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Corrosion performance of structural alloys.

Description: Component reliability and long-term trouble-free performance of structural materials are essential in power-generating and gasification processes that utilize coal as a feedstock. During combustion and conversion of coal, the environments encompass a wide range of oxygen partial pressures, from excess-air conditions in conventional boilers to air-deficient conditions in 10W-NO{sub x} and gasification systems. Apart from the environmental aspects of the effluent from coal combustion and conversion, one concern from the systems standpoint is the aggressiveness of the gaseous/deposit environment toward structural components such as waterwall tubes, steam superheaters, syngas coolers, and hot-gas filters. The corrosion tests in the program described in this paper address the individual and combined effects of oxygen, sulfur, and chlorine on the corrosion response of several ASME-coded and noncoded structural alloys that were exposed to air-deficient and excess-air environments typical of coal-combustion and gasification processes. Data in this paper address the effects of preoxidation on the subsequent corrosion performance of structural materials such as 9Cr-1Mo ferritic steel, Type 347 austenitic stainless steel, Alloys 800, 825, 625, 214, Hastelloy X, and iron aluminide when exposed at 650 C to various mixed-gas environments with and without HCI. Results are presented for scaling kinetics, microstructural characteristics of corrosion products, detailed evaluations of near-surface regions of the exposed specimens, gains in our mechanistic understanding of the roles of S and Cl in the corrosion process, and the effect of preoxidation on subsequent corrosion.
Date: July 15, 1999
Creator: Natesan, K.
Partner: UNT Libraries Government Documents Department

Effects of material and loading variables on fatigue life of carbon and low-alloy steels in LWR environments

Description: The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Section III of the Code specifies fatigue design curves for structural materials. While effects of reactor coolant environments are not explicitly addressed by the design curves, test data suggest that the Code fatigue curves may not always be adequate in coolant environments. This paper reports the results of recent fatigue tests that examine the effects of steel type, strain rate, dissolved oxygen level, strain range, loading waveform, and surface morphology on the fatigue life of A106-Gr B carbon steel and A533-Gr B low-alloy steel in water.
Date: March 1, 1995
Creator: Chopra, O.K. & Shack, W.J.
Partner: UNT Libraries Government Documents Department

Effects of LWR coolant environments on fatigue lives of austenitic stainless steels.

Description: Fatigue tests have been conducted on Types 304 and 316NG stainless steels to evaluate the effects of various material and loading variables, e.g., steel type, strain rate, dissolved oxygen (DO) in water, and strain range, on the fatigue lives of these steels. The results confirm significant decreases in fatigue life in water. Unlike the situation with ferritic steels, environmental effects on Types 304 and 316NG stainless steel are more pronounced in low-DO than in high-DO water. Experimental results have been compared with estimates of fatigue life based on a statistical model. The formation and growth of fatigue cracks in air and water environments are discussed.
Date: January 13, 1998
Creator: Chopra, O. K.
Partner: UNT Libraries Government Documents Department

The Stress-Relief Cracking Susceptibility of a New Ferritic Steel - Part I: Single-Pass Heat-Affected Zone Simulations

Description: The stress-relief cracking susceptibility of single-pass welds in a new ferritic steel, HCM2S, has been evaluated and compared to 2.25Cr-1Mo steel using Gleeble techniques. Simulated coarse-grained heat-affected zones (CGHAZ) were produced under a range of energy inputs and tested at various post-weld heat treatment (PWHT) temperatures. Both alloys were tested at a stress of 325 MPa. The 2.25 Cr-1Mo steel was also tested at 270 MPa to normalize for the difference in yield strength between the two materials. Light optical and scanning electron microscopy were used to characterize the CGHAZ microstructure. The ''as-welded'' CGHAZ of each alloy consisted of lath martensite or bainite and had approximately equal prior austenite grain sizes. The as-welded hardness of the 2.25Cr-1Mo steel CGHAZ was significantly higher than that of the HCM2S alloy. Over the range studied energy input had no effect on the as-welded microstructure or hardness of either alloy. The energy input also had no effect on the stress-relief cracking susceptibility of either material. Both alloys failed intergranularly along prior austenite grain boundaries under all test conditions. The 2.25Cr-1Mo steel samples experienced significant macroductility and some microductility when tested at 325 MPa. The ductility decreased significantly when tested at 270 MPa but was still higher that than of HCM2S at each test condition. The time to failure decreased with increasing PWHT Temperature for each material. There was no significant difference in the times to failure between the two materials. Varying energy input and stress had no effect on the time-to failure. The ductility, as measured by reduction in are% increased with increasing PWHT temperature for 2.25 Cr-1Mo steel tested at both stresses. However, PWHT temperature had no effect on the ductility of HCM2S. The hardness of the CGHAZ for 2.25Cr-1Mo steel decreased significantly after PWHT, but remained constant for HCM2S. The differences in ...
Date: December 15, 1999
Creator: NAWROCKI,J.G.; DUPONT,J.N.; ROBINO,CHARLES V. & MARDER,A.R.
Partner: UNT Libraries Government Documents Department

The effect of aqueous environments upon the initiation and propagation of fatigue cracks in low-alloy steels

Description: The effect of elevated temperature aqueous environments upon the initiation and propagation of fatigue cracks in low-alloy steels is discussed in terms of the several parameters which influence such behavior. These parameters include water chemistry, impurities within the steels themselves, as well as factors such as the water flow rate, loading waveform and loading rates. Some of these parameters have similar effects upon both crack initiation and propagation, while others exhibit different effects in the two stages of cracking. In the case of environmentally-assisted crack (EAC) growth, the most important impurities within the steel are metallurgical sulfide inclusions which dissolve upon contact with the water. A ``critical`` concentration of sulfide ions at the crack tip can then induce environmentally-assisted cracking which proceeds at significantly increased crack growth rates over those observed in air. The occurrence, or non-occurrence, of EAC is governed by the mass-transport of sulfide ions to and from the crack-tip region, and the mass-transport is discussed in terms of diffusion, ion migration, and convection induced within the crack enclave. Examples are given of convective mass-transport within the crack enclave resulting from external free stream flow. The initiation of fatigue cracks in elevated temperature aqueous environments, as measured by the S-N fatigue lifetimes, is also strongly influenced by the parameters identified above. The influence of sulfide inclusions does not appear to be as strong on the crack initiation process as it is on crack propagation. The oxygen content of the environment appears to be the dominant factor, although loading frequency (strain rate) and temperature are also important factors.
Date: January 1, 1996
Creator: James, L.A. & Van Der Sluys, W.A.
Partner: UNT Libraries Government Documents Department

Piping inspection round robin

Description: The piping inspection round robin was conducted in 1981 at the Pacific Northwest National Laboratory (PNNL) to quantify the capability of ultrasonics for inservice inspection and to address some aspects of reliability for this type of nondestructive evaluation (NDE). The round robin measured the crack detection capabilities of seven field inspection teams who employed procedures that met or exceeded the 1977 edition through the 1978 addenda of the American Society of Mechanical Engineers (ASME) Section 11 Code requirements. Three different types of materials were employed in the study (cast stainless steel, clad ferritic, and wrought stainless steel), and two different types of flaws were implanted into the specimens (intergranular stress corrosion cracks (IGSCCs) and thermal fatigue cracks (TFCs)). When considering near-side inspection, far-side inspection, and false call rate, the overall performance was found to be best in clad ferritic, less effective in wrought stainless steel and the worst in cast stainless steel. Depth sizing performance showed little correlation with the true crack depths.
Date: April 1996
Creator: Heasler, P. G. & Doctor, S. R.
Partner: UNT Libraries Government Documents Department

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

Description: Reduced activation ferritic/martensitic steel, F-82H (Fe-8Cr-2W-V-Ta), was irradiated in the High Flux Isotope Reactor (HFIR) to doses between 11 and 34 dpa at 400 and 500 C. Post irradiation tensile tests were performed at the nominal irradiation temperature in vacuum. Some specimens included {sup 10}B or natural boron (nB) to estimate the helium effect on tensile properties. Tensile properties including the 0.2% offset yield stress, the ultimate tensile strength, the uniform elongation and the total elongation were measured. The tensile properties were not dependent on helium content in specimens irradiated to 34 dpa, however {sup 10}B-doped specimens with the highest levels of helium showed slightly higher yield strength and less ductility than boron-free specimens. Strength appears to go through a peak, and ductility through a trough at about 11 dpa. The irradiation to more than 21 dpa reduced the strength and increased the elongation to the unirradiated levels. Ferritic steels are one of the candidate alloys for nuclear fusion reactors because of their good thermophysical properties, their superior swelling resistance, and the low corrosion rate in contact with potential breeder and coolant materials.
Date: December 31, 1994
Creator: Shiba, Kiyoyuki; Suzuki, Masahide; Hishinuma, Akimichi & Pawel, J.E.
Partner: UNT Libraries Government Documents Department

Materials for breeding blankets

Description: There are several candidate concepts for tritium breeding blankets that make use of a number of special materials. These materials can be classified as Primary Blanket Materials, which have the greatest influence in determining the overall design and performance, and Secondary Blanket Materials, which have key functions in the operation of the blanket but are less important in establishing the overall design and performance. The issues associated with the blanket materials are specified and several examples of materials performance are given. Critical data needs are identified.
Date: September 1, 1995
Creator: Mattas, R.F. & Billone, M.C.
Partner: UNT Libraries Government Documents Department

Technical progress summary for the period, March 1, 1992--January 1, 1993

Description: During this period work has encompassed: (1) development of electropotential drop techniques to monitor the growth of cracks in steel specimens for a variety of specimen geometries; (2) micromechanical modeling of fracture using finite element calculations of crack and notch-tip stress and strain fields; (3) examining helium effects on radiation damage in austenitic and ferritic stainless steels; (4) analysis of the degradation of the mechanical properties of austenitic stainless steels for the purpose of assessing the feasibility of using these steels in ITER; (5) development of an integrated approach to integrity assessment; and (6) development of advanced methods of measuring fracture properties.
Date: February 1, 1996
Partner: UNT Libraries Government Documents Department

Fusion materials semiannual progress report for the period ending March 31, 1995

Description: This is the eighteenth in a series of semiannual technical progress reports on fusion materials. This report combines research and development activities which were previously reported separately in the following progress reports: {sm_bullet} Alloy Development for Irradiation Performance. {sm_bullet} Damage Analysis and Fundamental Studies. {sm_bullet} Special Purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. The Fusion Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide. This report has been compiled and edited under the guidance of A.F. Rowcliffe by Gabrielle Burn, Oak Ridge National Laboratory. Their efforts, and the efforts of the many persons who made technical contributions, are gratefully acknowledged.
Date: July 1, 1995
Partner: UNT Libraries Government Documents Department

Role of phase transformations in residual stress development in multipass ferritic steel welds and Gleeble test bars

Description: Neutron strain scanning has proven very effective in non-destructive mapping of the distribution of residual stresses in weldments. Strain scanning of Gleeble test bars of 2 1/4 Cr-1 Mo steel has been carried out in conjunction with strain scanning investigations of a multipass weld in 0.5-in. plate of the same alloy. The residual stresses in the Cleeble bars depend on the time spent at the maximum temperature and the rate of cooling. The longitudinal strains on the Gleeble bar center-line are tensile with a maximum on either side of the central hot zone. The transverse strains are compressive but vary with thermal treatment to a higher degree than variations in the longitudinal strains. The difference between strains at the center-line and off the center-line can be significantly greater than statistical error in aircooled Gleeble bars. The strains in the Gleeble bar have a high tensile component parallel to the direction of maximum heat transfer (viz. along the bar axis). By contrast, the large tensile strains in the heat-affected zone (HAZ) of the weldment are along the weld line which is essentially perpendicular to the direction of maximum heat transfer. The simulated conditions present in Gleeble bar test specimens are different from that observed in weld HAZ.
Date: December 31, 1995
Creator: Spooner, S.; David, S.A. & Hubbard, C.R.
Partner: UNT Libraries Government Documents Department

Technical progress report, March 1, 1991--December 12, 1991

Description: During this period work was focussed on three major topics: (1) developing electropotential drop techniques to monitor the growth of part-through cracks; (2) developing combined micromechanical and finite element crack tip field models for failure assessment for ferritic and martensitic steels; (3) analysis and assessment of radiation-induced degradation of the mechanical properties of austenitic stainless steels pertinent to near term fusion machines. These activities are part of a broad effort to characterize failure criteria of austenitic and ferritic steels anticipated for use in fusion reactor structures.
Date: February 1, 1996
Partner: UNT Libraries Government Documents Department

Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures

Description: Disk compact specimens of candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at nominal irradiation temperatures of either 90 or 250{degrees}C. These specimens have been tested over a temperature range from 20 to 250{degrees}C to determine J-integral values and tearing moduli. The results show that irradiation at these temperatures reduces the fracture toughness of austenitic stainless steels, but the toughness remains quite high. The toughness decreases as the test temperature increases. Irradiation at 250{degrees}C is more damaging than at 90{degrees}C, causing larger decreases in the fracture toughness. Ferritic-martensitic steels are embrittled by the irradiation, and show the lowest toughness at room temperature.
Date: December 31, 1994
Creator: Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F. & Shiba, Kiyoyuki
Partner: UNT Libraries Government Documents Department

Environmentally assisted cracking in Light Water Reactors: Semiannual report, October 1994--March 1995. Volume 20

Description: This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) from October 1994 to March 1995. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, (b) EAC of Alloy 600 and 690, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water with several dissolvedoxygen (DO) concentrations to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Tensile properties and microstructures of several heats of Alloy 600 and 690 were characterized for correlation with EAC of the alloys in simulated LWR environments. Effects of DO and electrochemical potential on susceptibility to intergranular cracking of high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath irradiated in boiling water reactors were determined in slow-strain-rate-tensile tests at 289{degrees}C. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.
Date: January 1996
Creator: Chung, H. M.; Chopra, O. K.; Gavenda, D. J.; Hins, A. G.; Kassner, T. F.; Ruther, W. E. et al.
Partner: UNT Libraries Government Documents Department

Environmentally assisted cracking in light water reactors

Description: This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.
Date: July 1, 1996
Creator: Chopra, O.K.; Chung, H.M. & Gruber, E.E.
Partner: UNT Libraries Government Documents Department

Effect of neutron irradiation on mechanical properties of ferritic steels

Description: Effect of neutron radiation exposure was investigated in various ferritic steels with the main emphasis being the effects of thermal neutrons on radiation hardening. Pure iron of varied grain sizes was also used for characterizing the grain size effects on the source hardening before and after neutron irradiation. While many steels are considered in the overall study, the results on 1020, A516 and A588 steels are emphasized. Radiation hardening due to fast neutrons was seen to be sensitive to the composition of the steels with A354 being the least resistant and A490 the least sensitive. Majority of the radiation hardening stems from friction hardening, and source hardening term decreased with exposure to neutron radiation apparently due to the interaction of interstitial impurities with radiation produced defects. Inclusion of thermal neutrons along with fast resulted in further decrease in the source hardening with a slight increase in the friction hardening which revealed a critical grain size below which exposure to total (fast and thermal) neutron spectrum resulted in a slight reduction in the yield stress compared to the exposure to only fast neutrons. This is the first time such a grain size effect is reported and this is shown to be consistent with known radiation effects on friction and source hardening terms along with the observation that low energy neutrons have a nonnegligible effect on the mechanical properties of steels. In ferritic steels, however, despite their small grain size, exposure to total neutron spectrum yielded higher strengths than exposure to only fast neutrons. This behavior is consistent with the fact that the source hardening is small in these alloys and radiation effect is due only to friction stress.
Date: December 31, 1995
Creator: Kass, S.B. & Murty, K.L.
Partner: UNT Libraries Government Documents Department

A comparison of low-chromium and high-chromium reduced-activation steels for fusion applications

Description: Ferritic steels have been considered candidate structural materials for first wall and blanket structures for fusion power plants since the late 1970s. The first steels considered in the United States were the conventional Cr-Mo steels Sandvik HT9 (nominally 12Cr-1Mo-0.25V-0.5W-0.5Ni-0.2C, here designated l2Cr-1MoVW), modified 9Cr-1Mo steel (9Cr-1Mo-0.2V-0.06Nb-0. IC, designated 9Cr-1MoVNb) and, to a lesser extent, 2 1/4Cr-1Mo steel (2.25Cr-Mo-0.1C). All compositions are in wt. %. The normalized-and-tempered 9 and 12Cr steels had a tempered martensite microstructure, and the normalized-and-tempered 2 1/4 Cr steel had a tempered bainite microstructure. This report describes chromium steels tested in normalized and tempered conditions. Miniature tensile and Charpy specimens were tested.
Date: November 1, 1996
Creator: Klueh, R.L.; Maziasz, P.J. & Alexander, D.J.
Partner: UNT Libraries Government Documents Department

Fusion materials semiannual progress report for the period ending June 30, 1996

Description: This report combines the full spectrum of research and development activities on both metallic and non-metallic materials with primary emphasis on the effects of the neutronic and chemical environment on the properties and performance of materials for in-vessel components. It is divided into the following chapters: vanadium alloys; silicon carbide components; ferritic-martensitic steels; copper alloys and high heat flux materials; austenitic stainless steels; insulating ceramics and optical materials; radiation effects, mechanistic studies, and experimental methods; dosimetry, damage parameters, and activation calculations; and irradiation facilities, test matrices, and experimental methods. There were no papers for the chapters on solid breeding materials and materials engineering and design requirement. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.
Date: October 1, 1996
Partner: UNT Libraries Government Documents Department

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

Description: Tensile tests were conducted on 8 reduced-activation Cr-W steels after irradiation to 15-17 and 26-29 dpa, and Charpy impact tests were conducted on steels irradiated to 26-29 dpa. Irradiation was in Fast Flux Test Facility at 365 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 20,000 h at 365 C. Thermal aging had little effect on tensile properties or ductile-brittle transition temperature (DBTT), but several steels showed a slight increase in upper-shelf energy (USE). After 7 dpa, strength increased (hardened) and then remained relatively unchanged through 26-29 dpa (ie, strength saturated with fluence). Post-irradiation Charpy impact tests after 26-29 dpa showed that the loss of impact toughness (increased DBTT, decreased 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 had the most irradiation resistance.
Date: October 1, 1996
Creator: Klueh, R.L. & Alexander, D.J.
Partner: UNT Libraries Government Documents Department

Estimation of lower-bound K{sub Jc} on pressure vessel steels from invalid data

Description: Statistical methods are currently being introduced into the transition temperature characterization of ferritic steels. Objective is to replace imprecise correlations between empirical impact test methods and universal K{sub Ic} or K{sub Ia} lower-bound curves with direct use of material-specific fracture mechanics data. This paper introduces a computational procedure that couples order statistics, weakest-link statistical theory, and a constraint model to arrive at estimates of lower-bound K{sub Jc} values. All of the above concepts have been used before to meet various objectives. In the present case, scheme is to make a best estimate of lower-bound fracture toughness when resource K{sub Jc} data are too few to use conventional statistical analyses. Utility of the procedure is of greatest value in the middle-to-high toughness part of the transition range where specimen constraint loss and elevated lower-bound toughness interfere with conventional statistical analysis methods.
Date: October 1, 1996
Creator: McCable, D. E. & Merkle, J. G.
Partner: UNT Libraries Government Documents Department

Recommendations for protecting against failure by brittle fracture: Category II and III ferritic steel shipping containers with wall thickness greater than four inches

Description: This report provides criteria for selecting ferritic steels that would prevent brittle fracture in Category II and III shipping containers with wall thickness greater than 4 inches. These methods are extensions of those previously used for Category II and III containers less than 4 inches thick and Category I containers more than 4 inches thick.
Date: August 1, 1996
Creator: Schwartz, M.W. & Fischer, L.E.
Partner: UNT Libraries Government Documents Department

Irradiation creep at temperatures of 400 {degrees}C and below for application to near-term fusion devices

Description: To study irradiation creep at 400{degrees}C and below, a series of six austenitic stainless steels and two ferritic alloys was irradiated sequentially in two research reactors where the neutron spectrum was tailored to produce a He production rate typical of a fusion device. Irradiation began in the Oak Ridge Research Reactor; and, after an atomic displacement level of 7.4 dpa, the specimens were moved to the High Flux Isotope Reactor for the remainder of the 19 dpa accumulated. Irradiation temperatures of 60, 200, 330, and 400{degrees}C were studied with internally pressurized tubes of type 316 stainless steel, PCA, HT 9, and a series of four laboratory heats of: Fe-13.5Cr-15Ni, Fe-13.5Cr-35Ni, Fe-1 3.5Cr-1 W-0.18Ti, and Fe-16Cr. At 330{degrees}C, irradiation creep was shown to be linear in fluence and stress. There was little or no effect of cold-work on creep under these conditions at all temperatures investigated. The HT9 demonstrated a large deviation from linearity at high stress levels, and a minimum in irradiation creep with increasing stress was observed in the Fe-Cr-Ni ternary alloys.
Date: December 31, 1996
Creator: Grossbeck, M.L.; Gibson, L.T. & Mansur, L.K.
Partner: UNT Libraries Government Documents Department

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

Description: 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.
Date: June 1, 1997
Creator: Klueh, R.L.
Partner: UNT Libraries Government Documents Department