skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Evaluation of irradiation effects of 16 MeV proton-irradiated 12Cr-1MoV steel by small punch (SP) tests

Abstract

Recently, interest in small-scale specimens for testing irradiated materials has arisen in conjunction with the need to develop materials for fusion reactor materials and to study irradiation effects using an ion irradiation facility. Several attempts have been made to evaluate material property changes due to irradiation using a small specimen technique. The SP (small punch) test is an example of small-scale specimen test techniques, originally developed by Baik et al. to estimate DBTT (ductile-to-brittle transition temperature) using broken standard CVN (Charpy 5-notch) specimens. The objective of the present study is to evaluate 16 MeV proton irradiation effects on a fusion reactor candidate material in terms of changes in energy up to failure and J[sub IC] fracture toughness (SP J[sub IC]) by using a SP test technique and a J[sub IC] - [bar [epsilon]][sub qf] relationship. It has been known that protons at 16MeV accurately simulate the 14 MeV neutron-damage energy spectrum over most of the PKA energy range.

Authors:
;  [1];  [2]
  1. (Korea Atomic Energy Research Inst., Taejon (Korea, Republic of))
  2. (Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of). Dept. of Nuclear Engineering)
Publication Date:
OSTI Identifier:
7303243
Resource Type:
Journal Article
Resource Relation:
Journal Name: Scripta Metallurgica et Materialia; (United States); Journal Volume: 30:12
Country of Publication:
United States
Language:
English
Subject:
36 MATERIALS SCIENCE; 70 PLASMA PHYSICS AND FUSION TECHNOLOGY; Ma; STEEL-CRMOV; PHYSICAL RADIATION EFFECTS; THERMONUCLEAR REACTOR MATERIALS; EXPERIMENTAL DATA; FRACTURE PROPERTIES; MECHANICAL TESTS; ALLOYS; CHROMIUM ALLOYS; COPPER ADDITIONS; COPPER ALLOYS; DATA; INFORMATION; IRON ALLOYS; IRON BASE ALLOYS; LOW ALLOY STEELS; MATERIALS; MATERIALS TESTING; MECHANICAL PROPERTIES; MOLYBDENUM ADDITIONS; MOLYBDENUM ALLOYS; NICKEL ADDITIONS; NICKEL ALLOYS; NUMERICAL DATA; RADIATION EFFECTS; STEELS; TESTING; VANADIUM ADDITIONS; VANADIUM ALLOYS; 360106* - Metals & Alloys- Radiation Effects; 700480 - Fusion Technology- Component Development; Materials Studies- (1992-)

Citation Formats

Chi, S.H., Hong, J.H., and Kim, I.S. Evaluation of irradiation effects of 16 MeV proton-irradiated 12Cr-1MoV steel by small punch (SP) tests. United States: N. p., 1994. Web. doi:10.1016/0956-716X(94)90301-8.
Chi, S.H., Hong, J.H., & Kim, I.S. Evaluation of irradiation effects of 16 MeV proton-irradiated 12Cr-1MoV steel by small punch (SP) tests. United States. doi:10.1016/0956-716X(94)90301-8.
Chi, S.H., Hong, J.H., and Kim, I.S. 1994. "Evaluation of irradiation effects of 16 MeV proton-irradiated 12Cr-1MoV steel by small punch (SP) tests". United States. doi:10.1016/0956-716X(94)90301-8.
@article{osti_7303243,
title = {Evaluation of irradiation effects of 16 MeV proton-irradiated 12Cr-1MoV steel by small punch (SP) tests},
author = {Chi, S.H. and Hong, J.H. and Kim, I.S.},
abstractNote = {Recently, interest in small-scale specimens for testing irradiated materials has arisen in conjunction with the need to develop materials for fusion reactor materials and to study irradiation effects using an ion irradiation facility. Several attempts have been made to evaluate material property changes due to irradiation using a small specimen technique. The SP (small punch) test is an example of small-scale specimen test techniques, originally developed by Baik et al. to estimate DBTT (ductile-to-brittle transition temperature) using broken standard CVN (Charpy 5-notch) specimens. The objective of the present study is to evaluate 16 MeV proton irradiation effects on a fusion reactor candidate material in terms of changes in energy up to failure and J[sub IC] fracture toughness (SP J[sub IC]) by using a SP test technique and a J[sub IC] - [bar [epsilon]][sub qf] relationship. It has been known that protons at 16MeV accurately simulate the 14 MeV neutron-damage energy spectrum over most of the PKA energy range.},
doi = {10.1016/0956-716X(94)90301-8},
journal = {Scripta Metallurgica et Materialia; (United States)},
number = ,
volume = 30:12,
place = {United States},
year = 1994,
month = 6
}
  • Neutron-irradiation-induced embrittlement of a 2.25Cr1Mo steel is investigated by means of small punch testing along with scanning electron microscopy. There is an apparent irradiation-induced embrittlement effect after the steel is irradiated at about 400 deg. C for 86 days with a neutron dose rate of 1.75x10{sup -8} dpa/s. The embrittlement is mainly nonhardening embrittlement caused by impurity grain boundary segregation.
  • This paper reports on the 2-1/4 Cr-1M{sub 0} steel that has been selected as the material for the reactor pressure vessel (RPV) of a multipurpose experimental high temperature gas cooled reactor designed by JAERI. The 2-1/4 Cr-1M{sub 0} steel has successful records for high temperature pressure vessels in the petrochemical industries and the ASME Code Case authorizes the use of the steel in these pressure vessels. However, the steel has not been used to nuclear reactor pressure vessels so far. Since the material in the so-called belt line region of the nuclear pressure vessels undergo changes in toughness and strengthmore » due to neutron irradiation, it is quite urgent to collect the fracture toughness and strength data of the irradiated steel for the evaluation of the structural intergravity of the reactor pressure vessel of high radiation resistance. In order to study irradiation damage of 2-1/4 Cr-1M{sub 0} steel, small specimens are required because of the severe limitations on specimen size in irradiated-material testing facilities (e.g. the limited space available for testing in nuclear reactors and the narrow damage zone produced by charged particle accelerators). In order to obtain more information about fracture properties of the 2-1/4 Cr- 1M{sub 0} steel from specimens, a subsized compact tensile (CT) specimen, a small punch (SP) specimen and tensile specimen of the irradiated 2-1/4 Cr-1M{sub 0} steel were used to provide radiation effects on fracture toughness, yield strength and ultimate strength. The small punch test, which has been developed recently provides information of the yield and ultimate strength as well as fracture toughness. This report describes the behavior of the neutron irradiation embrittlement of the nuclear reactor pressure vessel steel 2-1/4 Cr-1M{sub 0} by use of new testing approach - subsized specimen techniques.« less
  • To minimize waste disposal problems associated with the residual radioactivity of the first wall material of a fusion reactor, fast induced radioactive decay (FIRD) alloys based on the Fe-Cr-Mn system are being investigated. The objective of this research was to evaluate the effects of irradiation on cyclic strain localization and fatigue crack initiation in a FIRD Fe-12Cr-20Mn alloy and to compare the response to commercially available 316 stainless steel. The alloys were irradiated with 200 keV Fe ions to a dose of 1 {times} 10{sup 16} ions/cm{sup 2} and 15.5 keV He ions to a dose of 7 {times} 10{supmore » 15} ions/cm{sup 2} to simulate the irradiation-induced defect structure and helium concentration that would be produced in a fusion reactor. Irradiated specimens were fatigued in a cantilever beam fatigue testing machine with the deflection set to produce a fully reversed total strain amplitude of 0.25% on the surface of the specimen. Acetate replicas were obtained during the fatigue tests to provide a record of surface fatigue damage. Transmission electron microscopy (TEM) analyses were performed to characterize the microstructural changes resulting from the irradiations and interactions between fatigue-induced glide dislocations and the irradiation-induced defects. Results indicate that the irradiated Fe-Cr-Mn alloy exhibits fatigue properties similar to 316 stainless steel. Glide dislocations produced by fatigue cycling annihilate irradiation-induced defects. The defect annihilation causes the formation of cleared channels in which the cyclic plastic strain is localized. Subsurface slip bands penetrate the irradiated regions through the cleared channels and serve as fatigue crack initiation sites.« less
  • The ductile-brittle transition temperature in steel is commonly determined using Charpy V-notch impact specimens as specified by ASTM E23-81. In some specific cases, however, the use of this standardized test specimen may be impractical, if not impossible. For instance, it is well known that ferritic steels show a substantial degradation of the mechanical properties after long time exposure to an irradiation environment. Because of the increase in strength and the reduction in ductility due to neutron irradiation, the Charpy V-notch transition temperature is raised causing concern from a safety point of view. To study these radiation effects, a test specimenmore » much smaller than the standard Charpy V-notch specimen would be extremely desirable for two reasons. First, to study neutron damage small specimens take less space within a reactor. Secondly, the damage achieved in simulation experiments, such as proton or electron accelerators, is limited to small penetration depths. Several efforts on the development of such a small test specimen, similar to that used to determine the ductility of sheet metal, as recommended by ASTM E643-78, have been described in the literature. The paper reports on correlations between small punch (SP) and Charpy V-notch (CVN) test results obtained on temper-embrittled NiCr steel. The ductile-brittle transition temperature (DBTT) with intergranular embrittlement being induced by grain boundary segregation of specific impurities was determined. The relation between test results discussed in terms of the micromechanisms of intergranular cracking. It is suggested that in radiation embrittlement investigations similar correlations may be obtained.« less
  • Experimentally quantifying the mechanical effects of radiation damage in reactor materials is necessary for the development and qualification of new materials for improved performance and safety. This can be achieved in a high-throughput fashion through a combination of ion beam irradiation and small scale mechanical testing in contrast to the high cost and laborious nature of bulk testing of reactor irradiated samples. The current paper focuses on using spherical nanoindentation stress-strain curves on unirradiated and proton irradiated (10 dpa at 360 °C) 304 stainless steel to quantify the mechanical effects of radiation damage. Spherical nanoindentation stress-strain measurements show a radiation-inducedmore » increase in indentation yield strength from 1.36 GPa to 2.72 GPa and a radiation-induced increase in indentation work hardening rate of 10 GPa–30 GPa. These measurements are critically compared against Berkovich nanohardness, micropillar compression, and micro-tension measurements on the same material and similar grain orientations. The ratio of irradiated to unirradiated yield strength increases by a similar factor of 2 when measured via spherical nanoindentation or Berkovich nanohardness testing. A comparison of spherical indentation stress-strain curves to uniaxial (micropillar and micro-tension) stress-strain curves was achieved using a simple scaling relationship which shows good agreement for the unirradiated condition and poor agreement in post-yield behavior for the irradiated condition. Finally, the disagreement between spherical nanoindentation and uniaxial stress-strain curves is likely due to the plastic instability that occurs during uniaxial tests but is absent during spherical nanoindentation tests.« less