.; Garner, F.A.; Hamilton, M.L.

1992-01-01

Currently available correlations for the effects of specimen size on the USE were developed for relatively ductile steels and will not serve as well when the steels become embrittled. Size effects correlations were developed recently for the impact properties of less ductile HT9 to be applied to other initially more ductile steels as they lose their ductility during irradiation. These new correlations successfully predict the ductile brittle transition temperature (DBTT) and the upper shelf energy (USE) of full size Charpy specimens based on subsize specimen data. The new DBTT and the USE correlations were tested against published experimental data on other ferritic steels and shown to perform successfully at lower USE particularly when both precracked and notched only specimens were employed

Design and use of nonstandard tensile specimens for irradiated materials testing

International Nuclear Information System (INIS)

Panayotou, N.F.

1984-10-01

Miniature, nonstandard, tensile-type specimens have been developed for use in radiation effects experiments at high energy neutron sources where the useful radiation volume is as small as a few cubic centimeters. The end result of our development is a sheet-type specimen, 12.7 mm long with a 5.1 mm long, 1.0 mm wide gage section, which is typically fabricated from 0.25 mm thick sheet stock by a punching technique. Despite this miniature geometry, it has been determined that the data obtained using these miniature specimens are in good agreement with data obtained using much larger specimens. This finding indicates that miniature tensile specimen data may by used for engineering design purposes. Furthermore, it is clear that miniature tensile specimen technology is applicable to fields other than the study of radiation effects. This paper describes the miniature specimen technology which was developed and compares the data obtained from these miniature specimens to data obtained from much larger specimens. 9 figures

Different approaches to estimation of reactor pressure vessel material embrittlement

Directory of Open Access Journals (Sweden)

V. M. Revka

2013-03-01

Full Text Available The surveillance test data for the nuclear power plant which is under operation in Ukraine have been used to estimate WWER-1000 reactor pressure vessel (RPV material embrittlement. The beltline materials (base and weld metal were characterized using Charpy impact and fracture toughness test methods. The fracture toughness test data were analyzed according to the standard ASTM 1921-05. The pre-cracked Charpy specimens were tested to estimate a shift of reference temperature T0 due to neutron irradiation. The maximum shift of reference temperature T0 is 84 °C. A radiation embrittlement rate AF for the RPV material was estimated using fracture toughness test data. In addition the AF factor based on the Charpy curve shift (ΔTF has been evaluated. A comparison of the AF values estimated according to different approaches has shown there is a good agreement between the radiation shift of Charpy impact and fracture toughness curves for weld metal with high nickel content (1,88 % wt. Therefore Charpy impact test data can be successfully applied to estimate the fracture toughness curve shift and therefore embrittlement rate. Furthermore it was revealed that radiation embrittlement rate for weld metal is higher than predicted by a design relationship. The enhanced embrittlement is most probably related to simultaneously high nickel and high manganese content in weld metal.

Development of PIE techniques for irradiated LWR pressure vessel steels

International Nuclear Information System (INIS)

Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide

1999-01-01

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)

Vanderbilt University Gamma Irradiation of Nano-modified Concrete (2017 Milestone Report)

Energy Technology Data Exchange (ETDEWEB)

Deichert, Geoffrey G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Selby, Aaron P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Reches, Yonathan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

This document outlines the irradiation of concrete specimens in the Gamma Irradiation Facility in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Two gamma irradiation runs were performed in July of 2017 on 18 reference mortar bar specimens, 26 reference cement paste bar specimens, and 28 reference cement paste tab specimens to determine the dose and temperature response of the specimens in the gamma irradiation environment. Specimens from the first two gamma irradiations were surveyed and released to Vanderbilt University. The temperature and dose information obtained informs the test parameters of the final two gamma irradiations of nano-modified concrete planned for FY 2018.

On the Thermal Conductivity Change of Matrix Graphite Materials after Neutron Irradiation

Energy Technology Data Exchange (ETDEWEB)

Lee, Young-Woo; Yeo, Seunghwan; Kim, Eung-Seon; Sah, Injin; Park, Daegyu; Kim, Youngjun; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

In this work, the variations of the thermal conductivity of the A3-3 matrix graphite after neutron irradiation is discussed as well as of the IG-110 graphite for comparison. Neutron irradiation of the graphite specimens was carried out as a part of the first irradiation test of KAERI's coated particle fuel specimens by use of Hanaro research reactor. This work can be summarized as follows: 1) In the evaluation of the specific heat of the graphite materials, various literature data were used and the variations of the specific heat data of all the graphite specimens are observed well agreed, irrespectively of the difference in specimens (graphite and matrix graphite and irradiated and un-irradiated). 2) This implies that it should be reasonable that for both structural graphite and fuel matrix graphite, and even for the neuron-irradiated graphite, any of these specific heat data set be used in the calculation of the thermal conductivity. 3) For the irradiated A3-3 matrix graphite specimens, the thermal conductivity decreased on both directions. On the radial direction, the tendency of variation upon temperature is similar to that of unirradiated specimen, i.e., decreasing as the temperature increases. 4) In the German irradiation experiments with A3-27 matrix graphite specimens, the thermal conductivity of the un-irradiated specimen shows a decrease and that of irradiated specimen is nearly constant as the temperature increases. 5) The thermal conductivity of the irradiated IG-110 was considerably decreased compared with that of un-irradiated specimens The difference of the thermal conductivity of un-irradiated and irradiated IG-110 graphite specimens is much larger than that of un-irradiated and irradiated A3-3 matrix graphite specimens.

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.

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.

Technology development on production of test specimens from irradiated capsule outer-tube and mechanical evaluation test of stainless steel with high dose carried out by the technology

International Nuclear Information System (INIS)

Hayashi, Koji; Shibata, Akira; Iwamatsu, Shigemi; Sozawa, Shizuo; Takada, Fumiki; Ohmi, Masao; Nakagawa, Tetsuya

2008-03-01

The irradiation capsule 74M-52J was irradiated during total 136 cycles at reactor core of JMTR and the maximum neutron dose reached on 3.9x10 26 n/m 2 at the capsule outer-tube made of a type 304 stainless steel. In order to produce mechanical test specimens from the outer-tube, a punching technique was developed as a simple remote-handling method in a hot-cell. From comparison between the punching and the mechanical cutting methods, it was clarified that the punching technique was applicable to practical use. Moreover, an evaluation test of mechanical properties using specimens sampled from the 74M-52 was performed with in-water high temperature condition, less than 288degC. The result shows that the residual elongation is 18% at 150degC and 13% at 288degC. It was confirmed that the type 304 stainless steel irradiated up to such high dose shows enough ductility. (author)

Present status of mechanical testing technology at the Research Hot Laboratory

International Nuclear Information System (INIS)

Kizaki, M.; Tobita, T.; Koya, T.; Kikuchi, T.

1993-01-01

Mechanical tests of irradiated metallic materials at the Research Hot Laboratory(RHL) have been carried out for 30 years to support material research in JAERI and to evaluate the irradiation integrity of pressure vessel steel in commercial power plant. Two tensile testing machines and one Charpy impact testing machine are available for the examinations. One of the tensile testing machines has 1000 kgf load capacity under the vacuum of ∼ 10 -7 torr at the temperature of 1300degC max.. The other one has 10 tonf load capacity, and is utilized for the multi-purpose tests such as tensile and compressive tests in air atmosphere at the temperature between -160 and 900degC. Examinations cover tensile test, bending test, J ic fracture toughness test, low cycle fatigue test and so on. Charpy impact testing machine with notched-bar specimen is instrumented with 30 kgf-m capacity in the temperature range of -140 - 240 degC. To support these mechanical tests in RHL, special jigs, devices and instruments have been developed. (author)

Effect of irradiation spectrum on the microstructure of ion-irradiated Al2O3

International Nuclear Information System (INIS)

Zinkle, S.J.

1994-01-01

Polycrystalline samples of alpha-alumina have been irradiated with various ions ranging from 3.6 MeV Fe + to 1 MeV H + ions at 650 C. Cross-section transmission electron microscopy was used to investigate the depth-dependent microstructure of the irradiated specimens. The microstructure following irradiation was observed to be dependent on the irradiation spectrum. In particular, defect cluster nucleation was effectively suppressed in specimens irradiated with light ions such as 1 MeV H + ions. On the other hand, light ion irradiation tended to accelerate the growth rate of dislocation loops. The microstructural observations are discussed in terms of ionization enhanced diffusion processes

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

International Nuclear Information System (INIS)

Odette, G.R.; Lucas, G.E.; Wirth, B.; Liu, C.L.

1997-01-01

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

Size effect in radiation damage

International Nuclear Information System (INIS)

Brumovsky, M.

1979-01-01

Radiation embrittlement of nuclear reactor pressure vessel steels is mostly measured using small standard specimens in dynamic bend tests. Their dimensions are much smaller than those of the reactor. The increase in the critical temperature (transition temperature from the brittle-to-ductile fracture) is normally measured using standard Charpy-V type specimens or small CT-type specimens. This increase is then used as the main parameter for the pressure vessel safety evaluation. The philosophy of experiments is discussed used for the nonirradiated and irradiated pressure vessel steels. A comparison of the increase in the transition temperature measured in different types of specimens using various testing methods (static and dynamic bend tests with notch or crack) is also made. The results of this comparison and another study showed a relatively good agreement. (author)

Development of piezoelectric ceramics driven fatigue testing machine for small specimens

International Nuclear Information System (INIS)

Saito, S.; Kikuchi, K.; Onishi, Y.; Nishino, T.

2002-01-01

A new fatigue testing machine with piezoelectric ceramics actuators was developed and a prototype was manufactured for high-cycle fatigue tests with small specimens. The machine has a simple mechanism and is compact. These features make it easy to set up and to maintain the machine in a hot cell. The excitation of the actuator can be transmitted to the specimen using a lever-type testing jig. More than 100 μm of displacement could be prescribed precisely to the specimen at a frequency of 50 Hz. This was sufficient performance for high-cycle bend fatigue tests on specimens irradiated at the SINQ target in Paul Scherrer Institute. The relationship of a displacement applied to the specimen and the strain of the necking part were obtained by experimental methods and by finite element method (FEM) calculations. Both results showed good agreement. This fact makes it possible to evaluate the strain of irradiated specimens by FEM simulations

Effect of neutron irradiation and post-irradiation annealing on microstructure and mechanical properties of OFHC-copper

International Nuclear Information System (INIS)

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

2001-01-01

Specimens of oxygen-free high conductivity (OFHC) copper were irradiated in the DR-3 reactor at Risoe at 100 deg. C to doses in the range 0.01-0.3 dpa (NRT). Some of the specimens were tensile tested in the as-irradiated condition at 100 deg. C whereas others were given a post-irradiation annealing treatment at 300 deg. C for 50 h and subsequently tested at 100 deg. C. The microstructure of specimens was characterized in the as-irradiated as well as irradiated and annealed conditions both before and after tensile deformation. While the interstitial loop microstructure coarsens with irradiation dose, no significant changes were observed in the population of stacking fault tetrahedra (SFT). The post-irradiation annealing leads to only a partial recovery and the level of recovery depends on the irradiation dose level. However, the post-irradiation annealing eliminates the yield drop and reinstates enough uniform elongation to render the material useful again. These results are discussed in terms of the cascade-induced source hardening (CISH) model

Effect of neutron irradiation and post-irradiation annealing on microstructure and mechanical properties of OFHC-copper

Energy Technology Data Exchange (ETDEWEB)

Singh, B.N. E-mail: bachu.singh@risoe.dk; Edwards, D.J.; Toft, P

2001-12-01

Specimens of oxygen-free high conductivity (OFHC) copper were irradiated in the DR-3 reactor at Risoe at 100 deg. C to doses in the range 0.01-0.3 dpa (NRT). Some of the specimens were tensile tested in the as-irradiated condition at 100 deg. C whereas others were given a post-irradiation annealing treatment at 300 deg. C for 50 h and subsequently tested at 100 deg. C. The microstructure of specimens was characterized in the as-irradiated as well as irradiated and annealed conditions both before and after tensile deformation. While the interstitial loop microstructure coarsens with irradiation dose, no significant changes were observed in the population of stacking fault tetrahedra (SFT). The post-irradiation annealing leads to only a partial recovery and the level of recovery depends on the irradiation dose level. However, the post-irradiation annealing eliminates the yield drop and reinstates enough uniform elongation to render the material useful again. These results are discussed in terms of the cascade-induced source hardening (CISH) model.

Effect of neutron irradiation and post-irradiation annealing on microstructure and mechanical properties of OFHC-copper

Science.gov (United States)

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

2001-12-01

Specimens of oxygen-free high conductivity (OFHC) copper were irradiated in the DR-3 reactor at Risø at 100 °C to doses in the range 0.01-0.3 dpa (NRT). Some of the specimens were tensile tested in the as-irradiated condition at 100 °C whereas others were given a post-irradiation annealing treatment at 300 °C for 50 h and subsequently tested at 100 °C. The microstructure of specimens was characterized in the as-irradiated as well as irradiated and annealed conditions both before and after tensile deformation. While the interstitial loop microstructure coarsens with irradiation dose, no significant changes were observed in the population of stacking fault tetrahedra (SFT). The post-irradiation annealing leads to only a partial recovery and the level of recovery depends on the irradiation dose level. However, the post-irradiation annealing eliminates the yield drop and reinstates enough uniform elongation to render the material useful again. These results are discussed in terms of the cascade-induced source hardening (CISH) model.

Correlation of irradiation-induced transition temperature increases from C sub v and K sub Jc /K sub Ic data

Energy Technology Data Exchange (ETDEWEB)

Hiser, A.L. (Materials Engineering Associates, Inc., Lanham, MD (USA))

1990-03-01

Reactor pressure vessel (RPV) surveillance capsules contain Charpy-V (C{sub v}) specimens, but many do not contain fracture toughness specimens; accordingly, the radiation-induced shift (increase) in the brittle-to-ductile transition region ({Delta}T) is based upon the {Delta}T determined from notch ductility (C{sub v}) tests. Since the ASME K{sub Ic} and K{sub IR} reference fracture toughness curves are shifted by the {Delta}T from C{sub v}, assurance that this {Delta}T does not underestimate {Delta}T associated with the actual irradiated fracture toughness is required to provide confidence that safety margins do not fall below assumed levels. To assess this behavior, comparisons of {Delta}T's defined by elastic-plastic fracture toughness and C{sub v} tests have been made using data from RPV base and weld metals in which irradiations were made under test reactor conditions. Using as-measure'' fracture toughness values (K{sub Jc}), average comparisons between {Delta}T(C{sub v}) and {Delta}T(K{sub Jc}) are: (a) All data: {Delta}T(K{sub Jc} 100 MPa{radical}{bar m}) = {Delta}T(C{sub v} 41 J) +10{degree}C; (b) Plates only: {Delta}T(K{sub Jc} 100 MPa{radical}{bar m}) = {Delta}T(C{sub v} 41 J) +15{degree}C; and (c) Welds only: {Delta}T(K{sub Jc} 100 MPa{radical}{bar m}) = {Delta}T(C{sub v} 41 J) {minus}1{degree}C. Fluence rate is found to have no significant effect on the relationship between {Delta}T(C{sub v}) and {Delta}T(K{sub Jc}). 12 refs., 12 figs., 5 tabs.

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.

Evaluation of Ion Irradiation Behavior of ODS Alloys

International Nuclear Information System (INIS)

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

2006-08-01

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

Study on creep-fatigue life of irradiated austenitic stainless steel

International Nuclear Information System (INIS)

Ioka, Ikuo; Miwa, Yukio; Tsuji, Hirokazu; Yonekawa, Minoru; Takada, Fumiki; Hoshiya, Taiji

2001-01-01

The low cycle creep-fatigue test with tensile strain hold of the austenitic stainless steel irradiated to 2 dpa was carried out at 823K in vacuum. The applicability of creep-fatigue life prediction methods to the irradiated specimen was examined. The fatigue life on the irradiated specimen without tensile strain hold time was reduced by a factor of 2-5 in comparison with the unirradiated specimen. The decline in fatigue life of the irradiated specimen with tensile strain hold was almost equal to that of the unirradiated specimen. The creep damage of both unirradiated and irradiated specimens was underestimated by the time fraction rule or the ductility exhaustion rule. The creep damage calculated by the time fraction rule or the ductility exhaustion rule increased by the irradiation. The predictions derived from the linear damage rule are unsafe as compared with the experimental fatigue lives. (author)

Irradiation effects on organic insulators

International Nuclear Information System (INIS)

Kasen, M.B.

1986-01-01

The overall objective of this work is to contribute to development of organic insulators having the cryogenic neutron irradiation resistance required for MFE systems utilizing superconducting magnet confinement. The system for producing standard 3.2-mm (0.125-in) diameter rod specimens discussed in previous reports has been further refined to permit the fabrication of both fiber-reinforced and heat-resin specimens from hot-melt resin systems. The method has been successfully used to produce very high quality specimens duplicating the G-11CR system and specimens from a variant of that system eliminating a boron-containing additive. We have also produced specimens from an epoxy system suitable for impregnation or potting operations and from a bismaleimide polyimide system. These materials will be used in the first irradiation program in the National Low Temperature Neutron Irradiation Facility (NLTNIF) reactor at Oak Ridge. We have refined the 4-K torsional shear test method for evaluating radiation degradation of the fiber-matrix interface and have developed a method of quantitatively measuring changes in fracture energy as a function of radiation dose. Cooperative work with laboratories in Japan and England in this area is continuing and plans are being formulated for joint production, irradiation, and testing of specimens

SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material Irradiation, Annealing and Re-Embrittlement

International Nuclear Information System (INIS)

Van Walle, E.; Chaouadi, R.; Puzzolante, J.L.; Fabry, A.; Van de Velde, J.

1998-01-01

The contribution of the Belgian Nuclear Research Centre SCK-CEN to the IAEA Round Robin Exercise on WWER-440 RPV weld material is reported. The objective of this contribution is twofold: (1) to gain experience in the field of the testing of WWER-440 steels; (2) to analyse the round-robin data according to in-house developed on used models in order to check their validity and applicability. Results from testing on unirradiated material are reported including data obtained from chemical analysis, Charpy-V impact testing, tensile testing and fracture toughness determination. Finally, irradiation strategies that can be used in the program to obtain irradiated, irradiated-annealed and irradiated-annealed-reirradiated conditions are outlined

Design of a radiation facility for very small specimens used in radiobiology studies

Science.gov (United States)

Rodriguez, Manuel; Jeraj, Robert

2008-06-01

A design of a radiation facility for very small specimens used in radiobiology is presented. This micro-irradiator has been primarily designed to irradiate partial bodies in zebrafish embryos 3-4 mm in length. A miniature x-ray, 50 kV photon beam, is used as a radiation source. The source is inserted in a cylindrical brass collimator that has a pinhole of 1.0 mm in diameter along the central axis to produce a pencil photon beam. The collimator with the source is attached underneath a computer-controlled movable table which holds the specimens. Using a 45° tilted mirror, a digital camera, connected to the computer, takes pictures of the specimen and the pinhole collimator. From the image provided by the camera, the relative distance from the specimen to the pinhole axis is calculated and coordinates are sent to the movable table to properly position the samples in the beam path. Due to its monitoring system, characteristic of the radiation beam, accuracy and precision of specimen positioning, and automatic image-based specimen recognition, this radiation facility is a suitable tool to irradiate partial bodies in zebrafish embryos, cell cultures or any other small specimen used in radiobiology research.

Transition Fracture Toughness Characterization of Eurofer 97 Steel using Pre-Cracked Miniature Multi-notch Bend Bar Specimens

Energy Technology Data Exchange (ETDEWEB)

Chen, Xiang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sokolov, Mikhail A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Clowers, Logan N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-11-01

In this report, we present the feasibility study of using pre-cracked miniature multi-notch bend bar specimens (M4CVN) with a dimension of 45mm (length) x 3.3mm (width) x 1.65mm (thickness) to characterize the transition fracture toughness of Eurofer97 based on the ASTM E1921 Master Curve method. From literature survey results, we did not find any obvious specimen size effects on the measured fracture toughness of unirradiated Eurofer97. Nonetheless, in order to exclude the specimen size effect on the measured fracture toughness of neutron irradiated Eurofer97, comparison of results obtained from larger size specimens with those from smaller size specimens after neutron irradiation is necessary, which is not practical and can be formidably expensive. However, limited literature results indicate that the transition fracture toughness of Eurofer97 obtained from different specimen sizes and geometries followed the similar irradiation embrittlement trend. We then described the newly designed experimental setup to be used for testing neutron irradiated Eurofer97 pre-cracked M4CVN bend bars in the hot cell. We recently used the same setup for testing neutron irradiated F82H pre-cracked miniature multi-notch bend bars with great success. Considering the similarity in materials, specimen types, and the nature of tests between Eurofer97 and F82H, we believe the newly designed experimental setup can be used successfully in fracture toughness testing of Eurofer97 pre-cracked M4CVN specimens.

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

Energy Technology Data Exchange (ETDEWEB)

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

2002-12-01

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

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.

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

International Nuclear Information System (INIS)

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

2001-02-01

Tensile specimens of OFHC-copper were irradiated with fission neutrons in the DR-3 reactor at Risoe 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 in the as-irradiated condition at 100 deg. C whereas other were given a post-irradiation annealing at 300 deg. C for 50 h and subsequently tested at 100 deg. C. Transmission electron microscopy was used to characterize the microstructure of specimens in the as-irradiation as well as irradiation and annealed conditions both before and after tensile deformation. The results show that while the interstitial loop microstructure coarsens with irradiation dose, no significant changes are observed in the population of stacking fault tetrahedra. The results also illustrates that the post-irradiation annealing leads to only a partial recovery and that the level of recovery depends on the irradiation dose level. However, the post-irradiation annealing eliminates the problem of yield drop and reinstates enough uniform elongation to render the material useful again. These results are discussed in terms of the cascade induced source hardening (CISH) and the dispersed barrier hardening (DBH) models. Both technological and scientific implications of these results are considered. (au)

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

International Nuclear Information System (INIS)

Hawthorne, J.R.

1996-01-01

The Charpy-V (C 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 x 10 19 n/cm 2 and 4.18 x 10 19 n/cm 2 , E > 1 MeV. The irradiation portion of the IAR condition represents an incremental fluence increase of 1. 05 x 10 19 n/cm 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 V upper shelf recovery and full or nearly full recovery in the C V 41 J (30 ft-lb) transition temperature. The C 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

Neutron irradiation effect on the strength of jointed Ti-6Al-4V alloy

International Nuclear Information System (INIS)

Ishiyama, Shintaro; Miya, Naoyuki

2002-01-01

In order to investigate applicability of Ti alloy to large scaled structural material for fusion reactors, irradiation effect on the mechanical properties of Ti-6Al-4V alloy and its TIG welded material was investigated after neutron irradiation (temperature: 746-788K, fluence: 2.8 x 10 23 n/m 2 (>0.18 MeV). The following results were obtained. (1) Irradiated Ti alloy shows about 20-30% increase of its tensile strength and large degradation of fracture elongation, comparing with those of unirradiated Ti alloy. (2) TIG welded material behaves as Ti alloy in its tensile test, however, shows 30% increase of area reduction in 373-473K, whereas 1/2 degradation of area reduction over 600K. (3) Irradiated TIG welded material behaves heavier embrittlement than that of irradiated Ti alloy. (4) Charpy impact properties of un- and irradiated Ti alloys shift to ductile from brittle fracture and transition temperature shift, ΔT was estimated as about 100K. (5) Remarkable increase of hardness was found, especially in HAZ of TIG welded material after irradiation. (author)

Flux effect on neutron irradiation embrittlement of reactor pressure vessel steels irradiated to high fluences

International Nuclear Information System (INIS)

Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A.; Iwasaki, M.; Tsuno, S.; Akiyama, T.; Watanabe, S.; Ohta, T.

2011-01-01

Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels is of great concern for the long term operation of light water reactors. In particular, the embrittlement of the RPV steels of pressurized water reactors (PWRs) at very high fluences beyond 6*10 19 n/cm 2 , E > 1 MeV, needs to be understood in more depth because materials irradiated in material test reactors (MTRs) to such high fluences show larger shifts than predicted by current embrittlement correlation equations available worldwide. The primary difference between the irradiation conditions of MTRs and surveillance capsules is the neutron flux. The neutron flux of MTR is typically more than one order of magnitude higher than that of surveillance capsule, but it is not necessarily clear if this difference in neutron flux causes difference in mechanical properties of RPV. In this paper, we perform direct comparison, in terms of mechanical property and microstructure, between the materials irradiated in surveillance capsules and MTRs to clarify the effect of flux at very high fluences and fluxes. We irradiate the archive materials of some of the commercial reactors in Japan in the MTR, LVR-15, of NRI Rez, Czech Republic. Charpy impact test results of the MTR-irradiated materials are compared with the data from surveillance tests. The comparison of the results of microstructural analyses by means of atom probe tomography is also described to demonstrate the similarity / differences in surveillance and MTR-irradiated materials in terms of solute atom behavior. It appears that high Cu material irradiated in a MTR presents larger shifts than those of surveillance data, while low Cu materials present similar embrittlement. The microstructural changes caused by MTR irradiation and surveillance irradiation are clearly different

Neutron metrology in the HFR. Steel irradiation. R139-801 (SINAS)

International Nuclear Information System (INIS)

Ketema, D.J.

1999-02-01

The R139-80 series irradiation experiments is part of the NRG materials test programme to evaluate the irradiation behaviour of several types of austenitic stainless steel. Within this programme five R139-80 specimen holders were irradiated in the HFR Petten to different dose levels. This report presents the final metrology results obtained from activation monitors in a specimen holder, coded as R139-801, containing 12 Compact Tension (CT-10 mm) specimens made from the austenitic stainless steel types 308LSXB/TIG and 304-SXB. The R139-801 assembly was irradiated in channel 1 of a TRIO type facility placed in HFR core-position F8. The aim of this irradiation of specimen holder R139-801 was to reach a minimum target damage level of 7.5 dpa for the specimens at a temperature of 335C. The monitor sets are used to calculate the thermal and fast neutron fluences, displacements per atom and the generated helium content. Additionally detailed information concerning an estimation of the fluence and damage doses received by each specimen and its temperature during irradiation are presented. The main results of the thermal and fast neutron fluence measurements are presented. The results indicate that the obtained damage levels in the steel specimens loaded in this specimen holder vary from 5.8 to 7.9 dpa. The temperatures of the specimens during irradiation varied between 304 and 337C. 14 refs

Isolation of chlamydia in irradiated and non-irradiated McCoy cells

International Nuclear Information System (INIS)

Johnson, L.; Harper, I.A.

1975-01-01

Specimens from eye and genital tract were cultured in parallel in irradiated and non-irradiated McCoy cells and the frequency of isolation of chlamydia using these culture methods was compared. There was a significant difference between the frequencies of isolation; irradiated McCoy cells produced a greater number of positive results. (author)

Preparation of TEM specimen by cross-section technique

International Nuclear Information System (INIS)

Hamada, Shozo

1986-01-01

Transmission electron microscopy (TEM) is applied to the direct observation of the depth dependent damage structure in ion-irradiated stainless steel by using the cross-section technique; obtaining the TEM specimen from a slice of the irradiated stainless steel with thick Ni plating. Here has been developed the specimen preparation method of cross-section technique without heat treatment, which was necessary in the conventional method to strengthen the bonding between Ni and stainless steel. Nickel plating with good bonding to stainless steel is enabled by the following manner. First, the irradiated stainless steel is immersed in the Wood's nickel solution at room temperature for 60s to activate the surface, followed by the stricking for 300s at a current density of 300 A/m 2 in the solution to make fine and homogeneous nucleation of Ni on the stainless steel. Then, the sample is plated with Ni in the Watt's nickel plating solution at 333 K with current density of 900 ∼ 1,000 A/m 2 . The TEM disc is obtained by mechanical slicing from the specimen with Ni plating of more than 3 mm thickness. Electropolishing is accomplished by using both Ballmann method and jet electropolishing to perforate the disc accurately at the aimed point for the observation of the damage structure. (author)

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

International Nuclear Information System (INIS)

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

2007-03-01

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

Computer model for ductile fracture

International Nuclear Information System (INIS)

Moran, B.; Reaugh, J. E.

1979-01-01

A computer model is described for predicting ductile fracture initiation and propagation. The computer fracture model is calibrated by simple and notched round-bar tension tests and a precracked compact tension test. The model is used to predict fracture initiation and propagation in a Charpy specimen and compare the results with experiments. The calibrated model provides a correlation between Charpy V-notch (CVN) fracture energy and any measure of fracture toughness, such as J/sub Ic/. A second simpler empirical correlation was obtained using the energy to initiate fracture in the Charpy specimen rather than total energy CVN, and compared the results with the empirical correlation of Rolfe and Novak

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.

Survey of irradiation embrittlement effects on the mechanical properties of alloyed steels

International Nuclear Information System (INIS)

Gillemot, F.

1992-01-01

In the everyday engineering practice the neutron irradiation embrittlement of the PWR wall materials is measured by empirical methods like Charpy impact testing. New developments in fracture mechanics are given better material characteristics. The use of Absorbed Specific Fracture Energy Measured on tensile bars is a promising way to solve the problem. On the other hand the IAEA runs coordinated research program to correlate the chemical analysis with the rate of the neutron embrittlement. Better understanding of the physics of neutron embrittlement should help the life time management of the PWR vessels

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.

Feasibility of and methodology for thermal annealing an embrittled reactor vessel. Volume 2. Detailed technical description of the work. Final report

International Nuclear Information System (INIS)

Mager, T.R.

1982-11-01

Program materials were three weldments fabricated from A533 Grade B class 1 plate material and Mn Mo Ni weld wire. Specimens fabricated from the three submerged arc weldments included Type A Charpy V-notch impact, small size tensile, and 1/2T compact tension specimens. After encapsulation, the specimens were irradiated at the UVAR to two fluence levels, 8 x 10 18 n/cm 2 and 1.5 x 10 19 n/cm 2 (E > 1 MeV). Specimens were subjected to sequences of irradiation and anneals and then tested. Metallurgial/mechanistic analyses were also performed. It was concluded that excellent recovery of all properties could be achieved by annealing at greater than or equal to 850 0 F (454 0 C) for 168 hours. Such an annealing resulted in ductile-brittle transition temperature shift recovery of 80 to 100%, and reirradiation after this annealing indicated that the ductile-brittle transition temperature shift appears to continue at the expected rate. Several drawbacks were identified for wet thermal annealing. A conceptual dry in-situ thermal annealing procedure was developed for thermal annealing embrittled reactor vessels

Mechanical and irradiation properties of zirconium alloys irradiated in HANARO

International Nuclear Information System (INIS)

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

2011-01-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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 ∼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 ∼25% δ-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

Fracture toughness measurements with subsize disk compact specimens

International Nuclear Information System (INIS)

Alexander, D.J.

1992-01-01

Special fixtures and test methods have been developed for testing small disk compact specimens (12.5 mm diam by 4.6 mm thick). Both unloading compliance and potential drop methods have been used to monitor crack extension during the J-integral resistance (J-R) curve testing. Provisions have been made to allow the necessary probes and instrumentation to be installed remotely using manipulators for testing of irradiated specimens in a hat cell. Laboratory trials showed that both unloading compliance and potential drop gave useful results. Both techniques gave similar data, and predicted the final crack extension within allowable limits. The results from the small disk compact specimens were similar to results from conventional compact specimens 12.7-mm thick. However, the slopes of the J-R curves from the larger specimens were lower, suggesting that the smaller disk compact specimens may have lost some constraint due to their size. The testing shows that it should be possible to generate useful J-R curve fracture toughness data from the small disk compact specimens

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)

Investigation on the effects of gamma irradiation on bitumen

International Nuclear Information System (INIS)

Mello, M.S.; Braz, D.; Motta, L.M.G.

2011-01-01

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 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)

Investigations of X-ray irradiation of marine fish aboard

International Nuclear Information System (INIS)

Karnop, G.; Reinacher, E.; Antonacopoulos, N.; Meyer, V.

1976-01-01

Studies on X-ray irradiation of ocean perch, cod and coley (at doses of 50-150 krad) are described. The results show that irradiation within this dose range has no significant effect on the shelf-life of fish stored in ice. Although irradiation positively influenced bacteriological and chemical characteristics (e.g. reduction of total aerobic count, and inhibition of decomposition of N-containing compounds), the organoleptically-limited shelf-life of irradiated specimens was similar to that of non-irradiated specimens. Organoleptic changes in irradiated and in non-irradiated samples differed; this is attributed to the abnormal spoilage flora (mainly radiation-resistant Moraxella spp.) in the irradiated samples. (orig./HP) [de

Multispecimen dual-beam irradiation damage chamber

International Nuclear Information System (INIS)

Packan, N.H.; Buhl, R.A.

1980-06-01

An irradiation damage chamber that can be used to rapidly simulate fast neutron damage in fission or fusion materials has been designed and constructed. The chamber operates in conjunction with dual Van de Graaff accelerators at ORNL to simulate a wide range of irradiation conditions, including pulsed irradiation. Up to six experiments, each with up to nine 3-mm disk specimens, can be loaded into the ultrahigh vacuum chamber. Specimen holders are heated with individual electron guns, and the temperature of each specimen can be monitored during bombardment by an infrared pyrometer. Three different dose levels may be obtained during any single bombardment, and the heavy-ion flux on each of the nine specimens can be measured independently with only a brief interruption of the beam. The chamber has been in service for nearly three years, during which time approximately 250 bombardments have been successfully carried out. An appendix contains detailed procedures for operating the chamber

Effect of lead factors on the embrittlement of RPV SA-508 cl 3 steel

Energy Technology Data Exchange (ETDEWEB)

Kempf, Rodolfo, E-mail: kempf@cnea.gov.ar [CNEA, Unidad Actividad Combustibles Nucleares, División Caracterización, Avda. Gral Paz 1499, C.P.B1650KNA, San Martín, Buenos Aires (Argentina); Troiani, Horacio, E-mail: troiani@cab.cnea.gov.ar [Centro Atómico Bariloche (CNEA) e Instituto Balseiro (UNCU), CONICET, Av. Bustillo 9500, CP 8400, Rio Negro (Argentina); Fortis, Ana Maria, E-mail: fortis@cnea.gov.ar [CNEA, Departamento Estructura y Comportamiento, UNSAM, Avda. Gral Paz 1499, C.P.B1650KNA, San Martín, Buenos Aires (Argentina)

2013-03-15

This paper presents a project to study the effect of lead factors on the mechanical behaviour of the SA-508 type 3 Reactor Pressure Vessel (RPV) steel used in the reactor under construction Atucha II in Argentina. Charpy-V notch specimens of this steel were irradiated at the RA1 experimental reactor at a temperature of 275 °C with two lead factors (186 and 93). The neutron flux was 3.71 × 10{sup 15} n m{sup −2} s{sup −1} and 1.85 × 10{sup 15} n m{sup −2} s{sup −1} (E > 1 MeV) respectively. In both cases, the fluence was 6.6 × 10{sup 21} n m{sup −2}, which is equivalent to that received by the PHWR Atucha II RPV in 10 years of full power irradiation. The results of Charpy tests revealed significant embrittlement both in the ΔT = 14 °C and ΔT = 21 °C shifts of the ductile–brittle transition temperatures (DBTT) and in the reduction of the maximum energy absorbed. This result shows that the shift of the DBTT with a lead factor of 93 is larger than that obtained with a lead factor of 186. Then, the results of irradiation in experimental reactors (MTR) with high lead factors may not be conservative with respect to the actual RPV embrittlement.

Irradiation hardening and localized deformation of neutron-irradiated α-iron single crystals

International Nuclear Information System (INIS)

Mughrabi, H.; Stroehle, D.; Wilkens, M.

1981-01-01

The early yielding behaviour of neutron-irradiated α iron single crystals orientated for single slip was investigated as a function of neutron dose. In the range of neutron doses between approx. equal to 10 18 and approx. equal to 10 19 n/cm 2 , the irradiation hardening increment was found to be almost constant. Qualitative modifications of this behaviour were observed in the case of predeformed specimens. The localized deformation of the neutron-irradiated specimens by dislocation channelling was investigated by slip-line observations, transmission electron microscopy and X-ray topography. A model of localized deformation is proposed in order to explain the development of the observed asymmetric dislocation double layers which bound the channels and transmit characteristic misorientations. (orig.)

Generic analyses for evaluation of low Charpy upper-shelf energy effects on safety margins against fracture of reactor pressure vessel materials

International Nuclear Information System (INIS)

Dickson, T.L.

1993-07-01

Appendix G to 10 CFR Part 50 requires that reactor pressure vessel beltline material maintain Charpy upper-shelf energies of no less than 50 ft-lb during the plant operating life, unless it is demonstrated in a manner approved by the Nuclear Regulatory Commission (NRC), that lower values of Charpy upper-shelf energy provide margins of safety against fracture equivalent to those in Appendix G to Section XI of the ASME Code. Analyses based on acceptance criteria and analysis methods adopted in the ASME Code Case N-512 are described herein. Additional information on material properties was provided by the NRC, Office of Nuclear Regulatory Research, Materials Engineering Branch. These cases, specified by the NRC, represent generic applications to boiling water reactor and pressurized water reactor vessels. This report is designated as HSST Report No. 140

Elemental microanalysis of biological and medical specimens with a scanning proton microprobe

International Nuclear Information System (INIS)

Legge, G.J.F.; Mazzolini, A.P.

1979-01-01

The scanning proton microprobe is shown to be a sensitive instrument for elemental microanalysis of cells and tissues in biological and medical specimens. The preparation of specimens and their behaviour under irradiation are crucial and the application of quantitative scanning analysis to the monitoring of such problems is illustrated

Irradiation Microstructure of Austenitic Steels and Cast Steels Irradiated in the BOR-60 Reactor at 320°C

Science.gov (United States)

Yang, Yong; Chen, Yiren; Huang, Yina; Allen, Todd; Rao, Appajosula

Reactor internal components are subjected to neutron irradiation in light water reactors, and with the aging of nuclear power plants around the world, irradiation-induced material degradations are of concern for reactor internals. Irradiation-induced defects resulting from displacement damage are critical for understanding degradation in structural materials. In the present work, microstructural changes due to irradiation in austenitic stainless steels and cast steels were characterized using transmission electron microscopy. The specimens were irradiated in the BOR-60 reactor, a fast breeder reactor, up to 40 dpa at 320°C. The dose rate was approximately 9.4x10-7 dpa/s. Void swelling and irradiation defects were analyzed for these specimens. A high density of faulted loops dominated the irradiated-altered microstructures. Along with previous TEM results, a dose dependence of the defect structure was established at 320°C.

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.

Remote replacement of materials open-test assembly specimens at the FFTF/IEM cell

International Nuclear Information System (INIS)

Gibbons, P.W.; Ramsey, E.B.

1990-01-01

The Fast Flux Test Facility (FFTF) interim examination and maintenance (IEM) cell is used for the remote disassembly of irradiated fuel and materials experiments. The materials open-test assembly (MOTA) is brought to the IEM cell for materials test specimen removal. The specimens are shipped to the materials laboratory for sorting and installation in new specimen holders and then returned within 10 days to the IEM cell where they are installed in a new MOTA vehicle for further irradiation. Reconstituting a MOTA is a challenging remote operation involving dozens of steps and two separate facilities. Handling and disassembling sodium-wetted components pose interesting handling, cleaning, and disposal challenges. The success of this system is evidenced by its timely completion in the critical path of FFTF outage schedules

Heavy-Section Steel Irradiation Program

International Nuclear Information System (INIS)

Anon.

1991-01-01

In FY1990 the Heavy-Section Steel Irradiation (HSSI) Program was arranged into 8 tasks: (1) program management, (2) K Ic curve shift in high-copper welds, (3) K Ia curve shift in high-copper welds, (4) irradiation effects on cladding, (5) K Ic and K Ia curve shifts in low upper-shelf (LUS) welds, (6) irradiation effects in a commercial LUS weld, (7) microstructural analysis of irradiation effects, and (8) in-service aged material evaluations. Of particular interest are the efforts in FY1990 concerning the shifts in fracture toughness and crack arrest toughness in high-copper welds, the unirradiated examination of a LUS weld from the Midland reactor, and the continued investigation into the causes of accelerated low-temperature embrittlement recently observed in RPV support steels. In the Fifth and Sixth Irradiation Series, designed to examine the shifts and possible changes in shape in the ASME K Ic and K Ia curves for two irradiated high-copper welds, it was seen that both the lower bound and mean fracture toughness shifts were greater than those of the associated Charpy-impact energies, whereas the shifts in crack arrest toughness were comparable. The irradiation-shifted fracture toughness data fell slightly below the appropriately indexed ASME K Ic curve even when it was shifted according to Revision 2 of Regulatory Guide 1.99 including its margins. The beltline weld, which was removed from the Midland reactor, fabricated by Babcock and Wilcox, Co. using Linde 80 flux, is being examined in the Tenth Irradiation Series to establish the effects of irradiation on a commercial LUS weld. A wide variation in the unirradiated fracture properties of the Midland weld were measured with values of RT NDT ranging from -22 to 54F through its thickness. In addition, a wide range of copper content from 0.21 to 0.45 wt % was found, compared to the 0.42 wt % previously reported

Some recent innovations in small specimen testing

International Nuclear Information System (INIS)

Odette, G.R.; He, M.; Gragg, D.; Klingensmith, D.; Lucas, G.E.

2002-01-01

New innovative small specimen test techniques are described. Finite element simulations show that combinations of cone indentation pile-up geometry and load-penetration depth relations can be used to determine both the yield stress and strain-hardening behavior of a material. Techniques for pre-cracking and testing sub-miniaturized fracture toughness bend bars, with dimensions of 1.65x1.65x9 mm 3 , or less, are described. The corresponding toughness-temperature curves have a very steep transition slope, primarily due to rapid loss of constraint, which has advantages in some experiments to characterize the effects of specified irradiation variables. As one example of using composite specimens, an approach to evaluating helium effects is proposed, involving diffusion bonding small wires of a 54 Fe-based ferritic-martensitic alloy to a surrounding fracture specimen composed of an elemental Fe-based alloy. Finally, we briefly outline some potential approaches to multipurpose specimens and test automation

Effects of residual stress on irradiation hardening in stainless steels

Energy Technology Data Exchange (ETDEWEB)

Okubo, N.; Kondo, K.; Kaji, Y. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Miwa, Y. [Nuclear Energy and Science Directorate, Japan Atomic Energy Agency, Tokai-mura, Ibaraki-ken (Japan)

2007-07-01

Full text of publication follows: Structural materials in fusion reactor with water cooling system will undergo corrosion in aqueous environment and heavier irradiation than that in LWR. Irradiation assisted stress corrosion (IASCC) may be induced in stainless steels exposed in these environment for a long term of reactor operation. The IASCC is considered to be caused in a welding zone. It is difficult to predict and estimate the IASCC, because several irradiation effects (irradiation hardening, swelling, irradiation induced stress relaxation, etc) work intricately. Firstly, effects of residual stress on irradiation hardening were investigated in stainless steels. Specimens used in this study were SUS316 and SUS316L. By bending deformation, the specimens with several % plastic strain, which corresponds to weld residual stress, were prepared. Ion irradiations of 12 MeV Ni{sup 3+} were performed at 330, 400 and 550 deg. C to 45 dpa in TIARA facility at JAEA. No bent specimen was simultaneously irradiated with the bent specimen. The residual stress was estimated by X-ray residual stress measurements before and after the irradiation. The micro-hardness was measured by using nano-indenter. The irradiation hardening and the stress relaxation were changed by irradiation under bending deformation. The residual stress did not relax even for the case of the higher temperature aging at 500 deg. C for the same time of irradiation. The residual stress after ion irradiation, however, relaxed at these experimental temperatures in SUS316L. The hardness was obviously suppressed in bent SUS316L irradiated at 300 deg. C to 6 or 12 dpa. It was evident that irradiation induced stress relaxation occasionally suppressed the irradiation hardening in SUS316L. (authors)

Evaluation of the mechanical properties of carbon fiber after electron beam irradiation

International Nuclear Information System (INIS)

Giovedi, Claudia; Diva Brocardo Machado, Luci; Augusto, Marcos; Segura Pino, Eddy; Radino, Patricia

2005-01-01

Carbon fibers are used as reinforcement material in epoxy matrix in advanced composites. An important aspect of the mechanical properties of composites is associated to the adhesion between the surface of the carbon fiber and the epoxy matrix. This paper aimed to the evaluation of the effects of EB irradiation on the tensile properties of two different carbon fibers prepared as resin-impregnated specimens. The fibers were EB irradiated before the preparation of the resin-impregnated specimens for mechanical tests. Observations of the specimens after breakage have shown that EB irradiation promoted significant changes in the failure mode. Furthermore, the tensile strength data obtained for resin-impregnated specimens prepared with carbons fibers previously irradiated presented a slight tendency to be higher than those obtained from non-irradiated carbon fibers

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.

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

International Nuclear Information System (INIS)

Jin, Y. G.; Kim, G. S.; Baik, S. J.; Baek, S. Y.

2016-01-01

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

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

In situ transmission electron microscope studies of ion irradiation-induced and irradiation-enhanced phase changes

International Nuclear Information System (INIS)

Allen, C.W.

1992-01-01

Motivated at least initially by materials needs for nuclear reactor development, extensive irradiation effects studies employing transmission electron microscopes (TEM) have been performed for several decades, involving irradiation-induced and irradiation-enhanced microstructural changes, including phase transformations such as precipitation, dissolution, crystallization, amorphization, and order-disorder phenomena. From the introduction of commercial high voltage electron microscopes (HVEM) in the mid-1960s, studies of electron irradiation effects have constituted a major aspect of HVEM application in materials science. For irradiation effects studies two additional developments have had particularly significant impact; the development of TEM specimen holder sin which specimen temperature can be controlled in the range 10-2200 K and the interfacing of ion accelerators which allows in situ TEM studies of irradiation effects and the ion beam modification of materials within this broad temperature range. This paper treats several aspects of in situ studies of electron and ion beam-induced and enhanced phase changes and presents two case studies involving in situ experiments performed in an HVEM to illustrate the strategies of such an approach of the materials research of irradiation effects

Nuclear reactor pressure vessel surveillance capsule examinations. Application of American Society for Testing and Materials Standards

International Nuclear Information System (INIS)

Perrin, J.S.

1978-01-01

A series of pressure vessel surveillance capsules is installed in each commercial nuclear power plant in the United States. A capsule typically contains neutron dose meters, thermal monitors, tensile specimens, and Charpy V-notch impact specimens. In order to determine property changes of the pressure vessel resulting from irradiation, surveillance capsules are periodically removed during the life of a reactor and examined. There are numerous standards, regulations, and codes governing US pressure vessel surveillance capsule programmes. These are put out by the US Nuclear Regulatory Commission, the Boiler and Pressure Vessel Committee of the American Society of Mechanical Engineers, and the American Society for Testing and Materials (ASTM). A majority of the pertinent ASTM standards are under the jurisdiction of ASTM Committee E-10 on Nuclear Applications and Measurements of Radiation Effects. The standards, regulations, and codes pertaining to pressure vessel surveillance play an important role in ensuring reliability of the nuclear pressure vessels. ASTM E 185-73 is the Standard Recommended Practice for Surveillance Tests for Nuclear Reactors. This standard recommends procedures for both the irradiation and subsequent testing of surveillance capsules. ASTM E 185-73 references many additional specialized ASTM standards to be followed in specific areas of a surveillance capsule examination. A key element of surveillance capsule programmes is the Charpy V-notch impact test, used to define curves of fracture behaviour over a range of temperatures. The data from these tests are used to define the adjusted reference temperature used in determining pressure-temperature operating curves for a nuclear power plant. (author)

Experimental data base for assessment of irradiation induced ageing effects in pre-irradiated RPV materials of German PWR

Energy Technology Data Exchange (ETDEWEB)

Hein, H.; Gundermann, A.; Keim, E.; Schnabel, H. [AREVA NP GmbH (Germany); Ganswind, J. [VGB PowerTech e.V (Germany)

2011-07-01

The 5 year research program CARISMA which ended in 2008 has produced a data base to characterize the fracture toughness of pre-irradiated original RPV (Reactor Pressure Vessel) materials being representative for all four German PWR construction lines of former Siemens/KWU company. For this purpose tensile, Charpy-V impact, crack initiation and crack arrest tests have been performed for three base materials and four weld metals irradiated to neutron fluences beyond the designed EoL range. RPV steels with optimized chemical composition and with high copper as well as high nickel content were examined in this study. The RTNDT concept and the Master Curve approach were applied for the assessment of the generated data in order to compare both approaches. A further objective was to clarify in which extent crack arrest curves can be generated for irradiated materials and how crack arrest can be integrated into the Master Curve approach. By the ongoing follow-up project CARINA the experimental data base will be extended by additional representative materials irradiated under different conditions and with respect to the accumulated neutron fluences and specific impact parameters such as neutron flux and manufacturing effects. The irradiation data cover also the long term irradiation behavior of the RPV steels concerned. Moreover, most of the irradiated materials were and will be used for microstructural examinations to get a deeper insight in the irradiation embrittlement mechanisms and their causal relationship to the material property changes. By evaluation of the data base the applicability of the Master Curve approach for both crack initiation and arrest was confirmed to a large extent. Moreover, within both research programs progress was made in the development of crack arrest test techniques and in specific issues of RPV integrity assessment. (authors)

Specimen preparation of irradiated materials for examination in the atom probe field ion microscope

International Nuclear Information System (INIS)

Russell, K.F.; Miller, M.K.

1994-01-01

The atom probe field ion microscope (APFIM) requires specimens in the form of ultrasharp needles. Basic protective measures used to reduce exposure druing specimen preparation are discussed. The low-level radioactive specimen blanks may be made using a two-stage electropolishing process using a thin layer of electrolyte floating on a denser inert liquid; this produces a necked region and eventually two specimens from each single blank. The amount of material handled may also be reduced using a micropolishing technique to repolish blunt or fractured specimens. Control of contamination and possible spills is discussed

Elemental microanalysis of botanical specimens using the Melbourne Proton Microprobe

International Nuclear Information System (INIS)

Mazzolini, A.P.J.; Legge, G.J.F.

1978-01-01

A proton microprobe has been used to obtain the distribution of elements of various botanical specimens. This paper presents preliminary results obtained by the irradiation of certain organs of the wheat plant

Kinetics of annealing of irradiated surveillance pressure vessel steel

International Nuclear Information System (INIS)

Harvey, D.J.; Wechsler, M.S.

1982-01-01

Indentation hardness measurements as a function of annealing were made on broken halves of Charpy impact surveillance samples. The samples had been irradiated in commercial power reactors to a neutron fluence of approximately 1 x 10 18 neutrons per cm 2 , E > 1 MeV, at a temperature of about 300 0 C (570 0 F). Results are reported for the weld metal, which showed greater radiation hardening than the base plate or heat-affected zone material. Isochronal and isothermal anneals were conducted on the irradiated surveillance samples and on unirradiated control samples. No hardness changes upon annealing occurred for the control samples. The recovery in hardness for the irradiated samples took place mostly between 400 and 500 0 C. Based on the Meechan-Brinkman method of analysis, the activation energy for annealing was found to be 0.60 +- 0.06 eV. According to computer simulation calculations of Beeler, the activation energy for migration of vacancies in alpha iron is about 0.67 eV. Therefore, the results of this preliminary study appear to be consistent with a mechanism of annealing of radiation damage in pressure vessel steels based on the migration of radiation-produced lattice vacancies

Fracture toughness evaluation of Eurofer'97 by testing small specimens

International Nuclear Information System (INIS)

Serrano, M.; Fernandez, P.; Lapena, J.

2006-01-01

The Eurofer'97 is the structural reference material that will be tested in the ITER modules. Its metallurgical properties have been well characterized during the last years. However, more investigations related with the fracture toughness of this material are necessary because this property is one of the most important to design structural components and to study their integrity assessment. In the case of structural materials for fusion reactor the small specimen technology (SSTT) are being actively developed to investigate the fracture toughness among other mechanical properties. The use of small specimens is due to the small available irradiation volume of IFMIF and also due to the high fluence expected in the fusion reactor. The aim of this paper is to determine the fracture toughness of the Eurofer'97 steel by testing small specimens of different geometry in the ductile to brittle transition region, with the application of the Master Curve methodology, and to evaluate this method to assess the decrease in fracture toughness due to neutron irradiation. The tests and data analysis have been performed following the Master Curve approach included in the ASTM Standard E1921-05. Specimen size effect and comparison of the fracture toughness results with data available in the literature are also considered. (author)

A study on the Charpy-V impact energy and impact properties of stainless steel

International Nuclear Information System (INIS)

Han, Gill Young

1988-01-01

It has been thought that by analyzing and considering the shock strength and fracture of impact load. We can accurtely determine strength and thus it will be helpful in optimum design. In this experimental study the following results were obtained by using the instrumented impact test for SUS 316 1) The total charpy impact energy is progressively decreased by increasing the shock pressure. 2) The dynamic yield strength is increased by increasing the shock pressure for all test temperatures. 3) The ratio of dynamic yield to static yield strength was found to decrease with increasing shock pressure. (Author)

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Production of iodine-123 radiobiological specimen on 25 MeV electron beam

International Nuclear Information System (INIS)

Oganesyan, Yu.Ts.; Starodub, G.Ya.; Buklanov, G.V.; Korotkin, Yu.S.; Belov, A.G.

1988-01-01

The technique is described and experimental results are presented for production of radioactive specimen-iodine-123 for medical biological investigations. It is shown that in ten hour irradiation of 124 Xe enriched target of 10 g weight by the 25 MeV electron beam at MT-25 microtron short lived 123 I with activity of about 200 mCl can be accumulated. The procedure was developed for extraction of radioactive atoms and preparing the solution that permits to obtain during 1-1.5 h after the end of irradiation the specimen which satisfies all pharmacopeia requirements. It follows from the results that using small-size electron accelerators with the beam energy up to 25 MeV permits to organize economical and large-scale production of high quality radioactive specimen of 123 I for servicing a large region of this country. 14 refs.; 4 figs.; 1 tab

Computational evaluation of the constraint loss on the fracture toughness of reactor pressure vessel steels

International Nuclear Information System (INIS)

Serrano Garcia, M.

2007-01-01

The Master Curve approach is included on the ASME Code through some Code Cases to assess the reactor pressure vessel integrity. However, the margin definition to be added is not defined as is the margin to be added when the Master Curve reference temperature T 0 is obtained by testing pre-cracked Charpy specimens. The reason is that the T 0 value obtained with this specimen geometry is less conservative than the value obtained by testing compact tension specimens possible due to a loss of constraint. The two parameter fracture mechanics, considered as an extension of the classical fracture mechanics, coupled to a micromechanical fracture models is a valuable tool to assess the effect of constraint loss on fracture toughness. The definition of a parameter able to connect the fracture toughens value to the constraint level on the crack tip will allow to quantify margin to be added to the T 0 value when this value is obtained testing the pre-cracked Charpy specimens included in the surveillance capsule of the reactor pressure vessel. The Nuclear Regulatory Commission (NRC) define on the To value obtained by testing compact tension specimens and ben specimens (as pre-cracked Charpy are) bias. the NRC do not approved any of the direct applications of the Master Curve the reactor pressure vessel integrity assessment until this bias will be quantified in a reliable way. the inclusion of the bias on the integrity assessment is done through a margin to be added. In this thesis the bias is demonstrated an quantified empirical and numerically and a generic value is suggested for reactor pressure vessel materials, so that it can be used as a margin to be added to the T 0 value obtained by testing the Charpy specimens included in the surveillance capsules. (Author) 111 ref

Structural changes induced by electron irradiation

International Nuclear Information System (INIS)

Koike, J.; Pedraza, D.F.

1993-01-01

Highly oriented pyrolytic graphite was irradiated at room temperature with 300 kV electrons. Transmission electron microscopy and electron energy loss spectroscopy were employed to study the structural changes produced by irradiation. The occurrence of a continuous ring intensity in the selected area diffraction (SAD) pattern obtained on a specimen irradiated with the electron beam parallel to the c-crystallographic axis indicated that microstructural changes had occurred. However, from the SAD pattern obtained for the specimens tilted relative to the irradiation direction, it was found that up to a fluence of 1.1x10 27 e/m 2 graphite remained crystalline. An SAD pattern of a specimen irradiated with the electron beam perpendicular to the c-axis confirmed the persistence of crystalline order. High resolution electron microscopy showed that ordering along the c-axis direction remained. A density reduction of 8.9% due to irradiation was determined from the plasmon frequency shift. A qualitative model is proposed to explain these observations. A new determination of the threshold displacement energy, Ed, of carbon atoms in graphite was done by examining the appearance of a continuous ring in the SAD pattern at various electron energies. A value of 30 eV was obtained whether the incident electron beam was parallel or perpendicular to the c-axis, demonstrating that Ed is independent of the displacement direction

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).

Visual interface for the automation of the instrumented pendulum of Charpy tests used in the surveillance program of reactors vessel of nuclear power plants

International Nuclear Information System (INIS)

Rojas S, A.S.; Sainz M, E.; Ruiz E, J.A.

2004-01-01

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)

A new materials irradiation facility at the Kyoto university reactor

International Nuclear Information System (INIS)

Yoshiie, T.; Hayashi, Y.; Yanagita, S.; Xu, Q.; Satoh, Y.; Tsujimoto, H.; Kozuka, T.; Kamae, K.; Mishima, K.; Shiroya, S.; Kobayashi, K.; Utsuro, M.; Fujita, Y.

2003-01-01

A new materials irradiation facility with improved control capabilities has been installed at the Kyoto University Reactor (KUR). Several deficiencies of conventional fission neutron material irradiation systems have been corrected. The specimen temperature is controlled both by an electric heater and by the helium pressure in the irradiation tube without exposure to neutrons at temperatures different from the design test conditions. The neutron spectrum is varied by the irradiation position. Irradiation dose is changed by pulling the irradiation capsule up and down during irradiation. Several characteristics of the irradiation field were measured. The typical irradiation intensity is 9.4x10 12 n/cm 2 s (>0.1 MeV) and the irradiation temperature of specimens is controllable from 363 to 773 K with a precision of ±2 K

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Irradiation damage of ferritic/martensitic steels: Fusion program data applied to a spallation neutron source

International Nuclear Information System (INIS)

Klueh, R.L.

1997-01-01

Ferritic/martensitic steels were chosen as candidates for future fusion power plants because of their superior swelling resistance and better thermal properties than austenitic stainless steels. For the same reasons, these steels are being considered for the target structure of a spallation neutron source, where the structural materials will experience even more extreme irradiation conditions than expected in a fusion power plant first wall (i.e., high-energy neutrons that produce large amounts of displacement damage and transmutation helium). Extensive studies on the effects of neutron irradiation on the mechanical properties of ferritic/martensitic steels indicate that the major problem involves the effect of irradiation on fracture, as determined by a Charpy impact test. There are indications that helium can affect the impact behavior. Even more helium will be produced in a spallation neutron target material than in the first wall of a fusion power plant, making helium effects a prime concern for both applications. 39 refs., 10 figs

Characterisation of gamma-irradiated crystalline polymer: Pt. 3

International Nuclear Information System (INIS)

Jinhua Feng; Lihua Zhang; Donglin Chen

1991-01-01

Thermal behaviour of γ-irradiated plain PA1010 containing different amounts of difunctional cross-linking agent BMI was investigated. In DSC endo- and exotherm, it was found that during irradiation the presence of BMI markedly changes the melting and crystallisation characteristics of PA1010. A supposition that the network of BMI-containing specimens is rather loose in structure was proposed to explain the discrepancy in thermal behaviour between these two kinds of specimens. The supposition was further ascertained by the less brittleness in mechanical property of specimens containing BMI. Besides, the complexity of the thermal behaviour of γ-irradiated PA1010 was discussed and attributed mainly to the increase in σ e , the fold surface free energy of chain fold crystals. (author)

Irradiation proctitis

International Nuclear Information System (INIS)

Minami, Akira

1977-01-01

Literatures on late rectal injuries are discussed, referring to two patients with uterine cervical cancer in whom irradiation proctitis occurred after telecobalt irradiation following uterine extirpation. To one patients, a total of 5000 rads was irradiated, dividing into 250 rads at one time, and after 3 months, irradiation with a total of 2000 rads, dividing into 200 rads at one time, was further given. In another one patient, two parallel opposing portal irradiation with a total of 6000 rads was given. About a year after the irradiation, rectal injuries and cystitis, accompanying with hemorrhage, were found in both of the patients. Rectal amputation and proctotoreusis were performed. Cystitis was treated by cystic irradiation in the urological department. Pathohistological studies of the rectal specimen revealed atrophic mucosa, and dilatation of the blood vessels and edema in the colonic submucosa. Incidence of this disease, term when the disease occurs, irradiation dose, type of the disease, treatment and prevention are described on the basis of the literatures. (Kanao, N.)

Irradiation proctitis

Energy Technology Data Exchange (ETDEWEB)

Minami, A [Osaka Kita Tsishin Hospital (Japan)

1977-06-01

Literatures on late rectal injuries are discussed, referring to two patients with uterine cervical cancer in whom irradiation proctitis occurred after telecobalt irradiation following uterine extirpation. To one patients, a total of 5000 rads was irradiated, dividing into 250 rads at one time, and after 3 months, irradiation with a total of 2000 rads, dividing into 200 rads at one time, was further given. In another one patient, two parallel opposing portal irradiation with a total of 6000 rads was given. About a year after the irradiation, rectal injuries and cystitis, accompanying with hemorrhage, were found in both of the patients. Rectal amputation and proctotoreusis were performed. Cystitis was treated by cystic irradiation in the urological department. Pathohistological studies of the rectal specimen revealed atrophic mucosa, and dilatation of the blood vessels and edema in the colonic submucosa. Incidence of this disease, term when the disease occurs, irradiation dose, type of the disease, treatment and prevention are described on the basis of the literatures.

Specimen rotation system of the WSU TRIGA-fueled reactor

International Nuclear Information System (INIS)

Lovas, Thomas A.

1976-01-01

The specimen rotation system presently in use at the WSU reactor has been designed to provide maximum utilization of the irradiation capabilities achieved through use of TRIGA-type fuel. This paper describes the system with particular emphasis on characteristics which are advantageous to experimenters. (author)

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.

Design, Fabrication, and Initial Operation of a Reusable Irradiation Facility

International Nuclear Information System (INIS)

Heatherly, D.W.; Thoms, K.R.; Siman-Tov, I.I.; Hurst, M.T.

1999-01-01

A Heavy-Section Steel Irradiation (HSSI) Program project, funded by the US Nuclear Regulatory Commission, was initiated at Oak Ridge National Laboratory to develop reusable materials irradiation facilities in which metallurgical specimens of reactor pressure vessel steels could be irradiated. As a consequence, two new, identical, reusable materials irradiation facilities have been designed, fabricated, installed, and are now operating at the Ford Nuclear Reactor at the University of Michigan. The facilities are referred to as the HSSI-IAR facilities with the individual facilities being designated as IAR-1 and IAR-2. This new and unique facility design requires no cutting or grinding operations to retrieve irradiated specimens, all capsule hardware is totally reusable, and materials transported from site to site are limited to specimens only. At the time of this letter report, the facilities have operated successfully for approximately 2500 effective full-power hours

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

International Nuclear Information System (INIS)

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

1980-02-01

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

Fabrication and operation of HFIR-MFE RB* spectrally tailored irradiation capsules

International Nuclear Information System (INIS)

Longest, A.W.; Pawel, J.E.; Heatherly, D.W.; Sitterson, R.G.; Wallace, R.L.

1993-01-01

Fabrication and operation of four HFIR-MFE RB * capsules (60, 200, 330, and 400 degrees C) to accommodate MFE specimens previously irradiated in spectrally tailored experiments in the ORR are proceeding satisfactorily. With the exception of the 60 degrees C capsule, where the test specimens were in direct contact with the reactor cooling water, specimen temperatures (monitored by 21 thermocouples) are controlled by varying the thermal conductance of a thin gap region between the specimen holder outer sleeve and containment tube. Irradiation of the 60 and 330 degrees C capsules, which started on July 17, 1990, was completed on November 14, 1992, after 24 cycles of irradiation to an incremental damage level of approximately 10.9 displacements per atom (dpa). Assembly of the follow-up 200 and 400 degrees C capsules was completed in November 1992, and their planned 20-cycle irradiation to approximately 9.1 incremental dpa was started on November 21, 1992. As of February 11, 1993, the 200 and 400 degrees C capsules had successfully completed three cycles of irradiation to approximately 1.4 incremental dpa

Electron-beam irradiation effects on mechanical properties of PEEK/CF composite

International Nuclear Information System (INIS)

Sasuga, Tsuneo; Seguchi, Tadao

1989-01-01

Carbon fibre-reinforced composite (PEEK/CF) using polyarylether-ether-ketone (PEEK) as a matrix material was prepared and electron-beam irradiation effects on the mechanical properties at low and high temperatures were studied. The flexural strength and modulus of the unirradiated PEEK/CF were almost the same as those of carbon fibre-reinforced composites with epoxide resin. The mechanical properties at room temperature were little affected by irradiation up to 180 MGy, but in the test at 77K the strength of the specimens irradiated over 100 MGy was slightly decreased. The mechanical properties of the unirradiated specimen decreased with increasing testing temperature, but the high-temperature properties were improved by irradiation, i.e. the strength measured at 413K for the specimen irradiated with 120 MGy almost reached the value for the unirradiated specimen measured at room temperature. It was apparent from the viscoelastic measurement that the improvement of mechanical properties at high temperature resulted from the high-temperature shift of the glass transition of the matrix PEEK caused by radiation-induced cross-linking. (author)

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.)

Elevated-temperature tensile properties of 2 1/4 Cr-1 Mo steel irradiated in the EBR-II, AD-2 experiment

International Nuclear Information System (INIS)

Klueh, R.L.; Vitek, J.M.

1984-01-01

The effect of irradiated on the tensile properties of 2 1/4 Cr-1 Mo steel was determined for specimens irradiation in EBR-II at 390 to 550 0 C. Unirradiated control specimens and specimens aged for 5000 h at the irradiation temperatures were also tested. Irradiation to approximately 9 dpa at 390 0 C increased the strength and decreased the ductility compared with the unirradiated and aged specimens. Softening occurred in samples irradiated and tested at 450, 500, and 550 0 C

Miniature specimen technology for postirradiation fatigue crack growth testing

International Nuclear Information System (INIS)

Mervyn, D.A.; Ermi, A.M.

1979-01-01

Current magnetic fusion reactor design concepts require that the fatigue behavior of candidate first wall materials be characterized. Fatigue crack growth may, in fact, be the design limiting factor in these cyclic reactor concepts given the inevitable presence of crack-like flaws in fabricated sheet structures. Miniature specimen technology has been developed to provide the large data base necessary to characterize irradiation effects on the fatigue crack growth behavior. An electrical potential method of measuring crack growth rates is employed on miniature center-cracked-tension specimens (1.27 cm x 2.54 cm x 0.061 cm). Results of a baseline study on 20% cold-worked 316 stainless steel, which was tested in an in-cell prototypic fatigue machine, are presented. The miniature fatigue machine is designed for low cost, on-line, real time testing of irradiated fusion candidate alloys. It will enable large scale characterization and development of candidate first wall alloys

Study on silk yellowing induced by gamma-irradiation

International Nuclear Information System (INIS)

Tsukada, Masuhiro; Aoki, Akira

1985-01-01

The changes in the yellow color of silk threads with total dose of irradiation applied were described and studied by a colorimetric method and by monochrome photography. The change into a yellow color of the specimen in the course of irradiation was clearly detected in photographs using filters, 2B and SC 56 under light conditions at the wavelength of 366 nm. The b/L value measured by colorimetry in undegummed and degummed silk fibers sharply increased in the early stage of irradiation. Yellow color indices (b/L) of the specimen subjected to gamma-irradiation continued to increase and the yellow color of the silk threads became more pronounced above a total dose of irradiation of 21 Mrad. The b/L value of the undegummed silk fiber which had deen irradiated was about 2 times that of the degummed silk fiber. (author)

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)

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)

An ESR study of radicals induced in irradiated fresh mango

International Nuclear Information System (INIS)

Kikuchi, Masahiro; Hussain, Mohammed S.; Morishita, Norio; Kobayashi, Yasuhiko; Ukai, Mitsuko; Shimoyama, Yuhei

2009-01-01

An electron spin resonance (ESR) spectroscopic study was performed on the radicals induced irradiated fresh mangoes. Fresh Philippine mangoes were irradiated by the γ-rays, lyophilized and powdered. The ESR spectrum of the dry specimen showed a strong main peak at g=2.004 and a pair of peaks at both magnetic fields of the main peak. The main peak detected from flesh and skin specimens faded away in a few days after the irradiation. On the other hand, the side peaks showed a well-defined dose response even 9 days after the irradiation. The side-peak is a useful mean to define the irradiation on fresh mangoes. (author)

Plastic zone size for nanoindentation of irradiated Fe–9%Cr ODS

Energy Technology Data Exchange (ETDEWEB)

Dolph, Corey K. [Boise State University, 1910 University Drive, Boise, ID 83725 (United States); Silva, Douglas J. da [Boise State University, 1910 University Drive, Boise, ID 83725 (United States); Federal University of São Carlos, Rodovia Washington Luís, km 235 - SP-310, São Carlos, São Paulo (Brazil); Swenson, Matthew J. [Boise State University, 1910 University Drive, Boise, ID 83725 (United States); Wharry, Janelle P., E-mail: jwharry@purdue.edu [Boise State University, 1910 University Drive, Boise, ID 83725 (United States); Purdue University, 400 Central Drive, West Lafayette, IN 47907 (United States)

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{sup ++} 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.

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.

Irradiation embrittlement and optimisation of annealing

International Nuclear Information System (INIS)

1993-01-01

This conference is composed of 30 papers grouped in 6 sessions related to the following themes: neutron irradiation effects in pressure vessel steels and weldments used in PWR, WWER and BWR nuclear plants; results from surveillance programmes (irradiation induced damage and annealing processes); studies on the influence of variations in irradiation conditions and mechanisms, and modelling; mitigation of irradiation effects, especially through thermal annealing; mechanical test procedures and specimen size effects

Irradiation embrittlement and optimisation of annealing

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-12-31

This conference is composed of 30 papers grouped in 6 sessions related to the following themes: neutron irradiation effects in pressure vessel steels and weldments used in PWR, WWER and BWR nuclear plants; results from surveillance programmes (irradiation induced damage and annealing processes); studies on the influence of variations in irradiation conditions and mechanisms, and modelling; mitigation of irradiation effects, especially through thermal annealing; mechanical test procedures and specimen size effects.

The technology development for surveillance test of RPV materials 2

International Nuclear Information System (INIS)

Chang, Kee Ok; Lee, Sam Lai; Kim, Byoung Chul; Choi, Sun Pil; Choi, Kwen Jai

1998-12-01

Irradiation-induced changes in mechanical properties and magnetic parameters were measured and compared to explore possible correlations for Mn-Mo-Ni low alloy steel surveillance specimens which were irradiated to a neutron fluence of 2.4 x 10 1 9n/cm 2 (E≥1.0 MeV) in a typical pressurized water reactor environment at about 288 deg C. For mechanical property parameters, microvickers hardness, tensile and Charpy impact test were performed and Barkhausen Noise(BN) amplitude, coercivity, maximum induction were measured for magnetic parameters, respectively. Results of mechanical property measurements showed an increase in yield and tensil strength, microvickers hardness 41J indexed RT NDT and a decrease in upper shelf energy irrespective of base and weld metals. In the case of magnetic measurements, it is found that magnetic remanence, BN amplitude, BN energy have dropped significantly but coercivity has increased rapidly after irradiation. For isothermally heat treated condition of irradiated specimen, BN energy has increased while Vickers microhardness has decreased. Results of BNE and Vickers microhardness are reversed to the results on irradiated condition. All these consistent changes in magnetic parameter and Vickers microhardness measurement, which are thought to be resulted from the interaction between irradiation-induced defects and dislocation, and magnetic domain, respectively, show a possibility that magnetic measurement may be used to the evaluation of material degradation and recovery due to neutron irradiation and heat treatment, respectively, if a relevant large database is prepared. (author). 49 refs., 7 tabs., 23 figs

Effect of fast-neutron irradiation on plastic deformation of Type 304 stainless steel

International Nuclear Information System (INIS)

Yamada, H.

1978-01-01

Plastic deformation of EBR-II-irradiated Type 304 stainless steel was investigated by a stress-relaxation method. The stress-strain-rate relationships for the irradiated specimens at room temperature are concave upward, which are similar to those for the unirradiated specimens. However, concave downward behavior in the stress-strain-rate relationships were observed at much lower temperatures for the irradiated specimens in contrast to the unirradiated specimens. These results were analyzed succccessfully using Hart's mechanical equation-of-state concept. It was found that the hardness sigma*, which is the minimum stress necessary for the dislocation to overcome obstacles without thermal activation, increases linearly with fast-neutron fluence. This increase in sigma* is consistent with so-called ''irradiation hardening.'' In addition, resistance to dislocation glide, which is quantitatively measured in terms of sigma 0 , was observed to decrease linearly with fast-neutron fluence. The decrease in sigma 0 can be attributed to a decrease of solute drag due to irradiation-induced solute segregation

Temperature and dose dependencies of microstructure and hardness of neutron irradiated OFHC copper

International Nuclear Information System (INIS)

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

1995-01-01

Tensile specimens of pure oxygen free high conductivity (OFHC) copper were irradiated with fission neutrons between 320 and 723 K to fluences in the range 5x10 21 to 1.5x10 24 n/m 2 (E>1 MeV) with a flux of 2.5x10 17 n/m 2 s. Irradiated specimens were investigated by transmission electron microscopy (TEM) and quantitative determinations were made of defect clusters and cavities. The dose dependence of tensile properties of specimens irradiated at 320 K was determined at 295 K. Hardness measurements were made at 295 K on specimens irradiated at different temperatures and doses. Microstructures of tensile tested specimens were also investigated by TEM. Results show that the increase in cluster density and hardening nearly saturate at a dose of similar 0.3 dpa. Irradiations at 320 K cause a drastic decrease in the uniform elongation already at ∼ =0.1 dpa. It is suggested that the irradiation-induced increase in the initial yield stress and a drastic decrease in the ability of copper to deform plastically in a homogeneous fashion are caused by a substantial reduction in the ability of grown-in dislocations to act as efficient dislocation sources. ((orig.))

Tensile properties of helium-injected V-15Cr-5Ti after irradiation in EBR-II

International Nuclear Information System (INIS)

Grossbeck, M.L.; Horak, J.A.

1985-01-01

Miniature specimens of V-15Cr-5Ti were prepared in the annealed condition and with 10, 20, and 30% cold work. The annealed specimens were cyclotron injected with helium and irradiated in sodium in EBR-II. The cold-worked specimens were irradiated in EBR-II but not helium injected. The specimens were irradiated at 400, 525, 625, and 700 0 C and received a fluence of 4.1 to 5.5 x 10 26 neutrons/m 2 (E > 0.1 meV). Tensile testing revealed very significant embrittlement as a result of the neutron irradiation but a much smaller change, mostly at 400 0 C, resulting from helium injection. 5 references, 9 figures, 2 tables

Tensile properties of irradiated TZM and tungsten

International Nuclear Information System (INIS)

Steichen, J.M.

1975-04-01

The effect of neutron irradiation on the elevated temperature tensile properties of TZM and tungsten has been experimentally determined. Specimens were irradiated at a temperature of approximately 720 0 F to fluences of 0.4 and 0.9 x 10 22 n/cm 2 (E greater than 0.1 MeV). Test parameters for both control and irradiated specimens included strain rates from 3 x 10 -4 to 1 s -1 and temperatures from 72 to 1700 0 F. The results of these tests were correlated with a rate-temperature parameter (T ln A/epsilon) to provide a concise description of material behavior over the range of deformation conditions of this study. The yield strength of the subject materials was significantly increased by decreasing temperature, increasing strain rate, and increasing fluence. Ductility was significantly reduced at any temperature or strain rate by increasing fluence. Cleavage fractures occurred in both unirradiated and irradiated specimens when the yield strength was elevated to the effective cleavage stress by temperature and/or strain rate. Neutron irradiation for the conditions of this study increased the ductile-to-brittle transition temperature of tungsten by approximately 300 0 F and TZM by approximately 420 0 F. (U.S.)

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

International Nuclear Information System (INIS)

Zinkle, S.J.; Snead, L.L.; Rowcliffe, A.F.; Alexander, D.J.; Gibson, L.T.

1998-01-01

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 ∼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 ∼1.5--15 dpa and tested at 200 C

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.

Irradiation creep of dispersion strengthened copper alloy

International Nuclear Information System (INIS)

Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A.

1997-01-01

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 2 O 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 21 n/cm 2 (E>0.1 MeV), which corresponds to ∼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 ∼2 x 10 -9 s -1 . These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys

Comparison of irradiated and hydrogen implanted German RPV steels using PAS technique

Energy Technology Data Exchange (ETDEWEB)

Pecko, Stanislav, E-mail: stanislav.pecko@stuba.sk; Sojak, Stanislav; Slugeň, Vladimír

2015-12-15

Highlights: • German RPV steels were originally studied by positron annihilation spectroscopy. • Neutron irradiated and hydrogen ion implanted specimens were studied. • Both irradiation ways caused to increase of defect size. • We determined that the defect size was higher in implanted specimens. - Abstract: Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This spectroscopic method is a really effective tool for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to irradiation. German commercial reactor pressure vessel steels, originally from CARISMA program, were used in our study. The German experimental reactor VAK was selected as the proper irradiation facility in the 1980s. A specimen in as-received state and 2 different irradiated cuts from the same material were measured by PALS and size of defects with their intensity was indentified. Afterwards there was prepared an experiment with concern in simulation of neutron irradiation by hydrogen ion implantation on a linear accelerator with energy of 100 keV. Results are concerning on comparison between defects caused by neutron irradiation and hydrogen implantation. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to hydrogen ions implantation.

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

International Nuclear Information System (INIS)

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

2008-03-01

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

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

International Nuclear Information System (INIS)

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

2008-03-01

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

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...

Mechanical properties of Mo and TZM alloy neutron-irradiated at high temperatures

International Nuclear Information System (INIS)

Ueda, Kazukiyo; Satou, Manabu; Hasegawa, Akira; Abe, Katsunori

1997-01-01

This work reports the mechanical properties of irradiated molybdenum (Mo) and its alloy, TZM. Recrystallized and stress-relieved specimens were irradiated at five temperatures between 373 and 800degC in FFTF/MOTA to fluence levels of 6.8 to 34 dpa. Irradiation embrittlement and hardening were evaluated by three-point bend test and Vickers hardness test, respectively. Stress-relieved materials showed the enough ductility even after high fluence irradiation. The role of layered structure of stress-relieved specimen was discussed. (author)

Low upper-shelf toughness, high transition temperature test insert in HSST [Heavy Section Steel Technology] PTSE-2 [Pressurized Thermal Shock Experiment-2] vessel and wide plate test specimens: Final report

International Nuclear Information System (INIS)

Domian, H.A.

1987-02-01

A piece of A387, Grade 22 Class 2 (2-1/4 Cr - 1 Mo) steel plate specially heat treated to produce low upper-shelf (LUS) toughness and high transition temperature was installed in the side wall of Heavy Section Steel Technology (HHST) vessel V-8. This vessel is to be tested by the Oak Ridge National Laboratory (ORNL) in the Pressurized Thermal Shock Experiment-2 (PTSE-2) project of the HSST program. Comparable pieces of the plate were made into six wide plate specimens and other samples. These samples underwent tensile tests, Charpy tests, and J-integral tests. The results of these tests are given in this report

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.

Relationship between irradiation hardening and embrittlement of pressure vessel steels

International Nuclear Information System (INIS)

Odette, G.R.; Lombrozo, P.M.; Wullaert, R.A.

1984-01-01

Based on a large body of test and power reactor data, empirical relationships between irradiation strengthening and embrittlement are derived. It is shown that the Charpy V-notch (C /SUB v/ ) 41-J indexed transition temperature increases and the upper-shelf energy decreases systematically with increases in the yield stress. The transition temperature shifts are related to two mechanisms: increases in the maximum temperature of elastic-cleavage fracture, and decreases in the slope of the C, energy versus test temperature curve associated with reductions in the upper-shelf energy. The cleavage shift contribution, which is usually dominant, can be predicted from the initial temperature of fracture at general yield and the change in ambient temperature static yield stress. In developing this simplified cleavage fracture model, it is shown that: (a) yield stress changes are independent of temperature and strain rate; (b) the increase in yield stress with decreasing temperature is independent of the strain rate, irradiation, and metallurgical state; and (c) the microcleavage fracture stress is independent of irradiation and temperature. A semi-empirical procedure for estimating the shift contribution due to upper-shelf energy decreases and the total temperature shift at 41 J, based on the observation of an approximately constant temperature interval of the transition regime, is proposed, along with a method for forecasting the entire irradiated C, curve

Ring ductility of irradiated Inconel 706 and Nimonic PE16

International Nuclear Information System (INIS)

Huang, F.H.; Fish, R.L.

1984-01-01

The tensile ductility of fast neutron-irradiated, precipitation-hardened alloys Inconel 706 and Nimonic PE16 has been observed to be very low for certain test conditions. Explanations for the low ductility behavior have been sought by examination of broken tensile specimens with microscopy and other similar techniques. A ring compression test provides a method of evaluating the ductility of irradiated cladding specimens. Unlike the conventional uniaxial tensile testing in which the tensile specimen is deformed uniformly, the ring specimen is subjected to localized bending where the crack is initiated. The ductility can be estimated through an analysis of the bending of a ring in terms of strain hardening. Ring sections from irradiated, solution-treated Inconel 706 and Nimonic PE16 were compressed in the diametral direction to provide load-deflection records over a wide range of irradiation and test temperatures. Results showed that ductility in both alloys decreased with increasing test temperatures. The poorest ductility was exhibited at different irradiation temperatures in the two alloys - near 550 0 C for PE16 and 460 to 520 0 C for Inconel 706. The ring ductility data indicate that the grain boundary strength is a major factor in controlling the ductility of the PE16 alloy

Dose Distribution of Gamma Irradiators

International Nuclear Information System (INIS)

Park, Seung Woo; Shin, Sang Hun; Son, Ki Hong; Lee, Chang Yeol; Kim, Kum Bae; Jung, Hai Jo; Ji, Young Hoon

2010-01-01

Gamma irradiator using Cs-137 have been widely utilized to the irradiation of cell, blood, and animal, and the dose measurement and education. The Gamma cell 3000 Elan (Nordion International, Kanata, Ontario, Canada) irradiator was installed in 2003 with Cs-137 and dose rate of 3.2 Gy/min. And the BioBeam 8000 (Gamma-Service Medical GmbH, Leipzig, Germany) irradiator was installed in 2008 with Cs-137 and dose rate of 3.5 Gy/min. Our purpose was to evaluate the practical dosimetric problems associated with inhomogeneous dose distribution within the irradiated volume in open air state using glass dosimeter and Gafchromic EBT film dosimeter for routine Gamma irradiator dosimetry applications at the KIRAMS and the measurements were compared with each other. In addition, an user guideline for useful utilization of the device based on practical dosimetry will be prepared. The measurement results of uniformity of delivered dose within the device showed variation more than 14% between middle point and the lowest position at central axis. Therefore, to maintain dose variation within 10%, the criteria of useful dose distribution, for research radiation effects, the irradiated specimen located at central axis of the container should be placed within 30 mm from top and bottom surface, respectively. In addition, for measurements using the film, the variations of dose distribution were more then 50% for the case of less than 10 second irradiation, mostly within 20% for the case of more than 20 second irradiation, respectively. Therefore, the irradiation experiments using the BioBeam 8000 irradiator are recommended to be used for specimen required at least more than 20 second irradiation time.

Fundamental irradiation studies on vanadium alloys

International Nuclear Information System (INIS)

Loomis, B.A.; Garner, F.A.; Ermi, A.M.

1985-05-01

A joint experiment on the irradiation response of simple vanadium alloys has been initiated under the auspices of the DAFS and BES progams. Specimen fabrication is nearly complete and the alloys are expected to be irradiated in lithium in FFTF-MOTA Cycles 7 and 8

Fracture toughness of ferritic alloys irradiated at FFTF

International Nuclear Information System (INIS)

Huang, F.H.

1986-05-01

Ferritic compact tension specimens loaded in the Material Open Test Assembly (MOTA) for irradiation during FFTF Cycle 4 were tested at temperatures ranging from room temperature to 428/degree/C. The electrical potential single specimen method was used to measure the fracture toughness of the specimens. Results showed that the fracture toughness of both HT-9 and 9Cr-1Mo decreases with increasing test temperature and that the toughness of HT-9 was about 30% higher than that of 9Cr-1Mo. In addition, increasing irradiation temperature resulted in an increase in tearing modulus for both alloys. 4 refs., 5 figs., 1 tab

Comparison of damage microstructures in neutron-irradiated vanadium and iron

International Nuclear Information System (INIS)

Horton, L.L.; Farrell, K.

1983-01-01

The cavity morphology and dislocation loop geometry in bcc vanadium are compared with the previously reported observations for neutron-irradiated iron. The specimens were vanadium (V) with 100 wppM of interstitial impurities and vanadium with boron carbide additions (V-B 4 C) which were irradiated to approx. 1 dpa in the same Oak Ridge Research Reactor capsules as the iron specimens

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)

Effect of bonding and bakeout thermal cycles on the properties of copper alloys irradiated at 350 deg. C

International Nuclear Information System (INIS)

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

1997-02-01

Screening experiments were carried out to investigate the effect of bonding and bakeout thermal cycles on microstructure, mechanical properties and electrical resistivity of the oxide dispersion strengthened (GlidCop, CuAl-25) and the precipitation hardened (CuCrZr, CuNiBe) cooper alloys. Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing, and bonding thermal treatment followed by re-ageing and the reactor bakeout treatment at 350 deg. C for 100 h. Tensile specimens of CuAl-25 were given the heat treatment corresponding to the bonding thermal cycle. A number of heat treated specimens were neutron irradiated at 350 deg. C to a dose level of ∼ 0.3 dpa in the DR-3 reactor at Risoe. Both unirradiated and irradiated specimens with various heat treatments were tensile tested at 350 deg. C. The microstructure and electrical resistivity of these specimens were determined in the unirradiated as well as irradiated conditions. The post-deformation microstructure of the irradiated specimens was also investigated. The fracture surfaces of both unirradiated and irradiated specimens were examined. Results of these investigations are reported in the present report. The results are briefly discussed in terms of thermal and irradiation stability of precipitates and particles and irradiation-induced segregation, precipitation and recovery of dislocation microstructure. (au) 6 tabs., 24 ills., 9 refs

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

Science.gov (United States)

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

2012-03-01

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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.

Initial decay process of radicals induced in irradiated food

International Nuclear Information System (INIS)

Kaimori, Yoshihiko; Sakamoto, Yuki; Nakamura, Hideo; Ukai, Mitsuko; Kikuchi, Masahiro; Shimoyama, Yuhei; Kobayashi, Yasuhiko; Kameya, Hiromi

2011-01-01

In order to determine radial decay behaviors of γ-irradiated food, we analyzed radicals in the food using ESR. We detected the ESR signal of specimens just several minutes after irradiation. The singlet signal intensity at g=2.0, originated from organic free radicals was increased as followed by the increasing radiation dose. Singlet signal intensity that increased by γ-irradiation was decreased with time. The phenomena of decay of the ESR singlet signal showed two phase that are rapid decay and slow decay. It was suggested that those two phase decay is due to at least the two radical species. Also we concluded that after three hours of radiation treatment long life radical as ESR signal intensity was detected in irradiated specimen; black pepper, green coffee bean and ginseng, showed the same decay phenomena. But the signal intensity of irradiated black pepper was three times larger than that of irradiated green coffee bean and irradiated ginseng. (author)

Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

International Nuclear Information System (INIS)

Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio

2002-10-01

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

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)

Property change of advanced tungsten alloys due to neutron irradiation

International Nuclear Information System (INIS)

Fukuda, Makoto; Hasegawa, Akira; Tanno, Takashi; Nogami, Shuhei; Kurishita, Hiroaki

2013-01-01

This study investigates the effect of neutron irradiation on the functional properties of pure tungsten (W) and advanced tungsten alloys (e.g., lanthanum (La)-doped W, potassium (K)-doped W, and ultra-fine-grained (UFG) W–TiC alloys) tested in the Japan Materials Testing Reactor (JMTR) or experimental fast reactor Joyo. The irradiation temperature and damage were in the range 804–1073 K and 0.15–0.47 dpa, respectively. TEM images of all samples after 0.42 dpa irradiation at 1023 K showed voids, black dots, and dislocation loops, indicating that similar damage structures were formed in pure W, La-doped W, K-doped W, and UFG W–0.5 wt% TiC. The electrical resistivity of all specimens increased following neutron irradiation. Nearly identical electrical resistivity and irradiation hardening were observed in pure W, La-doped W, and K-doped W. The electrical resistivity of UFG W–TiC was higher than that of other specimens before and after irradiation, which may be attributed to its ultra-fine-grain structure, as well as the presence of impurities introduced during the alloying process. Compared to the other specimens, the UFG W–TiC was more resistant to irradiation hardening

Final report on graphite irradiation test OG-3

International Nuclear Information System (INIS)

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 3 JHA 2 N, P 3 JHAN, and ASI2-500. Data were obtained in the temperature range 823 0 K to 1673 0 K. The peak fast neutron fluence in the experiment was 3 x 10 25 n/m 3 (E greater than 29 fJ)/sub HTGR/; the total accumulated fluence exceeded 9 x 10 25 n/m 2 on some H-451 specimens and 6 x 10 25 n/m 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 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 25 n/m 2 at 1275 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 3 JHAN and P 3 JHA 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 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 saturation conductivity for the new temperature

Effects of postpolymerization microwave irradiation on provisional dental acrylics: physical and mechanical properties.

Science.gov (United States)

Ozkomur, Ahmet; Fortes, Carmen Beatriz Borges

2016-07-26

This study aimed to evaluate the effects of microwave irradiation on the physical and mechanical properties of poly(methyl methacrylate) (PMMA) provisional resins. Twenty bars and 20 disc-shaped specimens were fabricated for each selected provisional restorative material (Dencor and Duralay). Test groups were subjected to microwave irradiation (3 minutes at 600 W) after polymerization. Bar specimens were subjected to a flexural strength test. Disc-shaped specimens were used to evaluate microhardness. Backscattered Raman spectroscopy was employed for each group to define the degree of conversion of the monomer/polymer. The frequency bands corresponding to C = C and C = O groups were used to determine the conversion of methyl methacrylate (MMA) monomers into polymers. Glass transition temperature was determined using a differential scanning calorimeter. Microwave irradiation of both tested autopolymerizing PMMA provisional materials resulted in a statistically significant increase in microhardness, degree of conversion and glass transition temperature values. Also, the results demonstrated a significant increase in flexural strength after postpolymerization microwave irradiation for the Dencor specimens. It is concluded that mechanical and physical properties are positively influenced by microwave irradiation.

Radiation annealing mechanisms of low-alloy reactor pressure vessel steels dependent on irradiation temperature and neutron fluence

International Nuclear Information System (INIS)

Pachur, D.

1982-01-01

Heat treatment after irradiation of reactor pressure vessel steels showed annealing of irradiation embrittlement. Depending on the irradiation temperature, the embrittlement started to anneal at about 220 0 C and was completely annealed at 500 0 C with 4 h of annealing time. The annealing behavior was normally measured in terms of the Vickers hardness increase produced by irradiation relative to the initial hardness as a function of the annealing temperature. Annealing results of other mechanical properties correspond to hardness results. During annealing, various recovery mechanisms occur in different temperature ranges. These are characterized by activation energies from 1.5 to 2.1 eV. The individual mechanisms were determined by the different time dependencies at various temperatures. The relative contributions of the mechanisms showed a neutron fluence dependence, with the lower activation energy mechanisms being predominant at low fluence and vice versa. In the temperature range where partial annealing of a mechanism took place during irradiation, an increase in activation energy was observed. Trend curves for the increase in transition temperature with irradiation, for the relative increase of Vickers hardness and yield strength, and for the relative decrease of Charpy-V upper shelf energy are interpreted by the behavior of different mechanisms

Synchrotron radiation microprobe quantitative analysis method for biomedical specimens

International Nuclear Information System (INIS)

Xu Qing; Shao Hanru

1994-01-01

Relative changes of trace elemental content in biomedical specimens are obtained easily by means of synchrotron radiation X-ray fluorescence microprobe analysis (SXRFM). However, the accurate assignment of concentration on a g/g basis is difficult. Because it is necessary to know both the trace elemental content and the specimen mass in the irradiated volume simultaneously. the specimen mass is a function of the spatial position and can not be weighed. It is possible to measure the specimen mass indirectly by measuring the intensity of Compton scattered peak for normal XRF analysis using a X-ray tube with Mo anode, if the matrix was consisted of light elements and the specimen was a thin sample. The Compton peak is not presented in fluorescence spectrum for white light SXRFM analysis. The continuous background in the spectrum was resulted from the Compton scattering with a linear polarization X-ray source. Biomedical specimens for SXRFM analysis, for example biological section and human hair, are always a thin sample for high energy X-ray, and they consist of H,C,N and O etc. light elements, which implies a linear relationship between the specimen mass and the Compton scattering background in the high energy region of spectrum. By this way , it is possible to carry out measurement of concentration for SXRFM analysis

Surface, structural and tensile properties of proton beam irradiated zirconium

Energy Technology Data Exchange (ETDEWEB)

Rafique, Mohsin; Chae, San; Kim, Yong-Soo, E-mail: yongskim@hanyang.ac.kr

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 × 10{sup 13} to 1 × 10{sup 16} protons/cm{sup 2}. 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.

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.

The results of the surveillance specimen program performed in the RPVs NPP V-2 in Jaslovske Bohunice

Energy Technology Data Exchange (ETDEWEB)

Kupca, L; Beno, P [Vyskumny Ustav Jadrovych Elektrarni, Trnava (Slovakia); Cepeek, S [Atomova Elektraren Bohunice, Jaslovske Bohunice (Slovakia); Tomasich, M [Slovak Nuclear Society, Bratislava (Slovakia)

1994-12-31

After a description of the mechanical and chemical characteristics of the materials (steels, welded joints) used in the pressure vessels of the WWER-440 V-213 type, the present status of the Bohunice NPP Unit 3 and 4 pressure vessel embrittlement assessment programme is presented: neutron flux monitoring and calculations, detector accuracy, irradiation temperature monitoring, reactor core fuel loading calculation, materials, number and types of surveillance specimens, specimen testing. Results are given for 5 years of irradiation: mechanical properties, transition temperatures, lifetime evaluation. 4 refs., 13 figs., 6 tabs.

Extra spots in the electron diffraction patterns of neutron irradiated zirconium and its alloys

International Nuclear Information System (INIS)

Madden, P.K.

1977-01-01

Specimens of neutron irradiated zirconium and its alloys were examined in the transmission electron microscope. Groups of extra spots, often exhibiting four-fold symmetry, were observed in thin foil electron diffraction patterns of these specimens. The 'extra-spot' structure, like the expected black-dot/small scale dislocation loop neutron irradiated damage, is approximately 100 A in size. Its nature is uncertain. It may be related to irradiation damage or to some artefact introduced during specimen preparation. If it is the latter, then published irradiation damage defect size distributions and determined irradiation growth strains of other investigators, may require modification. The present inconclusive results indicate that extra-spot structure is likely to consist of oxide particles, but may correspond to hydride precipitation or decoration effects, or even, to electron beam effects. (author)

EFFECT OF GAMMA RAY IRRADIATION ON INTERLAMINAR SHEAR STRENGTH OF GLASS FIBER REINFORCED PLASTICS AT 77 K

International Nuclear Information System (INIS)

Nishimura, A.; Nishijima, S.; Izumi, Y.

2008-01-01

It is known that an organic material is damaged by gamma ray irradiation, and the strength after irradiation has dependence on the gamma ray dose. These issues are important not only to make global understanding of electric insulating performance of glass fiber reinforced plastics (GFRP) under irradiation condition but also to develop new insulation materials. This paper presents the dependence of fracture mode and interlaminar shear strength (ILSS) on the material and the gamma ray irradiation effect on the fracture mode and the ILSS. 6 mm radius loading nose and supports were used to prompt ILS fracture for a short beam test. A 2.5 mm thick small specimen machined out of a 13 mm thick G-10CR GFRP plate (sliced specimen) showed lower ILSS and translaminar shear (TLS) fracture, although the same size specimen prepared from a 2.5 mm G-10CR GFRP plate (non-sliced specimen) showed ILS fracture and the higher ILSS. Both type of specimens showed the degradation of ILSS after gamma ray irradiation. The fracture mode of the non-sliced specimen changed from ILS to TLS fracture and no bending fracture was observed. The resistance to shear deformation of glass cloth/epoxy laminate structure would be damaged by the irradiation

Borosilicate glass for gamma irradiation fields

Science.gov (United States)

Baydogan, N.; Tugrul, A. B.

2012-11-01

Four different types of silicate glass specimens were irradiated with gamma radiation using a Co-60 radioisotope. Glass specimens, with four different chemical compositions, were exposed to neutron and mixed neutron/gamma doses in the central thimble and tangential beam tube of the nuclear research reactor. Optical variations were determined in accordance with standardisation concept. Changes in the direct solar absorbance (αe) of borosilicate glass were examined using the increase in gamma absorbed dose, and results were compared with the changes in the direct solar absorbance of the three different type silicate glass specimens. Solar absorption decreased due to decrease of penetration with absorbed dose. αe of borosilicate increased considerably when compared with other glass types. Changes in optical density were evaluated as an approach to create dose estimation. Mixed/thermal neutron irradiation on glass caused to increse αe.

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.

Argon laser irradiation of the otolithic organ

Energy Technology Data Exchange (ETDEWEB)

Okuno, T.; Nomura, Y.; Young, Y.H.; Hara, M. (Univ. of Tokyo (Japan))

1990-12-01

An argon laser was used to irradiate the otolithic organs of guinea pigs and cynomolgus monkeys. After stapedectomy, the argon laser (1.5 W x 0.5 sec/shot) irradiated the utricle or saccule without touching the sensory organs. The stapes was replaced over the oval window after irradiation. The animals used for acute observation were killed immediately for morphologic studies; those used for long-term observation were kept alive for 2, 4, or 10 weeks. Acute observation revealed that sensory and supporting cells were elevated from the basement membrane only in the irradiated area. No rupture of the membranous labyrinth was observed. Long-term observation revealed that the otolith of the macula utriculi had disappeared in 2-week specimens. The entire macula utricili had disappeared in 10-week specimens. No morphologic changes were observed in cochlea, semicircular canals, or membranous labyrinth. The saccule showed similar changes.

Determination of irradiation temperature using SiC temperature monitors

International Nuclear Information System (INIS)

Maruyama, Tadashi; Onose, Shoji

1999-01-01

This paper describes a method for detecting the change in length of SiC temperature monitors and a discussion is made on the relationship between irradiation temperature and the recovery in length of SiC temperature monitors. The SiC specimens were irradiated in the experimental fast reactor JOYO' at the irradiation temperatures around 417 to 645degC (design temperature). The change in length of irradiated specimens was detected using a dilatometer with SiO 2 glass push rod in an infrared image furnace. The temperature at which recovery in macroscopic length begins was obtained from the annealing intersection temperature. The results of measurements indicated that a difference between annealing intersection temperature and the design temperature sometimes reached well over ±100degC. A calibration method to obtain accurate irradiation temperature was presented and compared with the design temperature. (author)

A new shape specimens determined the J1c value of nuclear pressure vessel steel

International Nuclear Information System (INIS)

Xu, W.Q.

1989-01-01

The J integral has two basic definitions, a two-dimensional energy line integral definition and an energy rate definition. The line integral definition cannot be used for this experimental determination. The energy rate definition can be used but the procedure is somewhat laborious. Methods were developed for more easily determining J by approximation formulas. The first of these were where J could be estimated with reasonable accuracy for a deeply cracked bend-type specimen. This method is slightly inaccurate. This paper is concerned with a new shape specimen. It is called the W-shape specimen. The W-shape specimens are smaller volume than the compact specimens. It is convenient to operate the W-shape specimens in hot cell. It can be put into surveillance capsules and can also do specimen irradiation in engineering test reactor

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

Application of micromechanical models of ductile fracture initiation to reactor pressure vessel materials

International Nuclear Information System (INIS)

Chaouadi, R.; Walle, E. van; Fabry, A.; Velde, J. van de; Meester, P. de

1996-01-01

The aim of the current study is the application of local micromechanical models to predict crack initiation in ductile materials. Two reactor pressure vessel materials have been selected for this study: JRQ IAEA monitor base metal (A533B Cl.1) and Doel-IV weld material. Charpy impact tests have been performed in both un-irradiated and irradiated conditions. In addition to standard tensile tests, notched tensile specimens have been tested. The upper shelf energy of the weld material remains almost un-affected by irradiation, whereas a decrease of 20% is detected for the base metal. Accordingly, the tensile properties of the weld material do not reveal a clear irradiation effect on the yield and ultimate stresses, this in contrast to the base material flow properties. The tensile tests have been analyzed in terms of micromechanical models. A good correlation is found between the standard tests and the micromechanical models, that are able to predict the ductile damage evolution in these materials. Additional information on the ductility behavior of these materials is revealed by this micromechanical analysis

Vacancy defects in electron irradiated RPV steels studied by positron lifetime

Energy Technology Data Exchange (ETDEWEB)

Moser, P; Li, X H [CEA Centre d` Etudes de Grenoble, 38 (France). Dept. de Recherche Fondamentale sur la Matiere Condensee; Akamatsu, M; Van Duysen, J C [Electricite de France (EDF), 77 - Ecuelles (France)

1994-12-31

Specimens of French RPV (reactor pressure vessels) steels at different rates of segregation have been irradiated at 150 and 288 deg C with 3 MeV electrons (irradiation dose: 4*10{sup 19} e-/cm{sup 2}). Vacancy defects are studied by positron lifetime measurements before and after irradiation and at each step of isochronal annealing. After 150 deg C irradiation, a recovery step is observed in both specimens, for annealing treatments in the range 220-370 deg C and is attributed to the dissociation of vacancy-impurity complexes. The size of vacancy clusters never overcome 10 empty atomic volumes. If ``fresh`` dislocations are created just before irradiation, big vacancy clusters could be formed. After 288 deg C irradiation, small vacancy cluster of 4-10 empty atomic volumes are observed. (authors). 3 figs., 7 refs.

Infrared absorption studies of the annealing of irradiated diamonds

International Nuclear Information System (INIS)

Woods, G.S.

1984-01-01

Natural (types Ia and IIa) and synthetic (type Ib) diamonds have been irradiated with energetic electrons and neutrons and then heated at temperatures up to 1400 deg C. Attendant changes in the infrared absorption spectra, especially above the Raman frequency (1332 cm -1 ), have been monitored. The most prominent absorption to develop in the infrared region proper, on annealing both type Ia and type Ib specimens, whether electron- or neutron-irradiated is the H1a line at 1450 cm -1 . Measurements taken of neutron-irradiated type Ia specimens show that the strength of this line is specimen-dependent, and that it is a linear function of radiation dose. Isochronal annealing studies show that the onset of the line occurs during heating at 250 deg C for type Ia specimens and at 650 deg C for type Ib specimens. The absorption begins to weaken during heating at 1100 deg C, but it is very persistent, surviving an anneal of 4 hours at 1400 deg C, albeit with diminished intensity. Three other weaker lines at 1438, 1358 and 1355 cm -1 develop with the 1450 cm -1 line, but differ from it and from each other in subsequent annealing behaviour. Other lines were observed; these are reported and discussed. (author)

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.

A study of defect cluster formation in vanadium by heavy ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Sekimura, Naoto; Shirao, Yasuyuki; Morishita, Kazunori [Tokyo Univ. (Japan)

1996-10-01

Formation of defect clusters in thin foils of vanadium was investigated by heavy ion irradiation. In the very thin region of the specimens less than 20 nm, vacancy clusters were formed under gold ion irradiation, while very few clusters were detected in the specimens irradiated with 200 and 400 keV self-ions up to 1 x 10{sup 16} ions/m{sup 2}. The density of vacancy clusters were found to be strongly dependent on ion energy. Only above the critical value of kinetic energy transfer density in vanadium, vacancy clusters are considered to be formed in the cascade damage from which interstitials can escape to the specimen surface in the very thin region. (author)

Microstructural comparison of HT-9 irradiated in HFIR and EBR-II

International Nuclear Information System (INIS)

Gelles, D.S.

1985-05-01

A series of specimens of HT-9 heat 91354 have been examined following irradiation in HFIR to 39 dpa at 300, 400, 500 and 600 0 C and following irradiation in EBR-II to 29 dpa at 390 and 500 0 C. HFIR irradiation was found to have promoted helium bubble formation at all temperatures and voids at 400 0 C. Cavitation had not been observed at lower fluence, nor was it found in EBR-II irradiated specimens. The onset of void swelling in HFIR is attributed to helium generation. The observations provide an explanation for saturation of ductile-brittle transition temperature shifts with increasing fluence

Definition of the minimum longitude of insert in the rebuilding of Charpy test tubes for surveillance and life extension of vessels in Mexico

International Nuclear Information System (INIS)

Romero C, J.; Hernandez C, R.; Rocamontes A, M.

2011-11-01

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 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)

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.

Cytologic studies on irradiated gestric cancer cells

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Isono, S; Takeda, T; Amakasu, H; Asakawa, H; Yamada, S [Miyagi Prefectural Adult Disease Center, Natori (Japan)

1981-06-01

The smears of the biopsy and resected specimens obtained from 74 cases of irradiated gastric cancer were cytologically analyzed for effects of irradiation. Irradiation increased the amount of both necrotic materials and neutrophils in the smears. Cancer cells were decreased in number almost in inverse proportion to irradiation dose. Clusters of cancer cells shrank in size and cells were less stratified after irradiation. Irradiated cytoplasms were swollen, vacuolated and stained abnormally. Irradiation with less than 3,000 rads gave rise to swelling of cytoplasms in almost all cases. Nuclei became enlarged, multiple, pyknotic and/or stained pale after irradiation. Nuclear swelling was more remarkable in cancer cells of differentiated adenocarcinomas.

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Effects of helium on ductile brittle transition behavior of reduced activation ferritic steels after high concentration he implantation at high temperature

Energy Technology Data Exchange (ETDEWEB)

Hasegawa, A.; Ejiri, M.; Nogami, S.; Ishiga, M.; Abe, K. [Tohoku Univ., Dept. of Quantum Science and Energy Engr, Sendai (Japan); Kasada, R.; Kimura, A. [Kyoto Univ., Institute of Advanced Energy (Japan); Jitsukawa, S. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan)

2007-07-01

Full text of publication follows: Influence of Helium (He) on fracture behavior of reduced activation ferritic/martensitic steels including Oxide Dispersion Strengthening (ODS) steels and F82H were examined. To study the He effects on fracture behavior of these steels after He bubble formation conditions, higher concentration of He implantation at around 550 C were performed and examined the relationship between microstructure evolution and fracture behavior of the steels. The 1.5CVN mini size Charpy specimens were used to evaluate impact test behavior. Reduced activation ferritic ODS steels, 9Cr-ODS and 12Cr-ODS steels were examine. F82H was also examined as reference material. Helium implantation was performed by a cyclotron of Tohoku University with a beam of 50 MeV {alpha}-particles at temperature around 550 C. A tandem-type energy degrader system was used to implant He into the specimen from the irradiated surface to the range of 50 MeV {alpha}-particles, that was about 380 {mu}m in iron. Implanted He concentration were about 1000 appm. Charpy impact test was performed using a instrumented impact test apparatus in Oarai branch of IMR, Tohoku University. Analyses of absorbed energy change and fracture surface were carried out. Vickers hardness test was also carried out on He implanted area of the 1.5CVN specimen to estimate irradiation hardening. Microstructural observation was performed by TEM. In the case of F82H, DBTT increased by the 1000 appm He implantation condition was about 80 C and grain boundary fracture surface was only observed in the He implanted area of all the ruptured specimens in brittle manner. On the other hand, DBTT shift and fracture mode change of He implanted 9Cr-ODS steel was not observed after He implantation. Microstructural observation showed that He bubble formation on the lath boundaries and grain boundaries were significant in F82H, but the bubble segregation on grain boundary in ODS steel was not apparent. The bubble formation

Temperature dependence of the damage microstructures in neutron-irradiated vanadium

International Nuclear Information System (INIS)

Horton, L.L.; Farrell, K.

1983-01-01

Vanadium and vanadium with boron carbide additions (V-B 4 C) were irradiated to approx. 1 dpa in the Oak Ridge Research Reactor at controlled temperatures ranging from 455 to 925 K. The V-B 4 C alloy was enriched in 10 B, which produced approx. 3900 at. ppM helium. In the vanadium specimens, the dislocation microstructures varied from clusters of small ( . The V-B 4 C specimens contained only tangled dislocation segments. Cavities were observed in all specimens. The cavity concentration decrease and the average diameter increased with increasing irradiation temperature. At 725 K, the maximum swelling was observed in both the vanadium (0.1%) and V-B 4 C (1.4%). At comparable temperatures the cavities in the V-B 4 C specimens were smaller and more numerous than those in the vanadium specimens. Helium bubbles were found on the grain boundaries in all of the V-B 4 specimens

Influence of non-metallic inclusions on fatigue strength of high manganese steel