Heavy-Section Steel Irradiation Program on irradiation effects in light-water reactor pressure vessel materials

International Nuclear Information System (INIS)

Nanstad, R.K.; Corwin, W.R.; Alexander, D.J.; Haggag, F.M.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.; Stoller, R.E.

1995-01-01

The safety of commercial light-water nuclear plants is highly dependent on the structural integrity of the reactor pressure vessel (RPV). In the absence of radiation damage to the RPV, fracture of the vessel is difficult to postulate. Exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory, sponsored by the US Nuclear Regulatory Commission (USNRC), is assessing the effects of neutron irradiation on RPV material behavior, especially fracture toughness. The results of these and other studies are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety. In assessing the effects of irradiation, prototypic RPV materials are characterized in the unirradiated condition and exposed to radiation under varying conditions. Mechanical property tests are conducted to provide data which can be used in the development of guidelines for structural integrity evaluations, while metallurgical examinations and mechanistic modeling are performed to improve understanding of the mechanisms responsible for embrittlement. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. This irradiation-induced degradation of the materials can be mitigated by thermal annealing, i.e., heating the RPV to a temperature above that of normal operation. Thus, thermal annealing and evaluation of reirradiation behavior are major tasks of the HSSI Program. This paper describes the HSSI Program activities by summarizing some past and recent results, as well as current and planned studies. 30 refs., 8 figs., 1 tab

Heavy-section steel technology program. Semiannual progress report, October 1994--March 1995

International Nuclear Information System (INIS)

Pennell, W.E.

1996-07-01

The Heavy-Section Steel Technology (HSST) Program is conducted for the Nuclear Regulatory Commission (NRC) by Oak Ridge National Laboratory (ORNL). The program focus is on the development and validation of technology for the assessment of fracture-prevention margins in commercial nuclear reactor pressure vessels. The HSST Program is organized in seven tasks: (1) program management (2) constraint effects analytical development and validation, (3) evaluation of cladding effects, (4) ductile-to-cleavage fracture-mode conversion, (5) fracture analysis methods development and applications, (6) material property data and test methods, and (7) integration of results. The program tasks have been structured to place emphasis on the resolution of fracture issues with near-term licensing significance. Resources to execute the research tasks are drawn from ORNL with subcontract support from universities and other research laboratories. Close contact is maintained with the sister Heavy-Section Steel Irradiation Program at ORNL and with related research programs both in the United States and abroad. This report provides an overview of principal developments in each of the seven program tasks from October 1994-March 1995

Heavy-section steel technology program. Semiannual progress report, October 1994--March 1995

Energy Technology Data Exchange (ETDEWEB)

Pennell, W.E.

1996-07-01

The Heavy-Section Steel Technology (HSST) Program is conducted for the Nuclear Regulatory Commission (NRC) by Oak Ridge National Laboratory (ORNL). The program focus is on the development and validation of technology for the assessment of fracture-prevention margins in commercial nuclear reactor pressure vessels. The HSST Program is organized in seven tasks: (1) program management (2) constraint effects analytical development and validation, (3) evaluation of cladding effects, (4) ductile-to-cleavage fracture-mode conversion, (5) fracture analysis methods development and applications, (6) material property data and test methods, and (7) integration of results. The program tasks have been structured to place emphasis on the resolution of fracture issues with near-term licensing significance. Resources to execute the research tasks are drawn from ORNL with subcontract support from universities and other research laboratories. Close contact is maintained with the sister Heavy-Section Steel Irradiation Program at ORNL and with related research programs both in the United States and abroad. This report provides an overview of principal developments in each of the seven program tasks from October 1994-March 1995.

Heavy-Section Steel Technology program fracture issues

International Nuclear Information System (INIS)

Pennell, W.E.

1989-10-01

Large scale fracture mechanics tests have resulted in the identification of a number of fracture technology issues. Identification of additional issues has come from the reactor vessel materials irradiation test program and from reactor operating experience. This paper provides a review of fracture issues with an emphasis on their potential impact on a reactor vessel pressurized thermal shock (PTS) analysis. Mixed mode crack propagation emerges as a major issue, due in large measure to the poor performance of existing models for the prediction of ductile tearing. Rectification of ductile tearing technology deficiencies may require extending the technology to include a more complete treatment of stress state and loading history effects. The effect of cladding on vessel fracture remains uncertain to the point that it is not possible to determine at this time if the net effect will be positive or negative. Enhanced fracture toughness for shallow flaws has been demonstrated for low strength structural steels. Demonstration of a similar effect in reactor pressure vessel steels could have a significant beneficial effect on the probabilistic analysis of reactor vessel fracture. Further development of existing fracture mechanics models and concepts is required to meet the special requirements for fracture evaluation of circumferential flaws in the welds of ring forged vessels. Fracture technology advances required to address the issues discussed in this paper are the major objective for the ongoing Heavy Section Steel Technology (HSST) program at ORNL. 24 refs., 18 figs

Heavy-section steel technology program: Fracture issues

International Nuclear Information System (INIS)

Pennell, W.E.

1992-01-01

Large-scale fracture mechanics tests have resulted in the identification of a number of fracture technology issues. Identification of additional issues has come from the reactor vessel materials irradiation test program and from reactor operating experience. This paper provides a review of fracture issues with an emphasis on their potential impact on a reactor vessel pressurized thermal shock (PTS) analysis. Mixed mode crack propagation emerges as a major issue, due in large measure to the poor performance of existing models for the prediction of ductile tearing. Rectification of ductile tearing technology deficiencies may require extending the technology to include a more complete treatment of stress state and loading history effects. The effect of cladding on vessel fracture remains uncertain to the point that it is not possible to determine at this time if the net effect will be positive or negative. Enhanced fracture toughness for shallow flaws has been demonstrated for low-strength structural steels. Demonstration of a similar effect in reactor pressure vessel steels could have a significant beneficial effect on the probabilistic analysis of reactor vessel fracture. Further development of existing fracture mechanics models and concepts is required to meet the special requirements for fracture evaluation of circumferential flaws in the welds of ring-forged vessels. Fracture technology advances required to address the issues discussed in this paper are the major objective for the ongoing Heavy Section Steel Technology (HSST) program at ORNL

Heavy-Section Steel Technology Program Semiannual progress report, April--September 1993. Volume 10, No. 2

Energy Technology Data Exchange (ETDEWEB)

Pennell, W.E. [Oak Ridge National Lab., TN (United States)

1995-05-01

The Heavy-Section Steel Technology (HSST) Program is conducted for the Nuclear Regulatory Commission by Oak Ridge National Laboratory (ORNL). The program focuses on the development and validation of technology for the assessment of fracture-prevention margins in commercial nuclear reactor pressure vessels. The HSST Program is organized in 12 tasks: Program management, fracture methodology and analysis, material characterizations and properties, special technical assistance, fracture analysis computer programs, cleavage-crack initiation, cladding evaluations, pressurized-thermal-shock technology, analysis methods validation, fracture evaluation tests, warm prestressing, and biaxial loading effects on fracture toughness. The program tasks have been structured to emphasize the resolution fracture issues with near-term licensing significance. Resources to execute the research tasks are drawn from ORNL with subcontract support from universities and other research laboratories. Close contact is maintained with the sister Heavy-Section Steel Irradiation Program at ORNL and with related research programs both in the United States and abroad. This report provide s an overview of principal developments in each of the 12 program tasks from April -- September 1993.

Heavy-section steel technology program. Semiannual progress report, April--September 1995 Vol. 12, No. 2

Energy Technology Data Exchange (ETDEWEB)

Pennell, W.E.

1997-01-01

The Heavy-Section Steel Technology (HSST) Program is conducted for the Nuclear Regulatory Commission by Oak Ridge National Laboratory (ORNL). The program focus is on the development and validation of technology for the assessment of fracture-prevention margins in commercial nuclear reactor pressure vessels. The HSST Program is organized in seven tasks: (1) program management, (2) constraint effects analytical development and validation, (3) evaluation of cladding effects, (4) ductile-to-cleavage fracture-mode conversion, (5) fracture analysis methods development and applications, (6) material property data and test methods, and (7) integration of results. The program tasks have been structured to place emphasis on the resolution of fracture issues with near-term licensing significance. Resources to execute the research tasks are drawn from ORNL with subcontract support from universities and other research laboratories. Close contact is maintained with the sister Heavy-Section Steel Irradiation Program at ORNL and with related research programs both in the United States and abroad. This report provides an overview of principal developments in each of the seven program tasks from April 1995 to September 1995.

Heavy-Section Steel Technology Program Semiannual progress report, April--September 1993. Volume 10, No. 2

International Nuclear Information System (INIS)

Pennell, W.E.

1995-05-01

The Heavy-Section Steel Technology (HSST) Program is conducted for the Nuclear Regulatory Commission by Oak Ridge National Laboratory (ORNL). The program focuses on the development and validation of technology for the assessment of fracture-prevention margins in commercial nuclear reactor pressure vessels. The HSST Program is organized in 12 tasks: Program management, fracture methodology and analysis, material characterizations and properties, special technical assistance, fracture analysis computer programs, cleavage-crack initiation, cladding evaluations, pressurized-thermal-shock technology, analysis methods validation, fracture evaluation tests, warm prestressing, and biaxial loading effects on fracture toughness. The program tasks have been structured to emphasize the resolution fracture issues with near-term licensing significance. Resources to execute the research tasks are drawn from ORNL with subcontract support from universities and other research laboratories. Close contact is maintained with the sister Heavy-Section Steel Irradiation Program at ORNL and with related research programs both in the United States and abroad. This report provide s an overview of principal developments in each of the 12 program tasks from April -- September 1993

Study of irradiation damage structures in austenitic stainless steels

Energy Technology Data Exchange (ETDEWEB)

Hamada, Shozo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-08-01

The irradiation damage microstructures in austenitic stainless steels, which have been proposed to be a candidate of structural materials of a fusion reactor, under ions and neutrons irradiation have been studied. In ion irradiation experiments, cross-sectional observation of the depth distribution of damage formed due to ion irradiation became available. Comparison and discussion between experimental results with TEM and the calculated ones in the depth profiles of irradiation damage microstructures. Further, dual-phase stainless steels, consisted of ferritic/austenitic phases, showed irradiation-induced/enhanced precipitation during ion irradiation. High Flux Isotope Reactor with high neutron fluxes was employed in neutron-irradiation experiments. Swelling of 316 steel showed irradiation temperature dependence and this had strong correlation with phase instability under heavy damage level. Swelling resistance of Ti-modified austenitic stainless steel, which has good swelling resistance, decreased during high damage level. This might be caused by the instability of Ti-carbide particles. The preparation method to reduce higher radioactivity of neutron-irradiated TEM specimen was developed. (author). 176 refs.

Study of irradiation damage structures in austenitic stainless steels

International Nuclear Information System (INIS)

Hamada, Shozo

1997-08-01

The irradiation damage microstructures in austenitic stainless steels, which have been proposed to be a candidate of structural materials of a fusion reactor, under ions and neutrons irradiation have been studied. In ion irradiation experiments, cross-sectional observation of the depth distribution of damage formed due to ion irradiation became available. Comparison and discussion between experimental results with TEM and the calculated ones in the depth profiles of irradiation damage microstructures. Further, dual-phase stainless steels, consisted of ferritic/austenitic phases, showed irradiation-induced/enhanced precipitation during ion irradiation. High Flux Isotope Reactor with high neutron fluxes was employed in neutron-irradiation experiments. Swelling of 316 steel showed irradiation temperature dependence and this had strong correlation with phase instability under heavy damage level. Swelling resistance of Ti-modified austenitic stainless steel, which has good swelling resistance, decreased during high damage level. This might be caused by the instability of Ti-carbide particles. The preparation method to reduce higher radioactivity of neutron-irradiated TEM specimen was developed. (author). 176 refs

Heavy-Section Steel Technology Program: Semiannual progress report for April--September 1994. Volume 11, Number 2

Energy Technology Data Exchange (ETDEWEB)

Pennell, W.E. [Oak Ridge National Lab., TN (United States)

1996-04-01

The Heavy-Section Steel Technology (HSST) Program is conducted for the Nuclear Regulatory Commission (NRC) by Oak Ridge National Laboratory (ORNL). The program focus is on the development and validation of technology for the assessment of fracture-prevention margins in commercial nuclear reactor pressure vessels. The HSST Program is organized in seven tasks: (1) program management, (2) constraint effects analytical development and validation, (3) evaluation of cladding effects, (4) ductile-to-cleavage fracture-mode conversion, (5) fracture analysis methods development and applications, (6) material property data and test methods, and (7) integration of results. The program tasks have been structured to place emphasis on the resolution fracture issues with near-term licensing significance. Resources to execute the research tasks are drawn from ORNL with subcontract support from universities and other research laboratories. Close contact is maintained with the sister Heavy-Section Steel Irradiation (HSSI) Program at ORNL and with related research programs both in the US and abroad. This report provides an overview of principal developments in each of the seven program tasks from April 1994 to September 1994.

Heavy-Section Steel Technology Program: Semiannual progress report for April--September 1994. Volume 11, Number 2

International Nuclear Information System (INIS)

Pennell, W.E.

1996-04-01

The Heavy-Section Steel Technology (HSST) Program is conducted for the Nuclear Regulatory Commission (NRC) by Oak Ridge National Laboratory (ORNL). The program focus is on the development and validation of technology for the assessment of fracture-prevention margins in commercial nuclear reactor pressure vessels. The HSST Program is organized in seven tasks: (1) program management, (2) constraint effects analytical development and validation, (3) evaluation of cladding effects, (4) ductile-to-cleavage fracture-mode conversion, (5) fracture analysis methods development and applications, (6) material property data and test methods, and (7) integration of results. The program tasks have been structured to place emphasis on the resolution fracture issues with near-term licensing significance. Resources to execute the research tasks are drawn from ORNL with subcontract support from universities and other research laboratories. Close contact is maintained with the sister Heavy-Section Steel Irradiation (HSSI) Program at ORNL and with related research programs both in the US and abroad. This report provides an overview of principal developments in each of the seven program tasks from April 1994 to September 1994

Heavy-section steel technology program: Semiannual progress report, October 1993--March 1994. Volume 11, No. 1

Energy Technology Data Exchange (ETDEWEB)

Pennell, W.E. [Oak Ridge National Lab., TN (United States)

1995-11-01

The Heavy-Section Steel Technology (HSST) Program is conducted for the US Nuclear Regulatory Commission (NRC) by Oak Ridge National Laboratory (ORNL). The Program focus is on the development and validation of technology for the assessment Of fracture-prevention margins in commercial nuclear reactor pressure vessels. The HSST Program is organized in seven tasks: (1) program management (2) constraint effects analytical development and validation, (3) evaluation of cladding effects, (4) ductile to cleavage fracture mode conversion, (5) fracture analysis methods development and applications, (6) material Property data and test methods, and (7) integration of results into a state-of-the-art methodology. The program tasks have been structured to place emphasis on the resolution fracture issues with near-term licensing significance. Resources to execute the research tasks are drawn from ORNL with subcontract support from universities and other research laboratories. Close contact is maintained with the sister Heavy-Section Steel Irradiation Program at ORNL and with related research programs both in the United States and abroad. This report provides an overview of principal developments in each of the seven program tasks from October 1993--March 1994.

Heavy-section steel technology program: Semiannual progress report, October 1993--March 1994. Volume 11, No. 1

International Nuclear Information System (INIS)

Pennell, W.E.

1995-11-01

The Heavy-Section Steel Technology (HSST) Program is conducted for the US Nuclear Regulatory Commission (NRC) by Oak Ridge National Laboratory (ORNL). The Program focus is on the development and validation of technology for the assessment Of fracture-prevention margins in commercial nuclear reactor pressure vessels. The HSST Program is organized in seven tasks: (1) program management (2) constraint effects analytical development and validation, (3) evaluation of cladding effects, (4) ductile to cleavage fracture mode conversion, (5) fracture analysis methods development and applications, (6) material Property data and test methods, and (7) integration of results into a state-of-the-art methodology. The program tasks have been structured to place emphasis on the resolution fracture issues with near-term licensing significance. Resources to execute the research tasks are drawn from ORNL with subcontract support from universities and other research laboratories. Close contact is maintained with the sister Heavy-Section Steel Irradiation Program at ORNL and with related research programs both in the United States and abroad. This report provides an overview of principal developments in each of the seven program tasks from October 1993--March 1994

Historical summary of the heavy-section steel technology program and some related activities in light-water reactor pressure vessel safety research

International Nuclear Information System (INIS)

Whitman, G.D.

1986-03-01

The accomplishments of the Heavy-Section Steel Technology Program and other programs having a close relationship to the development of information used in the assessment of light-water reactor pressure vessel integrity are reviewed. The early Pressure Vessel Research Committee planning, the principals contributing to program formulation, the role of the US Atomic Energy Commission, and the developments under the US Nuclear Regulatory Commission sponsorship are identified. The need for major research and development accomplishments in fracture mechanics, heavy-section steel procurement, materials properties, irradiation effects, fatigue crack growth, and structural testing are summarized. The impact of program results on regulatory issues and the development of data used in the preparation of codes, standards, and guides are discussed. Continuing activities and recommendations for future research and development in support of pressure vessel integrity assessments are presented

Transmission electron microscopy study of the heavy-ion-irradiation-induced changes in the nanostructure of oxide dispersion strengthened steels

Science.gov (United States)

Rogozhkin, S. V.; Bogachev, A. A.; Orlov, N. N.; Korchuganova, O. A.; Nikitin, A. A.; Zaluzhnyi, A. G.; Kozodaev, M. A.; Kulevoy, T. V.; Kuibeda, R. P.; Fedin, P. A.; Chalykh, B. B.; Lindau, R.; Hoffman, Ya.; Möslang, A.; Vladimirov, P.; Klimenkov, M.

2017-07-01

Transmission electron microscopy was used to study the effect of heavy-ion irradiation on the structure and the phase state of three oxide dispersion strengthened (ODS) steels: ODS Eurofer, ODS 13.5Cr, and ODS 13.5Cr-0.3Ti (wt %). Samples were irradiated with iron and titanium ions to fluences of 1015 and 3 × 1015 cm-2 at 300, 573, and 773 K. The study of the region of maximum radiation damage shows that irradiation increases the number density of oxide particles in all samples. The fraction of fine inclusions increases in the particle size distribution. This effect is most pronounced in the ODS 13.5Cr steel irradiated with titanium ions at 300 K to a fluence of 3 × 1015 cm-2. It is demonstrated that oxide inclusions in ODS 13.5Cr-0.3Ti and ODS 13.5Cr steels are more stable upon irradiation at 573 and 773 K than upon irradiation at 300 K.

Effect of Al and N on the toughness of heavy section steel plates

International Nuclear Information System (INIS)

Kikutake, Tetsuo; Tokunaga, Yoshikuni; Nakao, Hitoji; Ito, Kametaro; Takaishi, Shogo.

1988-01-01

The effect of Al and N on the notch toughness and tensile strength of heavy section pressure vessel steel plates is investigated. Notch toughness of steel A533B (Mn-Mo-Ni), which has mixed microstructure of ferrite and bainite, is drastically changed by the ratio of sol.N/sol.Al. With metallurgical observations, it is revealed that AlN morphology is influenced by the ratio of sol.N/sol.Al through the level of solute Al(C Al ). At the heat treatment of heavy section steel plate, AlN shows OSTWALD ripening and its speed depends upon C Al . When Al is added (Al ≥ 0.010%) in steel and sol.N/sol.Al ≤ 0.5, C Al remains low. This prevents AlN ripening, and brings fine austenite grain size and high toughness. On the other hand, when sol.N/sol.Al Al becomes high and this gives poor toughness through coarse AlN precipitates and coarse austenite grain. Therefore, controll of sol.N/sol.Al over 0.5 is favorable to keep high toughness in A533B steel. In steel A387-22 (Cr-Mo) which has full bainitic microstructure, too fine austenite grain brings about poor hardenability, and polygonal ferrite, which brings about both poor strength and tughness, appears in microstructure. Then sol.N/sol.Al < 0.5 is better to give high hardenability in steel A387-22. (author)

Irradiation effect of different heavy ions and track section on the silkworm Bombyx mori

Energy Technology Data Exchange (ETDEWEB)

Tu Zhenli E-mail: tu514@yahoo.co.jp; Kobayashi, Yasuhiko; Kiguchi, Kenji; Watanabe, Hiroshi

2003-05-01

In order to compare the irradiation effects of different ions, wandering larvae were whole-body exposed or locally irradiated with 50-MeV {sup 4}He{sup 2+}, 220-MeV {sup 12}C{sup 5+}, and 350-MeV {sup 20}Ne{sup 8+} ions, respectively. For the whole-body-exposed individuals, the survival rates at the cocooning, pupation, and emergence stages all decreased as dose increased, and a range-dependent difference was clearly observed. For local irradiation of ovaries, irradiation effects depend very strongly on the projectile range. In the case of local irradiation of dermal cells by different track sections of heavy ions, the closer the target was to the high-LET section of the track, the more pronounced were the radiation effects. These results indicated that by selectively using ion species and adjusting the irradiation depth to the target, heavy-ion radiosurgery on particular tissues or organs of small experimental animals can be performed more accurately.

Irradiation effect of different heavy ions and track section on the silkworm Bombyx mori

International Nuclear Information System (INIS)

Tu Zhenli; Kobayashi, Yasuhiko; Kiguchi, Kenji; Watanabe, Hiroshi

2003-01-01

In order to compare the irradiation effects of different ions, wandering larvae were whole-body exposed or locally irradiated with 50-MeV 4 He 2+ , 220-MeV 12 C 5+ , and 350-MeV 20 Ne 8+ ions, respectively. For the whole-body-exposed individuals, the survival rates at the cocooning, pupation, and emergence stages all decreased as dose increased, and a range-dependent difference was clearly observed. For local irradiation of ovaries, irradiation effects depend very strongly on the projectile range. In the case of local irradiation of dermal cells by different track sections of heavy ions, the closer the target was to the high-LET section of the track, the more pronounced were the radiation effects. These results indicated that by selectively using ion species and adjusting the irradiation depth to the target, heavy-ion radiosurgery on particular tissues or organs of small experimental animals can be performed more accurately

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

High dose radiation damage in nuclear energy structural materials investigated by heavy ion irradiation simulation

International Nuclear Information System (INIS)

Zheng Yongnan; Xu Yongjun; Yuan Daqing

2014-01-01

Structural materials in ITER, ADS and fast reactor suffer high dose irradiations of neutrons and/or protons, that leads to severe displacement damage up to lOO dpa per year. Investigation of radiation damage induced by such a high dose irradiation has attracted great attention along with the development of nuclear energy facilities of new generation. However, it is deeply hampered for the lacking of high dose neutron and proton sources. Irradiation simulation of heavy ions produced by accelerators opens up an effective way for laboratory investigation of high dose irradiation induced radiation damage encountered in the ITER, ADS, etc. Radiation damage is caused mainly by atomic displacement in materials. The displacement rate of heavy ions is about lO 3 ∼10 7 orders higher than those of neutrons and protons. High displacement rate of heavy ions significantly reduces the irradiation time. The heavy ion irradiation simulation technique (HIIS) technique has been developed at China Institute of Atomic Energy and a series of the HIIS experiments have been performed to investigate radiation damage in stainless steels, tungsten and tantalum at irradiation temperatures from room temperature to 800 ℃ and in the irradiation dose region up to 100 dpa. The experimental results show that he radiation swelling peak for the modified stainless steel appears in the temperature region around 580 ℃ and the radiation damage is more sensitive to the temperature, the size of the radiation induced vacancy cluster or void increase with the increasing of the irradiation dose, and among the three materials the home-made modified stainless steel has the best radiation resistant property. (authors)

Hardening of ODS ferritic steels under irradiation with high-energy heavy ions

Science.gov (United States)

Ding, Z. N.; Zhang, C. H.; Yang, Y. T.; Song, Y.; Kimura, A.; Jang, J.

2017-09-01

Influence of the nanoscale oxide particles on mechanical properties and irradiation resistance of oxide-dispersion-strengthened (ODS) ferritic steels is of critical importance for the use of the material in fuel cladding or blanket components in advanced nuclear reactors. In the present work, impact of structures of oxide dispersoids on the irradiation hardening of ODS ferritic steels was studied. Specimens of three high-Cr ODS ferritic steels containing oxide dispersoids with different number density and average size were irradiated with high-energy Ni ions at about -50 °C. The energy of the incident Ni ions was varied from 12.73 MeV to 357.86 MeV by using an energy degrader at the terminal so that a plateau of atomic displacement damage (∼0.8 dpa) was produced from the near surface to a depth of 24 μm in the specimens. A nanoindentor (in constant stiffness mode with a diamond Berkovich indenter) and a Vickers micro-hardness tester were used to measure the hardeness of the specimens. The Nix-Gao model taking account of the indentation size effect (ISE) was used to fit the hardness data. It is observed that the soft substrate effect (SSE) can be diminished substantially in the irradiated specimens due to the thick damaged regions produced by the Ni ions. A linear correlation between the nano-hardeness and the micro-hardness was found. It is observed that a higher number density of oxide dispersoids with a smaller average diameter corresponds to an increased resistance to irradiation hardening, which can be ascribed to the increased sink strength of oxides/matrix interfaces to point defects. The rate equation approach and the conventional hardening model were used to analyze the influence of defect clusters on irradiation hardening in ODS ferritic steels. The numerical estimates show that the hardening caused by the interstitial type dislocation loops follows a similar trend with the experiment data.

Heavy Section Steel Technology Program. Part II. Intermediate vessel testing

International Nuclear Information System (INIS)

Whitman, G.D.

1975-01-01

The testing of the intermediate pressure vessels is a major activity under the Heavy Section Steel Technology Program. A primary objective of these tests is to develop or verify methods of fracture prediction, through the testing of selected structures and materials, in order that a valid basis can be established for evaluating the serviceability and safety of light-water reactor pressure vessels. These vessel tests were planned with sufficiently specific objectives that substantial quantitative weight could be given to the results. Each set of testing conditions was chosen so as to provide specific data by which analytical methods of predicting flaw growth, and in some cases crack arrest, could be evaluated. Every practical effort was made to assure that results would be relevant to some aspect of real reactor pressure vessel performance through careful control of material properties, selection of test temperatures, and design of prepared flaws. 5 references

Constraint effects in heavy-section steels

International Nuclear Information System (INIS)

Bass, B.R.; Shum, D.K.M.; Keeney-Walker, J.; Theiss, T.J.

1993-01-01

A focal point of the Nuclear Regulatory Commission-funded Heavy-Section Steel Technology (HSST) Program is the development of technology required for accurate assessment of fracture-prevention margins in commercial nuclear reactor pressure vessels (RPVs). In a series of investigations, the HSST Program is seeking to obtain an improved understanding of the relationships governing transfer of fracture toughness data from small-scale specimens to large-scale structures. This paper describes two analytical approaches to the transferability issues that are being evaluated in the HSST Program. One is a continuum correlative methodology based on two-parameter descriptions (K-T or J-Q) of the near crack-tip fields that incorporate effects of the higher-order T-stress for linear-elastic fracture mechanics conditions or the Q-stress for more general elastic-plastic fracture mechanics conditions. The second approach utilizes a micromechanical predictive methodology that relates cleavage crack initiation to the attainment of a critical volume enclosed within a selected maximum principal stress contour surrounding the crack tip. In preliminary evaluations, these methodologies were applied to experimental data taken from several intermediate- and large-scale testing programs. Results and conclusions from these applications are discussed in the paper. Applications of the methodologies to analytical studies concerning biaxial stress effects on fracture toughness and safety margin assessments of an RPV subjected to pressurized-thermal-shock transient loadings are also presented. While these fracture methodologies appear to show promise in being able to differentiate among crack-tip constraint levels, numerous issues were identified in the HSST studies that require further investigation. Recommendations are given concerning future work intended to resolve several of these issues. 42 refs., 26 figs., 1 tab

Defects and their inspectability by UT in current heavy section steels for nuclear power plant

International Nuclear Information System (INIS)

Onodera, S.; Ohkubo, Y.; Takeya, M.; Wataya, M.

1983-01-01

The ultrasonic examination (UT, hereinafter) techniques and their equipment have been improved in search of the defects in steels and structures for nuclear power plant components, while the acceptance standards of the defects became continually more stringent in a ''sword and armour'' race. Consequently, the steel making technique had to respond in minimizing the possible defects in steels with successful results in the past two decades. The conventional UT procedures cover basically the following categories of function. 1) Detection and location of defects. 2) Sizing of defects. 3) Characterization of defects. 4) Structure and residual stress effects in ultrasonic field. With proper considerations to the configuration of the steels under examination, the inspectability of the possible defects is further to be optimized. However, the final evaluation has often to be left to the discretion of a competent NDE engineer, well experienced in UT and knowledgeable in steel making. It is therefore the intention of the present paper to review the states-of-the-art of the defects found in the current heavy section steels for primary and secondary components of nuclear power plant, manufactured by the authors' plant. Typical defects, detectable size of them and inspectability of them are discussed

Determination of cross sections of nuclear reactions to use Al as monitoring foil in heavy ion irradiation with 20Ne projectile

International Nuclear Information System (INIS)

Chowdhury, D.P.; Datta, J.; Guin, R.; Verma, R.

2009-01-01

The beam current is generally accurately measured using monitoring foils during the irradiation of thick samples by high energy ion beams. The cross sections of many nuclear reactions induced by light particles are available in literature for use as monitoring foil. However, such cross sections of heavy ion induced reactions are not reported much for their use in applied works. We have determined cross sections of two nuclear reactions, 27 Al ( 20 Ne,2p2n) 43 Sc and 27 Al ( 20 Ne, 2pn) 44m Sc, to use Al as monitoring foil for the irradiation with 20 Ne heavy ion beam. (author)

Irradiation embrittlement of pressure vessel steels

International Nuclear Information System (INIS)

Brumovsky, M.; Vacek, M.

1975-01-01

A Standard Research Programme on Irradiation Embrittlement of Pressure Vessel Steels was approved by the Coordinating Meeting on the 12th May 1972 at the Working Group on Engineering Aspects of Irradiation Embrittlement of Pressure Vessel Steels. This Working Group was set up by the International Atomic Energy Agency in Vienna. Seven countries with their research institutes agreed on doing irradiation experiments according to the approved programme on steel A533 B from the U.S. HSST Programme. The Czechoslovak contribution covering tensile and impact testing of non-irradiated steel and steel irradiated at 280degC to 1.3 x 10 23 n/m 2 (E above 1 MeV) is presented in this report. As an additional part the same set of experiments was carried out on two additional steels - A 542 and A 543, made in SKODA Works for comparison of their irradiation embrittlement and hardening with A533 B steel. (author)

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)

Atom probe tomography of the evolution of the nanostructure of oxide dispersion strengthened steels under ion irradiation

Science.gov (United States)

Orlov, N. N.; Rogozhkin, S. V.; Bogachev, A. A.; Korchuganova, O. A.; Nikitin, A. A.; Zaluzhnyi, A. G.; Kozodaev, M. A.; Kulevoy, T. V.; Kuibeda, R. P.; Fedin, P. A.; Chalykh, B. B.; Lindau, R.; Hoffmann, Ya.; Möslang, A.; Vladimirov, P.

2017-09-01

The atom probe tomography of the nanostructure evolution in ODS1 Eurofer, ODS 13.5Cr, and ODS 13.5Cr-0.3Ti steels under heavy ion irradiation at 300 and 573 K is performed. The samples were irradiated by 5.6 MeV Fe2+ ions and 4.8 MeV Ti2+ ions to a fluence of 1015 cm-2. It is shown that the number of nanoclusters increases by a factor of 2-3 after irradiation. The chemical composition of the clusters in the steels changes after irradiation at 300 K, whereas the chemical composition of the clusters in the 13.5Cr-0.3Ti ODS steel remains the same after irradiation at 573 K.

Heavy-Section Steel Technology Program: Recent developments in crack initiation and arrest research

International Nuclear Information System (INIS)

Pennell, W.E.

1991-01-01

Technology for the analysis of crack initiation and arrest is central to the reactor pressure vessel fracture-margin-assessment process. Regulatory procedures for nuclear plants utilize this technology to assure the retention of adequate fracture-prevention margins throughout the plant operating license period. As nuclear plants age and regulatory procedures dictate that fracture-margin assessments be performed, interest in the fracture-mechanics technology incorporated into those procedures has heightened. This has led to proposals from a number of sources for development and refinement of the underlying crack-initiation and arrest-analysis technology. This paper presents an overview of ongoing Heavy-Section Steel Technology (HSST) Program research aimed at refining the fracture toughness data used in the analysis of fracture margins under pressurized-thermal-shock loading conditions. 33 refs., 13 figs

Austin: austenitic steel irradiation E 145-02 Irradiation Report

International Nuclear Information System (INIS)

Genet, F.; Konrad, J.

1987-01-01

Safety measures for nuclear reactors require that the energy which might be liberated in a reactor core during an accident should be contained within the reactor pressure vessel, even after very long irradiation periods. Hence the need to know the mechanical properties at high deformation velocity of structure materials that have received irradiation damage due to their utilization. The stainless steels used in the structures of reactors undergo damage by both thermal and fast neutrons, causing important changes in the mechanical properties of these materials. Various austenitic steels available as structural materials were irradiated or are under irradiation in various reactors in order to study the evolution of the mechanical properties at high deformation velocity as a function of the irradiation damage rate. The experiment called AUSTIN (AUstenitic STeel IrradiatioN) 02 was performed by the JRC Petten Establishment on behalf of Ispra in support of the reactor safety programme

Atom Probe Tomography Characterization of the Solute Distributions in a Neutron-Irradiated and Annealed Pressure Vessel Steel Weld

Energy Technology Data Exchange (ETDEWEB)

Miller, M.K.

2001-01-30

A combined atom probe tomography and atom probe field ion microscopy study has been performed on a submerged arc weld irradiated to high fluence in the Heavy-Section Steel irradiation (HSSI) fifth irradiation series (Weld 73W). The composition of this weld is Fe - 0.27 at. % Cu, 1.58% Mn, 0.57% Ni, 0.34% MO, 0.27% Cr, 0.58% Si, 0.003% V, 0.45% C, 0.009% P, and 0.009% S. The material was examined after five conditions: after a typical stress relief treatment of 40 h at 607 C, after neutron irradiation to a fluence of 2 x 10{sup 23} n m{sup {minus}2} (E > 1 MeV), and after irradiation and isothermal anneals of 0.5, 1, and 168 h at 454 C. This report describes the matrix composition and the size, composition, and number density of the ultrafine copper-enriched precipitates that formed under neutron irradiation and the change in these parameters with post-irradiation annealing treatments.

Neutron irradiation creep in stainless steel alloys

Energy Technology Data Exchange (ETDEWEB)

Schuele, Wolfgang (Commission of the European Union, Institute for Advanced Materials, I-21020 Ispra (Vatican City State, Holy See) (Italy)); Hausen, Hermann (Commission of the European Union, Institute for Advanced Materials, I-21020 Ispra (Vatican City State, Holy See) (Italy))

1994-09-01

Irradiation creep elongations were measured in the HFR at Petten on AMCR steels, on 316 CE-reference steels, and on US-316 and US-PCA steels varying the irradiation temperature between 300 C and 500 C and the stress between 25 and 300 MPa. At the beginning of an irradiation a type of primary'' creep stage is observed for doses up to 3-5 dpa after which dose the secondary'' creep stage begins. The primary'' creep strain decreases in cold-worked steel materials with decreasing stress and decreasing irradiation temperature achieving also negative creep strains depending also on the pre-treatment of the materials. These primary'' creep strains are mainly attributed to volume changes due to the formation of radiation-induced phases, e.g. to the formation of [alpha]-ferrite below about 400 C and of carbides below about 700 C, and not to irradiation creep. The secondary'' creep stage is found for doses larger than 3 to 5 dpa and is attributed mainly to irradiation creep. The irradiation creep rate is almost independent of the irradiation temperature (Q[sub irr]=0.132 eV) and linearly dependent on the stress. The total creep elongations normalized to about 8 dpa are equal for almost every type of steel irradiated in the HFR at Petten or in ORR or in EBR II. The negative creep elongations are more pronounced in PCA- and in AMCR-steels and for this reason the total creep elongation is slightly smaller at 8 dpa for these two steels than for the other steels. ((orig.))

Concomitant formation of different nature clusters and hardening in reactor pressure vessel steels irradiated by heavy ions

Energy Technology Data Exchange (ETDEWEB)

Fujii, K., E-mail: fujiik@inss.co.jp [Institute of Nuclear Safety System, Inc., Mihama 919-1205 (Japan); Fukuya, K. [Institute of Nuclear Safety System, Inc., Mihama 919-1205 (Japan); Hojo, T. [Japan Nuclear Energy Safety Organization, Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

2013-11-15

Specimens of A533B steels containing 0.04, 0.09 and 0.21 wt%Cu were irradiated at 290 °C to 3 dpa with 3 MeV Fe ions and subjected to atom probe analyses, transmission electron microscopy observations and hardness measurements. The atom probe analysis results showed that two types of solute clusters were formed: Cu-enriched clusters containing Mn, Ni and Si atoms as irradiation-enhanced solute atom clusters and Mn/Ni/Si-enriched clusters as irradiation-induced solute atom clusters. Both cluster types occurred in the highest Cu-content steel and the ratio of Mn/Ni/Si-enriched clusters to Cu-enriched clusters increased with irradiation doses. It was confirmed that the cluster formation was a key factor in the microstructure evolution until the high dose irradiation was reached even in the low Cu content steels though the dislocation loops with much lower density than that of the clusters were observed as matrix damage. The difference in the hardening efficiency due to the difference in the nature of the clusters was small. The irradiation-induced clustering of undersized Si atoms suggested that a clustering driving force other than vacancy-driven diffusion, probably an interstitial mechanism, may become important at higher dose rates.

Concomitant formation of different nature clusters and hardening in reactor pressure vessel steels irradiated by heavy ions

International Nuclear Information System (INIS)

Fujii, K.; Fukuya, K.; Hojo, T.

2013-01-01

Specimens of A533B steels containing 0.04, 0.09 and 0.21 wt%Cu were irradiated at 290 °C to 3 dpa with 3 MeV Fe ions and subjected to atom probe analyses, transmission electron microscopy observations and hardness measurements. The atom probe analysis results showed that two types of solute clusters were formed: Cu-enriched clusters containing Mn, Ni and Si atoms as irradiation-enhanced solute atom clusters and Mn/Ni/Si-enriched clusters as irradiation-induced solute atom clusters. Both cluster types occurred in the highest Cu-content steel and the ratio of Mn/Ni/Si-enriched clusters to Cu-enriched clusters increased with irradiation doses. It was confirmed that the cluster formation was a key factor in the microstructure evolution until the high dose irradiation was reached even in the low Cu content steels though the dislocation loops with much lower density than that of the clusters were observed as matrix damage. The difference in the hardening efficiency due to the difference in the nature of the clusters was small. The irradiation-induced clustering of undersized Si atoms suggested that a clustering driving force other than vacancy-driven diffusion, probably an interstitial mechanism, may become important at higher dose rates

Tensile behavior of irradiated manganese-stabilized stainless steel

Energy Technology Data Exchange (ETDEWEB)

Klueh, R.L. [Oak Ridge National Lab., TN (United States)

1996-10-01

Tensile tests were conducted on seven experimental, high-manganese austenitic stainless steels after irradiation up to 44 dpa in the FFTF. An Fe-20Mn-12Cr-0.25C base composition was used, to which various combinations of Ti, W, V, B, and P were added to improve strength. Nominal amounts added were 0.1% Ti, 1% W, 0.1% V, 0.005% B, and 0.03% P. Irradiation was carried out at 420, 520, and 600{degrees}C on the steels in the solution-annealed and 20% cold-worked conditions. Tensile tests were conducted at the irradiation temperature. Results were compared with type 316 SS. Neutron irradiation hardened all of the solution-annealed steels at 420, 520, and 600{degrees}C, as measured by the increase in yield stress and ultimate tensile strength. The steel to which all five elements were added to the base composition showed the least amount of hardening. It also showed a smaller loss of ductility (uniform and total elongation) than the other steels. The total and uniform elongations of this steel after irradiation at 420{degrees}C was over four times that of the other manganese-stabilized steels and 316 SS. There was much less difference in strength and ductility at the two higher irradiation temperatures, where there was considerably less hardening, and thus, less loss of ductility. In the cold-worked condition, hardening occured only after irradiation at 420{degrees}C, and there was much less difference in the properties of the steels after irradiation. At the 420{degrees}C irradiation temperature, most of the manganese-stabilized steels maintained more ductility than the 316 SS. After irradiation at 420{degrees}C, the temperature of maximum hardening, the steel to which all five of the elements were added had the best uniform elongation.

Mechanical properties of irradiated 9Cr-2WVTa steel

International Nuclear Information System (INIS)

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

1998-01-01

An Fe-9Cr-2W-0.25V-0.07Ta-0.1C (9Cr-2WVTa) steel has excellent strength and impact toughness before and after irradiation in the Fast Flux Test Facility and the High Flux Reactor (HFR). The ductile-brittle transition temperature (DBTT) increased only 32 C after 28 dpa at 365 C in FFTF, compared to a shift of ∼60 C for a 9Cr-2WV steel--the same as the 9Cr-2WVTa steel but without tantalum. This difference occurred despite the two steels having similar tensile but without tantalum. This difference occurred despite the two steels having similar tensile properties before and after irradiation. The 9Cr-2WVTa steel has a smaller prior-austenite grain size, but otherwise microstructures are similar before irradiation and show similar changes during irradiation. The irradiation behavior of the 9Cr-2WVTa steel differs from the 9Cr-2WV steel and other similar steels in two ways: (1) the shift in DBTT of the 9Cr-2WVTa steel irradiated in FFTF does not saturate with fluence by ∼28 dpa, whereas for the 9Cr-2WV steel and most similar steels, saturation occurs at <10 dpa, and (2) the shift in DBTT for 9Cr-2WVTa steel irradiated in FFTF and HFR increased with irradiation temperature, whereas it decreased for the 9Cr-2WV steel, as it does for most similar steels. The improved properties of the 9Cr-2WVTa steel and the differences with other steels were attributed to tantalum in solution

Some advances in fracture studies under the heavy-section steel technology program

International Nuclear Information System (INIS)

Pugh, C.E.; Corwin, W.R.; Bryan, R.H.; Bass, B.R.

1985-01-01

Recent results are summarized from HSST studies in three major areas that relate to assessing nuclear reactor pressure vessel integrity under pressurized-thermal-shock (PTS) conditions: irradiation effects on the fracture properties of stainless steel cladding, crack run-arrest behavior under nonisothermal conditions, and fracture behavior of a thick-wall vessel under combined thermal and pressure loadings

Hydrogen retention in ion irradiated steels

International Nuclear Information System (INIS)

Hunn, J.D.; Lewis, M.B.; Lee, E.H.

1998-01-01

In the future 1--5 MW Spallation Neutron Source, target radiation damage will be accompanied by high levels of hydrogen and helium transmutation products. The authors have recently carried out investigations using simultaneous Fe/He,H multiple-ion implantations into 316 LN stainless steel between 50 and 350 C to simulate the type of radiation damage expected in spallation neutron sources. Hydrogen and helium were injected at appropriate energy and rate, while displacement damage was introduced by nuclear stopping of 3.5 MeV Fe + , 1 microm below the surface. Nanoindentation measurements showed a cumulative increase in hardness as a result of hydrogen and helium injection over and above the hardness increase due to the displacement damage alone. TEM investigation indicated the presence of small bubbles of the injected gases in the irradiated area. In the current experiment, the retention of hydrogen in irradiated steel was studied in order to better understand its contribution to the observed hardening. To achieve this, the deuterium isotope ( 2 H) was injected in place of natural hydrogen ( 1 H) during the implantation. Trapped deuterium was then profiled, at room temperature, using the high cross-section nuclear resonance reaction with 3 He. Results showed a surprisingly high concentration of deuterium to be retained in the irradiated steel at low temperature, especially in the presence of helium. There is indication that hydrogen retention at spallation neutron source relevant target temperatures may reach as high as 10%

Hardness of AISI type 410 martensitic steels after high temperature irradiation via nanoindentation

Science.gov (United States)

Waseem, Owais Ahmed; Jeong, Jong-Ryul; Park, Byong-Guk; Maeng, Cheol-Soo; Lee, Myoung-Goo; Ryu, Ho Jin

2017-11-01

The hardness of irradiated AISI type 410 martensitic steel, which is utilized in structural and magnetic components of nuclear power plants, is investigated in this study. Proton irradiation of AISI type 410 martensitic steel samples was carried out by exposing the samples to 3 MeV protons up to a 1.0 × 1017 p/cm2 fluence level at a representative nuclear reactor coolant temperature of 350 °C. The assessment of deleterious effects of irradiation on the micro-structure and mechanical behavior of the AISI type 410 martensitic steel samples via transmission electron microscopy-energy dispersive spectroscopy and cross-sectional nano-indentation showed no significant variation in the microscopic or mechanical characteristics. These results ensure the integrity of the structural and magnetic components of nuclear reactors made of AISI type 410 martensitic steel under high-temperature irradiation damage levels up to approximately 5.2 × 10-3 dpa.

Heavy-Section Steel Technology Program: Recent developments in crack initiation and arrest research

International Nuclear Information System (INIS)

Pennell, W.E.

1991-01-01

Technology for the analysis of crack initiation and arrest is central to the reactor pressure vessel fracture-margin-assessment process. Regulatory procedures for nuclear plants utilize this technology to assure the retention of adequate fracture-prevention margins throughout the plant operating license period. As nuclear plants age and regulatory procedures dictate that fracture-margin assessments be performed, interest in the fracture-mechanics technology incorporated into those procedures has heightened. This has led to proposals from a number of sources for development and refinement of the underlying crack-initiation and arrest-analysis technology. An important element of the Heavy-Section Steel Technology (HSST) Program is devoted to the investigation and evaluation of these proposals. This paper presents the technological bases and fracture-margin assessment objectives for some of the recently proposed crack-initiation and arrest-technology developments. The HSST Program approach to the evaluation of the proposals is described and the results and conclusions obtained to date are presented

Weldability of neutron-irradiated stainless steel and nickel-base alloy

International Nuclear Information System (INIS)

Koyabu, Ken; Asano, Kyoichi; Takahashi, Hidenori; Sakamoto, Hiroshi; Kawano, Shohei; Nakamura, Tomomi; Hashimoto, Tsuneyuki; Koshiishi, Masato; Kato, Takahiko; Katsura, Ryoei; Nishimura, Seiji

2000-01-01

Degradation of of weldability caused by helium, which is generated by nuclear transmutation irradiated material, is an important issue to be addressed in planning of proactive maintenance of light water reactor core internal components. In this work, the weldability of neutron.irradiated stainless steel and nickel-base alloy, which are major constituting materials for components, was practically evaluated. The weldability was first examined by TIG welding in relation to the weld heat input and helium content using various specimens (made of SUS304 and SUS316L) sampled from reactor internal components. The specimens were neutron irradiated in a boiling water reactor to fluences from 4 x 10 24 to 1.4 x 10 26 n/ m 2 (E> l MeV ), and resulting helium generation ranged from 0.1 to 103 appm. The weld defects were characterized by dye penetrant test and cross-sectional metallography. The weldability of neutron-irradiated stainless steel was shown to be better at lower weld heat input and lower helium content. To evaluate mechanical properties of welded joints, thick plates (20 mm) specimens of SUS304 and Alloy 600 were prepared and irradiated in Japan Material Test Reactor (JMTR). The helium content of the specimens was controlled to range from 0.11 to 1.34 appm selected to determine threshold helium content to weld successfully. The welded joints had multiple passes by TIG welding process at 10 and 20 kJ/cm heat input. The welded joints of thick plate were characterized by dye penetrant test, cross-sectional metallography, tensile test, side bend test and root bend test. It was shown that irradiated stainless steel containing below 0.14 appm of helium could be welded with conventional TIG welding process (heat input below 20 kJ/cm). Nickel-base alloy, which contained as much helium as stainless steel could be welded successfully, could also be welded with conventional TIG welding process, These results served as basis to evaluate the applicability of repair welding to

Stress corrosion cracking of highly irradiated 316 stainless steels

Energy Technology Data Exchange (ETDEWEB)

Nakano, Morihito; Fukuya, Koji; Fujii, Katsuhiko; Nakajima, Nobuo; Furutani, Gen [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

Mechanical property tests, grain boundary (GB) composition analysis and slow strain rate test (SSRT) in simulated PWR primary water changing dissolved hydrogen (DH) and dissolved oxygen (DO) content were carried out on cold-worked (CW) 316 stainless steels which were irradiated to 1-8x10{sup 26} n/m{sup 2} (E>0.1 MeV) in a Japanese PWR in order to evaluate irradiation-assisted stress corrosion cracking (IASCC) susceptibility. Highly irradiated stainless steels were susceptible to intergranular stress corrosion cracking (IGSCC) in both hydrogenated water and oxygenated water and to intergranular cracking in inert gas atmosphere. IASCC susceptibility increased with increasing DH content (0-45 ccH{sub 2}/kgH{sub 2}O). Hydrogen content of the section containing fracture surface was higher than that of the section far from fracture surface. These results suggest that hydrogen would have an important role for IASCC. While mechanical property was saturated, GB segregation and IASCC susceptibility increased with an increase in fluence, suggesting that GB segregation would have a dominant role for an increase in IASCC susceptibility at this high fluence region. (author)

Effect of activation cross section uncertainties in the assessment of primary damage for MFE/IFE low-activation steels irradiated in IFMIF

International Nuclear Information System (INIS)

Cabellos, O.; Sanz, J.; Garcia-Herranz, N.; Otero, B.

2009-01-01

The present study is mainly aimed to provide the primary damage (displacements per atom, generation of solid transmutants and gas production rates) of structural materials irradiated in the high and medium flux test modules of the International Fusion Materials Irradiation Facility (IFMIF). We have investigated if the change of the composition during the irradiation time has effect on the prediction of the atomic displacements. The effect of the activation cross section uncertainties in the assessment of both solid transmutants and hydrogen and helium production is also analyzed. The results are provided element-by-element, so that the primary damage of any material irradiated in such neutron environments can be easily assessed; in this paper, we have predicted the primary damage of the low activation steel Eurofer.

Effect of activation cross section uncertainties in the assessment of primary damage for MFE/IFE low-activation steels irradiated in IFMIF

Energy Technology Data Exchange (ETDEWEB)

Cabellos, O. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid (UPM), C/Jose Gutierrez Abascal, n2, 28006 Madrid (Spain); Dept. de Ingenieria Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain)], E-mail: cabellos@din.upm.es; Sanz, J. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid (UPM), C/Jose Gutierrez Abascal, n2, 28006 Madrid (Spain); Dept. de Ingenieria Energetica, Universidad Nacional de Educacion a Distancia, 28045 Madrid (Spain); Garcia-Herranz, N. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid (UPM), C/Jose Gutierrez Abascal, n2, 28006 Madrid (Spain); Dept. de Ingenieria Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain); Otero, B. [Dept. de Ingenieria Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain)

2009-04-30

The present study is mainly aimed to provide the primary damage (displacements per atom, generation of solid transmutants and gas production rates) of structural materials irradiated in the high and medium flux test modules of the International Fusion Materials Irradiation Facility (IFMIF). We have investigated if the change of the composition during the irradiation time has effect on the prediction of the atomic displacements. The effect of the activation cross section uncertainties in the assessment of both solid transmutants and hydrogen and helium production is also analyzed. The results are provided element-by-element, so that the primary damage of any material irradiated in such neutron environments can be easily assessed; in this paper, we have predicted the primary damage of the low activation steel Eurofer.

Irradiation response of ODS Eurofer97 steel

Energy Technology Data Exchange (ETDEWEB)

Luzginova, N.V., E-mail: luzginova@nrg.eu [Nuclear Research and Consultancy Group, Petten (Netherlands); Nolles, H.S.; Pierick, P. ten; Bakker, T.; Mutnuru, R.K.; Jong, M.; Blagoeva, D.T. [Nuclear Research and Consultancy Group, Petten (Netherlands)

2012-09-15

Oxide dispersion strengthened (ODS) Eurofer97 steel (EU batch, 0.3 wt.% of Y{sub 2}O{sub 3} particles), produced by mechanical alloying followed by hot rolling, is irradiated in the High Flux Reactor in Petten, The Netherlands at three different irradiation temperatures (300, 450 and 550 Degree-Sign C) up to nominal doses of 1 dpa and 3 dpa. The effect of neutron irradiation on the mechanical properties of ODS Eurofer97 material is investigated. It is shown that the irradiation hardening of ODS Eurofer97 steel occurs at 300 Degree-Sign C, whereas during irradiation at 450 and 550 Degree-Sign C no changes in mechanical properties are observed compared to the unirradiated material. This effect is possibly a result of the annealing of the irradiation damage at temperatures higher than 300 Degree-Sign C. The observed shifts in the Ductile to Brittle Transition Temperatures due to irradiation at different temperatures are discussed and compared with non-ODS Eurofer97 steel.

Irradiation behavior evaluation of oxide dispersion strengthened ferritic steel cladding tubes irradiated in JOYO

Energy Technology Data Exchange (ETDEWEB)

Yamashita, Shinichiro, E-mail: yamashita.shinichiro@jaea.go.jp; Yano, Yasuhide; Ohtsuka, Satoshi; Yoshitake, Tsunemitsu; Kaito, Takeji; Koyama, Shin-ichi; Tanaka, Kenya

2013-11-15

Irradiation behavior of ODS steel cladding tubes was evaluated for the further progress in understanding of the neutron-irradiation effects on ODS steel. Two types of ODS (9Cr–ODS{sub F}/M, 12Cr–ODS{sub F}) steel cladding tubes with differences in basic compositions and matrix phases were irradiated in JOYO. Post-irradiation examination data concerning hardness, ring tensile property, and microstructure were obtained. Hardness measurement after irradiation showed that there was an apparent irradiation temperature dependence on hardness for 9Cr–ODS{sub F}/M steel whereas no distinct temperature dependence for 12Cr–ODS{sub F} steel. Also, there was no significant change in tensile strengths after irradiation below 923 K, but those above 1023 K up to 6.6 × 10{sup 26} n/m{sup 2} (E > 0.1 MeV) were decreased by about 20%. TEM observations showed that the radiation-induced defect cluster formation during irradiation was suppressed because of high density sink site for defect such as initially-existed dislocation, and precipitate interfaces. In addition, oxide particles were stable up to the maximum doses of this irradiation test.

Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

International Nuclear Information System (INIS)

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

1977-01-01

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

Analysis of Low Dose Irradiation Damages in Structural Ferritic/Martensitic Steels by Proton Irradiation and Nanoindentation

International Nuclear Information System (INIS)

Waseem, Owais A.; Ryu, Ho Jin; Park, Byong Guk; Jeong, Jong Ryul; Maeng, Cheol Soo; Lee, Myoung Goo

2016-01-01

As a result, ferritic-martensitic steels find applications in the in-core and out-of-core components which include ducts, piping, pressure vessel and cladding, etc. Due to ferromagnetism of F/M steel, it has been successfully employed in solenoid type fuel injector. Although the irradiation induced degradation in ferritic martensitic steels is lower as compare to (i) reduced activation steels, (ii) austenitic steels and (iii) martensitic steels, F/M steels are still prone to irradiation induced hardening and void swelling. The irradiation behavior may become more sophisticated due to transmutation and production of helium and hydrogen. The ductile to brittle transition temperature of F/M steels is also expected to increase due to irradiation. These irradiation induced degradations may deteriorate the integrity of F/M components. As a result of these investigations, it has found that the F/M steels experience no irradiation hardening above 400 .deg. C, but below this temperature, up to 350 .deg. C, weak hardening is observed. The irradiation hardening becomes more pronounced below 300 .deg. C. Moreover, the irradiation hardening has also found dependent upon radiation damage. The hardening was found increasing with increasing dose. Due to pronounced irradiation hardening below 300 .deg. C and increasing radiation damage with increasing dose (even at low dpa), it is required to investigate the post irradiation mechanical properties of F/M steel, in order to confirm its usefulness in structural and magnetic components which experience lifetime doses as low as 1x10"-"5 dpa.

Analysis of Low Dose Irradiation Damages in Structural Ferritic/Martensitic Steels by Proton Irradiation and Nanoindentation

Energy Technology Data Exchange (ETDEWEB)

Waseem, Owais A.; Ryu, Ho Jin; Park, Byong Guk [KAIST, Daejeon (Korea, Republic of); Jeong, Jong Ryul [Chungnam University, Daejeon (Korea, Republic of); Maeng, Cheol Soo; Lee, Myoung Goo [KEPCO, Daejeon (Korea, Republic of)

2016-05-15

As a result, ferritic-martensitic steels find applications in the in-core and out-of-core components which include ducts, piping, pressure vessel and cladding, etc. Due to ferromagnetism of F/M steel, it has been successfully employed in solenoid type fuel injector. Although the irradiation induced degradation in ferritic martensitic steels is lower as compare to (i) reduced activation steels, (ii) austenitic steels and (iii) martensitic steels, F/M steels are still prone to irradiation induced hardening and void swelling. The irradiation behavior may become more sophisticated due to transmutation and production of helium and hydrogen. The ductile to brittle transition temperature of F/M steels is also expected to increase due to irradiation. These irradiation induced degradations may deteriorate the integrity of F/M components. As a result of these investigations, it has found that the F/M steels experience no irradiation hardening above 400 .deg. C, but below this temperature, up to 350 .deg. C, weak hardening is observed. The irradiation hardening becomes more pronounced below 300 .deg. C. Moreover, the irradiation hardening has also found dependent upon radiation damage. The hardening was found increasing with increasing dose. Due to pronounced irradiation hardening below 300 .deg. C and increasing radiation damage with increasing dose (even at low dpa), it is required to investigate the post irradiation mechanical properties of F/M steel, in order to confirm its usefulness in structural and magnetic components which experience lifetime doses as low as 1x10{sup -5} dpa.

Phase stability in thermally-aged CASS CF8 under heavy ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Li, Meimei, E-mail: mli@anl.gov [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Miller, Michael K. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Chen, Wei-Ying [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

2015-07-15

Highlights: • Thermally-aged CF8 was irradiated with 1 MeV Kr ions at 400 °C. • Atom probe tomography revealed a strong dose dependence of G-phase precipitates. • Phase separation of α and α′ in ferrite was reduced after irradiation. - Abstract: The stability of the microstructure of a cast austenitic stainless steel (CASS), before and after heavy ion irradiation, was investigated by atom probe tomography (APT). A CF8 ferrite–austenite duplex alloy was thermally aged at 400 °C for 10,000 h. After this treatment, APT revealed nanometer-sized G-phase precipitates and Fe-rich α and Cr-enriched α′ phase separated regions in the ferrite. The thermally-aged CF8 specimen was irradiated with 1 MeV Kr ions to a fluence of 1.88 × 10{sup 19} ions/m{sup 2} at 400 °C. After irradiation, APT analysis revealed a strong spatial/dose dependence of the G-phase precipitates and the α–α′ spinodal decomposition in the ferrite. For the G-phase precipitates, the number density increased and the mean size decreased with increasing dose, and the particle size distribution changed considerably under irradiation. The inverse coarsening process can be described by recoil resolution. The amplitude of the α–α′ spinodal decomposition in the ferrite was apparently reduced after heavy ion irradiation.

Surface amorphization in Al2O3 induced by swift heavy ion irradiation

International Nuclear Information System (INIS)

Okubo, N.; Ishikawa, N.; Sataka, M.; Jitsukawa, S.

2013-01-01

Microstructure in single crystalline Al 2 O 3 developed during irradiation by swift heavy ions has been investigated. The specimens were irradiated by Xe ions with energies from 70 to 160 MeV at ambient temperature. The fluences were in the range from 1.0 × 10 13 to 1.0 × 10 15 ions/cm 2 . After irradiations, X-ray diffractometry (XRD) measurements and cross sectional transmission electron microscope (TEM) observations were conducted. The XRD results indicate that in the initial stage of amorphization in single crystalline Al 2 O 3 , high-density S e causes the formation of new planes and disordering. The new distorted lattice planes formed in the early stage of irradiation around the fluence of 5.0 × 10 13 ions/cm 2 for single crystalline Al 2 O 3 irradiated with 160 MeV-Xe ions. Energy dependence on structural modification was also examined in single crystalline Al 2 O 3 irradiated by swift heavy ions. The XRD results indicate that the swift heavy ion irradiation causes the lattice expansion and the structural modification leading to amorphization progresses above the energy around 100 MeV in this XRD study. The TEM observations demonstrated that amorphization was induced in surface region in single crystalline Al 2 O 3 irradiated by swift heavy ions above the fluence expected from the results of XRD. Obvious boundary was observed in the cross sectional TEM images. The crystal structure of surface region above the boundary was identified to be amorphous and deeper region to be single crystal. The threshold fluence of amorphization was found to be around 1.0 × 10 14 ions/cm 2 in the case over 80 MeV swift heavy ion irradiation and the fluence did not depend on the crystal structures

Surface amorphization in Al2O3 induced by swift heavy ion irradiation

Science.gov (United States)

Okubo, N.; Ishikawa, N.; Sataka, M.; Jitsukawa, S.

2013-11-01

Microstructure in single crystalline Al2O3 developed during irradiation by swift heavy ions has been investigated. The specimens were irradiated by Xe ions with energies from 70 to 160 MeV at ambient temperature. The fluences were in the range from 1.0 × 1013 to 1.0 × 1015 ions/cm2. After irradiations, X-ray diffractometry (XRD) measurements and cross sectional transmission electron microscope (TEM) observations were conducted. The XRD results indicate that in the initial stage of amorphization in single crystalline Al2O3, high-density Se causes the formation of new planes and disordering. The new distorted lattice planes formed in the early stage of irradiation around the fluence of 5.0 × 1013 ions/cm2 for single crystalline Al2O3 irradiated with 160 MeV-Xe ions. Energy dependence on structural modification was also examined in single crystalline Al2O3 irradiated by swift heavy ions. The XRD results indicate that the swift heavy ion irradiation causes the lattice expansion and the structural modification leading to amorphization progresses above the energy around 100 MeV in this XRD study. The TEM observations demonstrated that amorphization was induced in surface region in single crystalline Al2O3 irradiated by swift heavy ions above the fluence expected from the results of XRD. Obvious boundary was observed in the cross sectional TEM images. The crystal structure of surface region above the boundary was identified to be amorphous and deeper region to be single crystal. The threshold fluence of amorphization was found to be around 1.0 × 1014 ions/cm2 in the case over 80 MeV swift heavy ion irradiation and the fluence did not depend on the crystal structures.

Depth distribution of Frank loop defects formed in ion-irradiated stainless steel and its dependence on Si addition

Energy Technology Data Exchange (ETDEWEB)

Chen, Dongyue, E-mail: dychen@safety.n.t.u-tokyo.ac.jp [The University of Tokyo, Department of Nuclear Engineering and Management, School of Engineering, 7-3-1 Hongo Bunkyo-ku, Tokyo 113-8656 (Japan); Murakami, Kenta [The University of Tokyo, Nuclear Professional School, School of Engineering, 2-22 Shirakata-Shirane, Tokai-mura, Ibaraki 319-1188 (Japan); Dohi, Kenji; Nishida, Kenji; Soneda, Naoki [Central Research Institute of Electric Power Industry, 2-11-1 Iwado-kita, Komae, Tokyo 201-8511 (Japan); Li, Zhengcao, E-mail: zcli@tsinghua.edu.cn [Tsinghua University, School of Materials Science and Engineering, Beijing 100084 (China); Liu, Li; Sekimura, Naoto [The University of Tokyo, Department of Nuclear Engineering and Management, School of Engineering, 7-3-1 Hongo Bunkyo-ku, Tokyo 113-8656 (Japan)

2015-12-15

Although heavy ion irradiation is a good tool to simulate neutron irradiation-induced damages in light water reactor, it produces inhomogeneous defect distribution. Such difference in defect distribution brings difficulty in comparing the microstructure evolution and mechanical degradation between neutron and heavy ion irradiation, and thus needs to be understood. Stainless steel is the typical structural material used in reactor core, and could be taken as an example to study the inhomogeneous defect depth distribution in heavy ion irradiation and its influence on the tested irradiation hardening by nano-indentation. In this work, solution annealed stainless steel model alloys are irradiated by 3 MeV Fe{sup 2+} ions at 400 °C to 3 dpa to produce Frank loops that are mainly interstitial in nature. The silicon content of the model alloys is also tuned to change point defect diffusion, so that the loop depth distribution influenced by diffusion along the irradiation beam direction could be discussed. Results show that in low Si (0% Si) and base Si (0.42% Si) samples the depth distribution of Frank loop density quite well matches the dpa profile calculated by the SRIM code, but in high Si sample (0.95% Si), the loop number density in the near-surface region is very low. One possible explanation could be Si’s role in enhancing the effective vacancy diffusivity, promoting recombination and thus suppressing interstitial Frank loops, especially in the near-surface region, where vacancies concentrate. By considering the loop depth distribution, the tested irradiation hardening is successfully explained by the Orowan model. A hardening coefficient of around 0.30 is obtained for all the three samples. This attempt in interpreting hardening results may make it easier to compare the mechanical degradation between different irradiation experiments.

Characteristic effects of heavy ion irradiation on the rat brain

International Nuclear Information System (INIS)

Sun, X.Z.; Takahashi, S.; Kubota, Y.; Yoshida, S.; Takeda, H.; Zhang, R.; Fukui, Y.

2005-01-01

Heavy ion irradiation has the feature to administer a large radiation dose in the vicinity of the endpoint in the beam range, and its irradiation system and biophysical characteristics are different from ordinary irradiation instruments like X- or gamma-rays. Using this special feature, heavy ion irradiation has been applied for cancer treatment. The safety and efficacy of heavy ion irradiator have been demonstrated to a great extent. For instance, brain tumors treated by heavy-ion beams became smaller or disappearance. However, fundamental research related to such clinical phenotypes and their underlying mechanisms are little known. In order to clarify characteristic effects of heavy ion irradiation on the brain, we developed an experimental system for irradiating a restricted region of the rat brain using heavy ion beams. The characteristics of the heavy ion beams, histological, behavioral and elemental changes were studied in the rat following heavy ion irradiation. Adult male Sprague-Dawley rats, aged 12 weeks and weighing 260-340 g (Shizuoka Laboratory Animal Center, Hamamatsu, Japan) were used. Rats were deeply anesthetized 10-15 minutes before irradiation with ketamine (40 mg/kg) and xylazine (10 mg/kg), immobilized in a specifically designed jig, and irradiated with 290 MeV/nucleon charged carbon beams in a dorsal-to ventral direction, The left cerebral hemispheres of the brain were irradiated at doses of 100 Gy charged carbon particles. The depth-dose distribution of the heavy ion beams was modified to make a spread-out bragg peak of 5 mm wide with a range modulator. The characteristics of the heavy-ion beams (field and depth of the heavy-ion beams) were examined by a measuring paraffin section of rat brain at different thickness. That extensive necrosis was observed between 2.5 mm and 7.5 mm depth from the surface of the rat head, suggesting a relatively high dose and uniform dose was delivered among designed depths and the spread-out bragg peak used here

Heavy-Section Steel Irradiation Program. Volume 2, No. 2: Semiannual progress report, April--September 1991

International Nuclear Information System (INIS)

Corwin, W.R.

1994-10-01

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 stools 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 post-irradiation annealing are being examined on a wide range of fracture properties. The HSSI Program is into 10 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 welds, (6) irradiation effects in a commercial low upper-sheer weld, (7) microstructural analysis of irradiation effects, (8) in-service aged material evaluations, (9) correlation monitor materials, and (10) special technical assistance. This report provides an overview of the activities within each of these tasks from April to September 1991

Modeling irradiation embrittlement in reactor pressure vessel steels

International Nuclear Information System (INIS)

Odette, G.R.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 10, numerical modeling of irradiation embrittlement in reactor vessel steels are introduced. Physically-based models are developed and their role in advancing the state-of-the-art of predicting irradiation embrittlement of RPV steels is stressed

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.

Heavy section steel technology program technical report No. 38. Fracture toughness characterization of HSST intermediate pressure vessel material

International Nuclear Information System (INIS)

Mager, T.R.; Yanichko, S.E.; Singer, L.R.

1974-12-01

The primary objective of the Heavy Section Steel Technology (HSST) Program is to develop pertinent fracture technology to demonstrate the structural reliability of present and contemplated water-cooled nuclear reactor pressure vessels. In order to demonstrate the ability to predict failure of large, heavy-walled pressure vessels under service type loading conditions, the fracture toughness properties of the vessel's materials must be characterized. The sampling procedure and test results are presented for vessel material supplied by the Oak Ridge National Laboratory that were used to characterize the fracture toughness of the HSST Intermediate Test Vessels. The metallurgical condition and heat treatment of the test material was representative of the vessel simulated service test condition. Test specimen locations and orientations were selected by the Oak Ridge National Laboratory and are representative of flaw orientations incorporated in the test vessels. The fracture toughness is documented for the materials from each of the eight HSST Intermediate Pressure Vessels tested to date. 7 references. (U.S.)

Corrosion of carbon steel and low-alloy steel in diluted seawater containing hydrazine under gamma-rays irradiation

International Nuclear Information System (INIS)

Nakano, Junichi; Yamamoto, Masahiro; Tsukada, Takashi

2014-01-01

Seawater was injected into reactor cores of Units 1, 2, and 3 in the Fukushima Daiichi nuclear power station as an urgent coolant. It is considered that the injected seawater causes corrosion of steels of the reactor pressure vessel and primary containment vessel. To investigate the effects of gamma-rays irradiation on weight loss in carbon steel and low-alloy steel, corrosion tests were performed in diluted seawater at 50°C under gamma-rays irradiation. Specimens were irradiated with dose rates of 4.4 kGy/h and 0.2 kGy/h. To evaluate the effects of hydrazine (N 2 H 4 ) on the reduction of oxygen and hydrogen peroxide, N 2 H 4 was added to the diluted seawater. In the diluted seawater without N 2 H 4 , weight loss in the steels irradiated with 0.2 kGy/h was similar to that in the unirradiated steels, and weight loss in the steels irradiated with 4.4 kGy/h increased to approximate 1.7 times of those in the unirradiated steels. Weight loss in the steels irradiated in the diluted seawater containing N 2 H 4 was similar to that in the diluted seawater without N 2 H 4 . When N 2 was introduced into the gas phase in the flasks during gamma-rays irradiation, weight loss in the steels decreased. (author)

Heritable non-lethal damage to cultured human cells irradiated with heavy ions

International Nuclear Information System (INIS)

Walker, J.T.; Walker, O.A.

2002-01-01

During interplanetary flights the nuclei of all of a crew member's cells could be traversed by at least one high-LET (linear energy transfer) cosmic-ray particle. In mammalian cells irradiated in vitro about 1 in 10,000 of the surviving cells traversed by heavy particles is transformed to malignancy or mutated. What, if anything, happens to the remaining >99% of surviving cells? A retrospective analysis of archived data and samples from heavy-ion irradiation experiments with cultured human cells in vitro indicated that heavy ions caused a dose- and LET-dependent reduction in growth rates of progeny of irradiated cells, based on colony-size distributions. The maximum action cross section for this effect is between 100 and 300 μm 2 , at least as large as the cell nuclear area and up to 3 times the cross section for cell killing. Thus, heritable slow growth is the most prevalent effect of high-LET radiations on cultured animal cells, which may have implications for crew health during deep space travel. (author)

Electron beam welding of heavy section 3Cr-1.5Mo alloy

International Nuclear Information System (INIS)

King, J.F.; David, S.A.; Nasreldin, A.

1986-01-01

Welding of thick section steels is a common practice in the fabrication of pressure vessels for energy systems. The fabrication cost is strongly influenced by the speed at which these large components can be welded. Conventional welding processes such as shielded metal arc (SMA) and submerged arc (SA) are time-consuming and expensive. Hence there is a great need to reduce welding time and the tonnage of weld metal deposited. Electron beam welding (EBW) is a process that potentially could be used to achieve dramatic reduction in the welding time and costs. The penetrating ability of the beam produces welds with high depth-to-width ratios at relatively high travel speeds, making it possible to weld thick sections with one or two passes without filler metals and other consumables. The paper describes a study that was undertaken to investigate the feasibility of using a high power electron beam welding machine to weld heavy section steel. The main emphasis of this work was concentrated on determining the mechanical properties of the resulting weldment, characterizing the microstructure of the various weldment regions, and comparing these results with those from other processes. One of the steels selected for the heavy section electron beam welding study was a new 3 Cr-1.5 Mo-0.1 V alloy. The steel was developed at the AMAX Materials Research Center by Wada and co-workers for high temperature, high pressure hydrogen service as a possible improved replacement for 2-1/4 Cr-1 Mo steels. The excellent strength and toughness of this steel make it a promising candidate for future pressure vessels such as those for coal gasifiers. The work was conducted on 102 mm (4 in.) thick plates of this material in the normalized-and-tempered condition

Heavy ion irradiation of crystalline water ice. Cosmic ray amorphisation cross-section and sputtering yield

Science.gov (United States)

Dartois, E.; Augé, B.; Boduch, P.; Brunetto, R.; Chabot, M.; Domaracka, A.; Ding, J. J.; Kamalou, O.; Lv, X. Y.; Rothard, H.; da Silveira, E. F.; Thomas, J. C.

2015-04-01

Context. Under cosmic irradiation, the interstellar water ice mantles evolve towards a compact amorphous state. Crystalline ice amorphisation was previously monitored mainly in the keV to hundreds of keV ion energies. Aims: We experimentally investigate heavy ion irradiation amorphisation of crystalline ice, at high energies closer to true cosmic rays, and explore the water-ice sputtering yield. Methods: We irradiated thin crystalline ice films with MeV to GeV swift ion beams, produced at the GANIL accelerator. The ice infrared spectral evolution as a function of fluence is monitored with in-situ infrared spectroscopy (induced amorphisation of the initial crystalline state into a compact amorphous phase). Results: The crystalline ice amorphisation cross-section is measured in the high electronic stopping-power range for different temperatures. At large fluence, the ice sputtering is measured on the infrared spectra, and the fitted sputtering-yield dependence, combined with previous measurements, is quadratic over three decades of electronic stopping power. Conclusions: The final state of cosmic ray irradiation for porous amorphous and crystalline ice, as monitored by infrared spectroscopy, is the same, but with a large difference in cross-section, hence in time scale in an astrophysical context. The cosmic ray water-ice sputtering rates compete with the UV photodesorption yields reported in the literature. The prevalence of direct cosmic ray sputtering over cosmic-ray induced photons photodesorption may be particularly true for ices strongly bonded to the ice mantles surfaces, such as hydrogen-bonded ice structures or more generally the so-called polar ices. Experiments performed at the Grand Accélérateur National d'Ions Lourds (GANIL) Caen, France. Part of this work has been financed by the French INSU-CNRS programme "Physique et Chimie du Milieu Interstellaire" (PCMI) and the ANR IGLIAS.

Irradiation creep in ferritic steels

International Nuclear Information System (INIS)

Vandermeulen, W.; Bremaecker, A. de; Burbure, S. de; Huet, J.J.; Asbroeck, P. van

Pressurized and non-pressurized capsules of several ferritic steels have been irradiated in Rapsodie between 400 and 500 0 C up to 3.7 x 10 22 n/cm 2 (E>0.1 MeV). Results of the diameter measurements are presented and show that the total in-pile deformation is lower than for austenitic steels

Fatigue crack propagation in neutron-irradiated ferritic pressure-vessel steels

International Nuclear Information System (INIS)

James, L.A.

1977-01-01

The results of a number of experiments dealing with fatigue crack propagation in irradiated reactor pressure-vessel steels are reviewed. The steels included ASTM alloys A302B, A533B, A508-2, and A543, as well as weldments in A543 steel. Fluences and irradiation conditions were generally typical of those experienced by most power reactors. In general, the effect of neutron irradiation on the fatigue crack propagation behavior of these steels was neither significantly beneficial nor significantly detrimental

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

-thick (25-mm)] specimen. Thus, fracture toughness of the material can be described by a fracture toughness-based reference temperature rather that by a temperature derived from a combination of drop-weight and Charpy impact tests. A statistical size correction based upon weakest-link theory is used to adjust the measured fracture toughness to that expected from a 1T specimen. Although the details of a consensus procedure is still under development, the basic procedure is widely used now to characterize elastic-plastic K Jc values in the transition range. For application to commercial nuclear RPVs, however, various uncertainties are being investigated as part of the Heavy-Section Steel Irradiation (HSSI) Program managed by the Oak Ridge National laboratory (ORNL) for the U.S. Nuclear Regulatory Commission. These include the use of relatively small specimens, e.g., precracked CVN (PCVN) and smaller size specimens, the applicability of the master curve to highly irradiated steels, and the effects of intergranular fracture. (author)

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

Microstructure and mechanical behavior of neutron irradiated ultrafine grained ferritic steel

Energy Technology Data Exchange (ETDEWEB)

Alsabbagh, Ahmad, E-mail: ahalsabb@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Sarkar, Apu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Miller, Brandon [ATR National Scientific User Facility, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Burns, Jatuporn [Center for Advanced Energy Studies, Idaho Falls, ID 83401 (United States); Squires, Leah; Porter, Douglas; Cole, James I. [ATR National Scientific User Facility, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Murty, K.L. [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States)

2014-10-06

Neutron irradiation effects on ultra-fine grain (UFG) low carbon steel prepared by equal channel angular pressing (ECAP) have been examined. Counterpart samples with conventional grain (CG) sizes have been irradiated alongside with the UFG ones for comparison. Samples were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to 1.37 dpa. Atom probe tomography revealed manganese and silicon-enriched clusters in both UFG and CG steel after neutron irradiation. Mechanical properties were characterized using microhardness and tensile tests, and irradiation of UFG carbon steel revealed minute radiation effects in contrast to the distinct radiation hardening and reduction of ductility in its CG counterpart. After irradiation, micro hardness indicated increases of around 9% for UFG versus 62% for CG steel. Similarly, tensile strength revealed increases of 8% and 94% respectively for UFG and CG steels while corresponding decreases in ductility were 56% versus 82%. X-ray quantitative analysis showed that dislocation density in CG increased after irradiation while no significant change was observed in UFG steel, revealing better radiation tolerance. Quantitative correlations between experimental results and modeling were demonstrated based on irradiation induced precipitate strengthening and dislocation forest hardening mechanisms.

Defects investigation in neutron irradiated reactor steels by positron annihilation

International Nuclear Information System (INIS)

Slugen, V.

2003-01-01

Positron annihilation spectroscopy (PAS) based on positron lifetime measurements using the Pulsed Low Energy Positron System (PLEPS) was applied to the investigation of defects of irradiated and thermally treated reactor pressure vessel (RPV) steels. PLEPS results showed that the changes in microstructure of the RPV-steel properties caused by neutron irradiation and post-irradiation heat treatment can be well detected. From the lifetime measurements in the near-surface region (20-550 nm) the defect density in Russian types of RPV-steels was calculated using the diffusion trapping model. The post-irradiation heat treatment studies performed on non-irradiated specimens are also presented. (author)

Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

International Nuclear Information System (INIS)

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

1977-01-01

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

IRRADIATION CREEP AND MECHANICAL PROPERTIES OF TWO FERRITIC-MARTENSITIC STEELS IRRADIATED IN THE BN-350 FAST REACTOR

International Nuclear Information System (INIS)

Porollo, S. I.; Konobeev, Yu V.; Dvoriashin, A. M.; Budylkin, N. I.; Mironova, E. G.; Leontyeva-Smirnova, M. V.; Loltukhovsky, A. G.; Bochvar, A. A.; Garner, Francis A.

2002-01-01

Russian ferritic/martensitic steels EP-450 and EP-823 were irradiated to 20-60 dpa in the BN-350 fast reactor in the form of pressurized creep tubes and small rings used for mechanical property tests. Data derived from these steels serves to enhance our understanding of the general behavior of this class of steels. It appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures less then 420 degrees C, but may be camouflaged somewhat by precipitation-related densification. The irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels, and that the loss of strength at test temperatures above 500 degrees C is a problem generic to all F/M steels. This conclusion is supported by post-irradiation measurement of short-term mechanical properties. At temperatures below 500 degrees C both steels retain their high strength (yield stress 0.2=550-600 MPa), but at higher test temperatures a sharp decrease of strength properties occurs. However, the irradiated steels still retain high post-irradiation ductility at test temperatures in the range of 20-700 degrees C.

Evaluation of neutron irradiation embrittlement in the Korean reactor pressure vessel steels (Final Report)

Energy Technology Data Exchange (ETDEWEB)

Hong, J. H.; Lee, B. S.; Chi, S. H.; Kim, J. H.; Oh, Y. J.; Yoon, J. H.; Kwon, S. C.; Park, D. G.; Kang, Y. H.; Choo, K. N.; Oh, J. M.; Park, S. J.; Kim, B. K.; Shin, Y. T.; Cho, M. S.; Sohn, J. M.; Kim, D. S.; Choo, Y. S.; Ahn, S. B.; Oh, W. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2001-05-01

Reactor pressure vessel materials, which were produced by a domestic company, Doosan Heavy Industries and construction Co., Ltd., have been evaluated using the neutron irradiation facility HANARO. For this evaluation, instrumented capsules were used for neutron irradiation of various kinds of specimens made of different heats of steels, which are VCD(Y4), VCD+Al(U4), Si+Al(Y5), U4 weld metal, and U4 HAZ, respectively. The fast neutron fluence levels ranged 1 to 5 (x10{sup 19} n/cm{sup 2}, E>1MeV) depending on the specimens and the irradiation temperature was controlled within 290{+-}10 deg C. The test results showed that, in the ranking of the material properties of the base metals, both before and after neutron irradiation, Y5 is the best, U4 the next and Y4 the last. Y4 showed a substantial change by neutron irradiation as well as the properties was worse than others in the unirradiated state. However, Y5, which showed the best properties in unirradiated state, was also the best in the resistance for irradiation embrittlement and one can hardly detect the property change after irradiation. The weldment showed a reasonably good resistance to irradiation embrittlement while the unirradiated properties were worse than base metals. The RPV steels are all expected to meet the screening criteria of the USNRC codes and regulations during the end of plant life. 39 refs., 42 figs., 27 tabs. (Author)

Changing in tool steels wear resistance under electron irradiation

International Nuclear Information System (INIS)

Braginskaya, A.E.; Manin, V.N.; Makedonskij, A.V.; Mel'nikova, N.A.; Pakchanin, L.M.; Petrenko, P.V.

1983-01-01

The tool steels and alloys wear resistance under dry friction after electron irradiation has been studied. Electron irradiation of a wide variety of steels is shown to increase wear resistance. In this case phase composition and lattice parameters changes are observed both in matrix and carbides. The conclusion is drawn that an appreciable increase of steel wear resistance under electron irradiation can be explained both by carbide phase volume gain and changes in it's composition and the formation of carbide phase submicroscopic heterogeneities and, possibly, complexes of defects

Self-ion Irradiation Damage of F/M and ODS steels

International Nuclear Information System (INIS)

Kang, Suk Hoon; Chun, Young-Bum; Noh, Sanghoon; Jang, Jinsung; Kim, Tae Kyu

2014-01-01

Oxide dispersion strengthened (ODS) ferritic steels are potential high-temperature materials that are stabilized by dispersed particles at elevated temperatures. These dispersed particles improve the tensile strength and creep rupture strength, they are expected to increase the operation temperature up to approximately 650 .deg. C and also enhance the energy efficiency of the fusion reactor. Some reports described that the nano-clusters are strongly resistant to coarsening by annealing up to 1000 .deg. C, and nanoclusters do not change after ion irradiation up to 0.7 dpa at 300 .deg. C. ODS steels will be inevitably exposed to neutron irradiation condition; the irradiation damages, creep and swelling are always great concern. The dispersed oxide particles are believed to determine the performance of the steel, even the radiation resistance. In this study, F/M and ODS model alloys of Korea Atomic Energy Research Institute (KAERI) were irradiated by Fe 3+ self-ion to emulate the neutron irradiation effect. In this study, Fe 3+ self-ion irradiation is used as means of introducing radiation damage in F/M steel and ODS steel. The ion accelerator named DuET (in Kyoto University, Japan) was used for irradiation of Fe 3+ ion by 6.4 MeV at 300 .deg. C. The maximum damage rate in F/M and ODS steels were estimated roughly 6 dpa. After radiation, point or line defects were dominantly observed in F/M steel, on the other hands, small circular cavities were typically observed in ODS steel. Nanoindentation is a useful tool to determine the irradiationinduced hardness change in the damage layer of ionirradiated iron base alloys

Aging and Embrittlement of High Fluence Stainless Steels

Energy Technology Data Exchange (ETDEWEB)

Was, gary; Jiao, Zhijie; der ven, Anton Van; Bruemmer, Stephen; Edwards, Dan

2012-12-31

Irradiation of austenitic stainless steels results in the formation of dislocation loops, stacking fault tetrahedral, Ni-Si clusters and radiation-induced segregation (RIS). Of these features, it is the formation of precipitates which is most likely to impact the mechanical integrity at high dose. Unlike dislocation loops and RIS, precipitates exhibit an incubation period that can extend from 10 to 46 dpa, above which the cluster composition changes and a separate phase, (G-phase) forms. Both neutron and heavy ion irradiation showed that these clusters develop slowly and continue to evolve beyond 100 dpa. Overall, this work shows that the irradiated microstructure features produced by heavy ion irradiation are remarkably comparable in nature to those produced by neutron irradiation at much lower dose rates. The use of a temperature shift to account for the higher damage rate in heavy ion irradiation results in a fairly good match in the dislocation loop microstructure and the precipitate microstructure in austenitic stainless steels. Both irradiations also show segregation of the same elements and in the same directions, but to achieve comparable magnitudes, heavy ion irradiation must be conducted at a much higher temperature than that which produces a match with loops and precipitates. First-principles modeling has confirmed that the formation of Ni-Si precipitates under irradiation is likely caused by supersaturation of solute to defect sinks caused by highly correlated diffusion of Ni and Si. Thus, the formation and evolution of Ni-Si precipitates at high dose in austenitic stainless steels containing Si is inevitable.

An examination of the potential for 9%Cr1%Mo steel as thick section tubeplates in fast reactors

International Nuclear Information System (INIS)

Orr, J.; Sanderson, S.J.

1984-01-01

The steam generator units of future commercial demonstration fast reactors are likely to have a requirement for heavy section tubeplates (up to 500mm thick) with good elevated temperature strength and creep-fatigue resistance. A comparison of the mechanical properties available for ferritic steels has suggested that 9%Cr1%Mo steel would be a strong candidate material for this application. Although this steel is covered in some national specifications for tubes, pipes, plates and forgings and is also well established in the UK nuclear industry, international experience to date is confined to sections less than ca 150mm. The potential of 9%Cr1%Mo steel for use in thick sections has therefore been assessed in the present study by using simulation heat treatments. The work reported here involved the laboratory-scale cooling of bar samples to simulate water-quenching rates in cylindrical sections up to 720mm diameter (ie: equivalent to 500mm thick plate). The tensile properties at ambient and 525 0 C and impact fracture appearance transition temperatures were determined for material tempered after cooling at simulated thick section rates; the transformation characteristics as influenced by the net chromium equivalent were also established. The results of this work show that 9%Cr1%Mo steel may be fully hardened in the equivalent of the section sizes examined,and the mechanical properties of tempered material show only a small reduction from those of thin section normalised and tempered 9%Cr1%Mo steel. These findings support the potential usage of heavy section 9%Cr1%Mo steel envisaged for fast reactor steam generator tubeplates

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-10-01

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

Overview of microstructural evolution in neutron-irradiated austenitic stainless steels

International Nuclear Information System (INIS)

Maziasz, P.J.

1993-01-01

Austenitic stainless steels are important structural materials common to several different reactor systems, including light water and fast breeder fission, and magnetic fusion reactors (LWR, FBR, and MFR, respectively). The microstructures that develop in 300 series austenitic stainless steels during neutron irradiation at 60-700 C include combinations of dislocation loops and networks, bubbles and voids, and various kinds of precipitate phases (radiation-induced, or -enhanced or -modified thermal phases). Many property changes in these steels during neutron irradiation are directly or indirectly related to radiation-induced microstructural evolution. Even more important is the fact that radiation-resistance of such steels during either FBR or MFR irradiation is directly related to control of the evolving microstructure during such irradiation. The purpose of this paper is to provide an overview of the large and complex body of data accumulated from various fission reactor irradiation experiments conducted over the many years of research on microstructural evolution in this family of steels. The data can be organized into several different temperature regimes which then define the nature of the dominant microstructural components and their sensitivities to irradiation parameters (dose, helium/dpa ratio, dose rate) or metallurgical variables (alloy composition, pretreatment). The emphasis in this paper will be on the underlying mechanisms driving the microstructure to evolve during irradiation or those enabling microstructural stability related to radiation resistance. (orig.)

Moessbauer spectroscopy of He irradiated austenitic stainless steel SUS304 at low temperature

Energy Technology Data Exchange (ETDEWEB)

Horii, Kiyomasa; Ishibashi, Tetsu; Toriyama, Tamotsu; Wakabayashi, Hidehiko; Iijima, Hiroshi [Musashi Inst. of Tech., Tokyo (Japan); Kawasaki, Katsunori; Hayashi, Nobuyuki; Sakamoto, Isao

1996-04-01

SUS 304 austenitic stainless steel causes the magnetic transition at 60 K, and the Young`s modulus lowers. In addition, its composition elements have the large (n,{alpha}) reaction cross section to high energy neutrons, and helium is apt to be generated, and this is a factor that lowers the material strength. In the He-irradiated parts in austenitic stainless steel, the precursory state of martensite transformation should exist, and its effect is considered to be observable by carrying out low temperature Moessbauer spectroscopy. As to the preparation of He-irradiation samples, the SUS 304 foils used and the irradiation conditions are described. The measurement of low temperature Moessbauer spectra for the samples without irradiation and with irradiation is reported. In order to determine the magnetic transition point, the thermal scanning measurement was carried out for the samples without or with irradiation. The martensite transformation was measured by X-ray diffraction and transmission type Moessbauer spectroscopy. In order to observe the state of the sample surfaces, the measurement by internal conversion electron Moessbauer spectroscopy was performed. These results and the temperature dependence of the Moessbauer spectra for the irradiated parts are reported. (K.I.)

Moessbauer spectroscopy of He irradiated austenitic stainless steel SUS304 at low temperature

International Nuclear Information System (INIS)

Horii, Kiyomasa; Ishibashi, Tetsu; Toriyama, Tamotsu; Wakabayashi, Hidehiko; Iijima, Hiroshi; Kawasaki, Katsunori; Hayashi, Nobuyuki; Sakamoto, Isao.

1996-01-01

SUS 304 austenitic stainless steel causes the magnetic transition at 60 K, and the Young's modulus lowers. In addition, its composition elements have the large (n,α) reaction cross section to high energy neutrons, and helium is apt to be generated, and this is a factor that lowers the material strength. In the He-irradiated parts in austenitic stainless steel, the precursory state of martensite transformation should exist, and its effect is considered to be observable by carrying out low temperature Moessbauer spectroscopy. As to the preparation of He-irradiation samples, the SUS 304 foils used and the irradiation conditions are described. The measurement of low temperature Moessbauer spectra for the samples without irradiation and with irradiation is reported. In order to determine the magnetic transition point, the thermal scanning measurement was carried out for the samples without or with irradiation. The martensite transformation was measured by X-ray diffraction and transmission type Moessbauer spectroscopy. In order to observe the state of the sample surfaces, the measurement by internal conversion electron Moessbauer spectroscopy was performed. These results and the temperature dependence of the Moessbauer spectra for the irradiated parts are reported. (K.I.)

Irradiated accelerated corrosion of stainless steel

International Nuclear Information System (INIS)

Raiman, S.S.; Wang, P.; Was, G.S.

2015-01-01

Type 316L stainless steel was exposed to a simulated PWR environment with in-situ proton irradiation to investigate the effect of simultaneous irradiation and corrosion. To enable these experiments, a dedicated beamline was constructed to transport a 3.2 MeV proton beam from a tandem accelerator, through the sample that also acts as the window between the beamline vacuum and a corrosion cell designed to flow primary water at 320 C. degrees and 13.1 MPa. Experiments were conducted on 316L stainless steel samples which were irradiated for 24 hours in 320 C. degrees water with 3 ppm H 2 , at dose rates of 7*10 -6 dpa/s and 7*10 -7 dpa/s, for 4, 24, and 72 hours. A dual-layer oxide formed on the samples, with an inner layer rich in Cr with Fe and Ni content, and an outer layer of Fe oxides. Samples were characterized with TEM (Transmission Electron Microscopy), EDS, and Raman spectroscopy to determine the effect of irradiation. Irradiated samples were found to have a thinner and more porous inner oxide which was deficient in chromium. The outer oxide was found to have significant hematite content, suggesting that irradiation led to an increase in ECP (Electro-Chemical Potential) at the oxide-solution interface, causing accelerated dissolution of the oxide under irradiation. (authors)

Fractal characteristics of fracture morphology of steels irradiated with high-energy ions

Energy Technology Data Exchange (ETDEWEB)

Xian, Yongqiang; Liu, Juan [Institute of Modern Physics, Chinese Academy of Science, Lanzhou 730000 (China); University of Chinese Academy of Science, Beijing 100049 (China); Zhang, Chonghong, E-mail: c.h.zhang@impcas.ac.cn [Institute of Modern Physics, Chinese Academy of Science, Lanzhou 730000 (China); Chen, Jiachao [Paul Scherrer Institute, Villigen PSI (Switzerland); Yang, Yitao; Zhang, Liqing; Song, Yin [Institute of Modern Physics, Chinese Academy of Science, Lanzhou 730000 (China)

2015-06-15

Highlights: • Fractal dimensions of fracture surfaces of steels before and after irradiation were calculated. • Fractal dimension can effectively describe change of fracture surfaces induced by irradiation. • Correlation of change of fractal dimension with embrittlement of irradiated steels is discussed. - Abstract: A fractal analysis of fracture surfaces of steels (a ferritic/martensitic steel and an oxide-dispersion-strengthened ferritic steel) before and after the irradiation with high-energy ions is presented. Fracture surfaces were acquired from a tensile test and a small-ball punch test (SP). Digital images of the fracture surfaces obtained from scanning electron microscopy (SEM) were used to calculate the fractal dimension (FD) by using the pixel covering method. Boundary of binary image and fractal dimension were determined with a MATLAB program. The results indicate that fractal dimension can be an effective parameter to describe the characteristics of fracture surfaces before and after irradiation. The rougher the fracture surface, the larger the fractal dimension. Correlation of the change of fractal dimension with the embrittlement of the irradiated steels is discussed.

Neutron irradiation effects in pressure vessel steels and weldments

Energy Technology Data Exchange (ETDEWEB)

Ianko, L [International Atomic Energy Agency, Vienna (Austria). Div. of Nuclear Power; Davies, L M

1994-12-31

This paper deals with the effects of neutron irradiation on the steel and welds used for the pressure vessels which house the reactor cores in light water reactors: irradiation effects on mechanical properties and the shift in ductile-brittle transition temperature, importance of the knowledge of the neutron fluence and of the monitoring and surveillance programmes; empirical and mechanistic modelling of irradiation effects and the necessity of data extension to new operational limits; consequences on the manufacturing and structural design of materials and structures; mitigation of irradiation effects by annealing; international activities and programmes in the field of neutron irradiation effects on PV steels and welds. 37 refs., 22 figs.

Fracture toughness of irradiated and recovered vessel steels

International Nuclear Information System (INIS)

Perosanz, F.; Lapena, J.

1998-01-01

This paper presents the fracture toughness measurements carried out on three vessel steels in an irradiated condition and after a post-irradiation recovery treatment. A statistical approach and the fracture parameters corresponding to two theoretical models of the fracture tests are used for evaluating toughness. Test results show that the neutron fluence gradually transforms the fracture behaviour of the vessel steels from ductile to brittle and seriously reduces their fracture toughness. The effectiveness of the recovery treatment, as evaluated from the toughness measurements, is confirmed, although the efficiency is not the same for the steels and depends on the evaluation parameter except in the case of almost complete recovery. The recovery effect increases with the received neutron fluence if the toughness values after treatment are compared with those in the irradiated condition rather than those in the as received condition. (orig.)

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

Chemical coloring on stainless steel by ultrasonic irradiation.

Science.gov (United States)

Cheng, Zuohui; Xue, Yongqiang; Ju, Hongbin

2018-01-01

To solve the problems of high temperature and non-uniformity of coloring on stainless steel, a new chemical coloring process, applying ultrasonic irradiation to the traditional chemical coloring process, was developed in this paper. The effects of ultrasonic frequency and power density (sound intensity) on chemical coloring on stainless steel were studied. The uniformity of morphology and colors was observed with the help of polarizing microscope and scanning electron microscopy (SEM), and the surface compositions were characterized by X-ray photoelectric spectroscopy (XPS), meanwhile, the wear resistance and the corrosion resistance were investigated, and the effect mechanism of ultrasonic irradiation on chemical coloring was discussed. These results show that in the process of chemical coloring on stainless steel by ultrasonic irradiation, the film composition is the same as the traditional chemical coloring, and this method can significantly enhance the uniformity, the wear and corrosion resistances of the color film and accelerate the coloring rate which makes the coloring temperature reduced to 40°C. The effects of ultrasonic irradiation on the chemical coloring can be attributed to the coloring rate accelerated and the coloring temperature reduced by thermal-effect, the uniformity of coloring film improved by dispersion-effect, and the wear and corrosion resistances of coloring film enhanced by cavitation-effect. Ultrasonic irradiation not only has an extensive application prospect for chemical coloring on stainless steel but also provides an valuable reference for other chemical coloring. Copyright © 2017 Elsevier B.V. All rights reserved.

Behavior of ferritic steels irradiated by fast neutrons

International Nuclear Information System (INIS)

Erler, Jean; Maillard, Arlette; Brun, Gilbert; Lehmann, Jeanne; Dupouy, J.-M.

1979-01-01

Ferritic steels were irradiated in Rapsodie and Phenix at varying doses. The swelling and irradiation creep characteristics are reported below as are the mechanical characteristics of these materials [fr

Microstructure and nanoindentation of the CLAM steel with nanocrystalline grains under Xe irradiation

Energy Technology Data Exchange (ETDEWEB)

Chang, Yongqin, E-mail: chang@ustb.edu.cn; Zhang, Jing; Li, Xiaolin; Guo, Qiang; Wan, Farong; Long, Yi

2014-12-15

This work presents an early look at irradiation effects on China low activation martensitic (CLAM) steel with nanocrystalline grains (NC-CLAM steels) under 500 keV Xe-ion bombardment at room temperature to doses up to 5.3 displacements per atom (dpa). The microstructure in the topmost region of the steel is composed of nanocrystalline grains with an average diameter of 13 nm. As the samples were implanted at low dose, the nanocrystalline grains had martensite lath structure, and many dislocations and high density bubbles were introduced into the NC-CLAM steels. As the irradiation dose up to 5.3 dpa, a tangled dislocation network exists in the lath region, and the size of the bubbles increases. X-ray diffraction results show that the crystal quality decreases after irradiation, although the nanocrystals obviously coarsen. Grain growth under irradiation may be ascribed to the direct impact of the thermal spike on grain boundaries in the NC-CLAM steels. In irradiated samples, a compressive stress exists in the surface layer because of grain growth and irradiation-introduced defects, while the irradiation introduced grain-size coarsening and defects gradients from the surface to matrix result in a tensile stress in the irradiated NC-CLAM steels. Nanoindentation was used to estimate changes in mechanical properties during irradiation, and the results show that the hardness of the NC-CLAM steels increases with increasing irradiation dose, which was ascribed to the competition between the grain boundaries and the irradiation-introduced defects.

Effect of swift heavy ion irradiation on ethylene–chlorotrifluoroethylene copolymer

International Nuclear Information System (INIS)

Singh, Lakhwant; Devgan, Kusum; Samra, Kawaljeet Singh

2012-01-01

The swift heavy irradiation induced changes taking place in ethylene–chlorotrifluoroethylene (E–CTFE) copolymer films were investigated in correlation with the applied doses. Samples were irradiated in vacuum at room temperature by lithium (50 MeV), carbon (85 MeV), nickel (120 MeV) and silver (120 MeV) ions with the fluence in the range of 1×10 11 –3×10 12 ions cm −2 . Structural and thermal properties of the irradiated as well as pristine E–CTFE films were studied using FTIR, UV–visible, TGA, DSC and XRD techniques. Swift heavy ion irradiation was found to induce changes in E–CTFE depending upon the applied doses. - Highlights: ► Effect of swift heavy ion irradiation on E–CTFE films has been studied. ► Different structural changes in the original structure of E–CTFE are observed after irradiation with different ions. ► Swift heavy ion irradiation has made E–CTFE more prone to thermal degradation.

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

A correlation between micro- and nano-indentation on materials irradiated by high-energy heavy ions

Science.gov (United States)

Yang, Yitao; Zhang, Chonghong; Ding, Zhaonan; Su, Changhao; Yan, Tingxing; Song, Yin; Cheng, Yuguang

2018-01-01

Hardness testing is an efficient means of assessing the mechanical properties of materials due to the small sampling volume requirement. Previous studies have established the correlation between flow stress and Vickers hardness. However, the damage layer produced by ions irradiation with low energy is too thin to perform Vickers hardness test, which is usually measured by nano-indentation. Therefore, it is necessary to correlate the Vickers hardness and nano-hardness for the convenience of assessing mechanical properties of materials under irradiation. In this study, various materials (pure nickel, nickel base alloys and oxide dispersion strengthened steel) were irradiated with high-energy heavy ions to different damage levels. After irradiation, micro- and nano-indentation were performed to characterize the change in hardness. Due to indentation size effect (ISE), the hardness was dependent of load or depth. Therefore, Nix-Gao model was used to obtain the hardness without ISE (Hv0 and Hnano_0). The determined Hv0 was plotted as a function of the corresponding Hnano_0, then a good linear relation was found between Vickers hardness and nano-hardness, and a coefficient was determined to be 81.0 ± 10.5, namely, Hv 0 = 81.0Hnano _ 0 (Hv0 with unit of kgf/mm2, Hnano_0 with unit of GPa). This correlation was based on the data from various materials, therefore it was independent of materials. Based on the established correlation and nano-indentation results, the change fraction in yield stress of Inconel 718 and pure Ni with ion irradiation was compared with that with neutron irradiation. The data of Inconel 718 with heavy ion irradiation was in good agreement with the data with neutron irradiation, which was a good demonstration for the validation of the established correlation. However, a distinctive difference in change fraction of yield stress was seen for pure Ni under heavy ion irradiation and neutron irradiation, which was attributed to the difference in samples

Behavior of optical fibers under heavy irradiation

International Nuclear Information System (INIS)

Kakuta, T.; Sagawa, T.

1998-01-01

Several kinds of optical diagnostics are planned in a fusion reactor. Complicated optical systems such as periscopes are thought to be primary candidates for optical measurements, especially for visible wavelengths. However, optical fibers have several advantages over such optical systems. Also, the optical fibers could be a far better transmission line for signals under a high electromagnetic field. However, they have been considered vulnerable to heavy irradiation. In this study, several kinds of optical fibers were irradiated in the JMTR fission reactor. The optical transmissivity in fibers was measured in situ during fast neutron and gamma irradiation, up to doses of 2 x 10 24 n m -2 and 5 x 10 9 Gy, respectively. The irradiation temperature ranged from 300 to 700 K. For pure ionizing irradiation environments, some methods for improving the radiation resistance of optical fibers were indicated. The results showed that effects of the irradiation associated with fast neutrons would be different from the effects of pure ionizing irradiation. Some fibers were found to withstand the heavy irradiation, especially in an infrared wavelength range. (orig.)

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.)

Heavy ion irradiation of astrophysical ice analogs

International Nuclear Information System (INIS)

Duarte, Eduardo Seperuelo; Domaracka, Alicja; Boduch, Philippe; Rothard, Hermann; Balanzat, Emmanuel; Dartois, Emmanuel; Pilling, Sergio; Farenzena, Lucio; Frota da Silveira, Enio

2009-01-01

Icy grain mantles consist of small molecules containing hydrogen, carbon, oxygen and nitrogen atoms (e.g. H 2 O, GO, CO 2 , NH 3 ). Such ices, present in different astrophysical environments (giant planets satellites, comets, dense clouds, and protoplanetary disks), are subjected to irradiation of different energetic particles: UV radiation, ion bombardment (solar and stellar wind as well as galactic cosmic rays), and secondary electrons due to cosmic ray ionization of H 2 . The interaction of these particles with astrophysical ice analogs has been the object of research over the last decades. However, there is a lack of information on the effects induced by the heavy ion component of cosmic rays in the electronic energy loss regime. The aim of the present work is to simulate of the astrophysical environment where ice mantles are exposed to the heavy ion cosmic ray irradiation. Sample ice films at 13 K were irradiated by nickel ions with energies in the 1-10 MeV/u range and analyzed by means of FTIR spectrometry. Nickel ions were used because their energy deposition is similar to that deposited by iron ions, which are particularly abundant cosmic rays amongst the heaviest ones. In this work the effects caused by nickel ions on condensed gases are studied (destruction and production of molecules as well as associated cross sections, sputtering yields) and compared with respective values for light ions and UV photons. (authors)

Microstructural evolution in neutron irradiated reactor pressure vessel steels

International Nuclear Information System (INIS)

English, C.A.; Phythian, W.J.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

Annealing effect on restoration of irradiation steel properties

International Nuclear Information System (INIS)

Vishkarev, O.M.; Kolesova, T.N.; Myasnikova, K.P.; Pecherin, A.M.; Shamardin, V.K.

1986-01-01

The effect of temperature and annealing time on the restoration of properties of the 15Kh2NMFAA and 15Kh2MFA steels after irradiation at 285 deg with the fluence of 6x10 23 neutr/m 2 (E>0.5 MeV) is studied. Microhardness (H μ ) restoration in the irradiated 15Kh2NMFAA steel is shown to start from 350 deg C annealing temperature. The complete microhardness restoration is observed at the annealing temperature of 500 deg C for 10 hours

Post-irradiation characterization of PH13-8Mo martensitic stainless steel

Energy Technology Data Exchange (ETDEWEB)

Jong, M.; Schmalz, F.; Rensman, J.W. [Nuclear Research and consultancy Group, Westerduinweg 3, 1755 ZG Petten (Netherlands); Luzginova, N.V., E-mail: luzginova@nrg.eu [Nuclear Research and consultancy Group, Westerduinweg 3, 1755 ZG Petten (Netherlands); Wouters, O.; Hegeman, J.B.J.; Laan, J.G. van der [Nuclear Research and consultancy Group, Westerduinweg 3, 1755 ZG Petten (Netherlands)

2011-10-01

The irradiation response of PH13-8Mo stainless steel was measured up to 2.5 dpa at 200 and 300 deg. C irradiation temperatures. The PH13-8Mo, a martensitic precipitation-hardened steel, was produced by Hot Isostatic Pressing at 1030 deg. C. The fatigue tests (high cycle fatigue and fatigue crack propagation) showed a test temperature dependency but no irradiation effects. Tensile tests showed irradiation hardening (yield stress increase) of approximately 37% for 200 deg. C irradiated material tested at 60 deg. C and approximately 32% for 300 deg. C irradiated material tested at 60 deg. C. This contradicts the shift in reference temperature (T{sub 0}) measured in toughness tests (Master Curve approach), where the {Delta}T{sub 0} for 300 deg. C irradiated is approximately 170 deg. C and the {Delta}T{sub 0} for the 200 deg. C irradiated is approximately 160 deg. C. This means that the irradiation hardening of PH13-8Mo steel is not suitable to predict the shift in the reference temperature for the Master Curve approach.

Influence of laser shock peening on irradiation defects in austenitic stainless steels

Science.gov (United States)

Lu, Qiaofeng; Su, Qing; Wang, Fei; Zhang, Chenfei; Lu, Yongfeng; Nastasi, Michael; Cui, Bai

2017-06-01

The laser shock peening process can generate a dislocation network, stacking faults, and deformation twins in the near surface of austenitic stainless steels by the interaction of laser-driven shock waves with metals. In-situ transmission electron microscopy (TEM) irradiation studies suggest that these dislocations and incoherent twin boundaries can serve as effective sinks for the annihilation of irradiation defects. As a result, the irradiation resistance is improved as the density of irradiation defects in laser-peened stainless steels is much lower than that in untreated steels. After heating to 300 °C, a portion of the dislocations and stacking faults are annealed out while the deformation twins remain stable, which still provides improved irradiation resistance. These findings have important implications on the role of laser shock peening on the lifetime extension of austenitic stainless steel components in nuclear reactor environments.

The irradiation performance of austenitic stainless steel clade PWR fuel rods

International Nuclear Information System (INIS)

Teixeira e Silva, A.; Esteves, A.M.

1988-01-01

The steady state irradiation performance of austenitic stainless steel clad pressurized water reactor fuel rods is modeled with fuel performance codes of the FRAP series. These codes, originally developed to model the thermal-mechanical behavior of zircaloy clad fuel rods, are modified to model stainless steel clad fuel rods. The irradiation thermal-mechanical behavior of type 348 stainless steel and zircaloy fuel rods is compared. (author) [pt

Influence of laser shock peening on irradiation defects in austenitic stainless steels

Energy Technology Data Exchange (ETDEWEB)

Lu, Qiaofeng [Department of Mechanical & Materials Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Su, Qing [Nebraska Center for Energy Sciences Research, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Wang, Fei [Department of Mechanical & Materials Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Zhang, Chenfei; Lu, Yongfeng [Department of Electrical Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Nastasi, Michael [Department of Mechanical & Materials Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Nebraska Center for Energy Sciences Research, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Nebraska Center for Materials and Nanoscience, University of Nebraska-Lincoln, Lincoln, NE 68588 (United States); Cui, Bai, E-mail: bcui3@unl.edu [Department of Mechanical & Materials Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Nebraska Center for Materials and Nanoscience, University of Nebraska-Lincoln, Lincoln, NE 68588 (United States)

2017-06-15

The laser shock peening process can generate a dislocation network, stacking faults, and deformation twins in the near surface of austenitic stainless steels by the interaction of laser-driven shock waves with metals. In-situ transmission electron microscopy (TEM) irradiation studies suggest that these dislocations and incoherent twin boundaries can serve as effective sinks for the annihilation of irradiation defects. As a result, the irradiation resistance is improved as the density of irradiation defects in laser-peened stainless steels is much lower than that in untreated steels. After heating to 300 °C, a portion of the dislocations and stacking faults are annealed out while the deformation twins remain stable, which still provides improved irradiation resistance. These findings have important implications on the role of laser shock peening on the lifetime extension of austenitic stainless steel components in nuclear reactor environments. - Highlights: •Laser shock peening generates a dislocation network, stacking faults and deformation twins in stainless steels. •Dislocations and incoherent twin boundaries serve as effective sinks for the annihilation of irradiation defects. •Incoherent twin boundaries remain as stable and effective defect sinks at 300 °C.

Effect of ion irradiation-produced defects on the mobility of dislocations in 304 stainless steel

International Nuclear Information System (INIS)

Briceno, M.; Fenske, J.; Dadfarnia, M.; Sofronis, P.; Robertson, I.M.

2011-01-01

The impact of heavy-ion produced defects on the mobility of dislocations, dislocation sources and newly generated dislocations in 304 stainless steel are discovered by performing irradiation and deformation experiments in real time in the transmission electron microscope. Dislocations mobile prior to the irradiation are effectively locked in position by the irradiation, but the irradiation has no discernible impact on the ability of a source to generate dislocations. The motion and mobility of a dislocation is altered by the irradiation. It becomes irregular and jerky and the mobility increases slowly with time as the radiation-produced defects are annihilated locally. Channels created by dislocations ejected from grain boundary dislocation sources were found to have a natural width, as the emission sites within the boundary were spaced close together. Finally, the distribution of dislocations, basically, an inverse dislocation pile-up, within a cleared channel suggests a new mechanism for generating high local levels of stress at grain boundaries. The impact of these observations on the mechanical properties of irradiated materials is discussed briefly.

Effect of ion irradiation-produced defects on the mobility of dislocations in 304 stainless steel

Energy Technology Data Exchange (ETDEWEB)

Briceno, M.; Fenske, J. [Department of Materials Science and Engineering, University of Illinois, Urbana, IL 61801 (United States); Dadfarnia, M.; Sofronis, P. [Department of Mechanical Science and Engineering, University of Illinois, Urbana, IL 61801 (United States); Robertson, I.M., E-mail: ian.robertson@tcd.ie [Department of Materials Science and Engineering, University of Illinois, Urbana, IL 61801 (United States)

2011-02-01

The impact of heavy-ion produced defects on the mobility of dislocations, dislocation sources and newly generated dislocations in 304 stainless steel are discovered by performing irradiation and deformation experiments in real time in the transmission electron microscope. Dislocations mobile prior to the irradiation are effectively locked in position by the irradiation, but the irradiation has no discernible impact on the ability of a source to generate dislocations. The motion and mobility of a dislocation is altered by the irradiation. It becomes irregular and jerky and the mobility increases slowly with time as the radiation-produced defects are annihilated locally. Channels created by dislocations ejected from grain boundary dislocation sources were found to have a natural width, as the emission sites within the boundary were spaced close together. Finally, the distribution of dislocations, basically, an inverse dislocation pile-up, within a cleared channel suggests a new mechanism for generating high local levels of stress at grain boundaries. The impact of these observations on the mechanical properties of irradiated materials is discussed briefly.

Special heavy plates and steel solutions for bridge building

Science.gov (United States)

Lehnert, Tobias

2017-09-01

In many European countries infrastructure, -road as well as railway infrastructure-, needs intensive investments to follow the growing demands of mobility and goods traffic. Steel or steel composite bridges offer in this context viable and very sustainable solutions. Due to its unlimited recyclability steel can in general be seen as the ideal material for such sustainable constructions, but especially when designers or fabricators exploit the nowadays available possibilities of steel industry very cost-efficient and remarkable constructions are realizable. This paper will highlight some of these newest developments in heavy plates for bridge building. For example, for small span railway bridges the so-called thick plate trough bridges have proven to be a favourable concept. Very heavy plates with single plate weights up to 42 t allow building these bridges very efficiently out of one or very few single plates. Another interesting development is the so-called longitudinally profiled plates which allow a varying plate thickness along the actual loading profile. As last point the rising entry of higher strength steels in bridge building will be discussed and it will be shown why thermomechanically rolled plates are the ideal solution for these demands.

Heavy ion irradiation induces autophagy in irradiated C2C12 myoblasts and their bystander cells

International Nuclear Information System (INIS)

Hino, Mizuki; Tajika, Yuki; Hamada, Nobuyuki

2010-01-01

Autophagy is one of the major processes involved in the degradation of intracellular materials. Here, we examined the potential impact of heavy ion irradiation on the induction of autophagy in irradiated C2C12 mouse myoblasts and their non-targeted bystander cells. In irradiated cells, ultrastructural analysis revealed the accumulation of autophagic structures at various stages of autophagy (id est (i.e.) phagophores, autophagosomes and autolysosomes) within 20 min after irradiation. Multivesicular bodies (MVBs) and autolysosomes containing MVBs (amphisomes) were also observed. Heavy ion irradiation increased the staining of microtubule-associated protein 1 light chain 3 and LysoTracker Red (LTR). Such enhanced staining was suppressed by an autophagy inhibitor 3-methyladenine. In addition to irradiated cells, bystander cells were also positive with LTR staining. Altogether, these results suggest that heavy ion irradiation induces autophagy not only in irradiated myoblasts but also in their bystander cells. (author)

Heavy-Section Steel Technology program overview

International Nuclear Information System (INIS)

Pennell, W.E.

1990-01-01

This paper presents a status review of ongoing HSST program tasks aimed at refining the technology used in analysis of reactor pressure vessel fracture margins under pressurized thermal-shock (PTS) loading. Specific fracture-technology issues addressed include vessel flaw density and distribution, shallow flaws, fracture-toughness data transfer, circumferential cracks, ductile tearing and the influence of low-tearing toughness in stainless steel cladding. Preliminary results from the analysis and test programs are presented, together with interim assessments of their potential impact on a reactor vessel PTS analysis. 31 refs., 23 figs., 1 tab

Effect of irradiation temperature on microstructural changes in self-ion irradiated austenitic stainless steel

Science.gov (United States)

Jin, Hyung-Ha; Ko, Eunsol; Lim, Sangyeob; Kwon, Junhyun; Shin, Chansun

2017-09-01

We investigated the microstructural and hardness changes in austenitic stainless steel after Fe ion irradiation at 400, 300, and 200 °C using transmission electron microscopy (TEM) and nanoindentation. The size of the Frank loops increased and the density decreased with increasing irradiation temperature. Radiation-induced segregation (RIS) was detected across high-angle grain boundaries, and the degree of RIS increases with increasing irradiation temperature. Ni-Si clusters were observed using high-resolution TEM in the sample irradiated at 400 °C. The results of this work are compared with the literature data of self-ion and proton irradiation at comparable temperatures and damage levels on stainless steels with a similar material composition with this study. Despite the differences in dose rate, alloy composition and incident ion energy, the irradiation temperature dependence of RIS and the size and density of radiation defects followed the same trends, and were very comparable in magnitude.

Abrasive Wear of Alloyed Cast Steels Applied for Heavy Machinery

Directory of Open Access Journals (Sweden)

Studnicki A.

2015-03-01

Full Text Available In the paper the results and analysis of abrasive wear studies were shown for two grades of cast steels: low-alloyed cast steel applied for heavy machinery parts such as housing, covers etc. and chromium cast steels applied for kinetic nodes of pin-sleeve type. Studies were performed using the modified in Department of Foundry pin-on-disc method.

Cell cycle delays in synchronized cell populations following irradiation with heavy ions

International Nuclear Information System (INIS)

Scholz, M.

1992-11-01

Mammalian cells subjected to irradiation with heavy ions were investigated for cell cycle delays. The ions used for this purpose included Ne ions in the LET range of 400 keV/μm just as well as uranium ions of 16225 keV/μm. The qualitative changes in cell cycle progression seen after irradiation with Ne ions (400 keV/μm) were similar to those observed in connection with X-rays. Following irradiation with extremely heavy ions (lead, uranium) the majority of cells were even at 45 hours still found to be in the S phase or G 2 M phase of the first cycle. The delay cross section 'σ-delay' was introduced as a quantity that would permit quantitative comparisons to be carried out between the changes in cell progression and other effects of radiation. In order to evaluate the influence of the number of hits on the radiation effect observed, the size of the cell nucleus was precisely determined with reference to the cycle phase and local cell density. A model to simulate those delay effects was designed in such a way that account is taken of this probability of hit and that the results can be extrapolated from the delay effects after X-irradiation. On the basis of the various probabilities of hit for cells at different cycle stages a model was developed to ascertain the intensified effect following fractionated irradiation with heavy ions. (orig./MG) [de

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)

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

Microstructural evolution in austenitic stainless steel irradiated with triple-beam

Energy Technology Data Exchange (ETDEWEB)

Hamada, Shozo; Miwa, Yukio; Yamaki, Daiju [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yichuan, Zhang

1997-03-01

An austenitic stainless steel was simultaneously irradiated with nickel, helium and hydrogen ions at the temperature range of 573-673 K. The damage level and injected concentration of He and H ions in the triple-beam irradiated region are 57 dpa, 19000 and 18000 at.ppm, respectively. Following to irradiation, the cross sectional observation normal to the incident surface of the specimen was carried out with a transmission electron microscope. Two bands parallel to the incident surface were observed in the irradiated specimen, which consist of dislocation loops and lines of high number density. These locate in the range of the depth of 0.4 to 1.3 {mu}m and 1.8 to 2.4 {mu}m from the incident surface, respectively. The region between two bands, which corresponds to the triple beam irradiated region, shows very low number density of dislocations than that in each band. Observation with higher magnification of this region shows that fine cavities with high number density uniformly distribute in the matrix. (author)

Influence of the austenitic stainless steel microstructure on the void swelling under ion irradiation

Directory of Open Access Journals (Sweden)

Rouxel Baptiste

2016-01-01

Full Text Available To understand the role of different metallurgical parameters on the void formation mechanisms, various austenitic stainless steels were elaborated and irradiated with heavy ions. Two alloys, in several metallurgical conditions (15Cr/15Ni–Ti and 15Cr/25Ni–Ti, were irradiated in the JANNUS-Saclay facility at 600 °C with 2 MeV Fe2+ ions up to 150 dpa. Resulting microstructures were observed by Transmission Electron Microscopy (TEM. Different effects on void swelling are highlighted. Only the pre-aged samples, which were consequently solute and especially titanium depleted, show cavities. The nickel-enriched matrix shows more voids with a smaller size. Finally, the presence of nano-precipitates combined with a dense dislocation network decreases strongly the number of cavities.

Modeling of irradiation embrittlement and annealing/recovery in pressure vessel steels

International Nuclear Information System (INIS)

Lott, R.G.; Freyer, P.D.

1996-01-01

The results of reactor pressure vessel (RPV) annealing studies are interpreted in light of the current understanding of radiation embrittlement phenomena in RPV steels. An extensive RPV irradiation embrittlement and annealing database has been compiled and the data reveal that the majority of annealing studies completed to date have employed test reactor irradiated weldments. Although test reactor and power reactor irradiations result in similar embrittlement trends, subtle differences between these two damage states can become important in the interpretation of annealing results. Microstructural studies of irradiated steels suggest that there are several different irradiation-induced microstructural features that contribute to embrittlement. The amount of annealing recovery and the post-anneal re-embrittlement behavior of a steel are determined by the annealing response of these microstructural defects. The active embrittlement mechanisms are determined largely by the irradiation temperature and the material composition. Interpretation and thorough understanding of annealing results require a model that considers the underlying physical mechanisms of embrittlement. This paper presents a framework for the construction of a physically based mechanistic model of irradiation embrittlement and annealing behavior

The behaviour of ferritic steels under fast neutron irradiation

International Nuclear Information System (INIS)

Erler, J.; Maillard, A.; Brun, G.; Lehmann, J.; Dupouy, J.M.

1979-07-01

Ferritic steels have been irradiated in Rapsodie and Phenix to doses up to 150 dpa F. The swelling and irradiation creep characteristics and the mechanical properties of these materials are reported. (author)

Damage behavior in helium-irradiated reduced-activation martensitic steels at elevated temperatures

Energy Technology Data Exchange (ETDEWEB)

Luo, Fengfeng [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Chen, Jihong; Li, Tiecheng; Zheng, Zhongcheng [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Yao, Z. [Department of Mechanical and Materials Engineering, Queen’s University, Kingston K7L 3N6, ON (Canada); Suo, Jinping [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

2014-12-15

Dislocation loops induced by helium irradiation at elevated temperatures in reduced-activation martensitic steels were investigated using transmission electron microscopy. Steels were irradiated with 100 keV helium ions to 0.8 dpa between 300 K and 723 K. At irradiation temperatures T{sub irr} ⩽ 573 K, small defects with both Burger vectors b = 1/2〈1 1 1〉 and b = 〈1 0 0〉 were observed, while at T{sub irr} ⩾ 623 K, the microstructure was dominated by large convoluted interstitial dislocation loops with b = 〈1 0 0〉. Only small cavities were found in the steels irradiated at 723 K.

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

Genetic effects of heavy ion irradiation in maize and soybean

International Nuclear Information System (INIS)

Yatou, Osamu; Amano, Etsuo; Takahashi, Tan.

1992-01-01

Somatic mutation on leaves of maize and soybean were observed to investigate genetic effects of heavy ion irradiation. Maize seeds were irradiated with N, Fe and U ions and soybean seeds were irradiated with N ions. This is a preliminary report of the experiment, 1) to examine the mutagenic effects of the heavy ion irradiation, and 2) to evaluate the genetic effects of cosmic ray exposure in a space ship outside the earth. (author)

Structural characterization of swift heavy ion irradiated polycarbonate

International Nuclear Information System (INIS)

Singh, Lakhwant; Samra, Kawaljeet Singh

2007-01-01

Makrofol-N polycarbonate thin films were irradiated with copper (50 MeV) and nickel (86 MeV) ions. The modified films were analyzed by UV-VIS, FTIR and XRD techniques. The experimental data was used to evaluate the formation of chromophore groups (conjugated system of bonds), degradation cross-section of the special functional groups, the alkyne formation and the amorphization cross-section. The investigation of UV-VIS spectra shows that the formation of chromophore groups is reduced at larger wavelength, however its value increases with the increase of ion fluence. Degradation cross-section for the different chemical groups present in the polycarbonate chains was evaluated from the FTIR data. It was found that there was an increase of degradation cross-section of chemical groups with the increase of electronic energy loss in polycarbonate. The alkyne and alkene groups were found to be induced due to swift heavy ion irradiation in polycarbonate. The radii of the alkyne production of about 2.74 and 2.90 nm were deduced for nickel (86 MeV) and copper (50 MeV) ions respectively. XRD analysis shows the decrease of the main XRD peak intensity. Progressive amorphization process of Makrofol-N with increasing fluence was traced by XRD measurements

Evaluation of the cross-sections of threshold reactions leading to the production of long-lived radionuclides during irradiation of steels by thermonuclear spectrum neutrons

International Nuclear Information System (INIS)

Blokhin, A.I.; Buleeva, N.N.; Manokhin, V.N.; Mikhajlyukova, M.V.; Nasyrova, S.M.; Skripova, M.V.

2002-01-01

The present paper analyses and evaluates the cross-sections of threshold reactions leading to the production of long-lived radionuclides during the irradiation, by thermonuclear spectrum neutrons, of steels containing V, Ti, Cr, Fe and Ni. On the basis of empirical systematics. a new evaluation of the (n,2n), (n,p), (n,np), (n,α) and (n,nα) excitation functions is made for all isotopes of V, Ti, Cr, Fe and Ni and for intermediate isotopes produced in the chain from irradiated isotopes up to production of the long-lived radionuclides 39 Ar, 42 Ar, 41 Ca, 53 Mn, 60 Fe, 60 Co, 59 Ni and 63 Ni. A comparison is made with the experimental and other evaluated data. (author)

Hardness distribution and effect of irradiation in FSW-ODS ferritic steels

International Nuclear Information System (INIS)

Noh, Sanghoon; Kasada, Ryuta; Kimura, Akihiko; Nagasaka, Takuya; Sokolov, M.A.; Yamamoto, T.

2014-01-01

Oxide dispersion strengthened ferritic steels (ODS-FS) have been considered as one of the most promising structural materials for advanced nuclear systems such as fusion reactors and next generation fission reactors, because of its excellent elevated temperature strength, corrosion and radiation resistance. Especially, irradiation resistance is a critical issue for the high performance of ODS-FS. In this study, effects of the irradiation on hardness properties of friction stri processed (FSP) ODS-FS were investigated. FSP technique was employed on ODS-FS. A plate specimen was cut out from the cross section and irradiated to 1.2 dpa at 573K in the High Flux Isotope Reactor (HFIR). To investigate the effect of neutron irradiation on processed area, the hardness distributions were evaluated on the cross section. Hardness of FSP ODS-FS was various with each microstructure after irradiation to 1.2 dpa at 573K. The increase of Vickers hardness was significant in the stirred zone and heat affected zone. Base material exhibited the lowest hardening about 38HV. Since nano-oxide particles in stirred zone showed identical mean diameter and number density, it is considered that hardening differences between stirred zone and base material is due to differences in initial dislocation density. (author)

Study of irradiation effects in austenitic stainless steels

Energy Technology Data Exchange (ETDEWEB)

Etienne, A. [GPM UMR CNRS 6634, Universite et INSA de Rouen (France); Material Department, University of California, Santa Barbara (United States); Pareige, P.; Radiguet, B. [GPM UMR CNRS 6634, Universite et INSA de Rouen (France); Cunningham, N.J.; Odette, G.O. [Material Department, University of California, Santa Barbara (United States); Pokor, C. [EDF RD, departement MMC, site des Renardieres, Moret-sur-Loing (France)

2011-07-01

Chemical analyses using Atom Probe Tomography were performed on a bolt made of cold worked 316 austenitic stainless steel, extracted from the internal structures of a pressurized water reactor after seventeen years of reactor service. The irradiation temperature of these samples was 633 K and the irradiation dose was estimated to 12 dpa. These analyses have shown that neutron irradiation has a strong effect on the intragranular distribution of solute atoms. A very high number density (6.10{sup 23} m{sup -3}) of Ni-Si enriched and Cr-Fe depleted clusters was detected after irradiation. In order to bring complementary experimental results and to determine the mechanism of formation of these Ni-Si nano-clusters, Fe{sup 5+} ion irradiations have been performed on a 316 austenitic stainless steel. As after neutron irradiation, the formation of solute enriched features is observed. Linear features and two kinds of clusters, rounded and torus shaped, are present. Considering that solute enriched features are probably formed by radiation induced segregation on point defect sinks, these different shapes are due to the nature of the sinks where segregation occurs. (authors)

The effects of fast-neutron irradiation on the mechanical properties of austenitic stainless steel

International Nuclear Information System (INIS)

Dalton, J.H.

1978-01-01

The paper reviews the effects of fast-neutron irradiation on the tensile properties of austenitic stainless steels at irradiation temperatures of less than 400 degrees Celcius, using as an example, work carried out at Pelindaba on an AISI 316 type steel produced in South Africa. Damage produced in these steels at higher irradiation temperatures and fluences is also briefly discussed. The paper concludes with a discussion of some methods of overcoming or decreasing the effects of irradiation damage [af

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.

Investigation of neutron irradiated reactor vessel steels using post-irradiation annealing techniques

Energy Technology Data Exchange (ETDEWEB)

Nakata, Hayato; Fukuya, Koji [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

The matrix damage is known to be a major factor that contributes to embrittlement and hardening of irradiated reactor vessel steels, and is assumed to be composed of the point defect clusters. However field emission gun scanning transmission electron microscopy (FEGSTEM) and atom probe (AP) could not detect any evidence of the matrix damage. In this study, post irradiation annealing experiments combining positron annihilation lineshape analysis (PALA) and hardness experiments were applied to an actual surveillance test specimen and a sample of reactor vessel steel irradiated in a material test reactor (MTR), in order to investigate the matrix damage recovery behavior and its contribution to hardening. It was confirmed that higher fluence increased the hardness and the volume fraction of open volume defects and that higher flux decreased the thermal stability of matrix damage and the effect on hardening. The contribution of matrix damage to hardening could be estimated to be below 30%. (author)

Damage induced in semiconductors by swift heavy ion irradiation

International Nuclear Information System (INIS)

Levalois, M.; Marie, P.

1999-01-01

The behaviour of semiconductors under swift heavy ion irradiation is different from that of metals or insulators: no spectacular effect induced by the inelastic energy loss has been reported in these materials. We present here a review of irradiation effects in the usual semiconductors (silicon, germanium and gallium arsenide). The damage is investigated by means of electrical measurements. The usual mechanisms of point defect creation can account for the experimental results. Besides, some results obtained on the wide gap semiconductor silicon carbide are reported. Concerning the irradiation effects induced by heavy ions in particle detectors, based on silicon substrate, we show that the deterioration of the detector performances can be explained from the knowledge of the substrate properties which are strongly perturbed after high doses of irradiation. Finally, some future ways of investigation are proposed. The silicon substrate is a good example to compare the irradiation effects with different particles such as electrons, neutrons and heavy ions. It is then necessary to use parameters which account for the local energy deposition, in order to describe the damage in the material

Progress and tendency in heavy ion irradiation mutation breeding

International Nuclear Information System (INIS)

Zhou Libin; Li Wenjian; Qu Ying; Li Ping

2008-01-01

In recent years, the intermediate energy heavy ion biology has been concerned rarely comparing to that of the low-energy ions. In this paper, we summarized the advantage of a new mutation breeding method mediated by intermediate energy heavy ion irradiations. Meanwhile, the present state of this mutation technique in applications of the breeding in grain crops, cash crops and model plants were introduced. And the preview of the heavy ion irradiations in gene-transfer, molecular marker assisted selection and spaceflight mutation breeding operations were also presented. (authors)

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

Positron annihilation lifetime measurements of austenitic stainless and ferritic/martensitic steels irradiated in the SINQ target irradiation program

Science.gov (United States)

Sato, K.; Xu, Q.; Yoshiie, T.; Dai, Y.; Kikuchi, K.

2012-12-01

Titanium-doped austenitic stainless steel (JPCA) and reduced activated ferritic/martensitic steel (F82H) irradiated with high-energy protons and spallation neutrons were investigated by positron annihilation lifetime measurements. Subnanometer-sized (steel, the positron annihilation lifetime of the bubbles decreased with increasing irradiation dose and annealing temperature because the bubbles absorb additional He atoms. In the case of JPCA steel, the positron annihilation lifetime increased with increasing annealing temperature above 773 K, in which case the dissociation of complexes of vacancy clusters with He atoms and the growth of He bubbles was detected. He bubble size and density were also discussed.

Pressure vessel steels: influence of chemical composition on irradiation sensitivity

International Nuclear Information System (INIS)

Ghoniem, M.M.; Hammad, F.H.

1998-01-01

Neutron irradiation of the steels used in the construction of the nuclear reactor pressure vessels can lead to the embrittlement of these materials, increasing the ductile-to-brittle transition temperature and decreasing the fracture energy, which can limit the plant life. The knowledge of irradiation embrittlement and the means for minimizing such degradation is therefore important in the field of assuring the safety of the nuclear power plants. Irradiation embrittlement is quite a complex process. It involves many variables. The most important of these are irradiation temperature, neutron fluence (neutron dose), neutron flux (neutron dose rate), and chemical composition of the irradiated material. This paper is concerned with the effect of chemical composition, the role of residual and alloying elements in the irradiation embrittlement of nuclear reactor pressure vessel steels in light water reactors. It presents a critical review for the published work in this field through the last 25 years

Response of neutron-irradiated RPV steels to thermal annealing

International Nuclear Information System (INIS)

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

1997-01-01

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the fracture toughness properties that have been degraded by neutron irradiation. This paper summarizes experimental results of work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response of several irradiated RPV steels

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

Ductile fracture toughness of heavy section pressure vessel steel plate. A specimen-size study of ASTM A 533 steels

International Nuclear Information System (INIS)

Williams, J.A.

1979-09-01

The ductile fracture toughness, J/sub Ic/, of ASTM A 533, Grade B, Class 1 and ASTM A 533, heat treated to simulate irradiation, was determined for 10- to 100-mm thick compact specimens. The toughness at maximum specimen load was also measured to determine the conservatism of J/sub Ic/. The toughness of ASTM A 533, Grade B, Class 1 steel was 349 kJ/m 2 and at the equivalent upper shelf temperature, the heat treated material exhibited 87 kJ/m 2 . The maximum load fracture toughness was found to be linearly proportional to specimen size, and only specimens which failed to meet ASTM size criteria exhibited maximum load toughness less than J/sub Ic/

The coarsening effect of SA508-3 steel used as heavy forgings material

Directory of Open Access Journals (Sweden)

Dingqian Dong

2015-01-01

Full Text Available SA508Gr.3 steel is popularly used to produce core unit of nuclear power reactors due to its outstanding ability of anti-neutron irradiation and good fracture toughness. The forging process takes important role in manufacturing to refine the grain size and improve the material properties. But due to their huge size, heavy forgings cannot be cooled down quickly, and the refined grains usually have long time to grow in high temperature conditions. If the forging process is not adequately scheduled or implemented, very large grains up to millimetres in size may be found in this steel and cannot be eliminated in the subsequent heat treatment. To fix the condition which may causes the coarsening of the steel, hot upsetting experiments in the industrial production environment were performed under different working conditions and the corresponding grain sizes were measured and analysed. The observation showed that the grain will abnormally grow if the deformation is less than a critical value. The strain energy takes a critical role in the grain evolution. If dynamic recrystallization consumes the strain energy as much as possible, the normal grains will be obtained. While if not, the stored strain energy will promote abnormal growth of the grains.

Soil Contamination with Heavy Metals around Jinja Steel Rolling Mills in Jinja Municipality, Uganda

Directory of Open Access Journals (Sweden)

Noel Namuhani

2015-01-01

Conclusions. The concentration levels of heavy metals around the steel rolling mills did not appear to be of serious concern, except for copper and cadmium, which showed moderate pollution and moderate to strong pollution, respectively. All heavy metals were within the limits of the United States Environmental Protection Agency (USEPA residential soil standards and the Dutch intervention soil standards. Overall, soils around the Jinja steel rolling mills were slightly polluted with heavy metals, and measures therefore need to be taken to prevent further soil contamination with heavy metals.

Martensitic transformation induced by irradiation and deformation in stainless steels

International Nuclear Information System (INIS)

Maksimkin, O.P.

1997-01-01

In the present work the peculiarities of martensite γ → α , (γ → ε → α , ) transformation in the steels with a low stacking fault energy (12Cr18Ni10T, Cr15AG14) irradiated by neutrons, α-particles and electrons (pulse and stationary) and then deformed with the various strain rates in the temperature range - 20 - 1000 C are considered. It is established by the electron-microscope research that the phase γ → α ' transition in irradiated and deformed steels is observed on the definite stage of evolution of the dislocation structure (after the cell formation) and the martensite formation preferentially occurs on a stacking fault aggregation. The regularities of the irradiation by high energy particles effect on the formation parameters and martensite α , -phase accumulation kinetics ones and also their role in forming of the strength and ductile properties in steels are analysed. (A.A.D.)

Effects of heavy particle irradiation on central nervous system

International Nuclear Information System (INIS)

Nojima, Kumie; Nakadai, Taeko; Khono, Yukio

2006-01-01

Effects of low dose heavy particle radiation to central nervous system were studied using human embryonal carcinoma (Ntera2=NT2) and Human neuroblastoma cell (NB1). They exposed to heavy ions and X ray 80% confluent cells in culture bottles. The cells were different type about growth and differentiation in the neuron. The apoptosis profile was measured by AnnexinV-EGFP, PI stained and fluorescence-activated cell sorter (FACS). Memory and learning function of adult mice were studied using water maze test after carbon- or iron-ion irradiation. Memory functions were rapidly decreased after irradiation both ions. Iron -ion group were recovered 20 weeks after irradiation C-ion group were recovered 25 weeks after irradiation. Tier memory were still keep at over 100 weeks after irradiation. (author)

Radiation induced microstructural evolution in ferritic/martensitic steels

International Nuclear Information System (INIS)

Kohno, Y.; Kohyama, A.; Asakura, K.; Gelles, D.S.

1993-01-01

R and D of ferritic/martensitic steels as structural materials for fusion reactor is one of the most important issues of fusion technology. The efforts to characterize microstructural evolution under irradiation in the conventional Fe-Cr-Mo steels as well as newly developed Fe-Cr-Mn or Fe-Cr-W low activation ferritic/ martensitic steels have been continued. This paper provides some of the recent results of heavy irradiation effects on the microstructural evolution of ferritic/martensitic steels neutron irradiated in the FFTF/MOTA (Fast Flux Test Facility/Materials Open Test Assembly). Materials examined are Fe-10Cr-2Mo dual phase steel (JFMS: Japanese Ferritic/Martensitic Steel), Fe-12Cr-XMn-1Mo manganese stabilized martensitic steels and Fe-8Cr-2W Tungsten stabilized low activation martensitic steel (F82H). JFMS showed excellent void swelling resistance similar to 12Cr martensitic steel such as HT-9, while the manganese stabilized steels and F82H showed less void swelling resistance with small amount of void swelling at 640-700 K (F82H: 0.14% at 678 K). As for irradiation response of precipitate behavior, significant formation of intermetallic χ phase was observed in the manganese stabilized steels along grain boundaries which is though to cause mechanical property degradation. On the other hand, precipitates identified were the same type as those in unirradiated condition in F82H with no recognition of irradiation induced precipitates, which suggested satisfactory mechanical properties of F82H after the irradiation. (author)

Brittle and ductile rupture of 16MND5 steel. Irradiation effect

International Nuclear Information System (INIS)

Al Mundheri, M.; Soulat, P.; Pineau, A.

1986-06-01

Toughness tests have been made on 16MND5 steel (A508Cl3 steel) - before and after irradiation at 290 0 C (3.10 19 n/cm 2 , E > 1 MeV). It is shown that toughness is lowered following the irradiation and that it is a decreasing function of the thickness of the test pieces. In parallel, tests on three geometries of entailed specimens, prepared in the non-irradiated material, have been made at different temperatures to apply the methodology of local approach of ductile-brittle rupture [fr

Study of Irradiation Effects on the Fracture Properties of A533-Series Ferritic Steels

International Nuclear Information System (INIS)

Lee, Yong Bok; Lee, Gyeong Geun; Kwon, Jun Hyun

2011-01-01

Since the Kori nuclear power plant unit 3 (Kori-3) was founded in 1986, the surveillance tests have been conducted five times. One of the primary objectives of the surveillance test is to determine the effects of irradiation on reactor pressure vessel (RPV) steel embrittlement. The RPV is made out of ferritic steels such as SA533 type B class 1, which were used for early nuclear power plants industry including Kori-2, 3, 4 and Yonggwang-1, 2 units in Korea. The Westinghouse supplied Kori-3 with the RPV steels ASTM A533 grade B class 1, which is equivalent to SA533 type B class 1. The irradiation effects on tensile properties in ASTM A533 grade B class 1 steel had been studied by Steichen and Williams. They experimentally determined the effect of strain rate and temperature on the tensile properties of unirradiated and irradiated A533 grade B steel 1. The effects of neutron irradiation on ferritic steels could be determined from tensile properties, as well as the fracture strength and toughness measurements. Hunter and Williams have reported that the strength and ductility for unirradiated material at a low strain rate increase with decreasing test temperature. Also, neutron irradiation increases strength and decreases ductility. Crosley and Ripling revealed that the yield strength of unirradiated material rapidly increases with the strain rate. Therefore, yield strength for unirradiated and irradiated materials should be determined by test parameters along with strain rate and temperature. In this study we compare ASTM A533 grad B class 1 steel obtained from several papers with SA533 type B class 1 steel taken from the surveillance data of Kori-3 unit, whose mechanical property of unirradiated and irradiated materials was correlated with the rate-temperature parameter

Development of an intermediate energy heavy-ion micro-beam irradiation system

International Nuclear Information System (INIS)

Song Mingtao; Wang Zhiguang; He Yuan; Gao Daqing; Yang Xiaotian; Liu Jie; Su Hong; Man Kaidi; Sheng Li'na

2008-01-01

The micro-beam irradiation system, which focuses the beam down the micron order and precisely delivers a predefined number of ions to a predefined spot of micron order, is a powerful tool for radio-biology, radio-biomedicine and micromachining. The Institute of Modern Physics of Chinese Academy of Sciences is developing a heavy-ion micro-beam irradiation system up to intermediate energy. Based on the intermediate and low energy beam provided by Heavy Ion Research Facility of Lanzhou, the micro-beam system takes the form of the magnetic focusing. The heavy-ion beam is conducted to the basement by a symmetrical achromatic system consisting of two vertical bending magnets and a quadrupole in between. Then a beam spot of micron order is formed by magnetic triplet quadrupole of very high gradient. The sample can be irradiated either in vacuum or in the air. This system will be the first opening platform capable of providing heavy ion micro-beam, ranging from low (10 MeV/u) to intermediate energy (100 MeV/u), for irradiation experiment with positioning and counting accuracy. Target material may be biology cell, tissue or other non-biological materials. It will be a help for unveiling the essence of heavy-ion interaction with matter and also a new means for exploring the application of heavy-ion irradiation. (authors)

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.

Effect of cold work on tensile behavior of irradiated type 316 stainless steel

International Nuclear Information System (INIS)

Klueh, R.L.; Maziasz, P.J.

1986-01-01

Tensile specimens were irradiated in ORR at 250, 290, 450, and 500 0 C to produce a displacement damage of approx.5 dpa and 40 at. ppM He. Irradiation at 250 and 290 0 C caused an increase in yield stress and ultimate tensile strength and a decrease in ductility relative to unaged and thermally aged controls. The changes were greatest for the 20%-cold-worked steel and lowest for the 50%-cold-worked steel. Irradiation at 450 0 C caused a slight relative decrease in strength for all cold-worked conditions. A large decrease was observed at 500 0 C, with the largest decrease occurring for the 50%-cold-worked specimen. No bubble, void, or precipitate formation was observed for specimens examined by transmission electron microscopy (TEM). The irradiation hardening was correlated with Frank-loop and ''black-dot'' loop damage. A strength decrease at 500 0 C was correlated with dislocation network recovery. Comparison of tensile and TEM results from ORR-irradiated steel with those from steels irradiated in the High Flux Isotope Reactor and the Experimental Breeder Reactor indicated consistent strength and microstructure changes

Material property changes of stainless steels under PWR irradiation

International Nuclear Information System (INIS)

Fukuya, Koji; Nishioka, Hiromasa; Fujii, Katsuhiko; Kamaya, Masayuki; Miura, Terumitsu; Torimaru, Tadahiko

2009-01-01

Structural integrity of core structural materials is one of the key issues for long and safe operation of pressurized water reactors. The stainless steel components are exposed to neutron irradiation and high-temperature water, which cause significant property changes and irradiation assisted stress corrosion cracking (IASCC) in some cases. Understanding of irradiation induced material property changes is essential to predict integrity of core components. In the present study, microstructure and microchemistry, mechanical properties, and IASCC behavior were examined in 316 stainless steels irradiated to 1 - 73 dpa in a PWR. Dose-dependent changes of dislocation loops and cavities, grain boundary segregation, tensile properties and fracture mode, deformation behavior, and their interrelation were discussed. Tensile properties and deformation behavior were well coincident with microstructural changes. IASCC susceptibility under slow strain rate tensile tests, IASCC initiation under constant load tests in simulated PWR primary water, and their relationship to material changes were discussed. (author)

Microstructure and grain size effects on irradiation hardening of low carbon steel for reactor tanks

International Nuclear Information System (INIS)

Milasin, N.

1964-05-01

Irradiation hardening of steel for reactor pressure vessels has been studied extensively during the past few years. A great number of experimental results concerning the behaviour of these steels in the radiation field and several review papers (1,2) have been published. Most of the papers deal with the effects of specific metallurgical factors or irradiation conditions (temperature, flux) on irradiation hardening and embrittlement. In addition, a number of experiments are performed to give evidence on the mechanism of irradiation hardening of these steels. However, this mechanism is still unknown due to the complexity of steel as a system. Among different methods used in radiation damage studies, the changes of mechanical properties have been mainly investigated. By using Hall-Petch's empirical relation, σ y =σ i +k y d -1/2 between lower yield stress, σ y , and grain size, 2d, the information about the effect of irradiation on the parameters σ i and k y is obtained. Taking as a base interpretation of σ i and k y given by Petch and his co-workers it has been concluded that radiation does not change the stress to start slip but that it increase the friction that opposes the passage of free dislocations across a slip plane. In attempting to apply Hall-Petch's relation to one unirradiated ferritic steel with a carbon content higher than 0.15% some difficulties were encountered. The results obtained indicate that the influence of grain size can not be isolated from other factors introduced by the treatments used to produce different grain sizes. This paper deals with a similar problem in the case of irradiated steel. The results obtained give the changes of the mechanical properties of steel in neutron irradiation field as a function of microstructure and grain size. In addition, the mechanical properties of irradiated steel are measured after annealing at 150 deg C and 450 deg C. On the basis of the experimental results obtained the relative microstructure and

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.

Theoretical description of the influence of neutron irradiation on viscoplastic properties of mild steel

International Nuclear Information System (INIS)

Pecherski, R.

1978-01-01

The physical bases of plastic deformation of mild steel are described. The influence of neutron irradiation on the change of mechanisms of plastic deformation is discussed in detail. Constitutive equations of viscoplasticity for irradiated mild steel are given. The problem of thickwalled viscoplastic spherical tank irradiated by neutrons is studied. (Z.R.)

Irradiation behavior of German PWR RPV steels under operating conditions

Energy Technology Data Exchange (ETDEWEB)

May, J.; Hein, H. [AREVA NP Gmbh (Germany); Ganswind, J. [VGB PowerTech e.V. (Germany); Widera, M. [RWE Power AG (Germany)

2011-07-01

In 2007, the last standard surveillance capsule of the original RPV (Reactor Pressure Vessel) surveillance programs of the 11 currently operating German PWR has been evaluated. With it the standard irradiation surveillance programs of these plants was completed. In the present paper, irradiation data of these surveillance programs will be presented and a final assessment of the irradiation behavior of the German PWR RPV steels with respect to current standards KTA 3203 and Reg. Guide 1.99 Rev. 2 will be given. Data from two units which are currently under decommissioning will also be included, so that data from all 13 German PWR manufactured by the former Siemens/KWU company (now AREVA NP GmbH) are shown. It will be shown that all surveillance data within the approved area of chemical composition verify the limit curve RT(limit) of the KTA 3203, which is the relevant safety standard for these plants. An analysis of the data shows, that the prediction formulas of Reg. Guide 1.99 Rev. 2 Pos. 1 or from the TTS model tend to overestimate the irradiation behavior of the German PWR RPV steels. Possible reasons for this behavior are discussed. Additionally, the data will be compared to data from the research project CARISMA to demonstrate that these data are representative for the irradiation behavior of the German PWR RPV steels. Since the data of these research projects cover a larger neutron fluence range than the original surveillance data, they offer a future outlook into the irradiation behavior of the German PWR RPV steels under long term conditions. In general, as a consequence of the relatively large and beneficial water gap between core and RPV, especially in all Siemens/KWU 4-loop PWR, the EOL neutron fluence and therefore the irradiation induced changes in mechanical properties of the German PWR RPV materials are rather low. Moreover the irradiation data indicate that the optimized RPV materials specifications that have been applied in particular for the

Heavy irradiation effects in radiation-resistant optical fibers

Energy Technology Data Exchange (ETDEWEB)

Shikama, Tatsuo [Tohoku Univ., Oarai, Ibaraki (Japan). Oarai Branch, Inst. for Materials Research

1998-07-01

Development of a system for optical measurements in a nuclear reactor has been progressing to investigate dynamic changes in a material caused by heavy irradiation. In such system, transfer of optical signals to out-pile measuring systems is being attempted by the use of optical fibers. In this report, the characteristics of optical fibers in the heavy irradiation field were summarized. It has been known that amorphous silica might produce radiolysis and structural defects by the exposure to ionizing radiation. The effects of heavy irradiation on molten silica were extremely complicated. A large intensity of visible light absorption occurred from an early time during start-up of the reactor. The absorption range was limited below 700 nm for the radiation associating fast neutron and the absorption was mostly attributed to non-bridging oxygen hole center. The depletion of optical transferring capacity under the radiation might be related to the internal stress. Therefore, it seems desirable to use optical fibers in the conditions without leading too much stress. (M.N.)

Irradiation Creep of Ferritic-Martensitic Steels EP-450, EP-823 and EI-852 Irradiated in the BN-350 Reactor over Wide Ranges of Irradiation Temperature and Dose

International Nuclear Information System (INIS)

Porollo, S.I.; Konobeev, Y.V.; Ivanov, A.A.; Shulepin, S.V.; Garner, F.

2007-01-01

Full text of publication follows: Ferritic/martensitic (F/M) steels appear to be the most promising materials for advanced nuclear systems, especially for fusion reactors. Their main advantages are higher resistance to swelling and lower irradiation creep rate as has been repeatedly demonstrated in examinations of these materials after irradiation. Nevertheless, available experimental data on irradiation resistance of F/M steels are insufficient, with the greatest deficiency of data for high doses and for both low and high irradiation temperatures. From the very beginning of operation the BN-350 fast reactor has been used for irradiation of specimens of structural materials, including F/M steels. The most unique feature of BN-350 was its low inlet sodium temperature, allowing irradiation at temperatures over a very wide range of temperatures compared with the range in other fast reactors. In this paper data are presented on swelling and irradiation creep of three Russian F/M steels EP-450, EP-823 and EI-852, irradiated in experimental assemblies of the BN-350 reactor at temperatures in the range of 305-700 deg. C to doses ranging from 20 to 89 dpa. The investigation was performed using gas-pressurized creep tubes with hoop stresses in the range of 0 - 294 MPa. (authors)

Precipitation response of austenitic stainless steel to simulated fusion irradiation

International Nuclear Information System (INIS)

Maziasz, P.J.

1978-01-01

The precipitation response of annealed type 316 stainless steel irradiated in HFIR is studied and compared to previously observed thermal aging and fast reactor irradiation responses. Irradiation in HFIR simultaneously produces high levels of helium and displacement damage and partially simulates a fusion environment. Samples have been irradiated at temperatures from 550 to 680 0 C to fluences producing up to 3300 appm He and 47 dpa

Effects of irradiation of sewage sludge on heavy metal bioavailability

International Nuclear Information System (INIS)

Sheppard, S.C.; Mayoh, K.R.

1986-10-01

Sewage sludges are a valuable resource to agriculture, but their use is limited by the hazards of pathogens, toxic chemicals and heavy metals. Irradiation can control the pathogens and deactivate some of the toxic chemicals. The relative cost of industrial-scale irradiation using accelerators has decreased progressively. This, coupled with the increasing necessity to recycle wastes, has led to renewed interest in irradiation of sludges. In response to this renewed interest, this report examines what is known about the effects of irradiation on the bioavailability of heavy metals. Very few studies have addressed this topic, although workers in the U.S. have claimed decreased solubility of metals in irradiated sludges. We have also briefly reviewed the general literature on sludge to gain indirect evidence on the likely effects. The scant data, often based on less than ideal experimental methodologies, show no major consistent effects of irradiation on the availability of heavy metals from sludge. The data are not sufficient to rule out such effects entirely, but the effects appear to be fairly subtle and not likely to persist beyond one growth season. 85 refs

Precipitation behavior in austenitic and ferritic steels during fast neutron irradiation and thermal aging

International Nuclear Information System (INIS)

Kawanishi, H.; Hajima, R.; Sekimura, N.; Arai, Y.; Ishino, S.

1988-01-01

Precipitation behavior has been studied using a carbon extraction replica technique in Ti-modified Type 316 stainless steels (JPCA-2) and 9Cr-2Mo ferritic/martensitic steels (JFMS) irradiated to 8.1x10 24 n/m 2 at 873 and 673 K, respectively, in the experimental fast breeder reactor JOYO. Precipitate identification and compositional analysis were carried out on extracted replicas. The results were compared to those from the as-received steel and a control which had been given the same thermal as-treatment as the specimens received during irradiations. Carbides, Ti-sulphides and phosphides were precipitated in JPCA-2. Precipitate observed in JFMS included carbides, Laves-phases and phosphides. The precipitates in both steels were concluded to be stable under irradiation except for MC and M 6 C in JPCA-2. Small MC particles were found precipitated in JPCA-2 during both irradiation and aging. Irradiation proved to promote the precipitation of M 6 C in JPCA-2. (orig.)

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

Irradiation-induced creep in 316 and 304L stainless steels

International Nuclear Information System (INIS)

Walters, L.C.; McVay, G.L.; Hudman, G.D.

1977-01-01

Recent results are presented from the in-reactor creep experiments that are being conducted by Argonne National Laboratory. The experiments consist of four subassemblies that contain helium-pressurized as well as unstressed capsules of 316 and 304L stainless steels in several metallurgical conditions. Experiments are being irradiated in row 7 of the EBR-II sodium-cooled fast breeder reactor. Three of the subassemblies are being irradiated at temperatures near 400 0 C, and the fourth subassembly is being irradiated at a temperature of 550 0 C. Creep and swelling strains were determined by profilometer measurements on the full length of the capsules after each irradiation cycle. The accumulated neutron dose on the 304L capsules at 385 0 C was 45 dpa; on the 316 capsules at 400 0 C, 40 dpa; and on the 316 capsules at 550 0 C, 25 dpa. It was found that the in-reactor creep rates were linearly dependent on hoop stress, with the exception being capsules of 316 stainless steel that had been given long-term carbide aging treatment and then irradiated at 550 0 C. Those capsules exhibited much higher creep and swelling rates than their unaged counterparts. For the metallurgical conditions where significant swelling was observed (solution-annealed 304L and aged 316 stainless steels), it was found that the in-reactor creep rates were readily fit to a model that related the creep rates to accumulated swelling. Additionally, it was found that the stress-normalized creep rate for 20%-cold-worked 316 stainless steel at a temperature of 550 0 C was 1.6 times that observed at 400 0 C

Microstructure and grain size effects on irradiation hardening of low carbon steel for reactor tanks

Energy Technology Data Exchange (ETDEWEB)

Milasin, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1964-05-15

Irradiation hardening of steel for reactor pressure vessels has been studied extensively during the past few years. A great number of experimental results concerning the behaviour of these steels in the radiation field and several review papers (1,2) have been published. Most of the papers deal with the effects of specific metallurgical factors or irradiation conditions (temperature, flux) on irradiation hardening and embrittlement. In addition, a number of experiments are performed to give evidence on the mechanism of irradiation hardening of these steels. However, this mechanism is still unknown due to the complexity of steel as a system. Among different methods used in radiation damage studies, the changes of mechanical properties have been mainly investigated. By using Hall-Petch's empirical relation, {sigma}{sub y}={sigma}{sub i}+k{sub y} d{sup -1/2} between lower yield stress, {sigma}{sub y}, and grain size, 2d, the information about the effect of irradiation on the parameters {sigma}{sub i} and k{sub y} is obtained. Taking as a base interpretation of {sigma}{sub i} and k{sub y} given by Petch and his co-workers it has been concluded that radiation does not change the stress to start slip but that it increase the friction that opposes the passage of free dislocations across a slip plane. In attempting to apply Hall-Petch's relation to one unirradiated ferritic steel with a carbon content higher than 0.15% some difficulties were encountered. The results obtained indicate that the influence of grain size can not be isolated from other factors introduced by the treatments used to produce different grain sizes. This paper deals with a similar problem in the case of irradiated steel. The results obtained give the changes of the mechanical properties of steel in neutron irradiation field as a function of microstructure and grain size. In addition, the mechanical properties of irradiated steel are measured after annealing at 150 deg C and 450 deg C. On the basis of

Irradiation-induced precipitates in a neutron irradiated 304 stainless steel studied by three-dimensional atom probe

Energy Technology Data Exchange (ETDEWEB)

Toyama, T., E-mail: ttoyama@imr.tohoku.ac.jp [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Narita-cho 2145-2, Oarai, Ibaraki 311-1313 (Japan); Nozawa, Y. [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Narita-cho 2145-2, Oarai, Ibaraki 311-1313 (Japan); Van Renterghem, W. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, 2400 Mol (Belgium); Matsukawa, Y.; Hatakeyama, M.; Nagai, Y. [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Narita-cho 2145-2, Oarai, Ibaraki 311-1313 (Japan); Al Mazouzi, A. [EDF R and D, Avenue des Renardieres Ecuelles, 77818 Moret sur Loing Cedex (France); Van Dyck, S. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, 2400 Mol (Belgium)

2011-11-15

Highlights: > Irradiation-induced precipitates in a 304 stainless steel were investigated by three-dimensional atom probe. > The precipitates were found to be {gamma}' precipitates (Ni{sub 3}Si). > Post-irradiation annealing was performed to discuss the contribution of the precipitates to irradiation-hardening. - Abstract: Irradiation-induced precipitates in a 304 stainless steel, neutron-irradiated to a dose of 24 dpa at 300 deg. C in the fuel wrapper plates of a commercial pressurized water reactor, were investigated by laser-assisted three-dimensional atom probe. A high number density of 4 x 10{sup 23} m{sup -3} of Ni-Si rich precipitates was observed, which is one order of magnitude higher than that of Frank loops. The average diameter was {approx}10 nm and the average chemical composition was 40% Ni, 14% Si, 11% Cr and 32% Fe in atomic percent. Over a range of Si concentrations, the ratio of Ni to Si was {approx}3, close to that of {gamma}' precipitate (Ni{sub 3}Si). In some precipitates, Mn enrichment inside the precipitate and P segregation at the interface were observed. Post-irradiation annealing was performed to discuss the contribution of the precipitates to irradiation-hardening.

Dose dependence of the microstructural evolution in neutron-irradiated austenitic stainless steel

International Nuclear Information System (INIS)

Zinkle, S.J.; Maziasz, P.J.; Stoller, R.E.

1993-01-01

Microstructural data on the evolution of the dislocation loop, cavity, and precipitate populations in neutron-irradiated austenitic stainless steels are reviewed in order to estimate the displacement damage levels needed to achieve the 'steady state' condition. The microstructural data can be conveniently divided into two temperature regimes. In the low temperature regime (below about 200 degrees C) the microstructure of austenitic stainless steel is dominated by 'black spot' defect clusters and faulted interstitial dislocation loops. The dose needed to approach saturation of the loop and defect cluster densities is generally on the order of 1 displacement per atom (dpa) in this regime. In the high temperature regime (∼300 to 700 degrees C), cavities, precipitates, loops and network dislocations are all produced during irradiation; doses in excess of 10 dpa are generally required to approach a 'steady state' microstructural condition. Due to complex interactions between the various microstructural components that form during irradiation, a secondary transient regime is typically observed in commercial stainless steels during irradiation at elevated temperatures. This slowly evolving secondary transient may extend to damage levels in excess of 50 dpa in typical 300-series stainless steels, and to >100 dpa in radiation-resistant developmental steels. The detailed evolution of any given microstructural component in the high-temperature regime is sensitive to slight variations in numerous experimental variables, including heat-to-heat composition changes and neutron spectrum

Evaluation of the cross-sections of threshold reactions leading to the production of long-lived radionuclides during irradiation of steels by thermonuclear spectrum neutrons

CERN Document Server

Blokhin, A I; Manokhin, V N; Mikhajlyukova, M V; Nasyrova, S M; Skripova, M V

2001-01-01

The present paper analyses and evaluates the cross-sections of threshold reactions leading to the production of long-lived radionuclides during the irradiation, by thermonuclear spectrum neutrons, of steels containing V, Ti, Cr, Fe and Ni. On the basis of empirical systematics. a new evaluation of the (n,2n), (n,p), (n,np), (n,alpha) and (n,n alpha) excitation functions is made for all isotopes of V, Ti, Cr, Fe and Ni and for intermediate isotopes produced in the chain from irradiated isotopes up to production of the long-lived radionuclides sup 3 sup 9 Ar, sup 4 sup 2 Ar, sup 4 sup 1 Ca, sup 5 sup 3 Mn, sup 6 sup 0 Fe, sup 6 sup 0 Co, sup 5 sup 9 Ni and sup 6 sup 3 Ni. A comparison is made with the experimental and other evaluated data.

Electron-microscopic investigation of a pressure vessel steel after neutron irradiation

International Nuclear Information System (INIS)

Klaar, H.J.

1975-01-01

As an introduction, changes in the mechanical properties of pressure vessel steels on neutron irradiation and the causes of radiation embrittlement are discussed. After this, the author describes his own experiments with steel of the composition 0.19% C; 3.88% Ni; 1.57% Cr; 0.51% Mo; 0.2% V. Samples of this material were irradiated in-pile at 300 0 C with various neutron doses. To study the influence of neutron dose, irradiation temperature, and heat treatment on the mechanical properties, tensile tests, notched bar impact bending tests, hardness tests and structural analyses were carried out. The findings are reported. (GSC) [de

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

Irradiation effects of 11 MeV protons on ferritic steels

International Nuclear Information System (INIS)

Hamaguchi, Yoshikazu; Kuwano, Hisashi; Misawa, Toshihei

1985-01-01

It is considered that ferritic/martensitic steels are the candidate of the first wall materials for future fusion reactors. The most serious problem in the candidate materials is the loss of ductility due to the elevation of ductile-brittle transition temperature by the high dpa irradiation of neutrons. 14 MeV neutrons produced by D-T reaction cause high dpa damage and also produce large quantity of helium and hydrogen atoms in first wall materials. Those gas atoms also play an important role in the embrittlement of steels. The main purpose of this work was to simulate the behavior of hydrogen produced by the transmutation in the mechanical properties of ferritic steels when they were irradiated with 11 MeV protons. The experimental procedure and the results of hardness, the broadening of x-ray diffraction lines, Moessbauer spectroscopy and small punch test are reported. High energy protons of 10 - 20 MeV are suitable to the simulation experiment of 14 MeV neutron radiation damage. But the production of the active nuclei emitting high energy gamma ray and having long life, Co-56, is the most serious problem. Another difficulty is the control of irradiation temperature. A small irradiation chamber must be developed. (Kako, I.)

Study of 316 stainless steel swelling due to neutron irradiation

International Nuclear Information System (INIS)

Furutani, Gen; Konishi, Takao

2000-01-01

Large stresses will be generated in the austenitic stainless steel core internals of pressurized water reactors (PWRs) if excessive swelling occurs after long periods of operation. As a result, deformation or stress corrosion cracking (SCC) could occur in the core internals. However, data on the swelling of irradiated austenitic stainless steel in actual PWRs is limited. In this study, mechanical tests, measurement of produced helium amount and analysis using transmission electron microscopes were carried out on a cold-worked (CW) 316 stainless steel flux thimble tube irradiated up to approximately 35 dpa in a Japanese PWR. The swelling was evaluated to be approximately 0.02%. This level of swelling was much lower than the swelling of the more than several percent that has been observed in fast breeder reactors. (author)

Characterization and understanding of ion irradiation effect on the microstructure of austenitic stainless steels

International Nuclear Information System (INIS)

Volgin, Alexandre

2012-01-01

Austenitic stainless steels are widely used in nuclear industry for internal structures. These structures are located close to the fuel assemblies, inside the pressure vessel. The exposure of these elements to high irradiation doses (the accumulated dose, after 40 years of operation, can reach 80 dpa), at temperature close to 350 C, modifies the macroscopic behavior of the steel: hardening, swelling, creep and corrosion are observed. Moreover, in-service inspections of some of the reactor internal structures have revealed the cracking of some baffle bolts. This cracking has been attributed to Irradiation Assisted Stress Corrosion Cracking (IASCC). In order to understand this complex phenomenon, a first step is to identify the microstructural changes occurring during irradiation, and to understand the mechanisms at the origin of this evolution. In this framework, a large part of the European project 'PERFORM 60' is dedicated to the study of the irradiation damage in austenitic stainless steels. The objective of this PhD work is to bring comprehensive data on the irradiation effects on microstructure. To reach this goal, two model alloys (FeNiCr and FeNiCrSi) and an industrial austenitic stainless steel (316 steel) are studied using Atom Probe Tomography (APT), Transmission Electron Microscope (TEM) and Positron Annihilation Spectroscopy (PAS). They are irradiated by Ni ions in CSNSM (Orsay) at two temperatures (200 and 450 C) and three doses (0.5, 1 and 5 dpa). TEM observations have shown the appearance of dislocation loops, cavities and staking fault tetrahedra. The dislocation loops in 316 steel were preferentially situated in the vicinity of dislocations, while they were randomly distributed in the FeNiCr alloy. APT study has shown the redistribution of Ni and Si under irradiation in FeNiCrSi model alloy and 316 steel, leading to the appearance of (a) Cottrell clouds along dislocation lines, dislocation loops and other non-identified crystalline defects and (b

Review of production status of heavy steel castings and key technologies for their manufacture in China

Directory of Open Access Journals (Sweden)

Liu Baicheng

2008-02-01

Full Text Available This paper expatiates on domestic status of heavy steel casting production, with a special focus on hydraulic turbine castings for Three Gorges Project. In China, there is magnificent demand for heavy castings with the rapid growth of the national economy in recent years and the expected high growth in the coming 10 to 20 years. Some heavy and large castings such as mill housing and hydraulic turbine runner crown, blade and band for Three Gorges Project have been successfully made. However, the domestic production capability is still far from meeting the gigantic requirements. The domestic capability still lags behind the world class level, and a lot of heavy castings still depend on import. The paper also gives a particular introduction of the key technologies in the manufacturing of heavy steel castings like metal melting, foundry technology, heat treatment technology and numerical simulation technique, etc. In addition, several case studies on the application of numerical simulation in the production of heavy steel castings are presented.

In situ TEM study of G-phase precipitates under heavy ion irradiation in CF8 cast austenitic stainless steel

Science.gov (United States)

Chen, Wei-Ying; Li, Meimei; Zhang, Xuan; Kirk, Marquis A.; Baldo, Peter M.; Lian, Tiangan

2015-09-01

Thermally-aged cast austenitic stainless steels (CASS) CF8 was irradiated with 1 MeV Kr ions at 300, 350 and 400 °C to 1.88 × 1019 ions/m2 (∼3 dpa) at the IVEM-Tandem Facility at the Argonne National Laboratory. Before irradiation, the distribution of G-phase precipitates in the ferrite showed spatial variations, and both their size and density were affected by the ferrite-austenite phase boundary and presence of M23C6 carbides. Under 300 °C irradiation, in situ TEM observation showed G-phase precipitates were relatively unchanged in the vicinity of the phase boundary M23C6 carbides, while the density of G-phase precipitates increased with increasing dose within the ferrite matrix. Coarsening of G-phase precipitates was observed in the vicinity of phase boundary M23C6 carbides at 350 °C and 400 °C.

Structural stability of C60 films under irradiation with swift heavy ions

International Nuclear Information System (INIS)

Jin Yunfan; Yao Cunfeng; Wang Zhiguang; Xie Erqing; Song Yin; Sun Youmei; Zhang Chonghong; Liu Jie; Duan Jinglai

2005-01-01

In order to investigate the structural stability of fullerene (C 60 ) under swift heavy ion irradiation, the irradiation experiments of thin C 60 films were performed with 22 MeV/amu Fe 56 ions delivered by HIRFL at Lanzhou in China. The irradiated C 60 films were analyzed by means of Raman scattering and Fourier transform infrared (FTIR) spectroscopes. The analysis results indicated that the damage cross-sections σ of the C 60 molecule deduced from the data of the Raman spectra are between 1.1 and 4.5 x 10 -14 cm 2 for the electronic energy loss from 3.5 to 8.7 keV/nm and electronic energy transfer dominates the damage process of C 60 films. The partial recovery of the damage in irradiated C 60 films at certain electronic energy loss is attributed to an annealing effect of strong electronic excitation

Anti-biofilm activity of Fe heavy ion irradiated polycarbonate

Energy Technology Data Exchange (ETDEWEB)

Joshi, R.P. [Department of Physics, Savitribai Phule Pune University, Pune 411007 (India); Hareesh, K., E-mail: appi.2907@gmail.com [Department of Physics, Savitribai Phule Pune University, Pune 411007 (India); Bankar, A. [Department of Microbiology, Waghire College, Pune 412301 (India); Sanjeev, Ganesh [Microtron Centre, Department of Studies in Physics, Mangalore University, Mangalore 574166 (India); Asokan, K.; Kanjilal, D. [Inter University Accelerator Centre, Arun Asaf Ali Marg, New Delhi 110067 (India); Dahiwale, S.S.; Bhoraskar, V.N. [Department of Physics, Savitribai Phule Pune University, Pune 411007 (India); Dhole, S.D., E-mail: sanjay@physics.unipune.ac.in [Department of Physics, Savitribai Phule Pune University, Pune 411007 (India)

2016-10-01

Highlights: • PC films were irradiated by 60 and 120 MeV Fe ions. • Irradiated PC films showed changes in its physical and chemical properties. • Irradiated PC also showed more anti-biofilm activity compared to pristine PC. - Abstract: Polycarbonate (PC) polymers were investigated before and after high energy heavy ion irradiation for anti-bacterial properties. These PC films were irradiated by Fe heavy ions with two energies, viz, 60 and 120 MeV, at different fluences in the range from 1 × 10{sup 11} ions/cm{sup 2} to 1 × 10{sup 13} ions/cm{sup 2}. UV-Visible spectroscopic results showed optical band gap decreased with increase in ion fluences due to chain scission mainly at carbonyl group of PC which is also corroborated by Fourier transform infrared spectroscopic results. X-ray diffractogram results showed decrease in crystallinity of PC after irradiation which leads to decrease in molecular weight. This is confirmed by rheological studies and also by differential scanning calorimetric results. The irradiated PC samples showed modification in their surfaces prevents biofilm formation of human pathogen, Salmonella typhi.

High temperature deformation behavior, thermal stability and irradiation performance in Grade 92 steel

Science.gov (United States)

Alsagabi, Sultan

The 9Cr-2W ferritic-martensitic steel (i.e. Grade 92 steel) possesses excellent mechanical and thermophysical properties; therefore, it has been considered to suit more challenging applications where high temperature strength and creep-rupture properties are required. The high temperature deformation mechanism was investigated through a set of tensile testing at elevated temperatures. Hence, the threshold stress concept was applied to elucidate the operating high temperature deformation mechanism. It was identified as the high temperature climb of edge dislocations due to the particle-dislocation interactions and the appropriate constitutive equation was developed. In addition, the microstructural evolution at room and elevated temperatures was investigated. For instance, the microstructural evolution under loading was more pronounced and carbide precipitation showed more coarsening tendency. The growth of these carbide precipitates, by removing W and Mo from matrix, significantly deteriorates the solid solution strengthening. The MX type carbonitrides exhibited better coarsening resistance. To better understand the thermal microstructural stability, long tempering schedules up to 1000 hours was conducted at 560, 660 and 760°C after normalizing the steel. Still, the coarsening rate of M23C 6 carbides was higher than the MX-type particles. Moreover, the Laves phase particles were detected after tempering the steel for long periods before they dissolve back into the matrix at high temperature (i.e. 720°C). The influence of the tempering temperature and time was studied for Grade 92 steel via Hollomon-Jaffe parameter. Finally, the irradiation performance of Grade 92 steel was evaluated to examine the feasibility of its eventual reactor use. To that end, Grade 92 steel was irradiated with iron (Fe2+) ions to 10, 50 and 100 dpa at 30 and 500°C. Overall, the irradiated samples showed some irradiation-induced hardening which was more noticeable at 30°C. Additionally

R and D Developments. Research Programs on Irradiation Embrittlement of Reactor Vessel Steels

International Nuclear Information System (INIS)

Gomez Briceno, D.; Lapena, J.; Serrano, M.; Perosanz, F.

2000-01-01

Irradiation embrittlement of pressure vessel steels is a degradation mechanism time dependent that can lead to operational restrictions with adverse effects in the efficiency and life of a plant. For the last year, several research programs have been devoted to study thye evaluation of neutronic radiation effect on mechanical properties of pressure vessel steels. However, at the present, there is a growing interest on the development of new methodologies to optimize the surveillance program information, and the understanding of the irradiation damage mechanism. This paper give an overview of international research programs, and on the R+D activities carried out by the Structural Materials Project on irradiation embrittlement on pressure vessel steels. (Author)

Characterization of matrix damage in ion-irradiated reactor vessel steel

International Nuclear Information System (INIS)

Fujii, Katsuhiko; Fukuya, Koji

2004-01-01

Exact nature of the matrix damage, that is one of radiation-induced nano-scale microstructural features causing radiation embrittlement of reactor vessel, in irradiated commercial steels has not been clarified yet by direct characterization using transmission electron microscopy (TEM). We designed a new preparation method of TEM observation samples and applied it to the direct TEM observation of the matrix damage in the commercial steel samples irradiated by ions. The simulation irradiation was carried out by 3 MeV Ni 2+ ion to a dose of 1 dpa at 290degC. Thin foil specimens for TEM observation were prepared using the modified focused ion beam method. A weak-beam TEM study was carried out for the observation of matrix damage in the samples. Results of this first detailed observation of the matrix damage in the irradiated commercial steel show that it is consisted of small dislocation loops. The observed and analyzed dislocation loops have Burgers vectors b = a , and a mean image size and the number density are 2.5 nm and about 1 x 10 22 m -3 , respectively. In this experiment, all of the observed dislocation loops were too small to determine the vacancy or interstitial nature of the dislocation loops directly. Although it is an indirect method, post-irradiation annealing was used to infer the loop nature. Most of dislocation loops were stable after the annealing at 400degC for 30 min. This result suggests that their nature is interstitial. (author)

Positron annihilation lifetime measurements of austenitic stainless and ferritic/martensitic steels irradiated in the SINQ target irradiation program

Energy Technology Data Exchange (ETDEWEB)

Sato, K., E-mail: ksato@rri.kyoto-u.ac.jp [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan); Xu, Q.; Yoshiie, T. [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan); Dai, Y. [Spallation Neutron Source Division, Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Kikuchi, K. [Frontier Research Center for Applied Atomic Sciences, Ibaraki University, Tokai-mura, Naka-gun, Ibaraki 319-1106 (Japan)

2012-12-15

Titanium-doped austenitic stainless steel (JPCA) and reduced activated ferritic/martensitic steel (F82H) irradiated with high-energy protons and spallation neutrons were investigated by positron annihilation lifetime measurements. Subnanometer-sized (<{approx}0.8 nm) helium bubbles, which cannot be observed by transmission electron microscopy, were detected by positron annihilation lifetime measurements for the first time. For the F82H steel, the positron annihilation lifetime of the bubbles decreased with increasing irradiation dose and annealing temperature because the bubbles absorb additional He atoms. In the case of JPCA steel, the positron annihilation lifetime increased with increasing annealing temperature above 773 K, in which case the dissociation of complexes of vacancy clusters with He atoms and the growth of He bubbles was detected. He bubble size and density were also discussed.

Positron annihilation lifetime measurements of austenitic stainless and ferritic/martensitic steels irradiated in the SINQ target irradiation program

International Nuclear Information System (INIS)

Sato, K.; Xu, Q.; Yoshiie, T.; Dai, Y.; Kikuchi, K.

2012-01-01

Titanium-doped austenitic stainless steel (JPCA) and reduced activated ferritic/martensitic steel (F82H) irradiated with high-energy protons and spallation neutrons were investigated by positron annihilation lifetime measurements. Subnanometer-sized (<∼0.8 nm) helium bubbles, which cannot be observed by transmission electron microscopy, were detected by positron annihilation lifetime measurements for the first time. For the F82H steel, the positron annihilation lifetime of the bubbles decreased with increasing irradiation dose and annealing temperature because the bubbles absorb additional He atoms. In the case of JPCA steel, the positron annihilation lifetime increased with increasing annealing temperature above 773 K, in which case the dissociation of complexes of vacancy clusters with He atoms and the growth of He bubbles was detected. He bubble size and density were also discussed.

Fracture toughness of irradiated stainless steel alloys

International Nuclear Information System (INIS)

Mills, W.J.

1986-01-01

The postirradiation fracture toughness responses of Types 316 and 304 stainless steel (SS) wrought products, cast CF8 SS and Type 308 SS weld deposit were characterized at 427 0 C using J/sub R/-curve techniques. Fast-neutron irradiation of these alloys caused an order of magnitude reduction in J/sub c/ and two orders of magnitude reduction in tearing modulus at neutron exposures above 10 dpa, where radiation-induced losses in toughness appeared to saturate. Saturation J/sub c/ values for the wrought materials ranged from 28 to 31 kJ/m 2 ; the weld exhibited a saturation level of 11 kJ/m 2 . Maximum allowable flaw sizes for highly irradiated stainless steel components stressed to 90% of the unirradiated yield strength are on the order of 3 cm for the wrought material and 1 cm for the weld. Electron fractographic examination revealed that irradiation displacement damage brought about a transition from ductile microvoid coalescence to channel fracture, associated with local separation along planar deformation bands. The lower saturation toughness value for the weld relative to that for the wrought products was attributed to local failure of ferrite particles ahead of the advancing crack which prematurely initiated channel fracture

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)

Influence of Silicon on Swelling and Microstructure in Russian Austenitic Stainless Steels Irradiated to High Neutron Doses

International Nuclear Information System (INIS)

Porollo, S.I.; Shulepin, S.V.; Konobeev, Y.V.; Garner, F.

2007-01-01

Full text of publication follows: For some applications in fusion devices austenitic stainless steels are still considered to be candidates for use as structural components, but high neutron exposures must be endured by the steels. Operational experience of fast reactors in Western Europe, USA and Japan provides evidence of the possible use of austenitic steels up to ∼ 150 dpa. Studies aimed at improvement of existing Russian austenitic steels are being carried out in Russia. For improvement of irradiation resistance of Russian steels it is necessary to understand the basic mechanisms responsible for deterioration of steel properties. This understanding can be achieved by continuing detailed investigations of the microstructure of cladding steels after irradiation to high doses. By investigating the evolution of radiation-induced microstructure in neutron irradiated steels of different chemical composition one can study the effect of chemical variations on steel properties. Silicon is one of the most important chemical elements that strongly influence the behavior of austenitic steel properties under irradiation. In this paper results are presented of investigations of the effect of silicon additions on void swelling and microstructure of base austenitic stainless steel EI-847 (0.06C-16Cr-15Ni- 3Mo-Nb) irradiated as fuel pin cladding of both regular and experimental assemblies in the BOR-60, BN-350 and BN-600 fast reactors to neutron doses up to 49 dpa. The possible mechanisms of silicon's effect on void swelling in austenitic stainless steels are presented and analyzed. (authors)

2 CFR 176.140 - Award term-Required Use of American Iron, Steel, and Manufactured Goods-Section 1605 of the...

Science.gov (United States)

2010-01-01

..., tunnels, sewers, mains, power lines, pumping stations, heavy generators, railways, airports, terminals...) Domestic preference. (1) This award term and condition implements Section 1605 of the American Recovery and... the domestic iron, steel, and/or manufactured goods would be unreasonable. The cost of domestic iron...

Heavy Ion Irradiation Fluence Dependence for Single-Event Upsets of NAND Flash Memory

Science.gov (United States)

Chen, Dakai; Wilcox, Edward; Ladbury, Raymond; Kim, Hak; Phan, Anthony; Seidleck, Christina; LaBel, Kenneth

2016-01-01

We investigated the single-event effect (SEE) susceptibility of the Micron 16 nm NAND flash, and found the single-event upset (SEU) cross section varied inversely with fluence. The SEU cross section decreased with increasing fluence. We attribute the effect to the variable upset sensitivities of the memory cells. The current test standards and procedures assume that SEU follow a Poisson process and do not take into account the variability in the error rate with fluence. Therefore, heavy ion irradiation of devices with variable upset sensitivity distribution using typical fluence levels may underestimate the cross section and on-orbit event rate.

Ion irradiation-induced precipitation of Cr23C6 at dislocation loops in austenitic steel

International Nuclear Information System (INIS)

Jin, Shuoxue; Guo, Liping; Luo, Fengfeng; Yao, Zhongwen; Ma, Shuli; Tang, Rui

2013-01-01

The irradiation-induced precipitates in argon ion-irradiated austenitic stainless steel at 550 °C were examined via transmission electron microscopy. The selected-area electron diffraction patterns of precipitates indicated unambiguously that the precipitates were Cr 23 C 6 carbides. It was observed directly for the first time that irradiation-induced Cr 23 C 6 precipitates formed at dislocation loops in austenitic stainless steel, and coarsened with increasing irradiation dose.

Reference manual on the IAEA JRQ correlation monitor steel for irradiation damage studies

International Nuclear Information System (INIS)

2001-07-01

The objective of this report is to provide information on the mechanical properties of the ASTM A533 grade B class 1 steel that was designated as 'JRQ reference steel' and for many years served as a radiation/mechanical property correlation monitor in a number of international and national studies of irradiation embrittlement of reactor pressure vessel steel. This report provides the most comprehensive listing of material test data obtained on the JRQ manufacturing history and material properties in the initial, and as delivered condition during the implementation of two IAEA co-ordinated research projects (CRPs) on behaviour of reactor pressure vessel steels under neutron irradiation

Precipitation behavior in austenitic and ferritic steels during fast neutron irradiation and thermal aging*1

Science.gov (United States)

Kawanishi, H.; Hajima, R.; Sekimura, N.; Arai, Y.; Ishino, S.

1988-07-01

Precipitation behavior has been studied using a carbon extraction replica technique in Ti-modified Type 316 stainless steels (JPCA-2) and 9Cr-2Mo ferritic/martensitic steels (JFMS) irradiated to 8.1 × 10 24 n/m 2 at 873 and 673 K, respectively, in the experimental fast breeder reactor JOYO. Precipitate identification and compositional analysis were carried out on extracted replicas. The results were compared to those from the as-received steel and a control which had been given the same thermal as-treatment as the specimens received during irradiations. Carbides, Ti-sulphides and phosphides were precipitated in JPCA-2. Precipitate observed in JFMS included carbides, Laves-phases and phosphides. The precipitates in both steels were concluded to be stable under irradiation except for MC and M 6C in JPCA-2. Small MC particles were found precipitated in JPCA-2 during both irradiation and aging. Irradiation proved to promote the precipitation of M 6C in JPCA-2.

Microstructural stability of fast reactor irradiated 10 to 12% Cr ferritic-martensitic stainless steels

International Nuclear Information System (INIS)

Little, E.A.; Stoter, L.P.

1982-01-01

The strength and microstructural stability of three 10 to 12% Cr ferritic-martensitic stainless steels have been characterized following fast reactor irradiation to damage levels of 30 displacements per atom (dpa) at temperatures in the range 380 to 615 0 C. Irradiation results in either increases or decreases in room temperature hardness depending on the irradiation temperature. These strength changes can be qualitatively rationalized in terms of the combined effects of irradiation-induced interstitial dislocation loop formation and recovery of the dislocation networks comprising the initial tempered martensite structures. Precipitate evolution in the irradiated steels is associated with the nonequilibrium segregation of the elements nickel, silicon, molybdenum, chromium and phosphorus, brought about by solute-point defect interactions. The principal irradiation-induced precipitates identified are M 6 X, intermetallic chi and sigma phases and also α' (Cr-rich ferrite). The implications of the observed microstructural changes on the selection of martensitic stainless steels for fast reactor wrapper applications are briefly considered

Irradiation Assisted Stress Corrosion Cracking of austenitic stainless steels

Energy Technology Data Exchange (ETDEWEB)

Tsukada, Takashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

Irradiation Assisted Stress Corrosion Cracking (IASCC) of austenitic stainless steels in oxygenated high temperature water was studied. The IASCC failure has been considered as a degradation phenomenon potential not only in the present light water reactors but rather common in systems where the materials are exposed simultaneously to radiation and water environments. In this study, effects of the material and environmental factors on the IASCC of austenitic stainless steels were investigated in order to understand the underlying mechanism. The following three types of materials were examined: a series of model alloys irradiated at normal water-cooled research reactors (JRR-3M and JMTR), the material irradiated at a spectrally tailored mixed-spectrum research reactor (ORR), and the material sampled from a duct tube of a fuel assembly used in the experimental LMFBR (JOYO). Post-irradiation stress corrosion cracking tests in a high-temperature water, electrochemical corrosion tests, etc., were performed at hot laboratories. Based on the results obtained, analyses were made on the effects of alloying/impurity elements, irradiation/testing temperatures and material processing, (i.e., post-irradiation annealing and cold working) on the cracking behavior. On the basis of the analyses, possible remedies against IASCC in the core internals were discussed from viewpoints of complex combined effects among materials, environment and processing factors. (author). 156 refs.

Effects of nickel on irradiation embrittlement of light water reactor pressure vessel steels

International Nuclear Information System (INIS)

2005-06-01

This TECDOC was developed under the IAEA Coordinated Research Project (CRP) entitled Effects of Nickel on Irradiation Embrittlement of Light Water Reactor Pressure Vessel (RPV) Steels. This CRP is the sixth in a series of CRPs to determine the influence of the mechanism and quantify the influence of nickel content on the deterioration of irradiation embrittlement of reactor pressure vessel steels of the Ni-Cr-Mo-V or Mn-Ni-Cr-Mo types. The scientific scope of the programme includes procurement of materials, determination of mechanical properties, irradiation and testing of specimens in power and/or test reactors, and microstructural characterization. Eleven institutes from eight different countries and the European Union participated in this CRP and six institutes conducted the irradiation experiments of the CRP materials. In addition to the irradiation and testing of those materials, irradiation experiments of various national steels were also conducted. Moreover, some institutes performed microstructural investigations of both the CRP materials and national steels. This TECDOC presents and discusses all the results obtained and the analyses performed under the CRP. The results analysed are clear in showing the significantly higher radiation sensitivity of high nickel weld metal (1.7 wt%) compared with the lower nickel base metal (1.2 wt%). These results are supported by other similar results in the literature for both WWER-1000 RPV materials, pressurized water reactor (PWR) type materials, and model alloys. Regardless of the increased sensitivity of WWER-1000 high nickel weld metal (1.7 wt%), the transition temperature shift for the WWER-1000 RPV design fluence is still below the curve predicted by the Russian code (standard for strength calculations of components and piping in NPPs - PNAE G 7-002-86). For higher fluence, no data were available and the results should not be extrapolated. Although manganese content was not incorporated directly in this CRP

Folding two dimensional crystals by swift heavy ion irradiation

International Nuclear Information System (INIS)

Ochedowski, Oliver; Bukowska, Hanna; Freire Soler, Victor M.; Brökers, Lara; Ban-d'Etat, Brigitte; Lebius, Henning; Schleberger, Marika

2014-01-01

Ion irradiation of graphene, the showcase model of two dimensional crystals, has been successfully applied to induce various modifications in the graphene crystal. One of these modifications is the formation of origami like foldings in graphene which are created by swift heavy ion irradiation under glancing incidence angle. These foldings can be applied to locally alter the physical properties of graphene like mechanical strength or chemical reactivity. In this work we show that the formation of foldings in two dimensional crystals is not restricted to graphene but can be applied for other materials like MoS 2 and hexagonal BN as well. Further we show that chemical vapour deposited graphene forms foldings after swift heavy ion irradiation while chemical vapour deposited MoS 2 does not

Microstructural evolution in reactor pressure vessel steel under neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Ohno, Katsumi; Fukuya, Koji [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

2000-09-01

Understanding microstructural changes in reactor pressure vessel steels is important in order to evaluate radiation-induced embrittlement, one of the major aging phenomena affecting the extension of plant life. In this study, actual surveillance test specimens and samples of rector vessel low-alloy steel (A533B steel) irradiated in a research reactor were examined using state-of-the-art techniques to clarify the neutron flux effect on the microstructural changes. These techniques included small angle neutron scattering and atom probes. Microstructural changes which are considered to be the main factors affecting embrittlement, including the production of copper-rich precipitates and the segregation of impurity elements, were confirmed by the results of the study. In addition, the mechanical properties were predicted based on the obtained quantitative data such as the diameters of precipitates. Consequently, the hardening due to irradiation was almost simulated. (author)

Heavy ion irradiation effects of polymer film on absorption of light

Energy Technology Data Exchange (ETDEWEB)

Kasai, Noboru; Seguchi, Tadao [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment; Arakawa, Tetsuhito

1997-03-01

Ion irradiation effects on the absorption of light for three types of polymer films; polyethylene-terephthalate (PET), polyethylene-naphthalate (PEN), and polyether-ether-ketone (PEEK) were investigated by irradiation of heavy ions with Ni{sup 4+}(15MeV), O{sup 6+}(160MeV), and Ar{sup 8+}(175MeV), and compared with electron beams(EB) irradiation. The change of absorption at 400nm by a photometer was almost proportional to total dose for ions and EB. The absorption per absorbed dose was much high in Ni{sup 4+}, but rather small in O{sup 6+} and Ar{sup 8+} irradiation, and the absorption by EB irradiation was accelerated by the temperature of polymer film during irradiation. The beam heating of materials during ion irradiation was assumed, especially for Ni ion irradiation. The heavy ion irradiation effect of polymers was thought to be much affected by the ion beam heating than the linear energy transfer(LET) of radiation source. (author)

Cavity nucleation and growth during helium implantation and neutron irradiation of Fe and steel

DEFF Research Database (Denmark)

Eldrup, Morten Mostgaard; Singh, Bachu Narain

In order to investigate the role of He in cavity nucleation in neutron irradiated iron and steel, pure iron and Eurofer-97 steel have been He implanted and neutron irradiated in a systematic way at different temperatures, to different He and neutron doses and with different He implantation rates....

In situ TEM study of G-phase precipitates under heavy ion irradiation in CF8 cast austenitic stainless steel

Energy Technology Data Exchange (ETDEWEB)

Chen, Wei-Ying [Argonne National Laboratory, Argonne, IL 60439 (United States); University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Li, Meimei; Zhang, Xuan; Kirk, Marquis A.; Baldo, Peter M. [Argonne National Laboratory, Argonne, IL 60439 (United States); Lian, Tiangan [Electric Power Research Institute, Palo Alto, CA 94304 (United States)

2015-09-15

Thermally-aged cast austenitic stainless steels (CASS) CF8 was irradiated with 1 MeV Kr ions at 300, 350 and 400 °C to 1.88 × 10{sup 19} ions/m{sup 2} (∼3 dpa) at the IVEM-Tandem Facility at the Argonne National Laboratory. Before irradiation, the distribution of G-phase precipitates in the ferrite showed spatial variations, and both their size and density were affected by the ferrite–austenite phase boundary and presence of M{sub 23}C{sub 6} carbides. Under 300 °C irradiation, in situ TEM observation showed G-phase precipitates were relatively unchanged in the vicinity of the phase boundary M{sub 23}C{sub 6} carbides, while the density of G-phase precipitates increased with increasing dose within the ferrite matrix. Coarsening of G-phase precipitates was observed in the vicinity of phase boundary M{sub 23}C{sub 6} carbides at 350 °C and 400 °C.

Development of advanced austenitic stainless steels resistant to void swelling under irradiation

International Nuclear Information System (INIS)

Rouxel, Baptiste

2016-01-01

In the framework of studies about Sodium Fast Reactors (SFR) of generation IV, the CEA is developing new austenitic steel grades for the fuel cladding. These steels demonstrate very good mechanical properties but their use is limited because of the void swelling under irradiation. Beyond a high irradiation dose, cavities appear in the alloys and weaken the material. The reference material in France is a 15Cr/15Ni steel, named AIM1, stabilized with titanium. This study try to understand the role played by various chemical elements and microstructural parameters on the formation of the cavities under irradiation, and contribute to the development of a new grade AIM2 more resistant to swelling. In an analytical approach, model materials were elaborated with various chemical compositions and microstructures. Ten grades were cast with chemical variations in Ti, Nb, Ni and P. Four specific microstructures for each alloy highlighted the effect of dislocations, solutes or nano-precipitates on the void swelling. These materials were characterized using TEM and SANS, before irradiation with Fe"2"+ (2 MeV) ions in the order to simulate the damages caused by neutrons. Comparing the irradiated microstructures, it is demonstrated that the solutes have a dominating effect on the formation of cavities. Specifically titanium in solid solution reduces the swelling whereas niobium does not show this effect. Finally, a matrix enriched by 15% to 25% of nickel is still favorable to limit swelling in these advanced austenitic stainless steels. (author) [fr

Irradiation creep of 316 and 316 Ti steels

International Nuclear Information System (INIS)

Lehmann, J.; Dupouy, J.M.; Boutard, J.L.; Maillard, A.; Broudeur, R.

1979-07-01

Irradiation creep results for several 316 and 316 Ti steels show the effects of stress, dose and temperature in the range 400 to 550 0 C. The observed differences are related to the compositioning and metallurgical conditions of the materials. (author)

In-reactor precipitation and ferritic transformation in neutron--irradiated stainless steels

International Nuclear Information System (INIS)

Porter, D.L.; Wood, E.L.

1978-01-01

Ferritic transformation (γ → α) was observed in Type 304L, 20% cold-worked AISI 316, and solution-annealed AISI 316 stainless steels subjected to fast neutron irradiation. Each material demonstrated an increasing propensity for transformation with increasing irradiation temperature between 400 and 550 0 C. Irradiation-induced segregation of Ni solute to precipitates was found not to influence the transformation kinetics in 304L. Similar composition data from 316 materials demonstrates a much greater temperature dependence of precipitation reactions in the process of matrix Ni depletion during neutron irradiation. The 316 data establishes a strong link between such depletion and the observed γ → α transformation. Moreover, the lack of correlation between precipitate-related Ni depletion and the γ → α transformation in 304L can be related to the fact that irradiation-induced voids nucleate very quickly in 304L steel during irradiation. These voids present preferential sites for Ni segregation through a defect trapping mechanism, and hence Ni segregates to voids rather than to precipitates, as evidenced by observed stable γ shells around voids in areas of complete transformation

Effects of irradiation on the fracture behavior of austenitic stainless steels

International Nuclear Information System (INIS)

Grossbeck, M.L.; Stiegler, J.O.; Holmes, J.J.

1977-01-01

Fracture in irradiated materials occurs by mechanisms which occur in unirradiated materials in addition to mechanisms related to irradiation phenomena. The paper examines radiation effects in austenitic stainless steels for use as core structural materials in fast breeder reactors

Irradiation Creep and Swelling of Russian Ferritic-Martensitic Steels Irradiated to Very High Exposures in the BN-350 Fast Reactor at 305-335 degrees C

International Nuclear Information System (INIS)

Konobeev, Yury V.; Dvoriashin, Alexander M.; Porollo, S.I.; Shulepin, S.V.; Budylkin, N.I.; Mironova, Elena G.; Garner, Francis A.

2003-01-01

Russian ferritic/martensitic (F/M) steels EP-450, EP-852 and EP-823 were irradiated in the BN-350 fast reactor in the form of gas-pressurized creep tubes. The first steel is used in Russia for hexagonal wrappers in fast reactors. The other steels were developed for compatibility with Pb-Bi coolants and serve to enhance our understanding of the general behavior of this class of steels. In an earlier paper we published data on irradiation creep of EP-450 and EP-823 at temperatures between 390 and 520C, with dpa levels ranging from 20 to 60 dpa. In the current paper new data on the irradiation creep and swelling of EP-450 and EP-852 at temperatures between 305 and 335C and doses ranging from 61 to 89 dpa are presented. Where comparisons are possible, it appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures <420C, but may be camouflaged somewhat by precipitation-related densification. These irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels.

Positron annihilation and Moessbauer studies of neutron irradiated reactor pressure vessel steels

International Nuclear Information System (INIS)

Brauer, G.; Matz, W.; Liszkay, L.; Molnar, B.

1990-11-01

Positron annihilation (lifetime, Doppler broadening) and Moessbauer studies on unirradiated, neutron irradiated and neutron irradiated plus annealed reactor pressure vessel steels (Soviet type 15Kh2NMFA) are presented. The role of microstructural properties and the formation of irradiation-induced precipitates is discussed. (orig.) [de

Some elevated temperature tensile and strain-controlled fatigue properties for a 9%Cr1Mo steel heat treated to simulate thick section material

International Nuclear Information System (INIS)

Sanderson, S.J.; Jacques, S.

Current interest has been expressed in the usage of thick section 9%Cr1%Mo steel, particularly for UK Commercial Demonstration Fast Reactor (CDFR) steam generator tubeplates. This paper presents the results of some preliminary mechanical property test work on a single cast of the steel, heat treated to simulate heavy ruling sections encompassing thicknesses likely to be met in the CDFR context. The microstructures of the simulated thick section material were found to remain predominantly as tempered martensite even at the slowest transformation cooling rates used (50 deg. C/h). The effect of microstructure is reflected in the elevated temperature proof stress, tensile strength and strain-controlled fatigue endurance which were found to be comparable with the properties established for thin section normalised and tempered 9%Cr1%Mo steel. These results are extremely encouraging and, taken in conjunction with the results from other simulation work on this material, further demonstrate the potential of thick section 9%Cr1%Mo steel. (author)

Swelling analysis of austenitic stainless steels by means of ion irradiation and kinetic modeling

International Nuclear Information System (INIS)

Kohyama, Akira; Donomae, Takako

1999-03-01

The influences of irradiation environment on the swelling behavior of austenitic stainless steel has been studied, to aid understanding the origin of the difference in swelling response of PNC316 stainless steel in fuel-pin environment and in materials irradiation capsules, in terms of irradiation conditions, damage mechanism and material conditions. This work focused on the theoretical investigation of the influence of temperature variation on microstructural development of austenitic stainless steels during irradiation, using a kinetic rate theory model. A modeling and calculation on non-steady irradiation effects were first carried out. A fully dynamic model of point defect evolution and extended defect development, which accounts for cascade damage, was developed and successfully applied to simulate the interstitial loop evolution in low temperature regimes. The influence of cascade interstitial clustering on dislocation loop formation has also been assessed. The establishment of a basis for general assessment of non-steady irradiation effects in austenitic stainless steels was advanced. The developed model was applied to evaluate the influences of temperature variation in formerly carried out CMIR and FFTF/MFA-1 FBR irradiation experiments. The results suggested the gradual approach of microstructural features to equilibrium states in all the temperature variation conditions and no sign of anomalous behavior was noted. On the other hand, there is the influence of temperature variation on microstructural development under the neutron irradiation, like CMIR. So there are some possibilities of the work of mechanism which is not taken care on this model, for example the effect of the precipitate behavior which is sensitive to irradiation temperature. (author)

Results of neutron measurements in the spectral position of the Juelich FKS steel irradiation capsules

International Nuclear Information System (INIS)

Schneider, W.

1986-10-01

This is a report on the planning and results of neutron monitoring in the capsules of the Juelich steel irradiation for the research project on component safety (FKS). The table of results and their discussion is provided specifically for the spectral positions (for characterising the neutron spectrum) in each of the types of irradiation capsules used. The results are given for the reaction rates of the neutron measurement reactions used (activation or fission reactions), for the neutron flux densities and fluxes derived from them related to the actual or at least plausible neutron spectra and finally for the radiation damage (or exposure) of the irradiated material calculated from them, expressed as the atomic displacement figure (dpa) and its percentage in sections of the neutron spectrum. (orig.) [de

Biomonitoring of some heavy metal contaminations from a steel ...

African Journals Online (AJOL)

Soil and plants growing in the vicinity of industrial areas display increased concentrations of heavy metals and give an indication of the environmental quality. The contamination source for aluminum, iron, nickel and lead in the Botanical garden of Mobarakeh Steel Company was recognized by analyzing the leaves and ...

Folding two dimensional crystals by swift heavy ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Ochedowski, Oliver; Bukowska, Hanna [Fakultät für Physik and CENIDE, Universität Duisburg-Essen, D-47048 Duisburg (Germany); Freire Soler, Victor M. [Fakultät für Physik and CENIDE, Universität Duisburg-Essen, D-47048 Duisburg (Germany); Departament de Fisica Aplicada i Optica, Universitat de Barcelona, E08028 Barcelona (Spain); Brökers, Lara [Fakultät für Physik and CENIDE, Universität Duisburg-Essen, D-47048 Duisburg (Germany); Ban-d' Etat, Brigitte; Lebius, Henning [CIMAP (CEA-CNRS-ENSICAEN-UCBN), 14070 Caen Cedex 5 (France); Schleberger, Marika, E-mail: marika.schleberger@uni-due.de [Fakultät für Physik and CENIDE, Universität Duisburg-Essen, D-47048 Duisburg (Germany)

2014-12-01

Ion irradiation of graphene, the showcase model of two dimensional crystals, has been successfully applied to induce various modifications in the graphene crystal. One of these modifications is the formation of origami like foldings in graphene which are created by swift heavy ion irradiation under glancing incidence angle. These foldings can be applied to locally alter the physical properties of graphene like mechanical strength or chemical reactivity. In this work we show that the formation of foldings in two dimensional crystals is not restricted to graphene but can be applied for other materials like MoS{sub 2} and hexagonal BN as well. Further we show that chemical vapour deposited graphene forms foldings after swift heavy ion irradiation while chemical vapour deposited MoS{sub 2} does not.

Swift heavy ion irradiation effects on carbonyl and trans-vinylene groups in high and low density polyethylene

International Nuclear Information System (INIS)

Grosso, M.F. del; Chappa, V.C.; Arbeitman, C.R.; Garcia Bermudez, G.; Behar, M.

2009-01-01

In this work, we have studied the effects of swift heavy ion irradiation on the creation of new functional groups in high and low density polyethylene (HDPE and LDPE). Polymers were irradiated with different ions (6.77 MeV He and 47 MeV Li) and fluences. The induced changes were analyzed by Fourier transform infrared (FTIR) spectroscopy. Creation and damage cross sections for some groups were compared for two different types of PE.

Swift heavy ion irradiation effects on carbonyl and trans-vinylene groups in high and low density polyethylene

Energy Technology Data Exchange (ETDEWEB)

Grosso, M.F. del, E-mail: delgrosso@tandar.cnea.gov.a [Gerencia de Investigacion y Aplicaciones, TANDAR-CNEA (Argentina); Chappa, V.C. [Gerencia de Investigacion y Aplicaciones, TANDAR-CNEA (Argentina); CONICET (Argentina); Arbeitman, C.R. [Gerencia de Investigacion y Aplicaciones, TANDAR-CNEA (Argentina); Garcia Bermudez, G. [Gerencia de Investigacion y Aplicaciones, TANDAR-CNEA (Argentina); CONICET (Argentina); Escuela de Ciencia y Tecnologia, UNSAM (Argentina); Behar, M. [Instituto de Fisica, UFRGS, Porto Alegre (Brazil)

2009-10-01

In this work, we have studied the effects of swift heavy ion irradiation on the creation of new functional groups in high and low density polyethylene (HDPE and LDPE). Polymers were irradiated with different ions (6.77 MeV He and 47 MeV Li) and fluences. The induced changes were analyzed by Fourier transform infrared (FTIR) spectroscopy. Creation and damage cross sections for some groups were compared for two different types of PE.

Preliminary study on mutagenic effects of heavy ions irradiation on maize inbred lines

International Nuclear Information System (INIS)

Yu Lixia; Li Wenjian; Xie Hongmei; Chen Xuejun; Chen Jing

2010-01-01

In order to study mutagenic effects of different heavy ions irradiation on maize inbred lines,corn seeds of Zheng58, Lu9801, Jinxiang4C-1, CSR24001, 308 and 478 were irradiated with 12 C 6+ and 36 Ar 18+ ions. The experimental results showed that the germination rate and planting percent were different after irradiation. The wettish seeds had higher sensibility to heavy ion irradiation. The leaf type of the plant appeared visible changes in M 1 generation. In M 2 generation, great changes had taken place in economic traits, many of which are beneficial mutation. Some beneficia1 mutation could be stably inherited in M 3 generation. From the above, it can be predicted that heavy ions irradiation is an effective means of genetic improvement of maize. (authors)

Microchemical evolution of neutron-irradiated stainless steel

International Nuclear Information System (INIS)

Brager, H.R.; Garner, F.A.

1980-04-01

The precipitates that develop in AISI 316 stainless steel during irradiation play a dominant role in determining the dimensional and mechanical property changes of this alloy. This role is expressed primarily in a large change in matrix composition that alters the diffusional properties of the alloy matrix and also appears to alter the rate of acceptance of point defects at dislocations and voids. The major elemental participants in the evolution have been identified as nickel, silicon, and carbon. The exceptional sensitivity of this evolution to many variables accounts for much of the variability of response exhibited by this alloy in nominally similar irradiations

Effect of microstructure on radiation induced segregation and depletion in ion irradiated SS316 steel

International Nuclear Information System (INIS)

Jin, Hyung Ha; Kwon, Sang Chul; Kwon, Jun Hyun

2011-01-01

Irradiation assisted stress corrosion cracking (IASCC), void swelling and irradiation induced hardening are caused by change of characteristics of material by neutron irradiation, stress state of material and environmental situation. It has been known that chemical compositions varies at grain boundary (GB) significantly with fluence level and the depletion of Cr element at GB has been considered as one of important factors causing material degradation, especially, IASCC in austenitic stainless steel. However, experimental results of IASCC under PWR condition were directly not connected with Cr depletion phenomenon by neutron irradiation. Because the mechanism of IASCC under PWR has not yet been clearly understood in spite of many energetic researches, fundamental researches about radiation induced segregation and depletion in irradiated austenitic stainless steels have been attracted again. In this work, an effect of residual microstructure on radiation induced segregation and depletion of alloy elements at GB was investigated in ion irradiated SS316 steel using transmission electron microscope (TEM) with energy dispersive spectrometer (EDS)

Evaluation of AISI 316L stainless steel welded plates in heavy petroleum environment

International Nuclear Information System (INIS)

Carvalho Silva, Cleiton; Pereira Farias, Jesualdo; Batista de Sant'Ana, Hosiberto

2009-01-01

This work presents the study done on the effect of welding heating cycle on AISI 316L austenitic stainless steel corrosion resistance in a medium containing Brazilian heavy petroleum. AISI 316L stainless steel plates were welded using three levels of welding heat input. Thermal treatments were carried out at two levels of temperatures (200 and 300 deg. C). The period of treatment in all the trials was 30 h. Scanning electronic microscopy (SEM) and analysis of X-rays dispersive energy (EDX) were used to characterize the samples. Weight loss was evaluated to determine the corrosion rate. The results show that welding heating cycle is sufficient to cause susceptibility to corrosion caused by heavy petroleum to the heat affected zone (HAZ) of the AISI 316L austenitic stainless steel

A theoretical study on the influence of the homogeneity of heavy-ion irradiation field on the survival fraction of cells

International Nuclear Information System (INIS)

Wen Xiaoqiong; Li Qiang; Zhou Guangming; Li Wenjian; Wang Jufang; Wei Zengquan

2001-01-01

In order to provide theoretical basis for the homogeneity request of heavy-ion irradiation field, the most important design parameter of the heavy-ion radiotherapy facility planned in IMP (Institute of Modern Physics), the influence of the homogeneity of heavy-ion irradiation field on the survival fraction of cells was investigated theoretically. A formula for survival fraction of cells irradiated by the un-uniform heavy-ion irradiation field was deduced to estimate the influence of the homogeneity of heavy-ion irradiation field on the survival fraction of cells. The results show that the survival fraction of cells irradiation by the un-uniform irradiation field is larger than that of cells irradiated by the uniform irradiation field, and the survival fraction of cells increases as the homogeneity of heavy-ion irradiation field decreasing. Practically, the heavy-ion irradiation field can be treated as uniform irradiation field when its homogeneity is better than 95%. According to these results, design request for the homogeneity of heavy-ion irradiation field should be better than 95%. The present results also show that the agreement of homogeneity of heavy-ion irradiation field must be checked while comparing the survival fraction curves obtained by different laboratory

Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 deg. C

International Nuclear Information System (INIS)

Matijasevic, M.; Lucon, E.; Almazouzi, A.

2008-01-01

High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 deg. C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between -160 deg. C and 300 deg. C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 deg. C (up to 2.6 dpa), and tested between -170 deg. C and 300 deg. C. Irradiation effects at lower irradiation temperatures are more significant

Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 °C

Science.gov (United States)

Matijasevic, M.; Lucon, E.; Almazouzi, A.

2008-06-01

High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 °C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between -160 °C and 300 °C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 °C (up to 2.6 dpa), and tested between -170 °C and 300 °C. Irradiation effects at lower irradiation temperatures are more significant.

Impact of the nanostructuration on the corrosion resistance and hardness of irradiated 316 austenitic stainless steels

Energy Technology Data Exchange (ETDEWEB)

Hug, E., E-mail: eric.hug@ensicaen.fr [Laboratoire de Cristallographie et Sciences des Matériaux, Normandie Université, CNRS UMR 6508, 6 Bd Maréchal Juin, 14050 Caen (France); Prasath Babu, R. [School of Materials, University of Manchester, M13 9PL (United Kingdom); Groupe de Physique des Matériaux, UMR CNRS 6634, Université et INSA de Rouen, Normandie Université, Saint-Etienne du Rouvray Cedex (France); Monnet, I. [Centre de recherches sur les Ions, les Matériaux et la Photonique CEA-CNRS, Normandie Université, 6 Bd Maréchal Juin, 14050 Caen (France); Etienne, A. [Groupe de Physique des Matériaux, UMR CNRS 6634, Université et INSA de Rouen, Normandie Université, Saint-Etienne du Rouvray Cedex (France); Moisy, F. [Centre de recherches sur les Ions, les Matériaux et la Photonique CEA-CNRS, Normandie Université, 6 Bd Maréchal Juin, 14050 Caen (France); Pralong, V. [Laboratoire de Cristallographie et Sciences des Matériaux, Normandie Université, CNRS UMR 6508, 6 Bd Maréchal Juin, 14050 Caen (France); Enikeev, N. [Institute of Physics of Advanced Materials, Ufa (Russian Federation); Saint Petersburg State University, Laboratory of the Mechanics of Bulk Nanostructured Materials, 198504 St. Petersburg (Russian Federation); Abramova, M. [Institute of Physics of Advanced Materials, Ufa (Russian Federation); and others

2017-01-15

Highlights: • Impacts of nanostructuration and irradiation on the properties of 316 stainless steels are reported. • Irradiation of nanostructured samples implies chromium depletion as than depicted in coarse grain specimens. • Hardness of nanocrystalline steels is only weakly affected by irradiation. • Corrosion resistance of the nanostructured and irradiated samples is less affected by the chromium depletion. - Abstract: The influence of grain size and irradiation defects on the mechanical behavior and the corrosion resistance of a 316 stainless steel have been investigated. Nanostructured samples were obtained by severe plastic deformation using high pressure torsion. Both coarse grain and nanostructured samples were irradiated with 10 MeV {sup 56}Fe{sup 5+} ions. Microstructures were characterized using transmission electron microscopy and atom probe tomography. Surface mechanical properties were evaluated thanks to hardness measurements and the corrosion resistance was studied in chloride environment. Nanostructuration by high pressure torsion followed by annealing leads to enrichment in chromium at grain boundaries. However, irradiation of nanostructured samples implies a chromium depletion of the same order than depicted in coarse grain specimens but without metallurgical damage like segregated dislocation loops or clusters. Potentiodynamic polarization tests highlight a definitive deterioration of the corrosion resistance of coarse grain steel with irradiation. Downsizing the grain to a few hundred of nanometers enhances the corrosion resistance of irradiated samples, despite the fact that the hardness of nanocrystalline austenitic steel is only weakly affected by irradiation. These new experimental results are discussed in the basis of couplings between mechanical and electrical properties of the passivated layer thanks to impedance spectroscopy measurements, hardness properties of the surfaces and local microstructure evolutions.

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

In-reactor precipitation and ferritic transformation in neutron-irradiated stainless steels

International Nuclear Information System (INIS)

Porter, D.L.; Wood, E.L.

1979-01-01

Ferritic transformation (γ→α) was observed in type 304L, 20% cold-worked AISI 316, and solution-annealed AISI 316 stainless steels when subjected to fast neutron irradiation. Each material demonstrated an increasing propensity for transformation with increasing irradiation temperature between 40 and 550 0 C. Irradiation-induced segregation of Ni solute to precipitates was found not to be a controlling factor in the transformation kinetics in 304L. Similar composition data from 316 materials demonstrates a much greater dependence of matrix Ni depletion by precipitation reactions during neutron irradiation. The 316 data establishes a strong link between such depletion and the observed γ→α transformation. Moreover, the lack of correlation between precipitate-related Ni depletion and the γ→α transformation in 304L can be related to the fact that irradiation-induced voids nucleate very quickly in 304L steel during irradiation. These voids present competing sites for Ni segregation through a defect drag mechanism, and hence Ni segregates to voids rather than to precipitates, as evidenced by observed stable γ shells around voids in areas of complete transformation. (Auth.)

Effect of heavy ion irradiation on C 60

Science.gov (United States)

Lotha, S.; Ingale, A.; Avasthi, D. K.; Mittal, V. K.; Mishra, S.; Rustagi, K. C.; Gupta, A.; Kulkarni, V. N.; Khathing, D. T.

1999-06-01

Thin films of C 60 were subjected to swift heavy ion irradiation spanning the region from 2 to 11 keV/nm of electronic excitation. Studies of the irradiated films by Raman spectroscopy indicated polymerization and damage of the film with an ion fluence. The ion track radii are estimated for various ions using the Raman data. Photoluminescence spectroscopy of the irradiated film indicated a decrease in the C 60 phase with a dose, and an increase in the intensity at the 590 nm wavelength, which is attributed to an increase in the oxygen content.

Failure Analysis of Heavy-Ion-Irradiated Schottky Diodes

Science.gov (United States)

Casey, Megan C.; Lauenstein, Jean-Marie; Wilcox, Edward P.; Topper, Alyson D.; Campola, Michael J.; Label, Kenneth A.

2017-01-01

In this work, we use high- and low-magnitude optical microscope images, infrared camera images, and scanning electron microscope images to identify and describe the failure locations in heavy-ion-irradiated Schottky diodes.

The role of phosphorus in the irradiation embrittlement of PWR pressure vessel steels

International Nuclear Information System (INIS)

Jones, R.B.; Buswell, J.T.

1987-02-01

An analysis has been performed of the influence of phosphorus on post-irradiation materials properties and microstructures determined on a variety of PWR steels and variants following exposure to MTR or reactor surveillance irradiations to doses not exceeding 7 x 10 19 n.cm -2 (E>1.0MeV) at 250-290 0 C. The irradiation-induced shifts in impact transition temperature, matrix hardening and the relative small angle neutron scattering response were found to rise most rapidly with increasing phosphorus when the copper content of the steel was 0.03 w/o. The sensitivity of the changes in mechanical properties to phosphorus content decreased as the copper content was increased. At copper levels typical of modern PWR steel manufacture (Cu 3 P) produced by the irradiation induced segregation of phosphorus to defect sinks and the depletion of phosphorus in solid solution as detected by high sensitivity electron microscopy and other analytical techniques. At higher levels of copper (approx. 0.3 w/o) the effect of phosphorus on properties was reduced by a factor of three due to the observed incorporation of phosphorus into the small copper precipitates formed during irradiation. Grain boundary embrittlement by phosphorus under irradiation is not thought to be important but further evidence concerning the post-irradiation fracture mode and the development of the deleterious influence of phosphorus with irradiation dose is required for a comprehensive understanding of its action. Some suggestions for future work are made. (author)

Surface structure of Cr0.5 Ti0.5N coatings after heavy ions irradiation and annealing

International Nuclear Information System (INIS)

Kislitsin, Sergey; Gorlachev, Igor; Uglov, Vladimir

2015-01-01

Results of surface structure investigations of TiCrN coating on carbon steel after irradiation by helium, krypton and xenon heavy ions are reported in the present publication. The series of Cr50Ti50N coatings on carbon steel with thickness of 50,..., 300 nm were formed by vacuum arc deposition techniques. Specimens with TiCrN coating on carbon steel were irradiated by low energy 4 He +1 (22 keV) and 4 He +2 (40 keV) ions and high energy Xe +18 and Kr +14 ions with energy of 1.5 MeV/nucleon. Fluence of He ions was 1.0x10 17 ion.cm -2 , fluence of Xe and Kr ions was 5x10 14 -1.0x10 15 ion.cm -2 , irradiation temperature did not exceed 150 deg. C. Study of surface structure was performed by scanning electron microscopy (SEM) and atomic force microscopy (AFM). Methods of Roentgen diffractometry and Rutherford backscattering was applied for determination of structure and thickness of coating. In case of irradiation with Xe +18 and Kr +14 ions an investigation of surface morphology and structure was done after successive two hours vacuum annealing of irradiated samples at temperatures 400 deg. C, 500 deg. C and 600 deg. C. It was shown that after irradiation by Xe and Kr ions on the surface of coating convexities appear, surface density of which correlates with ion flux. In the case of Xe, ions irradiation generated convexities of spherical and elongated shape with dimensions ranging from ten to hundreds nm. In the case of Kr ions, only spherical globules were generated, dimensions of which are 10-30 nm. The most likely explanation of observed surface damage is that: convexities on the surface are generated at ion bombardment of specimens with coating. Convexities are the traces of ions passing through coating and they are due to structural reconstruction at energy release along a trajectory of ions braking. Convexities of elongated shape represent overlapping traces from two passing ions. When the projective range of Xe and Kr ions exceeds coating thickness, damage

Influence of He ions irradiation on the deuterium permeation and retention behavior in the CLF-1 steel

International Nuclear Information System (INIS)

Xu, Yu-Ping; Lu, Tao; Li, Xiao-Chun; Liu, Feng; Liu, Hao-Dong; Wang, Jing; An, Zhong-Qing; Ding, Fang; Hong, Suk-Ho; Zhou, Hai-Shan; Luo, Guang-Nan

2016-01-01

To evaluate the influence of He ions irradiation on the deuterium permeation and retention behavior in RAFM steels, samples made of the CLF-1 steel was irradiated with 3.5 MeV He ions. Gas driven permeation experiments were performed, and the permeability of virgin sample and pre-irradiated sample were obtained and compared. In order to characterize the effect of He ions irradiation on the deuterium retention behavior, deuterium gas exposure was carried out at 623 K, followed by thermal desorption spectra experiments. The total deuterium retention of the CLF-1 steel increased owing to He ions implantation, which could be attributed to the increase in trapping site for deuterium by the He pre-irradiation.

Resonant creep enhancement in austenitic stainless steels due to pulsed irradiation at low doses

International Nuclear Information System (INIS)

Kishimoto, N.; Amekura, H.; Saito, T.

1994-01-01

Steady-state irradiation creep of austenitic stainless steels has been extensively studied as one of the most important design parameters in fusion reactors. The steady-state irradiation creep has been evaluated using in-pile and light-ion experiments. Those creep compliances of various austenitic steels range in the vicinity of ε/Gσ = 10 -6 ∼10 -5 (dpa sm-bullet MPa) -1 , depending on chemical composition etc. The mechanism of steady-state irradiation creep has been elucidated, essentially in terms of stress-induced preferential absorption of point defects into dislocations, and their climb motion. From this standpoint, low doses such as 10 -3 ∼10 -1 dpa would not give rise to any serious creep, and the irradiation creep may not be a critical issue for the low-dose fusion devices including ITER. It is, however, possible that pulsed irradiation causes different creep behaviors from the steady-state one due to dynamic unbalance of interstitials and vacancies. The authors have actually observed anomalous creep enhancement due to pulsed irradiation in austenitic stainless steels. The resonant behavior of creep indicates that pulsed irradiation may cause significant deformation in austenitic steels even at such low doses and slow pulsing rates, especially for the SA-materials. The first-wall materials in plasma operation of ∼10 2 s may suffer from unexpected transient creep, even in the near-term fusion deices, such as ITER. Though this effect might be a transient effect for a relatively short period, it should be taken into account that the pulsed irradiation makes influences on stress relaxation of the fusion components and on the irradiation fatigue. The mechanism and the relevant behaviors of pulse-induced creep will be discussed in terms of a point-defect model based on the resonant interstitial enrichment

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

APFIM investigation of clustering in neutron-irradiated Fe-Cu alloys and pressure vessel steels

International Nuclear Information System (INIS)

Auger, P.; Pareige, P.; Blavette, D.

1996-01-01

Pressure vessel steels used in PWRs are known to be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are commonly supposed to result from the formation of point defects, dislocation loops, voids and copper-rich precipitates. However, the real nature of the irradiation induced damage, in these particularly low copper steels (>0,1 wt%), has not been clearly identify yet. A new experimental work has been carried out thanks to atom probe and field ion microscopy (APFIM) facilities and, more particularly with a new generation of atom probe recently developed, namely the tomographic atom probe (TAP), in order to improve: the understanding of the complex behavior of copper precipitation which occurs when low-alloyed Fe-Cu model alloys are irradiated with neutrons; the microstructural characterization of the pressure vessel steel of the CHOOZ A reactor under various fluences (French Surveillance Programme). The investigations clearly reveal the precipitation of copper-rich clusters in irradiated Fe-Cu alloys while more complicated Si, Ni, Mn and Cu-solute 'clouds' were observed to develop in the low-copper ferritic solid solution of the pressure vessel steel. (authors)

Characterization and Modeling of Grain Boundary Chemistry Evolution in Ferritic Steels under Irradiation

Energy Technology Data Exchange (ETDEWEB)

Marquis, Emmanuelle [Univ. of Michigan, Ann Arbor, MI (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States)

2016-03-28

Ferritic/martensitic (FM) steels such as HT-9, T-91 and NF12 with chromium concentrations in the range of 9-12 at.% Cr and high Cr ferritic steels (oxide dispersion strengthened steels with 12-18% Cr) are receiving increasing attention for advanced nuclear applications, e.g. cladding and duct materials for sodium fast reactors, pressure vessels in Generation IV reactors and first wall structures in fusion reactors, thanks to their advantages over austenitic alloys. Predicting the behavior of these alloys under radiation is an essential step towards the use of these alloys. Several radiation-induced phenomena need to be taken into account, including phase separation, solute clustering, and radiation-induced segregation or depletion (RIS) to point defect sinks. RIS at grain boundaries has raised significant interest because of its role in irradiation assisted stress corrosion cracking (IASCC) and corrosion of structural materials. Numerous observations of RIS have been reported on austenitic stainless steels where it is generally found that Cr depletes at grain boundaries, consistently with Cr atoms being oversized in the fcc Fe matrix. While FM and ferritic steels are also subject to RIS at grain boundaries, unlike austenitic steels, the behavior of Cr is less clear with significant scatter and no clear dependency on irradiation condition or alloy type. In addition to the lack of conclusive experimental evidence regarding RIS in F-M alloys, there have been relatively few efforts at modeling RIS behavior in these alloys. The need for predictability of materials behavior and mitigation routes for IASCC requires elucidating the origin of the variable Cr behavior. A systematic detailed high-resolution structural and chemical characterization approach was applied to ion-implanted and neutron-irradiated model Fe-Cr alloys containing from 3 to 18 at.% Cr. Atom probe tomography analyses of the microstructures revealed slight Cr clustering and segregation to dislocations and

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.

Evaluation of strain-rate sensitivity of ion-irradiated austenitic steel using strain-rate jump nanoindentation tests

Energy Technology Data Exchange (ETDEWEB)

Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University Gokasho, Uji 611-0011, Kyoto (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University Gokasho, Uji 611-0011, Kyoto (Japan); Hamaguchi, Dai; Ando, Masami; Tanigawa, Hiroyasu [Japan Atomic Energy Agency, Rokkasho, Aomori (Japan)

2016-11-01

Highlights: • We examined strain-rate jump nanoindentation on ion-irradiated stainless steel. • We observed irradiation hardening of the ion-irradiated stainless steel. • We found that strain-rate sensitivity parameter was slightly decreased after the ion-irradiation. - Abstract: The present study investigated strain-rate sensitivity (SRS) of a single crystal Fe–15Cr–20Ni austenitic steel before and after 10.5 MeV Fe{sup 3+} ion-irradiation up to 10 dpa at 300 °C using a strain-rate jump (SRJ) nanoindentation test. It was found that the SRJ nanoindentation test is suitable for evaluating the SRS at strain-rates from 0.001 to 0.2 s{sup −1}. Indentation size effect was observed for depth dependence of nanoindentation hardness but not the SRS. The ion-irradiation increased the hardness at the shallow depth region but decreased the SRS slightly.

Remediation of Steel Slag on Acidic Soil Contaminated by Heavy Metal

OpenAIRE

Gu, Haihong; Li, Fuping; Guan, Xiang; Li, Zhongwei; Yu, Qiang

2013-01-01

The technology of in situ immobilization with amendments is an important measure that remediates the soil contaminated by heavy metal, and selecting economical and effective modifier is the key. The effects and mechanism of steel slag, the silicon-rich alkaline by-product which can remediate acidic soil contaminated by heavy metal, are mainly introduced in this paper to provide theory inferences for future research. Firstly, the paper analyzes current research situation of in situ immobilizat...

Characterization of ion irradiation effects on the microstructure, hardness, deformation and crack initiation behavior of austenitic stainless steel:Heavy ions vs protons

Science.gov (United States)

Gupta, J.; Hure, J.; Tanguy, B.; Laffont, L.; Lafont, M.-C.; Andrieu, E.

2018-04-01

Irradiation Assisted Stress Corrosion Cracking (IASCC) is a complex phenomenon of degradation which can have a significant influence on maintenance time and cost of core internals of a Pressurized Water Reactor (PWR). Hence, it is an issue of concern, especially in the context of lifetime extension of PWRs. Proton irradiation is generally used as a representative alternative of neutron irradiation to improve the current understanding of the mechanisms involved in IASCC. This study assesses the possibility of using heavy ions irradiation to evaluate IASCC mechanisms by comparing the irradiation induced modifications (in microstructure and mechanical properties) and cracking susceptibility of SA 304 L after both type of irradiations: Fe irradiation at 450 °C and proton irradiation at 350 °C. Irradiation-induced defects are characterized and quantified along with nano-hardness measurements, showing a correlation between irradiation hardening and density of Frank loops that is well captured by Orowan's formula. Both irradiations (iron and proton) increase the susceptibility of SA 304 L to intergranular cracking on subjection to Constant Extension Rate Tensile tests (CERT) in simulated nominal PWR primary water environment at 340 °C. For these conditions, cracking susceptibility is found to be quantitatively similar for both irradiations, despite significant differences in hardening and degree of localization.

Low temperature fatigue crack propagation in neutron irradiated Type 316 steel and weld metal

International Nuclear Information System (INIS)

Lloyd, G.J.; Walls, J.D.; Gravenor, J.

1981-02-01

The fast cycling fatigue crack propagation characteristics of Type 316 steel and weld metal have been investigated at 380 0 C after irradiation to 1.72-1.92x10 20 n/cm 2 (E>1MeV) and 2.03x10 21 n/cm 2 (E>1MeV) at the same temperature. With mill-annealed Type 316 steel, modest decreases in the rates of crack propagation were observed for both dose levels considered, whereas for cold-worked Type 316 steel irradiation to 2.03x10 21 n/cm 2 (E>1MeV) caused increases in the rate of crack propagation. For Type 316 weld metal, increases in the rate of crack propagation were observed for both dose levels considered. The diverse influences of irradiation upon fatigue crack propagation in these materials are explained by considering a simple continuum mechanics model of crack propagation together with the results of control tensile experiments made on similarly irradiated materials. (author)

Studies of fragileness in steels of vessels of BWR reactors

International Nuclear Information System (INIS)

Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

2003-01-01

The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA MARK lll reactor and separately with Ni +3 ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A 2 . (Author)

Tensile behavior of EUROFER ODS steel after neutron irradiation up to 16.3 dpa between 250 and 450 °C

International Nuclear Information System (INIS)

Materna-Morris, Edeltraud; Lindau, Rainer; Schneider, Hans-Christian; Möslang, Anton

2015-01-01

Highlights: • The first 9%CrWVTa steel (0.5% Y_2O_3), EUROFER ODS HIP, have been neutron irradiated up to 16.3 dpa, between 250 and 450 °C, in the High Flux Reactor (HFR). • After post-irradiation tensile tests, there was not any increase of the upper yield strength or strain localization after irradiation which is typical of RAFM steels. • Initially higher yield strength, R_p_0_._2, and distinctive tensile strength, R_m, of EUROFER ODS HIP compared to EUROFER97 steel. • These values increased due to the neutron irradiation at lower irradiation temperatures. - Abstract: During the development of structural material for future fusion reactors, a 50 kg heat of reduced-activation ferritic-martensitic 9%CrWVTa steel with nanoscaled Y_2O_3-particles, EUROFER97 ODS HIP, was produced using powder metallurgy fabrication technology. This first batch of EUROFER97 ODS HIP and, for comparison, the steel EUROFER97 were prepared for a post-irradiation tensile test program. During neutron irradiation in the HFR (High Flux Reactor, The Netherlands), an accumulated dose of up to 16.3 dpa was reached for 771 days at full power, with the irradiation temperature ranging between 250 and 450 °C. During the post-examinations, all specimens showed the highest tensile strength at lower irradiation temperatures between 250 and 350 °C. However, ODS-alloy and steel were found to clearly differ in the mechanical behavior, which could be documented by fully instrumented tensile tests. In the un-irradiated state, tensile strength of the ODS-alloy already was increased considerably by about 60% compared to the steel. Strengthening was further increased by another 20% after neutron irradiation, but with a much better ductility than observed in the steel. The typical irradiation-induced strain localization of EUROFER97 or RAFM steels could not be observed in the EUROFER97 ODS HIP alloy.

Neutron irradiation effects on mechanical properties in SA508 Gr4N high strength low alloy steel

International Nuclear Information System (INIS)

Kim, Minchul; Lee, Kihyoung; Park, Sanggyu; Choi, Kwonjae; Lee, Bongsang

2012-01-01

The Reactor Pressure Vessel (RPV) is the key component in determining the lifetime of nuclear power plants because it is subject to the significant aging degradation by irradiation and thermal aging, and there is no practical method for replacing that component. Advanced reactors with much larger capacity than current reactor require the usage of higher strength materials inevitably. The SA508 Gr.4N Ni Cr Mo low alloy steel, in which Ni and Cr contents are larger than in conventional RPV steels, could be a promising RPV material offering improved strength and toughness from its tempered martensitic microstructure. For a structural integrity of RPV, the effect of neutron irradiation on the material property is one of the key issues. The RPV materials suffer from the significant degradation of transition properties by the irradiation embrittlement when its strength is increased by a hardening mechanism. Therefore, the potential for application of SA508 Gr.4N steel as the structural components for nuclear power reactors depends on its ability to maintain adequate transition properties against the operating neutron does. However, it is not easy to fine the data on the irradiation effect on the mechanical properties of SA508 Gr.4N steel. In this study, the irradiation embrittlement of SA508 Gr.4N Ni Cr Mo low alloy steel was evaluated by using specimens irradiated in research reactor. For comparison, the variations of mechanical properties by neutron irradiation for commercial SA508 Gr.3 Mn Mo Ni low alloy steel were also evaluated

Investigation of structural materials of reactors using high-energy heavy-ion irradiations

International Nuclear Information System (INIS)

Wang Zhiguang

2007-01-01

Radiation damage in structural materials of fission/fusion reactors is mainly attributed to the evolution of intensive atom displacement damage induced by energetic particles (n, α and/or fission fragments) and high-rate helium doping by direct α particle bombardments and/or (n, α) reactions. It can cause severe degradation of reactor structural materials such as surface blistering, bulk void swelling, deformation, fatigue, embrittlement, stress erosion corrosion and so on that will significantly affect the operation safety of reactors. However, up to now, behavior of structural materials at the end of their service can hardly be fully tested in a real reactor. In the present work, damage process in reactor structural materials is briefly introduced, then the advantages of energetic ion implantation/irradiation especially high-energy heavy ion irradiation are discussed, and several typical examples on simulation of radiation effects in reactor candidate structural materials using high-energy heavy ion irradiations are pronounced. Experimental results and theoretical analysis suggested that irradiation with energetic particles especially high-energy heavy ions is very useful technique for simulating the evolution of microstructures and macro-properties of reactor structural materials. Furthermore, an on-going plan of material irradiation experiments using high energy H- and He-ions based on the Heavy Ion Research Facilities in Lanzhou (HIRFL) is also briefly interpreted. (authors)

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)

Void shrinkage in stainless steel during high energy electron irradiation

International Nuclear Information System (INIS)

Singh, B.N.; Foreman, A.J.E.

1976-03-01

During irradiation of thin foils of an austenitic stainless steel in a high voltage electron microscope, steadily growing voids have been observed to suddenly shrink and disappear at the irradiation temperature of 650 0 Cthe phenomenon has been observed in specimens both with and withoutimplanted helium. Possible mechanisms for void shrinkage during irradiation are considered. It is suggested that the dislocation-pipe-diffusion of vacancies from or of self-interstitial atoms to the voids can explain the shrinkage behaviour of voids observed during our experiments. (author)

Heavy metal contamination and ecological risk of farmland soils adjoining steel plants in Tangshan, Hebei, China.

Science.gov (United States)

Yang, Liyun; Yang, Maomao; Wang, Liping; Peng, Fei; Li, Yuan; Bai, Hao

2018-01-01

The purpose of this study was to determine the heavy metal concentrations and ecological risks to farmland soils caused by atmospheric deposition adjoining five industrial steel districts in Tangshan, Hebei, China. A total of 39 topsoil samples from adjoining these plants were collected and analyzed for Pb, Zn, Cu, Cr, and As. The geo-accumulation index (Igeo) and potential ecological risk index (PERI) were calculated to assess the heavy metal pollution level in soils. The results showed that the levels of Pb and As in farmland soils adjoining all steel plants were more than the background value, with the As content being excessively high. The Cr and Cu contents of some samples were over the background values, but the Zn content was not. In all the research areas, the largest Igeo value of the heavy metals was for As, followed by Pb, and the largest monomial PERI ([Formula: see text]) was As, which showed that the pollution of As in farmland soils was significant and had considerable ecological risk. Additionally, the heavy metal sequential extraction experiments showed that Pb and Cr, which exceeded the background value, were present in about 20% of the exchangeable and carbonate-bound fractions in the soils surrounding some steel plants. This would imply the risk of these heavy metals being absorbed and accumulated by the crops. Therefore, the local government needs to control the pollution of heavy metals in the farmland soils adjoining the steel plant as soon as possible, in order to avoid possible ecological and food safety risks.

Improvement of the mechanical and frictional properties of steels by continuous and pulsed ion irradiation

International Nuclear Information System (INIS)

Romanov, I.G.

1992-01-01

Effect of continuous and powerful pulsed ion beams (PIB) on structural, mechanical, tribological properties and surface morphology of steels were investigated. The results obtained demonstrate the significant influence of ion irradiation type on microhardness, friction coefficient, wear resistance and surface roughness characteristics. Friction coefficient variation in irradiated steels is interpreted within the framework of an adhesion-deformation model

The natural aging of austenitic stainless steels irradiated with fast neutrons

Science.gov (United States)

Rofman, O. V.; Maksimkin, O. P.; Tsay, K. V.; Koyanbayev, Ye. T.; Short, M. P.

2018-02-01

Much of today's research in nuclear materials relies heavily on archived, historical specimens, as neutron irradiation facilities become ever more scarce. These materials are subject to many processes of stress- and irradiation-induced microstructural evolution, including those during and after irradiation. The latter of these, referring to specimens "naturally aged" in ambient laboratory conditions, receives far less attention. The long and slow set of rare defect migration and interaction events during natural aging can significantly change material properties over decadal timescales. This paper presents the results of natural aging carried out over 15 years on austenitic stainless steels from a BN-350 fast breeder reactor, each with its own irradiation, stress state, and natural aging history. Natural aging is shown to significantly reduce hardness in these steels by 10-25% and partially alleviate stress-induced hardening over this timescale, showing that materials evolve back towards equilibrium even at such a low temperature. The results in this study have significant implications to any nuclear materials research program which uses historical specimens from previous irradiations, challenging the commonly held assumption that materials "on the shelf" do not evolve.

A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

International Nuclear Information System (INIS)

Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky

1999-03-01

The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

Characterization of atom clusters in irradiated pressure vessel steels and model alloys

International Nuclear Information System (INIS)

Auger, P.; Pareige, P.; Akamatsu, M.; Van Duysen, J.C.

1993-12-01

In order to characterize the microstructural evolution of the iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions and, for comparison, low copper model alloys irradiated with neutrons and electrons have been studied. The characterization has been carried out mainly thanks to small angle neutron scattering and atom probe experiments. Both techniques lead to the conclusion that clusters develop with irradiations. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex. Solute atoms like Ni, Mn and Si, sometimes associated with Cu, segregate as ''clouds'' more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs

Evolution of precipitation in reactor pressure vessel steel welds under neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Lindgren, Kristina, E-mail: kristina.lindgren@chalmers.se [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden); Boåsen, Magnus [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Stiller, Krystyna [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden); Efsing, Pål [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Vattenfall Ringhals AB, SE-430 22 Väröbacka (Sweden); Thuvander, Mattias [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden)

2017-05-15

Reactor pressure vessel steel welds are affected by irradiation during operation. The irradiation results in nanometre cluster formation, which in turn affects the mechanical properties of the material, e.g. the ductile-to-brittle transition temperature is shifted to higher levels. In this study, cluster formation is characterised in high Ni (1.58%) low Cu (0.04%) steel welds identical to Ringhals R4 welds, using atom probe tomography in both surveillance material and in material irradiated at accelerated dose rates. Clusters containing mainly Ni and Mn, but also some Si and Cu were observed in all of the irradiated materials. Their evolution did not change drastically during irradiation; the clusters grew and new clusters were nucleated. Hence, both the cluster number density and the average size increased with irradiation time. Some flux effects were observed when comparing the high flux material and the surveillance material. The surveillance material has a lower cluster number density, but larger clusters. The resulting impact on the mechanical properties of these two effects cancel out, resulting in a measured hardness that seems to be on the same trend as the high flux material. The dispersed barrier hardening model with an obstacle strength factor of 0.15 was found to reproduce the increase in hardness. In the investigated high flux materials, the clusters' Cu content was higher. - Highlights: •Clustering in a low Cu, high Ni reactor pressure vessel steel weld is studied. •The clusters nucleate and grow during irradiation, and consist of Ni, Mn, Si, and Cu. •High flux neutron irradiated material is compared to surveillance material. •High flux was found to result in smaller clusters with a larger number density. •Hardness follows the same dependence on fluence, independent of flux.

Investigation of instability of M23C6 particles in F82H steel under electron and ion irradiation conditions

Science.gov (United States)

Kano, Sho; Yang, Huilong; Shen, Jingjie; Zhao, Zishou; McGrady, John; Hamaguchi, Dai; Ando, Mamami; Tanigawa, Hiroyasu; Abe, Hiroaki

2018-04-01

In order to clarify the instability of M23C6 in F82H steel under irradiation, both electron irradiation using a high voltage electron microscope (HVEM) and ion irradiation using an ion accelerator were performed. For the electron irradiation, in-situ observation under 2 MV electron irradiation and ex-situ high resolution electron microscopic (HREM) analysis were utilized to evaluate the response of M23C6 against irradiation. The temperature dependence of the irradiation induced instability of the carbide was first confirmed: 293 K indicating severe loss of crystallinity due to dissolution of the constituent atoms though irradiation-enhanced diffusion under the vacancy diffusion by the focused electron beam irradiation. For the ion irradiation, 10.5 MeV-Fe3+ ion was applied to bombard the F82H steel at 673 K to achieve the displacement damage of ≈20 dpa at the depth of 1.0 μm from surface. Cross-section TEM specimens were prepared by a focused ion beam technique. The shrinkage of carbide particles was observed especially near the irradiation surface. Besides, the lattice fringes at the periphery of carbide were observed in the irradiated M23C6 by the HREM analysis, which is different from that observed in the electron irradiation. It was clarified that the instability of M23C6 is dependent on the irradiation conditions, indicating that the flow rate of vacancy type defects might be the key factor to cause the dissolution of constituent atoms of carbide particles into matrix under irradiation.

Double-differential heavy-ion production cross sections

International Nuclear Information System (INIS)

Miller, T. M.; Townsend, L. W.

2004-01-01

Current computational tools used for space or accelerator shielding studies transport energetic heavy ions either using a one-dimensional straight-ahead approximation or by dissociating the nuclei into protons and neutrons and then performing neutron and proton transport using Monte Carlo techniques. Although the heavy secondary particles generally travel close to the beam direction, a proper treatment of the light ions produced in these reactions requires that double-differential cross sections should be utilised. Unfortunately, no fundamental nuclear model capable of serving as an event generator to provide these cross sections for all ions and energies of interest exists currently. Herein, we present a model for producing double-differential heavy-ion production cross sections that uses heavy-ion fragmentation yields produced by the NUCFRG2 fragmentation code coupled with a model of energy degradation in nucleus-nucleus collisions and systematics of momentum distributions to provide energy and angular dependences of the heavy-ion production. (authors)

Radiation blistering of stainless steel

International Nuclear Information System (INIS)

Yoshii, Naritsugu; Tanabe, Tetsuo; Imoto, Shosuke

1980-01-01

Surface blistering of stainless steels due to 20 keV He + ion bombardment has been investigated by examination of surface topography with a scanning electron microscope (SEM) and an optical microscope. Blisters of 0.1 to 2 μm in diameter are observed in all samples irradiated with fluence of about 1 x 10 18 He + /cm 2 at any temperature between -80 0 C and 500 0 C. With increasing the fluence blister covers are ruptured and exfoliated and finally the surface becomes rough surface without traces of blister formation. The surface effect is severer at 500 0 C than at 100 0 C irradiation. Also in double-phase stainless steel DP-3, similar surface topography to 316 SS is observed. But by the difference of the erosion rate by sputtering of the surface between α-phase and γ-phase, a striped pattern appears in DP-3 with heavy irradiation of about 2 x 10 19 He + /cm 2 . (author)

Fatigue behavior of Type 316 stainless steel following neutron irradiation inducing helium

International Nuclear Information System (INIS)

Grossbeck, M.L.; Liu, K.C.

1980-01-01

Since a tokamak reactor operates in a cyclic mode, thermal stresses will result in fatigue in structural components, especially in the first wall and blanket. There has been limited work on fatigue in irradiated alloys but none on irradiated materials containing significant amounts of irradiation-induced helium. To provide scoping data and to study the effects of irradiation on fatigue behavior, 20%-cold-worked type 316 stainless steel from the MFE reference heat was studied

Coolant compatibility studies. The effect of irradiation on tensile properties and stress corrosion cracking sensitivity of martensitic steels. MANET 4 - complementary studies

International Nuclear Information System (INIS)

Nystrand, A.C.

1994-02-01

Tensile and stress corrosion cracking tests have been carried out on MANET-type (1.4914 and FV448) and reduced activation (LA12TaLC) high-chromium martensitic steels. The materials had previously been exposed up to 5000 h at ∼275 degrees C in the core, above the core and remote from the core of a high pressure water loop in the Studsvik R2 reactor. After the mechanical testing the materials were examined visually and metallographically. The steel samples exposed in the core section showed large increases in tensile yield strengths when tested at 250 degrees C. However, the magnitude of the radiation hardening was considerably smaller in the reduced activation steel compared to the commercial steels; this observation is consistent with published data on other high-chromium martensitic steels and is associated with the lower chromium content of the LA12TaLC steel (8.9%) compared with those of the commercial steels (10.6 and 11.3%). Irradiation assisted stress corrosion cracking (IASCC) was not detected in any of the stressed steel samples after autoclave testing for times up to 1500 h at 250 degrees C in air-saturated high purity water. This apparent resistance to IASCC may be due to the high chromium martensitic steels not being sensitized by the irradiation in a comparable manner to that shown by the austenitic steels. However, additional studies are required to clarify some of the existing uncertainties with respect to IASCC of these martensitic steels

The radiation swelling effect on fracture properties and fracture mechanisms of irradiated austenitic steels. Part I. Ductility and fracture toughness

Energy Technology Data Exchange (ETDEWEB)

Margolin, B., E-mail: mail@crism.ru; Sorokin, A.; Shvetsova, V.; Minkin, A.; Potapova, V.; Smirnov, V.

2016-11-15

The radiation swelling effect on the fracture properties of irradiated austenitic steels under static loading has been studied and analyzed from the mechanical and physical viewpoints. Experimental data on the stress-strain curves, fracture strain, fracture toughness and fracture mechanisms have been represented for austenitic steel of 18Cr-10Ni-Ti grade (Russian analog of AISI 321 steel) irradiated up to neutron dose of 150 dpa with various swelling. Some phenomena in mechanical behaviour of irradiated austenitic steels have been revealed and explained as follows: a sharp decrease of fracture toughness with swelling growth; untypical large increase of fracture toughness with decrease of the test temperature; some increase of fracture toughness after preliminary cyclic loading. Role of channel deformation and channel fracture has been clarified in the properties of irradiated austenitic steel and different tendencies to channel deformation have been shown and explained for the same austenitic steel irradiated at different temperatures and neutron doses.

Formation of austenite in high Cr ferritic/martensitic steels by high fluence neutron irradiation

Science.gov (United States)

Lu, Z.; Faulkner, R. G.; Morgan, T. S.

2008-12-01

High Cr ferritic/martensitic steels are leading candidates for structural components of future fusion reactors and new generation fission reactors due to their excellent swelling resistance and thermal properties. A commercial grade 12%CrMoVNb ferritic/martensitic stainless steel in the form of parent plate and off-normal weld materials was fast neutron irradiated up to 33 dpa (1.1 × 10 -6 dpa/s) at 400 °C and 28 dpa (1.7 × 10 -6 dpa/s) at 465 °C, respectively. TEM investigation shows that the fully martensitic weld metal transformed to a duplex austenite/ferrite structure due to high fluence neutron irradiation, the austenite was heavily voided (˜15 vol.%) and the ferrite was relatively void-free; whilst no austenite phases were detected in plate steel. Thermodynamic and phase equilibria software MTDATA has been employed for the first time to investigate neutron irradiation-induced phase transformations. The neutron irradiation effect is introduced by adding additional Gibbs free energy into the system. This additional energy is produced by high energy neutron irradiation and can be estimated from the increased dislocation loop density caused by irradiation. Modelling results show that neutron irradiation reduces the ferrite/austenite transformation temperature, especially for high Ni weld metal. The calculated results exhibit good agreement with experimental observation.

Research on reconstruction of steel tube section from few projections

International Nuclear Information System (INIS)

Peng Shuaijun; Wu Haifeng; Wang Kai

2007-01-01

Most parameters of steel tube can be acquired from CT image of the section so as to evaluate its quality. But large numbers of projections are needed in order to reconstruct the section image, so the collection and calculation of the projections consume lots of time. In order to solve the problem, reconstruction algorithms of steel tube from few projections are researched and the results are validated with simulation data in the paper. Three iterative algorithms, ART, MAP and OSEM, are attempted to reconstruct the section of steel tube by using the simulation model. Considering the prior information distributing of steel tube, we improve the algorithms and get better reconstruction images. The results of simulation experiment indicate that ART, MAP and OSEM can reconstruct accurate section images of steel tube from less than 20 projections and approximate images from 10 projections. (authors)

Irradiation proposition of ferritic steels in a russian reactor

International Nuclear Information System (INIS)

Seran, J.L.; Decours, J.; Levy, L.

1987-04-01

Using the low temperatures of russian reactors, a sample irradiation is proposed to study mechanical properties and swelling of martensitic steels (EM10, T91, 1.4914, HT9), ferrito-martensitic (EM12) and ferritic (F17), at temperatures lower than 400 0 C [fr

Neutron irradiation test of copper alloy/stainless steel joint materials

International Nuclear Information System (INIS)

Yamada, Hirokazu; Kawamura, Hiroshi

2006-01-01

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

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)

Radiation Stability of Nanoclusters in Nano-structured Oxide Dispersion Strengthened (ODS) Steels

International Nuclear Information System (INIS)

Certain, Alicia G.; Kuchibhatla, Satyanarayana; Shutthanandan, V.; Allen, T. R.

2013-01-01

Nanostructured oxide dispersion strengthened (ODS) steels are considered candidates for nuclear fission and fusion applications at high temperature and dose. The complex oxide nanoclusters in these alloys provide high-temperature strength and are expected to afford better radiation resistance. Proton, heavy ion, and neutron irradiations have been performed to evaluate cluster stability in 14YWT and 9CrODS steel under a range of irradiation conditions. Energy-filtered transmission electron microscopy and atom probe tomography were used in this work to analyze the evolution of the oxide population.

Studies in heavy ion activation analysis Pt. 5

International Nuclear Information System (INIS)

Ojo, J.F.; Lass, B.D.; Schweikert, E.A.

1980-01-01

Nondestructive heavy ion activation analysis has been used to determine the carbon content in various NBS SRM steel samples with a 7.0 MeV 6 Li + beam. The reaction 12 C( 6 Li,αn) 13 N allows for carbon analysis with the only possible interference being beryllium, 9 Be( 6 Li,2n) 13 N. Under interference-free conditions, and employing a post-irradiation etch, the detection limit for carbon analysis in steel was 5 ppm. (author)

Quantitative analysis of genes regulating sensitivity to heavy ion irradiation in cultured cell lines of malignant choroid melanoma

International Nuclear Information System (INIS)

Kumagai, Ken; Adachi, Nanao; Nimura, Yoshinori

2004-01-01

As a treatment strategy for malignant melanoma, heavy ion irradiation has been planned in National Institute of Radiological Sciences (NIRS). However, the molecular biology of the malignant melanoma cell after irradiation of heavy ion is still unknown. In this study, we used resistant and sensitive cell lines of malignant melanoma to study the effects of heavy ion irradiation. Furthermore, gene expression profiling of early response genes for heavy ion irradiation was carried out on these cell lines using microarray technology. (author)

Quantitative analysis of genes regulating sensitivity to heavy ion irradiation in cultured cell lines of malignant choroid melanoma

International Nuclear Information System (INIS)

Kumagai, Ken; Nimura, Yoshinori; Kato, Masaki; Seki, Naohiko; Miyahara, Nobuyuki; Aoki, Mizuho; Shino, Yayoi; Furusawa, Yoshiya; Mizota, Atsushi

2005-01-01

As a treatment strategy for malignant melanoma, heavy ion irradiation has been planned in National Institute of Radiological Sciences (NIRS). However, the molecular biology of the malignant melanoma cell after irradiation of heavy ion is still unknown. In this study, we used resistant and sensitive cell lines of malignant melanoma to study the effects of heavy ion irradiation. Furthermore, gene expression profiling of early response genes for heavy ion irradiation was carried out on these cell lines using microarray technology. (author)

Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-08-01

The embrittlement trend curve development project for HFIR reactor pressure vessel (RPV) steels was carried out with three major tasks. Which are (1) data collection to match that used in HFIR steel embrittlement trend published in 1994 Journal Nuclear Material by Remec et. al, (2) new embrittlement data of A212B steel that are not included in earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with the consideration of the irradiation temperature effect. An updated HFIR RPV steel embrittlement trend curve was developed, as described below. NDTT( C) = 23.85 log(x) + 203.3 log (x) + 434.7, with 2- uncertainty of 34.6 C, where parameter x is referred to total dpa. The developed update HFIR RPV embrittlement trend curve has higher embrittlement rate compared to that of the trend curve developed in 1994.

Development of heavy-ion irradiation technique for single-event in semiconductor devices

Energy Technology Data Exchange (ETDEWEB)

Nemoto, Norio; Akutsu, Takao; Matsuda, Sumio [National Space Development Agency of Japan, Tsukuba, Ibaraki (Japan). Tsukuba Space Center; Naitoh, Ichiro; Itoh, Hisayoshi; Agematsu, Takashi; Kamiya, Tomihiro; Nashiyama, Isamu

1997-03-01

Heavy-ion irradiation technique has been developed for the evaluation of single-event effects on semiconductor devices. For the uniform irradiation of high energy heavy ions to device samples, we have designed and installed a magnetic beam-scanning system in a JAERI cyclotron beam course. It was found that scanned area was approximately 4 x 2 centimeters and that the deviation of ion fluence from the average value was less than 7%. (author)

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

Microstructural observation of ion-irradiated austenitic stainless steel

International Nuclear Information System (INIS)

Sawai, T.; Hamada, S.; Hishinuma, A.

1992-01-01

Type 316 stainless steel, base metal and weld metal obtained from an electron beam weld joint, was irradiated with 90 MeV Br +6 in the JAERI tandem accelerator. Cross-sectional TEM specimens were obtained by nickel plating. Microstructural observation revealed a band of tiny dislocation loops was observed around the mean projected range and the measured distance from the surface was 6.75±0.15μm. This is slightly larger than the calculated value using Ziegler's stopping power. Defect clusters were also observed around defect sinks within the mean projected range, suggesting cascade-sink interaction. These sinks are the grain boundary in the base metal specimen and preexisting dislocation lines in the weld metal specimen. Surface roughness of polished specimen was detected at the shallower side of the peak damage band, although no visible crystalline defect cluster was observed. This suggests radiation-induced microchemical evolution prior to sever microstructural evolution. (author)

Parametric Study of Fire Performance of Concrete Filled Hollow Steel Section Columns with Circular and Square Cross-Section

Science.gov (United States)

Nurfaidhi Rizalman, Ahmad; Tahir, Ng Seong Yap Mahmood Md; Mohammad, Shahrin

2018-03-01

Concrete filled hollow steel section column have been widely accepted by structural engineers and designers for high rise construction due to the benefits of combining steel and concrete. The advantages of concrete filled hollow steel section column include higher strength, ductility, energy absorption capacity, and good structural fire resistance. In this paper, comparison on the fire performance between circular and square concrete filled hollow steel section column is established. A three-dimensional finite element package, ABAQUS, was used to develop the numerical model to study the temperature development, critical temperature, and fire resistance time of the selected composite columns. Based on the analysis and comparison of typical parameters, the effect of equal cross-sectional size for both steel and concrete, concrete types, and thickness of external protection on temperature distribution and structural fire behaviour of the columns are discussed. The result showed that concrete filled hollow steel section column with circular cross-section generally has higher fire resistance than the square section.

Bootstrap calculation of ultimate strength temperature maxima for neutron irradiated ferritic/martensitic steels

Science.gov (United States)

Obraztsov, S. M.; Konobeev, Yu. V.; Birzhevoy, G. A.; Rachkov, V. I.

2006-12-01

The dependence of mechanical properties of ferritic/martensitic (F/M) steels on irradiation temperature is of interest because these steels are used as structural materials for fast, fusion reactors and accelerator driven systems. Experimental data demonstrating temperature peaks in physical and mechanical properties of neutron irradiated pure iron, nickel, vanadium, and austenitic stainless steels are available in the literature. A lack of such an information for F/M steels forces one to apply a computational mathematical-statistical modeling methods. The bootstrap procedure is one of such methods that allows us to obtain the necessary statistical characteristics using only a sample of limited size. In the present work this procedure is used for modeling the frequency distribution histograms of ultimate strength temperature peaks in pure iron and Russian F/M steels EP-450 and EP-823. Results of fitting the sums of Lorentz or Gauss functions to the calculated distributions are presented. It is concluded that there are two temperature (at 360 and 390 °C) peaks of the ultimate strength in EP-450 steel and single peak at 390 °C in EP-823.

Microstructural evolution of HFIR-irradiated low activation F82H and F82H-10B steels

International Nuclear Information System (INIS)

Wakai, E.; Shiba, K.; Sawai, T.; Hashimoto, N.; Robertson, J.P.; Klueh, R.L.

1998-01-01

Microstructures of reduced-activation F82H (8Cr-2W-0.2V-0.04Ta) and the F82H steels doped with 10 B, irradiated at 250 and 300 C to 3 and 57 dpa in the High Flux Isotope Reactor (HFIR), were examined by TEM. In the F82H irradiated at 250 C to 3 dpa, dislocation loops, small unidentified defect clusters with a high number density, and a few MC precipitates were observed in the matrix. The defect microstructure after 300 C irradiation to 57 dpa is dominated by the loops, and the number density of loops was lower than that of the F82H- 10 B steel. Cavities were observed in the F82H- 10 B steels, but the swelling value is insignificant. Small particles of M 6 C formed on the M 23 C 6 carbides that were present in both steels before the irradiation at 300 C to 57 dpa. A low number density of MC precipitate particles formed in the matrix during irradiation at 300 C to 57 dpa

A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

Energy Technology Data Exchange (ETDEWEB)

Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky [NRI Czech (Czech Republic)

1999-03-01

The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

Heavy ion induced mutation in arabidopsis

Energy Technology Data Exchange (ETDEWEB)

Tano, Shigemitsu [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment

1997-03-01

Heavy ions, He, C, Ar and Ne were irradiated to the seeds of Arabidopsis thaliana for inducing the new mutants. In the irradiated generation (M{sub 1}), germination and survival rate were observed to estimate the relative biological effectiveness in relation to the LET including the inactivation cross section. Mutation frequencies were compared by using three kinds of genetic loci after irradiation with C ions and electrons. Several interesting new mutants were selected in the selfed progenies of heavy ion irradiated seeds. (author)

Radiation response of ODS ferritic steels with different oxide particles under ion-irradiation at 550 °C

Science.gov (United States)

Song, Peng; Morrall, Daniel; Zhang, Zhexian; Yabuuchi, Kiyohiro; Kimura, Akihiko

2018-04-01

In order to investigate the effects of oxide particles on radiation response such as hardness change and microstructural evolution, three types of oxide dispersion strengthened (ODS) ferritic steels (named Y-Ti-ODS, Y-Al-ODS and Y-Al-Zr-ODS), mostly strengthened by Y-Ti-O, Y-Al-O and Y-Zr-O dispersoids, respectively, were simultaneously irradiated with iron and helium ions at 550 °C up to a damage of 30 dpa and a corresponding helium (He) concentration of ∼3500 appm to a depth of 1000-1300 nm. A single iron ion beam irradiation was also performed for reference. Transmission electron microscopy revealed that after the dual ion irradiation helium bubbles of 2.8, 6.6 and 4.5 nm in mean diameter with the corresponding number densities of 1.1 × 1023, 2.7 × 1022 and 3.6 × 1022 m-3 were observed in Y-Ti-ODS, Y-Al-ODS and Y-Al-Zr-ODS, respectively, while no such bubbles were observed after single ion irradiation. About 80% of intragranular He bubbles were adjacent to oxide particles in the ODS ferritic steels. Although the high number density He bubbles were observed in the ODS steels, the void swelling in Y-Ti-ODS, Y-Al-ODS and Y-Al-Zr-ODS was still small and estimated to be 0.13%, 0.53% and 0.20%, respectively. The excellent swelling resistance is dominantly attributed to the high sink strength of oxide particles that depends on the morphology of particle dispersion rather than the crystal structure of the particles. In contrast, no dislocation loops were produced in any of the irradiated steels. Nanoindentation measurements showed that no irradiation hardening but softening was found in the ODS ferritic steels, which was probably due to irradiation induced dislocation recovery. The helium bubbles in high number density never contributed to the irradiation hardening of the ODS steels at these irradiation conditions.

Irradiation hardening of Mod.9Cr-1Mo steel

International Nuclear Information System (INIS)

Ryu, Woo-Seog; Kim, Sung-Ho; Choo, Kee-Nam; Kim, Do-Sik

2009-01-01

An irradiation test of Mod.9Cr-1Mo steel was carried out in the OR5 test hole of HANARO of a 30 MW thermal power at 390±10degC up to a fast neutron fluence of 4.4x10 19 (n/cm 2 ) (E > 1.0 MeV). The dpa of the irradiated specimens was evaluated to be 0.034 - 0.07. Tensile and impact tests of the irradiated Mod.9Cr-1Mo were done in the hot cell of the IMEF. The change of the tensile strength by irradiation was similar to the change of the yield strength. The increase of the yield and tensile strengths was up to 18% and 10% respectively. The elongation reduction of the weldment was up to 65%. (author)

Radiation-induced segregation at grain boundaries in AL-6XN stainless steels irradiated by hydrogen ions

Science.gov (United States)

Long, Yunxiang; Zheng, Zhongcheng; Guo, Liping; Zhang, Weiping; Shen, Zhenyu; Tang, Rui

2018-04-01

The effect of high concentration of hydrogen on the segregation of radiation-induced segregation (RIS) in AL-6XN stainless steels has been investigated by transmission electron microscopy (TEM) with energy-dispersive X-ray spectroscopy. Specimens were irradiated with 100 keV H2+ ions from 1 dpa to 5 dpa at 380 °C to investigated the dose dependence of grain boundary RIS. A specimen was irradiated to 5 dpa at 290 °C to study the effect of irradiation temperature. The trends of Cr depletion and Ni enrichment with irradiation dose is similar to that of other austenitic steels reported in the literatures, but the higher concentration of hydrogen made the RIS profile wider. An abnormal phenomenon that the degree of RIS increased with decreasing irradiation temperature was found, indicating that with the retention of hydrogen in the steels, temperature dependence of RIS is dominated by the quantity of retained hydrogen, rather than by thermal segregation processes.

Mechanical properties and TEM examination of RAFM steels irradiated up to 70 dpa in BOR-60

Energy Technology Data Exchange (ETDEWEB)

Gaganidze, E., E-mail: Ermile.Gaganidze@kit.edu [Karlsruher Institut fuer Technologie, Institut fuer Angewandte Materialien, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Petersen, C.; Materna-Morris, E.; Dethloff, C.; Weiss, O.J.; Aktaa, J. [Karlsruher Institut fuer Technologie, Institut fuer Angewandte Materialien, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Povstyanko, A.; Fedoseev, A.; Makarov, O.; Prokhorov, V. [Joint Stock Company ' State Scientific Centre Research Institute of Atomic Reactors' , 433510 Dimitrovgrad-10, Ulyanovsk Region (Russian Federation)

2011-10-01

Mechanical properties of Reduced Activation Ferritic/Martensitic (RAFM) steels were studied after irradiation in BOR-60 reactor to a neutron displacement damage of 70 dpa at 330-340 deg. C. Yield stress and Ductile-to-Brittle-Transition-Temperature of EUROFER97 indicate saturation of hardening and embrittlement. The phenomenological models for description of microstructure evolution and resulting irradiation hardening and embrittlement are discussed. The evolution of yield stress with dose is qualitatively understood within a Whapham and Makin model. Dislocation loops examined in TEM are considered a main source for low-temperature irradiation hardening. The analysis of the fatigue data in terms of the inelastic strain reveals comparable fatigue behaviour both for unirradiated and irradiated conditions, which can be described by a common Manson-Coffin relation. The study of helium effects in B-doped model steels indicated progressive material embrittlement with helium content. Post-irradiation annealing of RAFM steels yielded substantial recovery of mechanical properties.

Radiation damage in heavy irradiated aluminum nitride

Energy Technology Data Exchange (ETDEWEB)

Atobe, Kozo; Honda, Makoto; Fukuoka, Noboru [Naruto Univ. of education, Tokushima (Japan); Okada, Moritami; Nakagawa, Masuo

1996-04-01

AlN, one of candidate for ceramic materials used in nuclear fusion reactor, was irradiated by fast and thermal neutrons. The high concentration of irradiated defects and the nuclear transformation elements were detected by electron spin resonance (ESR) and x-ray photoelectron spectroscopy (XPS) method. The exposure of fast neutron and thermal neutron were 1.2x10{sup 20}n/cm{sup 2} and 1.2x10{sup 21}n/cm{sup 2}, respectively. The spreads of ESR spectra of ultra hyperfine structure depending on interaction between {sup 27}Al nuclear spin and electron trapped in tetrahedron consisted of Al atoms was found in the spectra of heavy irradiated AlN. F type defects was estimated 10{sup 19}n/cm{sup 3}. Photoelectrons from 2s and 2p in {sup 28}Si which produced in process of {beta}-decay of {sup 27}Al(n,{gamma}){sup 28}Al were observed in XPS spectra of irradiated samples. (S.Y.)

Radiation damage in heavy irradiated aluminum nitride

International Nuclear Information System (INIS)

Atobe, Kozo; Honda, Makoto; Fukuoka, Noboru; Okada, Moritami; Nakagawa, Masuo.

1996-01-01

AlN, one of candidate for ceramic materials used in nuclear fusion reactor, was irradiated by fast and thermal neutrons. The high concentration of irradiated defects and the nuclear transformation elements were detected by electron spin resonance (ESR) and x-ray photoelectron spectroscopy (XPS) method. The exposure of fast neutron and thermal neutron were 1.2x10 20 n/cm 2 and 1.2x10 21 n/cm 2 , respectively. The spreads of ESR spectra of ultra hyperfine structure depending on interaction between 27 Al nuclear spin and electron trapped in tetrahedron consisted of Al atoms was found in the spectra of heavy irradiated AlN. F type defects was estimated 10 19 n/cm 3 . Photoelectrons from 2s and 2p in 28 Si which produced in process of β-decay of 27 Al(n,γ) 28 Al were observed in XPS spectra of irradiated samples. (S.Y.)

Atom probe study of the microstructural evolution induced by irradiation in Fe-Cu ferritic alloys and pressure vessel steels

International Nuclear Information System (INIS)

Pareige, P.

1996-04-01

Pressure vessel steels used in pressurized water reactors are low alloyed ferritic steels. They may be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are generally supposed to result from the formation of point defects, dislocation loops, voids and/or copper rich clusters. However, the real nature of the irradiation induced-damage in these steels has not been clearly identified yet. In order to improve our vision of this damage, we have characterized the microstructure of several steels and model alloys irradiated with electrons and neutrons. The study was performed with conventional and tomographic atom probes. The well known importance of the effects of copper upon pressure vessel steel embrittlement has led us to study Fe-Cu binary alloys. We have considered chemical aging as well as aging under electron and neutron irradiations. The resulting effects depend on whether electron or neutron irradiations ar used for thus. We carried out both kinds of irradiation concurrently so as to compare their effects. We have more particularly considered alloys with a low copper supersaturation representative of that met with the French vessel alloys (0.1% Cu). Then, we have examined steels used on French nuclear reactor pressure vessels. To characterize the microstructure of CHOOZ A steel and its evolution when exposed to neutrons, we have studied samples from the reactor surveillance program. The results achieved, especially the characterization of neutron-induced defects have been compared with those for another steel from the surveillance program of Dampierre 2. All the experiment results obtained on model and industrial steels have allowed us to consider an explanation of the way how the defects appear and grow, and to propose reasons for their influence upon steel embrittlement. (author). 3 appends

Heavy concrete shieldings made of recycled radio-active steel

International Nuclear Information System (INIS)

Holland, D.; Quade, U.; Sappok, M.; Heim, H.

1998-01-01

Maintenance and decommissioning of nuclear installations will generate increasing quantities of radioactively contaminated metallic residues. For many years, Siempelkamp has been melting low-level radioactive scrap in order to re-use it for containers of nuclear industry. Another new recycling path has recently been developed by producing steel granules from the melt. These granules are used as replacement for hematite (iron ore) in the production of heavy concrete shieldings. In the CARLA plant (central plant for the recycling of low-level radioactive waste) of Siempelkamp Nuklear- und Umwelttechnik GmbH and Co., the scrap is melted in a medium frequency induction furnace. The liquid iron is poured into a cooling basin through a water jet, which splits the iron into granules. The shape of these granules is determined by various factors, such as water jet speed, pouring rate of the liquid iron and different additives to the melt. In this process, massive spheres with diameters ranging from 1 to 8 mm can be produced which add to the density of heavy concrete elements for optimum shielding. In close cooperation with Boschert, which indeed is an expert for the production of concrete shieldings, a new technology for manufacturing heavy concrete shieldings, containing low-level radioactive steel granules, has been developed. The portion of steel granules in the concrete is approx. 50 weight-%. A concrete density between 2.4 kg/dm 3 and 4.0 kg/dm 3 is available. The compressive strength for the concrete reaches values up to 65 MPa. Different types of Granulate Shielding Casks (GSC) are offered by Siempelkamp. The most famous one is the GSC 200 for 200 1 drums, which has already been qualified for final storage of radioactive wastes at the German Morsleben final repository (ERAM). This newly developed recycling process further increases the quantities of low-level radioactive metallic wastes available for recycling. Expensive storage area can thus be saved respectively

The irradiation embrittlement of two pressure vessel steels -Contribution of local approach

Energy Technology Data Exchange (ETDEWEB)

Soulat, P; Marini, B [CEA Centre d` Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France). Service de Recherches Metallurgiques Appliquees; Miannay, D; Horowitz, H [CEA Centre d` Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire; Schill, R [CEA Centre d` Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie

1994-12-31

Within the IAEA Coordinated Research Programme on ``Optimizing the Reactor Pressure Vessel Surveillance Programmes and their Analyses``, the French participation has been focused on the contribution of the local approach to the determination of the sensitivity to radiation embrittlement of two different pressure vessel steels: a low sensitive French forging steel (FFA) and a high sensitive ``monitor`` Japanese plate steel (JRQ) were irradiated to a fluence of 3.10{sup 19} n/cm{sup 2} at 290 C. The irradiation embrittlement of the two steels measured by the shift of Charpy V transition curves is in good agreement with the estimated shifts given by theoretical prediction. The fracture toughness properties were examined at low temperature with brittle fracture, and at service temperature (290 C), with ductile tearing. The values of K{sub 1C} or K{sub JC} for the brittle fracture and J{sub 1C} for the ductile fracture are compared to predictions established using the local approach of cleavage fracture (Weibull analysis) and the critical rate of void growth respectively. 8 refs., 14 figs., 10 tabs.

Study on prediction model of irradiation embrittlement for reactor pressure vessel steel

International Nuclear Information System (INIS)

Wang Rongshan; Xu Chaoliang; Huang Ping; Liu Xiangbing; Ren Ai; Chen Jun; Li Chengliang

2014-01-01

The study on prediction model of irradiation embrittlement for reactor pres- sure vessel (RPV) steel is an important method for long term operation. According to the deep analysis of the previous prediction models developed worldwide, the drawbacks of these models were given and a new irradiation embrittlement prediction model PMIE-2012 was developed. A corresponding reliability assessment was carried out by irradiation surveillance data. The assessment results show that the PMIE-2012 have a high reliability and accuracy on irradiation embrittlement prediction. (authors)

Effects of irradiation at lower temperature on the microstructure of Cr-Mo-V-alloyed reactor pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

Grosse, M; Boehmert, J; Gilles, R [Hahn-Meitner-Institut Berlin GmbH (Germany)

1998-10-01

The microstructural damage process due to neutron irradiation [1] proceeds in two stages: - formation of displacement cascades - evolution of the microstructure by defect reactions. Continuing our systematic investigation about the microstructural changes of Russian reactor pressure vessel steel due to neutron irradiation the microstructure of two laboratory heats of the VVER 440-type reactor pressure vessel steel after irradiation at 60 C was studied by small angle neutron scattering (SANS). 60 C-irradiation differently changes the irradiation-induced microstructure in comparison with irradiation at reactor operation temperature and can, thus, provide new insights into the mechanisms of the irradiation damage. (orig.)

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

Science.gov (United States)

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

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

Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

Energy Technology Data Exchange (ETDEWEB)

Li, Zhangbo; Lo, Wei-Yang [Department of Materials Science and Engineering, Nuclear Engineering Program, University of Florida, Gainesville, FL 32611 (United States); Chen, Yiren [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL 60439 (United States); Pakarinen, Janne [Belgian Nuclear Research Center (SCK-CEN), Boeretang 200, B-2400 Mol (Belgium); Wu, Yaqiao [Department of Materials Science and Engineering, Boise State University, Boise, ID 83715 (United States); Center for Advanced Energy Studies, Idaho Falls, ID 83401 (United States); Allen, Todd [Engineering Physics Department, University of Wisconsin, Madison, WI 53706 (United States); Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Yang, Yong, E-mail: yongyang@ufl.edu [Department of Materials Science and Engineering, Nuclear Engineering Program, University of Florida, Gainesville, FL 32611 (United States)

2015-11-15

To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ∼315 °C to 0.08 dpa (5.6 × 10{sup 19} n/cm{sup 2}, E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10{sup −9} dpa/s was found to induce spinodal decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.

Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

International Nuclear Information System (INIS)

Li, Zhangbo; Lo, Wei-Yang; Chen, Yiren; Pakarinen, Janne; Wu, Yaqiao; Allen, Todd; Yang, Yong

2015-01-01

To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ∼315 °C to 0.08 dpa (5.6 × 10"1"9 n/cm"2, E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10"−"9 dpa/s was found to induce spinodal decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.

Behavior of implanted hydrogen in ferritic/martensitic steels under irradiation

Science.gov (United States)

Wan, F.; Takahashi, H.; Ohnuki, S.; Nagasaki, R.

1988-07-01

The aim of this study was to clarify the behavior of hydrogen under irradiation in ferritic/martensitic stainless steel Fe-10Cr-2Mo-1Ni. Hydrogen was implanted into the specimens by ion accelerator or chemical cathodic charging method, followed by electron irradiation in a HVEM at temperatures from room temperature to 773 K. Streaks in the electron diffraction patterns were observed only during electron irradiation at 623-723 K. From these results it is suggested that the occurrence of the streak pattern is due to the formation of radiation-induced complexes of Ni or Cr with hydrogen along directions.

Microstructural characterization of atom clusters in irradiated pressure vessel steels and model alloys

International Nuclear Information System (INIS)

Auger, P.; Pareige, P.; Akamatsu, M.; Van Duysen, J.C.

1993-01-01

In order to characterize the microstructural evolution of iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions, and, for comparison, low copper model alloys irradiated with neutrons and electrons, have been studied through small angle neutron scattering and atom probe experiments. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex; solute atoms such as Ni, Mn and Si, sometimes associated with Cu, segregate as ''clouds'' more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs

Microstructural characterization of atom clusters in irradiated pressure vessel steels and model alloys

Energy Technology Data Exchange (ETDEWEB)

Auger, P; Pareige, P [Rouen Univ., 76 - Mont-Saint-Aignan (France); Akamatsu, M; Van Duysen, J C [Electricite de France (EDF), 77 - Ecuelles (France)

1994-12-31

In order to characterize the microstructural evolution of iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions, and, for comparison, low copper model alloys irradiated with neutrons and electrons, have been studied through small angle neutron scattering and atom probe experiments. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex; solute atoms such as Ni, Mn and Si, sometimes associated with Cu, segregate as ``clouds`` more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs.

Swift heavy ions induced irradiation effects in monolayer graphene and highly oriented pyrolytic graphite

International Nuclear Information System (INIS)

Zeng, J.; Yao, H.J.; Zhang, S.X.; Zhai, P.F.; Duan, J.L.; Sun, Y.M.; Li, G.P.; Liu, J.

2014-01-01

Monolayer graphene and highly oriented pyrolytic graphite (HOPG) were irradiated by swift heavy ions ( 209 Bi and 112 Sn) with the fluence between 10 11 and 10 14 ions/cm 2 . Both pristine and irradiated samples were investigated by Raman spectroscopy. It was found that D and D′ peaks appear after irradiation, which indicated the ion irradiation introduced damage both in the graphene and graphite lattice. Due to the special single atomic layer structure of graphene, the irradiation fluence threshold Φ th of the D band of graphene is significantly lower ( 11 ions/cm 2 ) than that (2.5 × 10 12 ions/cm 2 ) of HOPG. The larger defect density in graphene than in HOPG indicates that the monolayer graphene is much easier to be damaged than bulk graphite by swift heavy ions. Moreover, different defect types in graphene and HOPG were detected by the different values of I D /I D′ . For the irradiation with the same electronic energy loss, the velocity effect was found in HOPG. However, in this experiment, the velocity effect was not observed in graphene samples irradiated by swift heavy ions

Evaluation of neutron irradiation effect on SCC crack growth behaviour for austenitic stainless steels

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

Austenitic stainless steels are widely used as structural components in reactor pressure vessel internals because of their high strength, ductility, and fracture toughness. However, exposure to neutron irradiation results in changes in microstructure, mechanical properties and microchemistry of the steels. Irradiation assisted stress corrosion cracking (IASCC) caused by the effect of neutron irradiation during long term plant operation in high temperature water environments is considered to take the form of intergranular stress corrosion cracking (IGSCC) and the critical fluence level has been reported to be about 5x10{sup 24}n/m{sup 2} (E>1MeV) in Type 304 stainless steel in BWR environment. JNES had been conducting IASCC project during the JFY (2000) - JFY (2008) period, and prepared an engineering database on IASCC. However, the data of Crack Growth Rate (CGR) below the critical fluence level are not sufficient. So, this project was initiated to obtain the CGR data below the critical fluence level. Test specimens have been irradiated in the Halden reactor, operating by the OECD Halden Reactor Project, and the post irradiation examination (PIE) will be conducted from JFY (2011) to JFY (2013), finally the modified IASCC guide will be prepared in JFY (2013). (author)

Effects of Ti element on the microstructural stability of 9Cr–WVTiN reduced activation martensitic steel under ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Luo, Fengfeng [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Jin, Shuoxue; Li, Tiecheng; Chen, Jihong [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Suo, Jinping; Yang, Feng [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Yao, Z. [Department of Mechanical and Materials Engineering, Queen’s University, Kingston K7L 3N6, ON (Canada)

2014-12-15

Microstructure of 9Cr–WVTiN reduced-activation martensitic steels with two different Ti concentrations irradiated with Fe{sup +}, He{sup +} and H{sup +} at 300 °C was studied with transmission electron microscopy. Small dislocation loops were observed in the irradiated steels. The mean size and number density of dislocation loops decreased with the increase of Ti concentration. The segregation of Cr and Fe in carbides was observed in both irradiated steels, and the enrichment of Cr and depletion of Fe were more severe in the low Ti-concentration 9Cr–WVTiN steel.

Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

International Nuclear Information System (INIS)

Dethloff, Christian; Gaganidze, Ermile; Svetukhin, Vyacheslav V.; Aktaa, Jarir

2012-01-01

Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different 10 B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

Energy Technology Data Exchange (ETDEWEB)

Dethloff, Christian, E-mail: christian.dethloff@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Gaganidze, Ermile [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Svetukhin, Vyacheslav V. [Ulyanovsk State University, Leo Tolstoy Str. 42, 432970 Ulyanovsk (Russian Federation); Aktaa, Jarir [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

2012-07-15

Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different {sup 10}B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

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.)

First results of laser welding of neutron irradiated stainless steel

International Nuclear Information System (INIS)

Osch, E.V. van; Hulst, D.S. d'; Laan, J.G. van der.

1994-10-01

First results of experimental investigations on the laser reweldability of neutron irradiated material are reported. These experiments include the manufacture of 'heterogeneous' joints, which means joining of irradiated stainless steel of type AISI 316L-SPH to 'fresh' unirradiated material. The newly developed laser welding facility in the ECN Hot Cell Laboratory and experimental procedures are described. Visual inspections of welded joints are reported as well as results of electron microscopy and preliminary metallographic examinations. (orig.)

Heavy steel casting components for power plants 'mega-components' made of high Cr-steels

Energy Technology Data Exchange (ETDEWEB)

Hanus, Reinhold [voestalpine Giesserei Linz GmbH, Linz (Austria)

2010-07-01

Steel castings of creep resistant steels play a key role in fossil fuel fired power plants for highly loaded components in the high and intermediate pressure section of the turbines. Inner and outer casings, valve casings, inlet connections and elbows are examples of such critical components. The most important characteristic in a power plant is the efficiency, which mainly drives the CO2-emission. As a consequence of steadily improving power plant efficiencies and ever stricter emission standards, steam parameters become more critical and the creep resistance of the cast materials must also be constantly improved. The foundries voestalpine Giesserei Linz and voestalpine Giesserei Traisen participated in the development of the new 9-10% Cr-steels for application up to 625 C/650 C and in the THERMIE project where Ni-base alloys for 700 C-power plants were developed. Beside the material development in the European research projects the commercial production had to be established for industrial processes and the newly developed materials have to be transferred from research into the commercial production of heavy cast components. After selecting the most promising alloy from the laboratory melts, welding tests were performed - mostly with matching electrodes also produced within COST/THERMIE. Base material and welds were investigated in respect of microstructure, creep resistance, mechanical properties and weldability. Heat treatment investigations were also necessary for optimization of the mechanical properties. Based on the results of these studies, pilot components and plates for testing welding processes were cast in order to verify the castability and weldability of larger parts and to make any necessary adjustments to chemical composition, heat treatment or welding parameters. Parallel to the ongoing creep tests within COST/THERMIE-program, the newly developed steel grades were introduced into the commercial production of large components. This involved finding

Crack initiation behavior of neutron irradiated model and commercial stainless steels in high temperature water

Energy Technology Data Exchange (ETDEWEB)

Stephenson, Kale J., E-mail: kalejs@umich.edu; Was, Gary S.

2014-01-15

Highlights: • Environmental constant extension rate tensile tests were performed on neutron irradiated steel. • Percentage of intergranular cracking quantified the cracking susceptibility. • Cracking susceptibility varied with test environment, solute addition, and cold work. • No singular microstructural change could explain increases in cracking susceptibility with irradiation dose. • The increment of yield strength due to irradiation correlated well with cracking susceptibility. -- Abstract: The objective of this study was to isolate key factors affecting the irradiation-assisted stress corrosion cracking (IASCC) susceptibility of eleven neutron-irradiated austenitic stainless steel alloys. Four commercial purity and seven high purity stainless steels were fabricated with specific changes in composition and microstructure, and irradiated in a fast reactor spectrum at 320 °C to doses between 4.4 and 47.5 dpa. Constant extension rate tensile (CERT) tests were performed in normal water chemistry (NWC), hydrogen water chemistry (HWC), or primary water (PW) environments to isolate the effects of environment, elemental solute addition, alloy purity, alloy heat, alloy type, cold work, and irradiation dose. The irradiated alloys showed a wide variation in IASCC susceptibility, as measured by the relative changes in mechanical properties and crack morphology. Cracking susceptibility measured by %IG was enhanced in oxidizing environments, although testing in the lowest potential environment caused an increase in surface crack density. Alloys containing solute addition of Ni or Ni + Cr exhibited no IASCC. Susceptibility was reduced in materials cold worked prior to irradiation, and increased with increasing irradiation dose. Irradiation-induced hardening was accounted for by the dislocation loop microstructure, however no relation between crack initiation and radiation hardening was found.

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)

Advance of investigation of irradiation embrittlement mechanism of nuclear reactor pressure vessel steels. History and future of irradiation embrittlement researches

International Nuclear Information System (INIS)

Ishino, Shiori

2007-01-01

The nuclear reactor pressure vessel is the most important component of LWR plants required to be safe. This paper describes contents of the title consisting of four chapters. The first chapter states the general theory of irradiation effects, irradiation embrittlement and decreasing of toughness, and some kinds of pressure vessel steels. The second chapter explains history of irradiation embrittlement investigations and the advance of research methods for experiments and calculation. The third chapter contains information of inner structure of irradiated materials and development of prediction equations, recent information of embrittlement mechanism and mechanism guided prediction method, USA model and Central Research Institute of Electric Power Industry (CRIEPI) model. The fourth chapter states recent problems from viewpoints of experimental and analytical approaches. Comparison of standards of LWR pressure vessel steels between Japan and USA, relation between the density of number of cluster and the copper content, effect of flux on clustering of copper atoms, and CRIEPI's way of approaching the prediction method are illustrated. (S.Y.)

Aluminum and steel adhesion with polyurethanes from castor oil adhesives submitted to gamma irradiation

International Nuclear Information System (INIS)

Azevedo, Elaine C.; Assumpcao, Roberto L.; Nascimento, Eduardo M. do; Claro Neto, Salvador; Soboll, Daniel S.

2009-01-01

Polyurethanes adhesive from castor oil is used to join aluminum and steel pieces. The effect of gamma radiation on the resistance to tension tests is investigated. The aluminum and steel pieces after being glued with the adhesive were submitted to gamma irradiation in doses of 1 kGy, 25 kGy and 100 kGy. The rupture strength of the joints after irradiation have a slightly increase or remains practically unchanged indicating that the adhesive properties is not affected by the gamma radiation. (author)

Effect of cyclic electron irradiation on mechanical properties of austenite steel

International Nuclear Information System (INIS)

Tsepelev, A.B.; Sadykhov, S.I.O.; Chernov, A.I.; Sevost'yanov, M.A.

2006-01-01

To check the supposition on the possibility of radiation-stimulated process enhancement under cyclic irradiation conditions an experimental investigation is carried out to elucidate the effect of the mode of irradiation (continuous or cyclic) on mechanical properties of chromium-manganese austenitic stainless steel type 10Kh12G20V. The effect of some radiation hardening is observed under cyclic irradiation, however, the data obtained cannot be considered as good evidence for the validity of proposed model of dynamic preference if the scatter in experimental data is taken into account [ru

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-03-15

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

Irradiation induced tensile property change of SA 508 Cl.3 reactor pressure vessel steels

International Nuclear Information System (INIS)

Chi, Se-Hwan; Hong, Jun-Hwa; Kuk, Il-Hiun

1998-01-01

Irradiation induced tensile property change of four kinds of reactor pressure vessel steels manufactured by different steel refining process was compared based on the differences in the unirradiated and irradiated microstructure. Microvickers hardness, indentation, and miniature tensile specimen tests were conducted for mechanical property measurement and optical microscope (OM) and transmission electron microscope (TEM) were used for microstructural characterization. Specimens were 2 irradiated to a neutron fluence of 2.7x10 19 n/cm 2 (E ≥ 1 MeV) at 288 deg. C. Investigation on the unirradiated microstructures showed largely a same microstructure in that tempered acicular bainite and ferrite with bainitic phase prevailing in the unirradiated condition. Band-shaped segregations were also clearly observed except a kind of materials. A large difference in the unirradiated microstructure appeared in the grain size and carbide microstructure. Of carbide microstructures, noticeable differences were observed in the size and distribution of cementite, and bainitic lath microstructures. No noticeable changes were observed in the optical and thin film TEM microstructures after irradiation. Complicated microstructural. state of heat treated bainitic low alloy microstructure prevents easy quantification of microstructural changes due to irradiation. Apparent differences, however, were observed in the results of mechanical testing. Results of tensile testing and hardness measurement show that a steel refined by vacuum carbon deoxidation(VCD) method exhibits the highest radiation hardening behavior. Some of mechanical testing results on irradiated materials were possible to understand based on the initial microstructure, but further investigations using a wide array of sophisticated tools (for example, SANS, APFIM) are required to understand and characterize irradiation induced defects that are responsible for irradiation hardening behavior but are not revealed by

Nitric oxide-mediated bystander signal transduction induced by heavy-ion microbeam irradiation

Science.gov (United States)

Tomita, Masanori; Matsumoto, Hideki; Funayama, Tomoo; Yokota, Yuichiro; Otsuka, Kensuke; Maeda, Munetoshi; Kobayashi, Yasuhiko

2015-07-01

In general, a radiation-induced bystander response is known to be a cellular response induced in non-irradiated cells after receiving bystander signaling factors released from directly irradiated cells within a cell population. Bystander responses induced by high-linear energy transfer (LET) heavy ions at low fluence are an important health problem for astronauts in space. Bystander responses are mediated via physical cell-cell contact, such as gap-junction intercellular communication (GJIC) and/or diffusive factors released into the medium in cell culture conditions. Nitric oxide (NO) is a well-known major initiator/mediator of intercellular signaling within culture medium during bystander responses. In this study, we investigated the NO-mediated bystander signal transduction induced by high-LET argon (Ar)-ion microbeam irradiation of normal human fibroblasts. Foci formation by DNA double-strand break repair proteins was induced in non-irradiated cells, which were co-cultured with those irradiated by high-LET Ar-ion microbeams in the same culture plate. Foci formation was suppressed significantly by pretreatment with an NO scavenger. Furthermore, NO-mediated reproductive cell death was also induced in bystander cells. Phosphorylation of NF-κB and Akt were induced during NO-mediated bystander signaling in the irradiated and bystander cells. However, the activation of these proteins depended on the incubation time after irradiation. The accumulation of cyclooxygenase-2 (COX-2), a downstream target of NO and NF-κB, was observed in the bystander cells 6 h after irradiation but not in the directly irradiated cells. Our findings suggest that Akt- and NF-κB-dependent signaling pathways involving COX-2 play important roles in NO-mediated high-LET heavy-ion-induced bystander responses. In addition, COX-2 may be used as a molecular marker of high-LET heavy-ion-induced bystander cells to distinguish them from directly irradiated cells, although this may depend on the time

Influence of ultraviolet light irradiation on the corrosion behavior of carbon steel AISI 1015

Science.gov (United States)

Riazi, H. R.; Danaee, I.; Peykari, M.

2013-03-01

Corrosion of carbon steel in sodium chloride solution was studied under ultraviolet illumination using weight loss, polarization, electrochemical impedance spectroscopy and current transient tests. The polarization test revealed an increase in the corrosion current density observed under UV illumination. The impedance spectroscopy indicated that the charge transfer resistance of the system was decreased by irradiation of UV light on a carbon steel electrode. The weight loss of carbon steel in solution increased under UV light, which confirms the results obtained from electrochemical measurements. We propose that the main effect of UV irradiation is on the oxide film, which forms on the surface. Thus, in presence of UV, the conductivity of oxide film might increase and lead to higher metal dissolution and corrosion rate.

Contributions from research on irradiated ferritic/martensitic steels to materials science and engineering

Science.gov (United States)

Gelles, D. S.

1990-05-01

Ferritic and martensitic steels are finding increased application for structural components in several reactor systems. Low-alloy steels have long been used for pressure vessels in light water fission reactors. Martensitic stainless steels are finding increasing usage in liquid metal fast breeder reactors and are being considered for fusion reactor applications when such systems become commercially viable. Recent efforts have evaluated the applicability of oxide dispersion-strengthened ferritic steels. Experiments on the effect of irradiation on these steels provide several examples where contributions are being made to materials science and engineering. Examples are given demonstrating improvements in basic understanding, small specimen test procedure development, and alloy development.

Structural modifications of swift heavy ion irradiated PEN probed by optical and thermal measurements

International Nuclear Information System (INIS)

Devgan, Kusum; Singh, Lakhwant; Samra, Kawaljeet Singh

2013-01-01

Highlights: • The present paper reports the effect of swift heavy ion irradiation on Polyethylene Naphthalate (PEN). • Swift heavy ion irradiation introduces structural modification and degradation of PEN at different doses. • Lower irradiation doses in PEN result in modification of structural properties and higher doses lead to complete degradation. • Strong correlation between structural, optical, and thermal properties. - Abstract: The effects of swift heavy ion irradiation on the structural characteristics of Polyethylene naphthalate (PEN) were studied. Samples were irradiated in vacuum at room temperature by lithium (50 MeV), carbon (85 MeV), nickel (120 MeV) and silver (120 MeV) ions with the fluence in the range of 1×10 11 –3×10 12 ions cm −2 . Ion induced changes were analyzed using X-ray diffraction (XRD), Fourier transform infra red (FT-IR), UV–visible spectroscopy, thermo-gravimetric analysis (TGA) and differential scanning calorimetry (DSC) techniques. Cross-linking was observed at lower doses resulting in modification of structural properties, however higher doses lead to the degradation of the investigated polymeric samples

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.

Irradiation of steel, molybdenum and tungsten - VISA-2f; Ozracivanje celika, molibdena i volframa - VISA-2f -

Energy Technology Data Exchange (ETDEWEB)

Veljkovic, S; Milasin, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1962-12-15

The objective of the experiment is to study the radiation damage of steel, molybdenum and tungsten after irradiation under fast neutron flux. The sample wires of steel Mo and W will be irradiated, integral fast neutron flux should be higher than 10{sup 18} neutrons/cm{sup 2}, the temperature should be as low as possible.

Swelling and irradiation creep of neutron irradiated 316Ti and 15-15Ti steels

International Nuclear Information System (INIS)

Maillard, A.; Touron, H.; Seran, J.L.; Chalony, A.

1992-01-01

The global behavior, the swelling and irradiation creep resistances of cold worked 316Ti and 15-15Ti, two variants of austenitic steels in use as core component materials of the French fast reactors, are compared. The 15-15Ti leads to a significant improvement due to an increase in the incubation dose swelling. The same phenomena observed on 316Ti are found on 15-15Ti. All species without fuel like samples, wrappers or empty clad swell and creep less than fuel pin cladding irradiated in the same conditions. To explain the swelling difference, as for 316Ti, thermal gradient is also invoked but the irradiation creep difference is not yet clearly understood. To predict the behavior of clads it is indispensable to study the species themselves and to use specific rules. All results confirm the good behavior of 15-15Ti, the best behavior being obtained with the 1% Si doped version irradiated up to 115 dpa

The occurrence of an ordered fcc phase in neutron irradiated M316 stainless steel

International Nuclear Information System (INIS)

Cawthorne, C.; Brown, C.

1977-01-01

A small precipitate giving a superlattice type diffraction pattern has been observed in M316 type stainless steel irradiated in the Dounreay Fast Reactor. The precipitate was observed in cold worked and solution treated samples which were unstressed and irradiated below 540 0 C, but not in those irradiated above this temperature or in the stressed samples. (B.D.)

High Ni austenite stainless steel resistant to neutron irradiation degradation

International Nuclear Information System (INIS)

Yonezawa, Toshio; Iwamura, Toshihiko; Kanasaki, Hiroshi; Fujimoto, Koji; Nakata, Shizuo; Ajiki, Kazuhide; Nakamura, Mitsuhiro.

1997-01-01

The composition of the stainless steel of the present invention comprises from 0.005 to 0.08% of C, up to 3% of Mn, up to 0.2% of Si+P+S, from 25 to 40% of Ni, from 25 to 40% of Cr, up to 3% of Mo, up to 0.3% of Nb+Ta, up to 0.3% of Ti, up to 0.001% of B and the balance of Fe. A solid solubilization treatment at a temperature of from 1,000 to 1,150degC is applied to the stainless steel having the composition. The stainless steel is excellent in stress corrosion cracking-resistance at a working circumstance of a LWR type reactor (high temperature and high pressure water at from 270 to 350degC/from 70 to 160 atm even after undergoing neutron irradiation of about 1 x 10 22 n/cm 2 (E>1 MeV) which is a maximum neutron irradiation amount undergone till the final stage of the working life of the LWR-type reactor. In addition, the average thermal expansion coefficient at from room temperature to 400degC ranges from 15x10 -6 - 19x10 -6 /K. (I.N.)

Sub-micron indent induced plastic deformation in copper and irradiated steel

International Nuclear Information System (INIS)

Robertson, Ch.

1998-09-01

In this work we aim to study the indent induced plastic deformation. For this purpose, we have developed a new approach, whereby the indentation curves provides the mechanical behaviour, while the deformation mechanisms are observed thanks to Transmission Electron Microscopy (TEM). In order to better understand how an indent induced dislocation microstructure forms, numerical modeling of the indentation process at the scale of discrete dislocations has been worked out as well. Validation of this modeling has been performed through direct comparison of the computed microstructures with TEM micrographs of actual indents in pure Cu [001]. Irradiation induced modifications of mechanical behaviour of ion irradiated 316L have been investigated, thanks to the mentioned approach. An important hardening effect was reported from indentation data (about 50%), on helium irradiated 316L steel. TEM observations of the damage zone clearly show that this behaviour is associated with the presence of He bubbles. TEM observations of the indent induced plastic zone also showed that the extent of the plastic zone is strongly correlated with hardness, that is to say: harder materials gets a smaller plastic zone. These results thus clearly established that the selected procedure can reveal any irradiation induced hardening in sub-micron thick ion irradiated layers. The behaviour of krypton irradiated 316L steel is somewhat more puzzling. In one hand indeed, a strong correlation between the defect cluster size and densities on the irradiation temperature is observed in the 350 deg C -600 deg C range, thanks to TEM observations of the damage zone. On the other hand, irradiation induced hardening reported from indentation data is relatively small (about 10%) and shows no dependence upon the irradiation temperature (within the mentioned range). In addition, it has been shown that the reported hardening vanishes following appropriate post-irradiation annealing, although most of the TEM

Degradation of austenitic stainless steel (SS) light water ractor (LWR) core internals due to neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Rao, Appajosula S., E-mail: Appajosula.Rao@nrc.gov

2014-04-01

Austenitic stainless steels (SSs) are extensively being used in the fabrication of light water reactor (LWR) core internal components. It is because these steels have relatively high ductility, fracture toughness and moderate strength. However, the LWR internal components exposure to neutron irradiation over an extended period of plant operation degrades the materials mechanical properties such as the fracture toughness. This paper summarizes some of the results of the existing open literature data on irradiation assisted stress corrosion cracking (IASCC) of 316 CW steels that have been published by the United States Nuclear Regulatory Commission (USNRC), industry, academia, and other research agencies.

Irradiation effects on material properties of steels used in nuclear reactors: a literature review

International Nuclear Information System (INIS)

Gerceker, N.; Dara, I. H.

2001-01-01

The structural materials of a nuclear power plant are of vital importance as they provide mechanical strength, structural support and physical containment for the primary reactor components as well as the nuclear power plant itself. These structural materials comprise mainly of metals and their alloys, ceramics and cermets. However, metals and their alloys are the most widely used materials and the irradiation effects are more pronounced on metallic materials as of their high temperature properties are more sensitive (with respect to ceramics and cermets) to any kind of external effects. The wholesale creation of effects on material properties has been studied for over four decades and it is not realistic to attempt to represent even a small part of the field in single poster paper. In the present contribution, a literature review of the irradiation effects on the material properties of different types of steel alloys will be given because steels are widely used as structural materials in reactors and therefore the irradiation effects on steels may be of paramount importance for reactor design, operation and safety concepts which will be discussed about radiation effects on material properties of steels will provide highlights to better understanding of the origins and development of radiation effects in materials

The radiation swelling effect on fracture properties and fracture mechanisms of irradiated austenitic steels. Part II. Fatigue crack growth rate

Energy Technology Data Exchange (ETDEWEB)

Margolin, B., E-mail: margolinbz@yandex.ru; Minkin, A.; Smirnov, V.; Sorokin, A.; Shvetsova, V.; Potapova, V.

2016-11-15

The experimental data on the fatigue crack growth rate (FCGR) have been obtained for austenitic steel of 18Cr-10Ni-Ti grade (Russian analog of AISI 321 steel) irradiated up to neutron dose of 150 dpa with various radiation swelling. The performed study of the fracture mechanisms for cracked specimens under cyclic loading has explained why radiation swelling affects weakly FCGR unlike its effect on fracture toughness. Mechanical modeling of fatigue crack growth has been carried out and the dependencies for prediction of FCGR in irradiated austenitic steel with and with no swelling are proposed and verified with the obtained experimental results. As input data for these dependencies, FCGR for unirradiated steel and the tensile mechanical properties for unirradiated and irradiated steels are used.

Fracture toughness behavior of irradiated stainless steel in PWR systems

Energy Technology Data Exchange (ETDEWEB)

Xu, H.; Fyfitch, S. [AREVA NP Inc., Lynchburg, Pennsylvania (United States); Tang, H.T. [Electric Power Research Inst., Palo Alto, California (United States)

2007-07-01

Data from available research programs were collected and evaluated by the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) to determine the relationship between fracture toughness and neutron fluence for conditions representative of pressurized water reactor (PWR) conditions. It is shown that the reduction of fracture toughness with increasing neutron dose in both boiling water reactors (BWRs) and PWRs is consistent with that observed in fast reactors. The lower bound fracture toughness observed for irradiated stainless steels in PWRs is 38 MPa{radical}m (34.6 ksi{radical}in) at neutron exposures greater than 6.7 X 10{sup 21} n/cm{sup 2} (E > 1.0 MeV) or approximately 10 dpa. For such levels of fracture toughness, it is recommended that linear-elastic fracture mechanics (LEFM) analyses be considered for design and operational analyses. The results from this study can be used by the nuclear industry to assess the effects of irradiation on stainless steels in PWR systems. (author)

Opto-chemical response of Makrofol-KG to swift heavy ion irradiation

Indian Academy of Sciences (India)

In the present study, the effects of swift heavy ion beam irradiation on the structural, chemical and optical properties of Makrofol solid-state nuclear track detector (SSNTD) were investigated. Makrofol-KG films of 40 m thickness were irradiated with oxygen beam (8+) with fluences ranging between 1010 ion/cm2 and 1012 ...

Irradiation performance of 9--12 Cr ferritic/martensitic stainless steels and their potential for in-core application in LWRs

International Nuclear Information System (INIS)

Jones, R.H.; Gelles, D.S.

1993-08-01

Ferritic-martensitic stainless steels exhibit radiation stability and stress corrosion resistance that make them attractive replacement materials for austenitic stainless steels for in-core applications. Recent radiation studies have demonstrated that 9% Cr ferritic/martensitic stainless steel had less than a 30C shift in ductile-to-brittle transition temperature (DBTT) following irradiation at 365C to a dose of 14 dpa. These steels also exhibit very low swelling rates, a result of the microstructural stability of these alloys during radiation. The 9 to 12% Cr alloys to also exhibit excellent corrosion and stress corrosion resistance in out-of-core applications. Demonstration of the applicability of ferritic/martensitic stainless steels for in-core LWR application will require verification of the irradiation assisted stress corrosion cracking behavior, measurement of DBTT following irradiation at 288C, and corrosion rates measurements for in-core water chemistry

Changes in grain boundary composition induced by neutron irradiation of austenitic stainless steels

International Nuclear Information System (INIS)

Asano, K.; Nakata, K.; Fukuya, K.; Kodama, M.

1992-01-01

The radiation induced segregation of solutes to the grain boundary in austenitic stainless steels were studied. Type 304 and type 316 steel samples neutron irradiated at 561K up to 9.2x10 25 n/m 2 were obtained and minute compositional profiles across grain boundaries were examined using an analytical scanning transmission electron microscope equipped with a field emission electron gun. Chromium was slightly enriched at grain boundaries at the lowest irradiation dose but decreased with increasing fluence. Higher fluence irradiation resulted in depletion in chromium and molybdenum, and enrichment in nickel, silicon and phosphorus. These changes in grain boundary chemistry were limited within about 5nm of the boundary. Significant depletion of chromium and enrichment of impurities on the grain boundary occurred at fluences roughly coincidental with that of SCC susceptibility change obtained from another project

Effects of irradiation on low cycle fatigue properties for reduced activation ferritic/martensitic steel

International Nuclear Information System (INIS)

Kim, S.W.; Tanigawa, H.; Hirose, T.; Kohyama, A.

2007-01-01

Full text of publication follows: In materials life decision for a commercial blanket, thermal fatigue property of materials is a particularly important. The loading of structural materials in fusion reactor is, besides the plasma surface interactions, a combined effect of high heat fluxes and neutron irradiation. Depending on the pulse lengths, the operating conditions, and the thermal conductivity, these oscillating temperature gradients will cause elastic and elastic-plastic cyclic deformation giving rise to (creep-) fatigue in structural first wall and blanket components. Especially, investigation of the fatigue property in Reduced Activation Ferritic/Martensitic (RAF/M) steel and establishment of the evaluation technology are demanded in particular immediately for design/manufacturing of ITER-TBM. And also, fatigue testing after irradiation will be carried out in hot cells with remote control system. Considering limited ability of specimen manipulation in the cells, the specimen and the test method need to be simple for operation. The existing data bases of RAF/M steel provide baseline data set including post-irradiation fatigue data. However, to perform the accurate fatigue lifetime assessment for ITER-TBM and beyond utilizing the existing data base, the mechanical understanding of fatigue fracture is mandatory. It has been previously reported by co-authors that dislocation cell structure was developed on low cycle fatigued RAF/M steel, and led the fatigue crack to develop along prior austenitic grain boundary. In this work, the effects of nuclear irradiation on low cycle fatigue properties for RAF/M steels and its fracture mechanisms were examined based on the flow stress analysis and detailed microstructure analysis. Fracture surfaces and crack initiation site were investigated by scanning electron microscope (SEM). Transmission electron microscopy (TEM) was also applied to clarify the microstructural features of fatigue behavior. It is also important to

Effects of irradiation on low cycle fatigue properties for reduced activation ferritic/martensitic steel

Energy Technology Data Exchange (ETDEWEB)

Kim, S.W. [Kyoto Univ., Graduate School of Energy Science (Japan); Tanigawa, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Kohyama, A. [Kyoto Univ., lnstitute of Advanced Energy (Japan)

2007-07-01

Full text of publication follows: In materials life decision for a commercial blanket, thermal fatigue property of materials is a particularly important. The loading of structural materials in fusion reactor is, besides the plasma surface interactions, a combined effect of high heat fluxes and neutron irradiation. Depending on the pulse lengths, the operating conditions, and the thermal conductivity, these oscillating temperature gradients will cause elastic and elastic-plastic cyclic deformation giving rise to (creep-) fatigue in structural first wall and blanket components. Especially, investigation of the fatigue property in Reduced Activation Ferritic/Martensitic (RAF/M) steel and establishment of the evaluation technology are demanded in particular immediately for design/manufacturing of ITER-TBM. And also, fatigue testing after irradiation will be carried out in hot cells with remote control system. Considering limited ability of specimen manipulation in the cells, the specimen and the test method need to be simple for operation. The existing data bases of RAF/M steel provide baseline data set including post-irradiation fatigue data. However, to perform the accurate fatigue lifetime assessment for ITER-TBM and beyond utilizing the existing data base, the mechanical understanding of fatigue fracture is mandatory. It has been previously reported by co-authors that dislocation cell structure was developed on low cycle fatigued RAF/M steel, and led the fatigue crack to develop along prior austenitic grain boundary. In this work, the effects of nuclear irradiation on low cycle fatigue properties for RAF/M steels and its fracture mechanisms were examined based on the flow stress analysis and detailed microstructure analysis. Fracture surfaces and crack initiation site were investigated by scanning electron microscope (SEM). Transmission electron microscopy (TEM) was also applied to clarify the microstructural features of fatigue behavior. It is also important to

Antiradiation Vaccine: Technology Development- Radiation Tolerance,Prophylaxis, Prevention And Treatment Of Clinical Presentation After Heavy Ion Irradiation.

Science.gov (United States)

Popov, Dmitri; Maliev, Slava; Jones, Jeffrey

Introduction: Research in the field of biological effects of heavy charged particles is necessary for both heavy-ion therapy (hadrontherapy) and protection from the exposure to galactic cosmic radiation in long-term manned space missions.[Durante M. 2004] In future crew of long-term manned missions could operate in exremely high hadronic radiation areas of space and will not survive without effective radiation protection. An Antiradiation Vaccine (AV) must be an important part of a countermeasures regimen for efficient radiation protection purposes of austronauts-cosmonauts-taukonauts: immune-prophylaxis and immune-therapy of acute radiation toxic syndromes developed after heavy ion irradiation. New technology developed (AV) for the purposes of radiological protection and improvement of radiation tolerance and it is quite important to create protective immune active status which prevent toxic reactions inside a human body irradiated by high energy hadrons.[Maliev V. et al. 2006, Popov D. et al.2008]. High energy hadrons produce a variety of secondary particles which play an important role in the energy deposition process, and characterise their radiation qualities [Sato T. et al. 2003] Antiradiation Vaccine with specific immune-prophylaxis by an anti-radiation vaccine should be an important part of medical management for long term space missions. Methods and experiments: 1. Antiradiation vaccine preparation standard, mixture of toxoid form of Radiation Toxins [SRD-group] which include Cerebrovascular RT Neurotoxin, Cardiovascular RT Neurotoxin, Gastrointestinal RT Neurotoxin, Hematopoietic RT Hematotoxin. Radiation Toxins of Radiation Determinant Group isolated from the central lymph of gamma-irradiated animals with Cerebrovascular, Cardiovascular, Gastro-intestinal, Hematopoietic forms of ARS. Devices for radiation are "Panorama", "Puma". 2. Heavy ion exposure was accomplished at Department of Research Institute of Nuclear Physics, Dubna, Russia. The heavy ions

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.

Effects of irradiation on tungsten stabilized martensitic steels*1

Science.gov (United States)

Gelles, D. S.; Hsu, C. Y.; Lechtenberg, T. A.

1988-07-01

Tungsten stabilized martensitic stainless steels are being developed for fusion reactor first wall applications in order to lower retained radioactivity so as to permit shallow land burial after reactor decommissioning. Two such alloys have been designed, fabricated, fast neutron irradiated in FFTF and examined by transmission electron microscopy. The two compositions were Fe-7.5Cr-2.0W-0.17 C and Fe-10.2Cr-1.7W-0.3V-0.02C. Conditions examined included irradiation temperatures of 365, 426, 520 and 600°C to doses as high as 34 dpa. Small amounts of void swelling are found at the two lowest temperatures. It is demonstrated that levels of tungsten on the order of 2 wt% do not result in excessive intermetallic precipitation under these irradiation conditions.

Irradiation effects of swift heavy ions in matter

Energy Technology Data Exchange (ETDEWEB)

Osmani, Orkhan

2011-12-22

In the this thesis irradiation effects of swift heavy ions in matter are studied. The focus lies on the projectiles charge exchange and energy loss processes. A commonly used computer code which employs rate equations is the so called ETACHA code. This computer code is capable to also calculate the required input cross-sections. Within this thesis a new model to compute the charge state of swift heavy ions is explored. This model, the so called matrix method, takes the form of a simple algebraic expression, which also requires cross-sections as input. In the present implementation of the matrix method, cross-sections are taken from the ETACHA code, while excitation and deexcitation processes are neglected. Charge fractions for selected ion/target combinations, computed by the ETACHA code and the matrix method are compared. It is shown, that for sufficient large ion energies, both methods agree very well with each other. However, for lower energies pronounced differences are observed. These differences are believed to stem from the fact, that no excited states as well as the decay of theses excited states are included in the present implementation of the matrix method. Both methods are then compared with experimental measurements, where significant deviations are observed for both methods. While the predicted equilibrium charge state by both methods is in good agreement with the experiments, the matrix method predicts a much too large equilibrium thickness compared to both the ETACHA calculation as well as the experiment. Again, these deviations are believed to stem from the fact, that excitation and the decay of excited states are not included in the matrix method. A possible way to include decay processes into the matrix method is presented, while the accuracy of the applied capture cross-sections is tested by comparison with scaling rules. Swift heavy ions penetrating a dielectric are known to induced structural modifications both on the surface and in the bulk

Irradiation effects of swift heavy ions in matter

International Nuclear Information System (INIS)

Osmani, Orkhan

2011-01-01

In the this thesis irradiation effects of swift heavy ions in matter are studied. The focus lies on the projectiles charge exchange and energy loss processes. A commonly used computer code which employs rate equations is the so called ETACHA code. This computer code is capable to also calculate the required input cross-sections. Within this thesis a new model to compute the charge state of swift heavy ions is explored. This model, the so called matrix method, takes the form of a simple algebraic expression, which also requires cross-sections as input. In the present implementation of the matrix method, cross-sections are taken from the ETACHA code, while excitation and deexcitation processes are neglected. Charge fractions for selected ion/target combinations, computed by the ETACHA code and the matrix method are compared. It is shown, that for sufficient large ion energies, both methods agree very well with each other. However, for lower energies pronounced differences are observed. These differences are believed to stem from the fact, that no excited states as well as the decay of theses excited states are included in the present implementation of the matrix method. Both methods are then compared with experimental measurements, where significant deviations are observed for both methods. While the predicted equilibrium charge state by both methods is in good agreement with the experiments, the matrix method predicts a much too large equilibrium thickness compared to both the ETACHA calculation as well as the experiment. Again, these deviations are believed to stem from the fact, that excitation and the decay of excited states are not included in the matrix method. A possible way to include decay processes into the matrix method is presented, while the accuracy of the applied capture cross-sections is tested by comparison with scaling rules. Swift heavy ions penetrating a dielectric are known to induced structural modifications both on the surface and in the bulk

Special neutron measurement results from the spectral positions of the Juelich FKS steel irradiation capsules

International Nuclear Information System (INIS)

Schneider, W.; Kuepper, H.; Pott, G.; Borchardt, G.; Segelhorst, G.; Thoene, L.; Weise, L.

1986-10-01

For the German project 'Forschungsvorhaben Komponentensicherheit' (FKS, i.e., Structural Integrity of Components) steel specimen irradiations have been carried out in the Juelich Merlin-type reactor (FRJ-1). The neutron monitoring to these irradiations is described in a German report (Juel-2087). In this context, some special considerations and results are given here, i.e., an experimental investigation of the fast neutron spectrum variation over a thick steel plate (in a special dosimetry test experiment); a comparison of the outcome of this investigation with the results from other FKS participants; and finally, the evaluation of the neutron exposure expressed in displacements per atom (dpa) in the centre of that steel plate. (orig.)

Irradiation induced tensile property change of SA 508 Cl. 3 reactor pressure vessel steels

International Nuclear Information System (INIS)

Chi, Se Hwan; Hong, Jun Hwa; Kuk, Il Hiun

1998-01-01

Irradiation induced tensile property change of four kinds of reactor pressure vessel steels manufactured by different steel refining process was compared based on the differences in the miniature tensile specimen tests were conducted for mechanical property measurement and optical microscope (OM) and transmission electron microscope (TEM) were used for microstructural characterization. Specimens were irradiated to a neutron fluence of 2.7 x 10 19 n/cm 2 (E ≥ 1 MeV) at 288 deg C. Investigation on the unirradiated microstructures showed largely a same microstructure in that tempered acicular bainite and ferrite with bainitic phase prevailing in the unirradiated condition. Ban-shaped segregations were also clearly observed except a kind of materials. A large difference in the unirradiated microstructure appeared in the grain size and carbide microstructure. Of carbide microstructures, noticeable differences were observed in the size and distribution of cementite, and bainitic lath microstructures. No noticeable changes were observed in the optical and thin film TEM microstructures after irradiation. Complicated microstructural state of heat treated bainitic low alloy microstructure prevents easy quantification of microstructural changes due to irradiation. Apparent differences, however, were observed in the results of mechanical testing. Results of tensile testing and hardness measurement show that a steel refined by vacuum carbon deoxidation (VCD) method exhibits the highest radiation hardening behavior. Some of mechanical testing results on irradiated materials were possible to understand based on the initial microstructure, but further investigations using a wide array of sophisticated tools (for example, SANS, APFIM) are required to understand and characterize irradiation induced defects that are responsible for irradiation hardening behavior but are not revealed by conventional TEM. (author)

Effect of additional minor elements on accumulation behavior of point defects under electron irradiation in austenitic stainless steels

International Nuclear Information System (INIS)

Sekio, Yoshihiro; Yamashita, Shinichiro; Takahashi, Heishichiro; Sakaguchi, Norihito

2014-01-01

Addition of minor elements to a base alloy is often applied with the aim of mitigating void swelling by decreasing the vacancy diffusivity and flux which influence vacancy accumulation behavior. However, the comparative evaluations of parameters, such as the diffusivity and flux, between a base alloy and modified alloys with specific additives have not been studied in detail. In this study, type 316 austenitic stainless steel as a base alloy and type 316 austenitic stainless steels modified with vanadium (V) or zirconium (Zr) additions were used to perform evaluations from the changes of widths of the void denuded zone (VDZ) formed near a random grain boundary during electron irradiation because these widths depend on vacancy diffusivity and flux. The formations of VDZs were observed in in-situ observations during electron irradiation at 723 K and the formed VDZ widths were measured from the transmission electron microscopic images after electron irradiation. As a result, the VDZs were formed in both steels without and with V, and respective widths were ∼119 and ∼100 nm. On the other hand, the VDZ formation was not observed clearly in the steel with Zr. From the measured VDZ widths in the steels without and with V addition, the estimated ratio of the vacancy diffusivity in the steel with V to that in the steel without V was about 0.50 and the estimated ratio of the vacancy flux in the steel with V to that in the steel without V was about 0.71. This result suggests that the effect of additional minor elements on vacancy accumulation behaviors under electron irradiation could be estimated from evaluations of the VDZ width changes among steels with and without minor elements. Especially, because void swelling is closely related with the vacancy diffusion process, the VDZ width changes would also be reflected on void swelling behavior. (author)

Impact of the nanostructuration on the corrosion resistance and hardness of irradiated 316 austenitic stainless steels

Science.gov (United States)

Hug, E.; Prasath Babu, R.; Monnet, I.; Etienne, A.; Moisy, F.; Pralong, V.; Enikeev, N.; Abramova, M.; Sauvage, X.; Radiguet, B.

2017-01-01

The influence of grain size and irradiation defects on the mechanical behavior and the corrosion resistance of a 316 stainless steel have been investigated. Nanostructured samples were obtained by severe plastic deformation using high pressure torsion. Both coarse grain and nanostructured samples were irradiated with 10 MeV 56Fe5+ ions. Microstructures were characterized using transmission electron microscopy and atom probe tomography. Surface mechanical properties were evaluated thanks to hardness measurements and the corrosion resistance was studied in chloride environment. Nanostructuration by high pressure torsion followed by annealing leads to enrichment in chromium at grain boundaries. However, irradiation of nanostructured samples implies a chromium depletion of the same order than depicted in coarse grain specimens but without metallurgical damage like segregated dislocation loops or clusters. Potentiodynamic polarization tests highlight a definitive deterioration of the corrosion resistance of coarse grain steel with irradiation. Downsizing the grain to a few hundred of nanometers enhances the corrosion resistance of irradiated samples, despite the fact that the hardness of nanocrystalline austenitic steel is only weakly affected by irradiation. These new experimental results are discussed in the basis of couplings between mechanical and electrical properties of the passivated layer thanks to impedance spectroscopy measurements, hardness properties of the surfaces and local microstructure evolutions.

Effects of Si and Ti on the phase stability and swelling behavior of AISI 316 stainless steel

International Nuclear Information System (INIS)

Lee, E.H.; Rowcliffe, A.F.; Kenik, E.A.

1979-01-01

The swelling behavior of neutron irradiated stainless steels is strongly influenced by solute segregation and precipitation phenomena. The extent to which in-reactor swelling behavior may be simulated by heavy ion irradiation depends upon the extent to which in-reactor phase changes are reproduced; this question is addressed by comparing the precipitation behavior under neutron irradiation with behavior during 4 MeV Ni ion irradiation for AISI 316 stainless steel and a related stainless steel containing additions of titanium and silicon. The results are discussed qualitatively in terms of the effects of damage rate on solute segregation and the effects of displacement cascades on the dissolution of particles. It is shown that the partitioning of elements into various phases during irradiation is not a sufficient condition for the iniatiation of swelling in stainless steels modified with silicon and titanium. It is also necessary for helium to be generated simultaneously with the breakdown of the matrix into various phases; it is believed that helium trapping at the growing particle-matrix interface is responsible for the observed physical association between voids and precipitates. (Auth.)

Effects of Si and Ti on the phase stability and swelling behavior of AISI 316 stainless steel

International Nuclear Information System (INIS)

Lee, E.H.; Rowcliffe, A.F.; Kenik, E.A.

1978-01-01

Swelling behavior of neutron irradiated stainless steels is influenced by solute segregation and preciptation phenomena. The extent to which in-reactor swelling behavior may be simulated by heavy ion irradiation depends upon the extent to which in-reactor phase changes are reproduced; this question is addressed by comparing the precipitation behavior under neutron irradiation with behavior during 4 MeV Ni ion irradiation for AISI 316 stainless steel and a related stainless steel containing additions of titanium and silicon. The results are discussed qualitatively in terms of the effects of damage rate on solute segregation and the effects of displacement cascades on the dissolution of particles. It is shown that the partitioning of elements into various phases during irradiation is not a sufficient condition for the initiation of swelling in stainless steels modified with silicon and titanium. It is also necessary for helium to be generated simultaneously with the breakdown of the matrix into various phases; it is believed that helium trapping at the growing particle-matrix interface is responsible for the observed physical association between voids and precipitates

Apparatus of irradiation of steel test pieces in the Marcoule pile G 1

International Nuclear Information System (INIS)

Marinot, R.; Wallet, Ph.

1960-01-01

Test pieces of steel were irradiated in the reactor G1 at Marcoule, in convectors replacing fuel elements, and in vertical channels in furnace-heated containers. The apparatus designed for this irradiation is described: containers, converter-rods, suspension fixtures and clamps, temperature measurement devices, lead castles and unloading set-ups. (author) [fr

Airborne heavy metal pollution in the environment of a danish steel plant

DEFF Research Database (Denmark)

Vestergaard, N. K.; Stephansen, U.; Rasmussen, L.

1986-01-01

A survey of heavy metal deposition was carried out in the vicinity of a Danish steel plant. Bulk precipitation and transplanted lichen (Hypogymnia physodes (L.) Nyl.) were sampled at 12 stations in the environment before and after the production had been converted from open-hearth furnaces...

Strain hardening and plastic instability properties of austenitic stainless steels after proton and neutron irradiation

International Nuclear Information System (INIS)

Byun, T.S.; Farrell, K.; Lee, E.H.; Hunn, J.D.; Mansur, L.K.

2001-01-01

Strain hardening and plastic instability properties were analyzed for EC316LN, HTUPS316, and AL6XN austenitic stainless steels after combined 800 MeV proton and spallation neutron irradiation to doses up to 10.7 dpa. The steels retained good strain-hardening rates after irradiation, which resulted in significant uniform strains. It was found that the instability stress, the stress at the onset of necking, had little dependence on the irradiation dose. Tensile fracture stress and strain were calculated from the stress-strain curve data and were used to estimate fracture toughness using an existing model. The doses to plastic instability and fracture, the accumulated doses at which the yield stress reaches instability stress or fracture stress, were predicted by extrapolation of the yield stress, instability stress, and fracture stress to higher dose. The EC316LN alloy required the highest doses for plastic instability and fracture. Plastic deformation mechanisms are discussed in relation to the strain-hardening properties of the austenitic stainless steels

Void swelling and phase stability in different heats of cold-drawn type 1.4970 stainless steel after heavy-ion irradiation

International Nuclear Information System (INIS)

Vaidya, W.V.; Knoblauch, G.; Ehrlich, K.

1982-01-01

The present investigations were undertaken with the aim to understand, to what extent variations of the tube fabrication parameters and slight modifications in the chemical composition might influence the swelling behaviour of Type 1.4970 stainless steel. The parameters varied were: Variations in the manufacturing parameters for coldworked tubes (type and degree of drawing, solution-annealing temperature and thermomechanical treatments), and variations in minor elements (C, Ti, Mo) within the specified range of chemical composition. In addition, the Si-content and the Ti/C ratio - the so-called stabilization - were changed within a broader range. The samples were irradiated with 46 MeV-Ni-ions to 64 dpa at 575 0 C and swelling as well as austenite stability, formation of precipitates and other microstructural changes were investigated by TEM. Though the austenite was stable under irradiation with respect to ferrite/martensite-transformation, the cold-drawn alloys showed a tendency to recrystallize during irradiation and exhibited lean precipitation. With respect to swelling, the only parameter that substantially reduced it, was the high Si addition; otherwise the alloys were practically insensitive to changes in the investigated parameters. These results are discussed in terms of the radiation-induced recrystallization and the high Si-effect, both of which are found to be beneficial in reducing swelling. (orig.)

Void swelling and phase stability in different heats of cold-drawn type 1.4970 stainless steel after heavy-ion irradiation

International Nuclear Information System (INIS)

Vaidya, W.V.; Knoblauch, G.; Ehrlich, K.

1982-01-01

The present investigations were undertaken with the aim to understand, to what extent variations of the tube fabrication parameters and slight modifications in the chemical composition might influence the swelling behavior of Type 1.4970 stainless steel. The parameters varied were: variations in the manufacturing parameters for cold-worked tubes (type and degree of drawing, solution-annealing temperature and thermomechanical treatments), and variations in minor elements (C, Ti, Mo) within the specified range of chemical composition. In addition, the Si-content and the Ti/C ratio - the so-called stabilization - were changed within a broader range. The samples were irradiated with 46 MeV-Ni-ions to 64 dpa at 575 0 C and swelling as well as austenite stability, formation of precipitates and other microstructural changes were investigated by TEM. Though the austenite was stable under irradiation with respect to ferrite/martensite-transformation, the cold-drawn alloys showed a tendency to recrystallize during irradiation and exhibited lean precipitation. With respect to swelling, the only parameter that substantially reduced it, was the high Si addition; otherwise the alloys were practically insensitive to changes in the investigated parameters. These results are discussed in terms of the radiation-induced recrystallization and the high-Si-effect, both of which are found to be beneficial in reducing swelling

ESR investigation of L-α-alanine and sucrose radicals produced by heavy-ion irradiation

International Nuclear Information System (INIS)

Nakagawa, K.; Sato, Y.

2005-01-01

We investigated sucrose and L-α-alanine radicals produced by heavy (particle) ion irradiation with various LETs (linear energy transfer). The impact of the heavy ions on the samples produced stable free radicals, which were analyzed by ESR (electron spin resonance). Identical spectra were measured after one year. The obtained spectral patterns were the same as those for helium (He), carbon (C), and neon (Ne) ions irradiation. The absorbed dose dependences for the irradiated sucrose and alanine samples were examined. The ESR response has a linear relation with the absorbed dose. The ESR response at 60 Gy was slightly lower than a linear line for sucrose; however, the response showed good linearity for the alanine. In addition, the total spin concentration obtained by heavy-ion irradiation correlated logarithmically with the LET. Qualitative ESR analyse showed that the production of sucrose and alanine radicals depended on both different particle irradiation and the LET under the same dose. Thus, the present ESR results imply that sucrose together with L-α-alanine can be used to monitor LET as well as the number of ionizing particle for the production of stable free radicals. (author)

Effect of heavy ion irradiation on thermodynamically equilibrium Zr-Excel alloy

Science.gov (United States)

Yu, Hongbing; Liang, Jianlie; Yao, Zhongwen; Kirk, Mark A.; Daymond, Mark R.

2017-05-01

The thermodynamically equilibrium state was achieved in a Zr-Sn-Nb-Mo alloy by long-term annealing at an intermediate temperature. The fcc intermetallic Zr(Mo, Nb)2 enriched with Fe was observed at the equilibrium state. In-situ 1 MeV Kr2+ heavy ion irradiation was performed in a TEM to study the stability of the intermetallic particles under irradiation and the effects of the intermetallic particle on the evolution of type dislocation loops at different temperatures from 80 to 550 °C. Chemi-STEM elemental maps were made at the same particles before and after irradiation up to 10 dpa. It was found that no elemental redistribution occurs at 200 °C and below. Selective depletion of Fe was observed from some precipitates under irradiation at higher temperatures. No change in the morphology of particles and no evidence showing a crystalline to amorphous transformation were observed at all irradiation temperatures. The formation of type dislocation loops was observed under irradiation at 80 and 200 °C, but not at 450 and 550 °C. The loops were non-uniformly distributed; a localized high density of type dislocation loops were observed near the second phase particles; we suggest that loop nucleation is favored as a result of the stress induced by the particles, rather than by elemental redistribution. The stability of the second phase particles and the formation of the type loops under heavy ion irradiation are discussed.

Microstructural evolution of reduced-activation martensitic steel under single and sequential ion irradiations

Energy Technology Data Exchange (ETDEWEB)

Luo, Fengfeng [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Jin, Shuoxue; Li, Tiecheng; Zheng, Zhongcheng [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Yang, Feng; Xiong, Xuesong; Suo, Jinping [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

2013-07-15

Microstructural evolution of super-clean reduced-activation martensitic steels irradiated with single-beam (Fe{sup +}) and sequential-beam (Fe{sup +} plus He{sup +}) at 350 °C and 550 °C was studied. Sequential-beam irradiation induced smaller size and larger number density of precipitates compared to single-beam irradiation at 350 °C. The largest size of cavities was observed after sequential-beam irradiation at 550 °C. The segregation of Cr and W and depletion of Fe in carbides were observed, and the maximum depletion of Fe and enrichment of Cr occurred under irradiation at 350 °C.

Parameter optimization for steel quenching by C02-laser irradiation

International Nuclear Information System (INIS)

Moryashchev, S.F.; Kislitsyn, A.A.; Kosyrev, F.K.

1984-01-01

The dependence of average absorption factor on maximal temperature of the article surface during quenching by CO 2 -laser irradiation was determined empirically. The calculations of depth of a hardening zone and process productivity in 40 Kh, 4Kh13 steels and Armco-iron with regard to this dependence were conducted

Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

International Nuclear Information System (INIS)

Chen, Y.; Chopra, O. K.; Gruber, Eugene E.; Shack, William J.

2010-01-01

The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC. The

Microstructural characterization and model of hardening for the irradiated austenitic stainless steels of the internals of pressurized water reactors

International Nuclear Information System (INIS)

Pokor, C.

2003-01-01

The core internals of Pressurized Water Reactors (PWR) are composed of SA 304 stainless steel plates and CW 316 stainless steel bolts. These internals undergo a neutron flux at a temperature between 280 deg C and 380 deg C which modifies their mechanical properties. These modifications are due to the changes in the microstructure of these materials under irradiation which depend on flux, dose and irradiation temperature. We have studied, by Transmission Electron Microscopy, the microstructure of stainless steels SA 304, CW 316 and CW 316Ti irradiated in a mixed flux reactor (OSIRIS at 330 deg C between 0,8 dpa et 3,4 dpa) and in a fast breeder reactor at 330 deg C (BOR-60) up to doses of 40 dpa. Moreover, samples have been irradiated at 375 deg C in a fast breeder reactor (EBR-II) up to doses of 10 dpa. The microstructure of the irradiated stainless steels consists in faulted Frank dislocation loops in the [111] planes of austenitic, with a Burgers vector of [111]. It is possible to find some voids in the solution annealed samples irradiated at 375 deg C. The evolution of the dislocations loops and voids has been simulated with a 'cluster dynamic' model. The fit of the model parameters has allowed us to have a quantitative description of our experimental results. This description of the microstructure after irradiation was coupled together with a hardening model by Frank loops that has permitted us to make a quantitative description of the hardening of SA 304, CW 316 and CW 316Ti stainless steels after irradiation at a certain dose, flux and temperature. The irradiation doses studied grow up to 90 dpa, dose of the end of life of PWR internals. (author)

A function of mutagenesis on rhodotorula RY strain irradiated by heavy ion

International Nuclear Information System (INIS)

Li Hongyu; Li Chenghua; Ding Xinchun; Wang Jufang; Zhou Guangming; Xie Hongmei; Li Qiang; Dang bingrong; Wen Xiaoqiong; Li Wenjian; Wei Zengquan

2004-01-01

In this paper, red yeast (Rhodotorula RY Strain) that produces carotene is irradiated by 50 MeV/u 12 C 6+ heavy ion from Heavy Ion Accelerator in IMP. Fermentation tests show that 50 MeV/u 12 C 6+ heavy ion has a mutagenesis effect on the red yeast. Some strains of red yeast with changed production of carotene were found by screening. Meanwhile, by RFLP and RAPD analysis, authors have a further evidence that heavy ion can cause mutagenesis in Rhodotorula RY Strain. This presents a new prospect for the mutagenesis breeding by heavy ion in industry

Characterization of Flame Cut Heavy Steel: Modeling of Temperature History and Residual Stress Formation

Science.gov (United States)

Jokiaho, T.; Laitinen, A.; Santa-aho, S.; Isakov, M.; Peura, P.; Saarinen, T.; Lehtovaara, A.; Vippola, M.

2017-12-01

Heavy steel plates are used in demanding applications that require both high strength and hardness. An important step in the production of such components is cutting the plates with a cost-effective thermal cutting method such as flame cutting. Flame cutting is performed with a controlled flame and oxygen jet, which burns the steel and forms a cutting edge. However, the thermal cutting of heavy steel plates causes several problems. A heat-affected zone (HAZ) is generated at the cut edge due to the steep temperature gradient. Consequently, volume changes, hardness variations, and microstructural changes occur in the HAZ. In addition, residual stresses are formed at the cut edge during the process. In the worst case, unsuitable flame cutting practices generate cracks at the cut edge. The flame cutting of thick steel plate was modeled using the commercial finite element software ABAQUS. The results of modeling were verified by X-ray diffraction-based residual stress measurements and microstructural analysis. The model provides several outcomes, such as obtaining more information related to the formation of residual stresses and the temperature history during the flame cutting process. In addition, an extensive series of flame cut samples was designed with the assistance of the model.

Microstructural and microchemical evolution in vanadium alloys by heavy ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Sekimura, Naoto; Kakiuchi, Hironori; Shirao, Yasuyuki; Iwai, Takeo [Tokyo Univ. (Japan)

1996-10-01

Microstructural and microchemical evolution in vanadium alloys were investigated using heavy ion irradiation. No cavities were observed in V-5Cr-5Ti alloys irradiated to 30 dpa at 520 and 600degC. Energy dispersive X-ray spectroscopy analyses showed that Ti peaks around grain boundaries. Segregation of Cr atoms was not clearly detected. Co-implanted helium was also found to enhance dislocation evolution in V-5Cr-5Ti. High density of matrix cavities were observed in V-5Fe alloys irradiated with dual ions, whereas cavities were formed only around grain boundaries in single ion irradiated V-5Fe. (author)

Numerical simulation of convection and inclusion distribution during solidification in a heavy steel ingot

International Nuclear Information System (INIS)

Lin, Rui; Shen, Houfa

2015-01-01

Inclusions content in the steel ingot is an important index for homogeneity, and it becomes more serious for heavy steel ingots which are used for major equipment. However, knowledge about the formation of inclusion in steel ingot is limited, and modeling of inclusion distribution is still challenging, so it is of great significance to research the behavior of inclusion. In this paper, fluid flow during solidification is numerically simulated based on the equilibrium equations of mass, momentum and energy, and then inclusion distribution is modeled according to the Lagrangian Stokes trajectory method. The Results show that the inclusion distribution in the steel ingot is influenced by the flow pattern which is affected by the solidification pattern. Therefore, inclusion distribution could be controlled by the solidification front with the optimization of heat transfer condition such as the hot top design of steel ingot for the high quality steel production. (paper)

Microstructure in HIP-bonded F82H steel and its mechanical properties after irradiation

International Nuclear Information System (INIS)

Furuya, K.; Wakai, E.

2006-01-01

A first primary blanket structure is composed of the low-activation steel, e.g. F82H, and is fabricated by using a solid hot isostatic pressing (HIP) bonding method. A partial mock-up of such a blanket structure was successfully fabricated. The tensile specimen including HIP-bonded region possessed a sufficient strength and elongation under a non-irradiated condition as reported in our previous studies. In this study, the microstructures of HIP interface before irradiation were observed by a TEM, and the effects of irradiation on mechanical properties of the HIP-bonded region were also examined. TEM observation and elemental analysis of the HIP-bonded region before the irradiation were performed by using a FE-TEM of HF-2000 equipped with EDX spectroscopy. Tensile specimens (type SS-3) were prepared from a HIP-bonded region and a plate region of the mock-up block. Neutron irradiation was performed up to about 1.9 dpa at about 523 K in JMTR. After the irradiation, tensile test was performed at temperatures of 295 and 523 K. After the tensile test, OM observation at the rupture region and SEM observation at the fracture surface were conducted, respectively. TEM observation and analytical results revealed that the HIP interface possessed many precipitates, and enriched peak spectrum of chromium was detected from the precipitates. In addition, aspect of the spectrum was qualitatively equivalent to that of M23C6 in grain boundaries of F82H steel. In result, the HIP boundary has many M23C6 which were generally seen in grain boundaries of F82H steel, and it can be mentioned that the HIP interface is, in this sense, a new grain boundary. Obvious HIP boundary was seen at rupture region of tensile specimens sampled from the HIP-bonded region, by the macroscopic observation. It means that rupture do not occur in the HIP interface. In result, it can be mentioned that bondability of the HIP interfaces is kept under the irradiation and testing conditions. The strength and

Cell survival in spheroids irradiated with heavy-ion beams

International Nuclear Information System (INIS)

Rodriguez, A.; Alpen, E.L.

1981-01-01

Biological investigations with accelerated heavy ions have been carried out regularly at the Lawrence Berkeley Laboratory Bevalac for the past four years. Most of the cellular investigations have been conducted on cell monolayer and suspension culture systems. The studies to date suggest that heavy charged particle beams may offer some radiotherapeutic advantages over conventional radiotherapy sources. The advantages are thought to lie primarily in an increased relative biological effectiveness (RBE), a decrease in the oxygen enhancement ratio (OER), and better tissue distribution dose. Experiments reported here were conducted with 400 MeV/amu carbon ions and 425 MeV/amu neon ions, using a rat brain gliosarcoma cell line grown as multicellular spheroids. Studies have been carried out with x-rays and high-energy carbon and neon ion beams. These studies evaluate high-LET (linear energy transfer) cell survival in terms of RBE and the possible contributions of intercellular communication. Comparisons were made of the post-irradiation survival characteristics for cells irradiated as multicellular spheroids (approximately 100 μm and 300 μm diameters) and for cells irradiated in suspension. These comparisons were made between 225-kVp x-rays, 400 MeV/amu carbon ions, and 425 MeV/amu neon ions

Irradiation embrittlement of reactor pressure vessel steels: Considerations for thermal annealing

International Nuclear Information System (INIS)

Burke, M.G.; Freyer, P.D.; Mager, T.R.

1993-01-01

In this paper, an overview of the irradiation embrittlement phenomenon is presented from a structure-properties viewpoint. Effects of irradiation conditions on embrittlement are first reviewed: irradiation temperature, fluence, flux, and steel or alloy composition. Then, the techniques for identifying/characterizing the irradiation-induced microstructural features are described: TEM/STEM (electron microscopy), small angle neutron scattering, atom probe field-ion microscopy, positron annihilation lifetime spectroscopy. Mechanisms of hardening and embrittlement generally consist of a ''precipitation-type'' and a ''damage-type'' component and the potential of annealing treatments for restoring the most of the original pressure vessel material toughness is examined; its conditions and mechanisms involved are discussed. Feasibility and economic evaluation of annealing costs is also carried out. 90 refs., 4 figs

Irradiation embrittlement of reactor pressure vessel steels: Considerations for thermal annealing

Energy Technology Data Exchange (ETDEWEB)

Burke, M G; Freyer, P D; Mager, T R

1994-12-31

In this paper, an overview of the irradiation embrittlement phenomenon is presented from a structure-properties viewpoint. Effects of irradiation conditions on embrittlement are first reviewed: irradiation temperature, fluence, flux, and steel or alloy composition. Then, the techniques for identifying/characterizing the irradiation-induced microstructural features are described: TEM/STEM (electron microscopy), small angle neutron scattering, atom probe field-ion microscopy, positron annihilation lifetime spectroscopy. Mechanisms of hardening and embrittlement generally consist of a ``precipitation-type`` and a ``damage-type`` component and the potential of annealing treatments for restoring the most of the original pressure vessel material toughness is examined; its conditions and mechanisms involved are discussed. Feasibility and economic evaluation of annealing costs is also carried out. 90 refs., 4 figs.

Attempts of local irradiation of cells by microbeam. From ultraviolet to heavy particles

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, Yasuhiko [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment

2002-03-01

This review describes the history of attempts of local irradiation of cells by microbeam and present status of the study. Local irradiation of cells was attempted as early as in 1912 with use of short {alpha}-particle range and of focused UV beams. After the war, laser microbeams were then developed for microsurgery in embryology. In addition, microbeams of electron generated from the gun and of X-ray collimated were developed. In 1950s, the electron microbeam was generated from Van de Graaff accelerator in Chicago University and proton, deuteron and He-ion microbeams from the cyclotron, in BNL. In 1980s, Gesellschaft fuer Schwerionenforshung (Germany) used heavy ion microbeams from C to U generated from the linear accelerator and PNL, proton to {sup 4}He-ion microbeams from the tandem-electrostatic accelerator. At present in 2002, the equipments for microbeam for cell irradiation are the Van de Graaff accelerators in Gray Cancer Institute (England) and in Columbia University, and the cyclotron in TIARA in Japan. The purpose of the study in TIARA is to develop a system to generate heavy particle microbeams for cell irradiation for analysis of the biological effect of ultra-low fluence, high LET heavy particles like the galactic cosmic ray. Recently, the CHO-KI cell nucleus is irradiated by {sup 40}Ar and {sup 20}Ne ions. (K.H.)

Attempts of local irradiation of cells by microbeam. From ultraviolet to heavy particles

International Nuclear Information System (INIS)

Kobayashi, Yasuhiko

2002-01-01

This review describes the history of attempts of local irradiation of cells by microbeam and present status of the study. Local irradiation of cells was attempted as early as in 1912 with use of short α-particle range and of focused UV beams. After the war, laser microbeams were then developed for microsurgery in embryology. In addition, microbeams of electron generated from the gun and of X-ray collimated were developed. In 1950s, the electron microbeam was generated from Van de Graaff accelerator in Chicago University and proton, deuteron and He-ion microbeams from the cyclotron, in BNL. In 1980s, Gesellschaft fuer Schwerionenforshung (Germany) used heavy ion microbeams from C to U generated from the linear accelerator and PNL, proton to 4 He-ion microbeams from the tandem-electrostatic accelerator. At present in 2002, the equipments for microbeam for cell irradiation are the Van de Graaff accelerators in Gray Cancer Institute (England) and in Columbia University, and the cyclotron in TIARA in Japan. The purpose of the study in TIARA is to develop a system to generate heavy particle microbeams for cell irradiation for analysis of the biological effect of ultra-low fluence, high LET heavy particles like the galactic cosmic ray. Recently, the CHO-KI cell nucleus is irradiated by 40 Ar and 20 Ne ions. (K.H.)

Specific Features of Structural-Phase State and Properties of Reactor Pressure Vessel Steel at Elevated Irradiation Temperature

Directory of Open Access Journals (Sweden)

E. A. Kuleshova

2017-01-01

Full Text Available This paper considers influence of elevated irradiation temperature on structure and properties of 15Kh2NMFAA reactor pressure vessel (RPV steel. The steel is investigated after accelerated irradiation at 300°C (operating temperature of VVER-1000-type RPV and 400°C supposed to be the operating temperature of advanced RPVs. Irradiation at 300°C leads to formation of radiation-induced precipitates and radiation defects-dislocation loops, while no carbide phase transformation is observed. Irradiation at a higher temperature (400°C neither causes formation of radiation-induced precipitates nor provides formation of dislocation loops, but it does increase the number density of the main initial hardening phase—of the carbonitrides. Increase of phosphorus concentration in grain boundaries is more pronounced for irradiation at 400°C as compared to irradiation at 300°C due to influence of thermally enhanced diffusion at a higher temperature. The structural-phase changes determine the changes of mechanical properties: at both irradiation temperatures irradiation embrittlement is mainly due to the hardening mechanism with some contribution of the nonhardening one for irradiation at 400°C. Lack of formation of radiation-induced precipitates at T = 400°C provides a small ΔTK shift (17°C. The obtained results demonstrate that the investigated 15Kh2NMFAA steel may be a promising material for advanced reactors with an elevated operating temperature.

Heavy-ion irradiation induced diamond formation in carbonaceous materials

International Nuclear Information System (INIS)

Daulton, T. L.

1999-01-01

The basic mechanisms of metastable phase formation produced under highly non-equilibrium thermodynamic conditions within high-energy particle tracks are investigated. In particular, the possible formation of diamond by heavy-ion irradiation of graphite at ambient temperature is examined. This work was motivated, in part, by earlier studies which discovered nanometer-grain polycrystalline diamond aggregates of submicron-size in uranium-rich carbonaceous mineral assemblages of Precambrian age. It was proposed that the radioactive decay of uranium formed diamond in the fission particle tracks produced in the carbonaceous minerals. To test the hypothesis that nanodiamonds can form by ion irradiation, fine-grain polycrystalline graphite sheets were irradiated with 400 MeV Kr ions. The ion irradiated graphite (and unirradiated graphite control) were then subjected to acid dissolution treatments to remove the graphite and isolate any diamonds that were produced. The acid residues were then characterized by analytical and high-resolution transmission electron microscopy. The acid residues of the ion-irradiated graphite were found to contain ppm concentrations of nanodiamonds, suggesting that ion irradiation of bulk graphite at ambient temperature can produce diamond

Neutron irradiation effects in reactor pressure vessel steels and weldments. Working document

International Nuclear Information System (INIS)

1998-10-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. A separate abstract was prepared for the introduction and for each of the eleven chapters, which are: 1. Reactor Pressure Vessel Design, 2. Reactor Pressure Materials, 3. WWER Pressure Vessels, 4. Determination of Mechanical Properties, 5. Neutron Exposure, 6. Methodology of Irradiation Experiments, 7. Effect of Irradiation on Mechanical Properties, 8. Mechanisms of Irradiation Embrittlement, 9. Modelling of Irradiation Damage, 10. Annealing of Irradiation Damage, 11. Safety Assessment using Surveillance Programmes and Data Bases

On the way to high resolution TEM characterization of dual ion beam irradiated ODS steels

International Nuclear Information System (INIS)

Hsiung, L.; Tumey, S.; Fluss, M. J.; King, W.; Marian, J.; Kuntz, J.; Dasher, B. El; Serruys, Y.; Willaime, F.; Kimura, A.

2009-01-01

Fission and fusion energy application of ODS steels while appearing promising requires that many key science issues be resolved. Among these issues are our incomplete understanding of the effect of irradiation on low-temperature fracture properties, the role of fusion relevant helium and hydrogen transmutation gases on the deformation and fracture of irradiated material at low and high temperatures, radiation-induced solute segregation and phase stability, mechanisms of swelling suppression in ODS steels, and the effects of radiation damage on localized deformation. While planning to focus on all these issues we are particularly interested in the atomic scale mechanism by which helium is mitigated by the nano scale particles. In order to obtain insight we are performing analytical transmission electron microscopy (AEM), high resolution electron microscopy (HRTEM) to investigate micro-structural and micro-compositional changes and property alterations of Fe-Cr ferritic/martensitic and ODS steels driven by temperature and ion-beam irradiation with Fe, H, and He. As a beginning to a collaboration between LLNL and CEA-Saclay, we have carried out an irradiation of four specimens, Fe, Fe14%Cr, and two ODS steels (14% Cr and 16% Cr) using the dual beam facility at CEA-Saclay (JANNuS). An Fe 8+ beam was implanted at 24 MeV and helium was implanted through a degrader wheel with energies between 1.7 MeV and 1.3 MeV. The nominal radiation parameters were 40 to 25 DPA, 10 to 25 appm He/DPA ratio, and specimen temperatures of ∼425 deg. C. Our goal is to compare the evolved microstructure with respect to the accumulation of helium at or near the particle matrix interface. Preparatory to this first study we have made many hi-resolution analyses of the nano-particles in the two ODS steels which serve as a base line for comparison with the TEM post irradiation examination reported here. These base line studies are reported separately at this conference. (author)

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.

Single-Event Effects in Power MOSFETs During Heavy Ion Irradiations Performed After Gamma-Ray Degradation

Science.gov (United States)

Busatto, G.; De Luca, V.; Iannuzzo, F.; Sanseverino, A.; Velardi, F.

2013-10-01

The robustness of commercial power metal-oxide semiconductor field-effect transistors to combined gamma-heavy ion irradiation has been investigated, evidence that the degradation of the gate oxide caused by the γ irradiation can severely corrupt the robustness to single-event effects and drastically modify the physical behavior of the device under test after the impact of a heavy ion. A decrease of the critical voltages at which destructive burnouts and gate ruptures for heavy ion impact appear, has been detected in the devices under test, which were previously irradiated with γ rays. In addition, the amount of critical voltage reduction is strictly related to the amount of the absorbed γ-ray dose. Furthermore, at the failure voltage, the behavior of the device is affected by the conduction of a current through the gate oxide. Moreover, the single-event gate rupture” of the device appears at lower voltages because of the reduction of the Fowler-Nordheim limit in the γ-irradiated devices.

Nuclear Data Processing for Generation of Stainless Steel Cross-Sections Data

International Nuclear Information System (INIS)

Suwoto; Zuhair

2007-01-01

Stainless steel has been used as important material in nuclear reactor and also in non nuclear industries. Nuclear data processing for generation of composite mixture cross-sections from several nuclides have been made. Provided evaluated nuclear data file (ENDF) such as ENDF/B- VI.8, JEFF-3.1 and JENDL-3.3 files were employed. Raw nuclear data cross-sections on file ENDF should be prepared and processed before it used in calculation. Sequence of nuclear data processing for generation of mixture cross-sections data from several nuclides is started from LINEAR, RECENT, SIGMA1 and MIXER codes taken from PREPR02000 utility code. Nuclear data processing is started from linearization of nuclear cross-sections data by using LINEAR code and counting background contribution of resonance parameter (MF2) with RECENT code (0 K) at energy ranges from 10 -5 to 10 7 eV. Afterward, the neutron cross-sections data should be processed and broadened to desire temperature (300 K) by using SIGMA1 code. Consistency of each cross-sections which used in nuclear data processing is checked and verified using FIXUP code. The next step is to define the composite mixture density (gr/cm 3 ) of stainless steel SUS-310 and weight fraction of each nuclide composition prior used it in MIXER code. All of the stainless steel SUS-310 cross sections are condensed to 650 energy groups structure (TART-energy structure) by using GROUPIE code to evaluate, analysis and review it more easily. The total, elastic scattering, non-elastic scattering and capture cross- sections of stainless steel SUS-310 have been made of ENDF/B-VI.8, JEFF-3.1 and JENDL-3.3 files. The stainless steel cross-sections made of ENDF/B- VI.8 file was taken as reference during validation process. The validation result of total cross-sections for stainless steel SUS-310 is clearly observed that the differences of total cross-sections error in nuclear data processing is relatively low than 0.01%. (author)