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Fatigue strength of tungsten-copper duplex structures for divertor plates
Fatigue crack growth is considered to be a limiting factor for the life-time of the first wall of a Tokamak type controlled thermonuclear reactor. As a result of the high energy neutron spectrum in the D-T fusion reactions, not only defects, but also large quantities of helium due to nuclear transmutation are produced in the structural materials. One way of stimulating the damage of high energy neutrons is to use light ions by means of a charged particle accelerator. Type 316 stainless steel specimens were irradiated in a suitable irradiation chamber using a variable energy cyclotron. Fatigue crack growth under cyclic tensile strength was measured under simultaneous 18 MeV proton irradiation producing a displacement damage rate of the order of 10 -~ dpa d - ' . The same type of specimens were implanted with 38 MeV (z-particles and tested after bombardment. It was found that light ion irradiation has only a slight influence on fatigue crack growth at 500"C in type 316 stainless steel 23 refs.
Post irradiation tensile and fatigue behaviour of austenitic PCA stainless steels irradiated in HFIR. Tanaka, M.t~, Hamada, S., Hishinum& A. and Grossbeck, M.L. J. Nucl. Mater. J u l y (11) 1988, 156--1§7B, 9 5 7 - 9 6 2 Mechanical properties were determined on solution annealed (SA) and cold worked (CW) JPCA (Ti-modified austenitic stainless steel) irradiated in the high flux isotope reactor (HFIR) at temperatures ranging from 3 0 0 ~ ' C . The irradiation produced damage levels from 16-56 dpa and helium concentration from 1020-4100 appm. The improved stability of MC precipitates which formed in the matrix during irradiation prevent loss of ductility at 500"C and below. Application of solution annealed JPCA is recommended for structural components of fusion reactors to be operated at 500"C and below. 15 refs.
Influence of neutron irradiation at 430"C on the fatigue properties of SA 3"16L s t e e l . Vandermeulen, W., Hendrix, W., Massau¢ V and Van de Velde, J. J. Nucl. Mater. J u l y (11) 1988, 1 5 6 - 1 5 " / B , 9 6 3 - 9 5 6 Fatigue tests have been carried out at 430°C on hour-glass shaped specimens of the CEC reference heat of SDA 316L stainless steel. The tests were performed under constant total axial strain control with a triangular fully reversed wave shape at frequencies of 0.5, 0,05 and 0,005 Hz, Specimens irradiated at 430"C to doses of 9-12 dpe and helium contents of 80-145 appm showed a fatigue life reduction by about a factor of two, compared to unirrediated specimens. The cyclic stress is found to be strongly increased by the irradiation, The test frequency influences the fatigue hardening slightly but it does not affect the fatigue life, 4 refs.
Failure of f i r s t w a l l structures by fatigue crack growth. Diegele, E., Fatt, T., Munz, D. and Stamm, H. J. Nucl. Mater. J u l y (11) 1988, 1 5 5 - 1 5 7 B , 6 7 9 - 6 8 2 A general procedure of lifetime calculations for first wall components of a fusion reactor is presented, the principal mechanical behaviour of plasma faced structures is demonstrated, and effects of irradiation on the material behaviour are considered. Failure by fatigue crack growth is outlined in the case of an unprotected first wall concept, Two material types are taken into consideration, an austenitic steel (SS 316) and a martenistic steel (MANET, 1.4914). Furthermore. the influence of the cooling media and of the choice of the material is shown for two first wall concepts with radiatively cooled protection tiles of graphite 4 refs
The effect of neutron irradiation on the fatigue and fatigue-creep behaviour of structural materiels, van der Schaaf, B. J. Nucl. Mater. J u l y (11) 1988, 188--lb'TA, 1 5 6 - 1 6 3 The primary circuit of a pulsed type fusion reactor will be subjected to cyclic loads at high temperature Deformation rates witl be in the range from < 10 -8 s TM (creep) to ls " ~ (fatigue) leading to creep-fatigue interaction in the neutron irradiated structural materials. The effects of neutron irradiation on fatigue and fatigue-creep interaction are reviewed The study of austenitic stainless steel is most advanced, but the test conditions are still far from the first wall operating conditions, leaving a lot of uncertainties. It is expected that irradiation reduces the fatigue-creep endurance of austenites to very low levels, because of enhanced intergranular cracking. Two classes of alloys in an early stage of development, low activation steels and vanadium base alloys, hold the promise to be more fatigue-creep resistant, due to their more ductile creep behaviour 43 refs
The effect of helium on the fatigue properties of structural materials. Trinkaus, H. and UIImaier, H. J. Nucl. Mater. July (11) 1988, 155-1b'TA, 1 4 8 - 1 5 5 Experimental and theoretical work on high temperature fatigue properties of metals (SS316, 316L. 1.4948. Alloy 8206) containing helium is reported. The most important experimental results are: above 500"C. the reduction in the fatigue life to be attributed to He increases with increasing temperature and decreasing frequency or strain rate. respectively. This reduction is generally accompanied by a transition from transgranular to intergranular failure. Details depend sensitively on the material's state. This embrittlement effect is attributed to the dynamic transformation of He bubbles, breathing on grain boundaries under cycling stress, into unstably growing voids. Cavity stability under cycling stress (and analogously under pulsed irradiation) may be analysed in terms of attractor solutions of the cavity growth equation The stability limit is defined by the appearance of a 'semi-stable' limiting cycle For given stress (strain) amplitude, the relaxed critical bubble size is significantly lower below a temperature dependent critical frequency than above it. Possible chemical and microstructural influences on the time available until the cavity stability limit is reached are discussed in reference to observed fatigue life times. 3 refs.
Metallic materials as plasma facing compormnts~a review. Whitiey, J.B., Wilson, K.L and Buchermu~, D.A. J. Nucl. Mater. J u l y (11) 1988, 1 5 5 - 1 b - / A , 8 2 - 9 4 The design and fabrication of suitable plasma-facing components for fusion devices is a difficult task, with often conflicting requirements. Most operating devices utilize stainless steel as the first wall material and graphite for the main lirniters. There are potential problems with the extrapolation of graphite to power producing reactors and, hence, the evaluation of metallic alternatives is important. Prime candidate metals are beryllium and tungsten for plasma interactive components. There are several potential advantages to metallic compo nents over graphite or ceramics. In general, metals have superior fracture tougllness, are inert chemically to the hydrogen atmosphere, have a better developed fabrication and joining technology, and better resistance to neutron damage. Metals are limited, however,
206
in their ability to withstand disruption damage and. with the exception of Be, require a very low edge temperature to avoid sputtering problems. The current status of research in the areas of erosion, tritium permeation and inventory, thermal stress and fatigue, offnormal events, neutron irradiation an fabrication of these and other metals (Ni, Fe, Mo) is reviewed and compared to the prime non-metal choice of graphite. 90 refs
Joining Leak-before-break: an integrated approach for high energy piping. Chexal, VK., Norris, D.M. alxt Ser~r, W.L. Int. J. Pressure Vessels Pip/rig 1988, 34, ( 1 - 5 ) , 2 3 7 - 2 5 4 An integrated approach that invoives system design, thermal hydraulics, materials, and fracture mechanics analyses to assure that pipe failure is highly unlikely is described. This approach is based on a leak-before-break (LBB) promise and includes through-wall flaw detectable leakage, through-wall flaw stability, and part-through-wall flaw fatigue crack propagation catculations. A successful application of LBB can reduce the amount of excessive pipe rupture restraint hardware, Assuring LBB not only reduces initial construction. future maintenance, and radiation exposure costs, but the overatl safety and integrity of the plant are improved. This last benefit comes about from gaining additional insight into the piping systems and their capabilities. Details of the LBB methodology are presented, with specific examples of two pressurized water reactor lines (one inside containment fabricated of stainless steel 304, 316, and the other outside containment made from ferritic steel SA106B), The application of this approach at Beaver Valley Power Station (Unit 2) indictes that pipe rupture hardware is not necessary for stainless steel lines inside containment as small as 152 mm nominal pipe size that have passed screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel lines as small as 76 mm diameter (outside containment) can qualify for pipe rupture hardware elimination. 19 refs,
Fatigue strength of tungsten-copper duplex structures for divertor plates. Seki, M., Horie, T. and Tone, T. J. NucL Mater. J u l y (11) 1988, 1 5 5 - 1 5 7 A , 3 9 2 - 3 9 7 A W-Cu duplex structure is specified in a conceptual design of the Japan fusion experimental reactor (FER). The evaluation of the fatigue and creep life of the interface region between tungsten and Cu is essential for design of the divertor plate. Fatigue crack initiation life and crack propagation behaviour at room temperature and 200"C were measured for fully-annealed OFHC Cu and For W-OFHC Cu joints brazed with amorphous Ni-base filler metal. The debonding fatigue strength for the brazed joints was relatively high, but less than that of the Cu. Fatigue crack growth rates in the braze layer were approximately similar to those of the Cu. Fatigue lives were estimated for the divertor plate with small defects, and a method for analysing the apparent K-values of interface cracks was presented. 20 refs,
The influence of radiation on the properties of welds and joints. Tavassoli, A. A. J. Nucl. Mater. J u l y (11) 1988, 1r:8-1b'TA, 1 0 5 - 1 1 2 The effect of radiation on mechanical properties of candidate structural materials for the first wall and breeder blanket of fusion reactors is reviewed. The emphasis is placed on austenitic stainless steel type 316L and its weld metals; design parameters considered are similar to those currently specified for the next European torus, irradiation doses ~<15 dpa, temperatures ~< 400°C., number of pulse cycles approx 105 and hold times ~<15 min The effect of irradiation on other materials, including austenitic stainless steel Type 304L, weld metal Type 308L and ferritic/martensitic steels (~12% CFMo), and other service conditions such as temperatures as high as 550"C are also briefly discussed. The data collected and presented are those usually measured before and after irradiation, through tensile, impact toughness, fracture toughness fatigue, creep-fatigue and fatigue crack propagation in each case, the influence of irradiation parameters on the observed changes are discussed and relative conclusions are drawn The most important observation made is the lack of medium dose irradiation data on the weld metal and, in particular, on the electron beam welded joints 66 refs
Welding consumables and equipment for Honshu-Shikoku bridge project. Ueda, K. Kobelco Technol. Rev. 2 7 - 3 0 Aug. 1 988, (4) Some of the high grade welding consumables and equipment used for the Honshu-Shikoru bridges are introduced, Ultra low hydrogen covered electrodes (UL Series) with extremely low H content and low moisture pick-up have been developed in an oblique Y-groove weld cracking test, good results have been obtained. The 'LBF-62' covered electrode has remarkably improved fatigue strength in fillet welds. Applications of flux- and metalcored wires have enabled good welding performance and efficiency. Mechanized welding equipment for fillet welding (MIGTRAIN) has been extensively used for inner welds of box column, etc 1 ref
Damage to an isotope container. Stange, J. andZaminer, C. Prakt. Matallogr. Aug. 1988, 25, (8), 4 0 2 - 4 0 8 An isotope container used for the storage of an (r source for radio-therapeutic purposes was damaged by the fracture of the isotope positioning wire (made of X12CrNi17 7 steel). Positioned in the upper part of the container is the angular shaped holder in conjunction with a plug and an auxiliary plate onto which the positioning wire is welded, the tr source being mounted onto the opposite end of the positioning wire, The fracture of the wire occurred at its point of exit from the auxiliary plate. It was the aim of the investigation to clarify the causes of the wire fracture. Of special interest was the question, if material or working failures were pre-existing or if only the design-dependent sharp notch at the point of exit from the auxiliary plate and the operational conditions caused the fracture
Use of radioactive tracers for detecting cracks in steel structural welds i n s e a w a t e r . Jones, J.E., Natalie, C.A. and Burns, M. Weld. J. Sept. 1988, 67, (9), 4 7 - 5 3 The feasibili W of a system is summarized where, during fabrication, a radioactive tracer is included in the interior passes of a multiple pass weldment, but not in the outer cap and root passes. The feasibility has been examined using only chemical tracers, When a stress corrosion or corrosion fatigue crack penetrates through these outer passes and
Int J Fatigue May 1989
Post irradiation tensile and fatigue behaviour of austenitic PCA stainless steels irradiated in HFIR. Tanaka, M.t~, Hamada, S., Hishinum& A. and Grossbeck, M.L. J. Nucl. Mater. J u l y (11) 1988, 156--1§7B, 9 5 7 - 9 6 2 Mechanical properties were determined on solution annealed (SA) and cold worked (CW) JPCA (Ti-modified austenitic stainless steel) irradiated in the high flux isotope reactor (HFIR) at temperatures ranging from 3 0 0 ~ ' C . The irradiation produced damage levels from 16-56 dpa and helium concentration from 1020-4100 appm. The improved stability of MC precipitates which formed in the matrix during irradiation prevent loss of ductility at 500"C and below. Application of solution annealed JPCA is recommended for structural components of fusion reactors to be operated at 500"C and below. 15 refs.
Influence of neutron irradiation at 430"C on the fatigue properties of SA 3"16L s t e e l . Vandermeulen, W., Hendrix, W., Massau¢ V and Van de Velde, J. J. Nucl. Mater. J u l y (11) 1988, 1 5 6 - 1 5 " / B , 9 6 3 - 9 5 6 Fatigue tests have been carried out at 430°C on hour-glass shaped specimens of the CEC reference heat of SDA 316L stainless steel. The tests were performed under constant total axial strain control with a triangular fully reversed wave shape at frequencies of 0.5, 0,05 and 0,005 Hz, Specimens irradiated at 430"C to doses of 9-12 dpe and helium contents of 80-145 appm showed a fatigue life reduction by about a factor of two, compared to unirrediated specimens. The cyclic stress is found to be strongly increased by the irradiation, The test frequency influences the fatigue hardening slightly but it does not affect the fatigue life, 4 refs.
Failure of f i r s t w a l l structures by fatigue crack growth. Diegele, E., Fatt, T., Munz, D. and Stamm, H. J. Nucl. Mater. J u l y (11) 1988, 1 5 5 - 1 5 7 B , 6 7 9 - 6 8 2 A general procedure of lifetime calculations for first wall components of a fusion reactor is presented, the principal mechanical behaviour of plasma faced structures is demonstrated, and effects of irradiation on the material behaviour are considered. Failure by fatigue crack growth is outlined in the case of an unprotected first wall concept, Two material types are taken into consideration, an austenitic steel (SS 316) and a martenistic steel (MANET, 1.4914). Furthermore. the influence of the cooling media and of the choice of the material is shown for two first wall concepts with radiatively cooled protection tiles of graphite 4 refs
The effect of neutron irradiation on the fatigue and fatigue-creep behaviour of structural materiels, van der Schaaf, B. J. Nucl. Mater. J u l y (11) 1988, 188--lb'TA, 1 5 6 - 1 6 3 The primary circuit of a pulsed type fusion reactor will be subjected to cyclic loads at high temperature Deformation rates witl be in the range from < 10 -8 s TM (creep) to ls " ~ (fatigue) leading to creep-fatigue interaction in the neutron irradiated structural materials. The effects of neutron irradiation on fatigue and fatigue-creep interaction are reviewed The study of austenitic stainless steel is most advanced, but the test conditions are still far from the first wall operating conditions, leaving a lot of uncertainties. It is expected that irradiation reduces the fatigue-creep endurance of austenites to very low levels, because of enhanced intergranular cracking. Two classes of alloys in an early stage of development, low activation steels and vanadium base alloys, hold the promise to be more fatigue-creep resistant, due to their more ductile creep behaviour 43 refs
The effect of helium on the fatigue properties of structural materials. Trinkaus, H. and UIImaier, H. J. Nucl. Mater. July (11) 1988, 155-1b'TA, 1 4 8 - 1 5 5 Experimental and theoretical work on high temperature fatigue properties of metals (SS316, 316L. 1.4948. Alloy 8206) containing helium is reported. The most important experimental results are: above 500"C. the reduction in the fatigue life to be attributed to He increases with increasing temperature and decreasing frequency or strain rate. respectively. This reduction is generally accompanied by a transition from transgranular to intergranular failure. Details depend sensitively on the material's state. This embrittlement effect is attributed to the dynamic transformation of He bubbles, breathing on grain boundaries under cycling stress, into unstably growing voids. Cavity stability under cycling stress (and analogously under pulsed irradiation) may be analysed in terms of attractor solutions of the cavity growth equation The stability limit is defined by the appearance of a 'semi-stable' limiting cycle For given stress (strain) amplitude, the relaxed critical bubble size is significantly lower below a temperature dependent critical frequency than above it. Possible chemical and microstructural influences on the time available until the cavity stability limit is reached are discussed in reference to observed fatigue life times. 3 refs.
Metallic materials as plasma facing compormnts~a review. Whitiey, J.B., Wilson, K.L and Buchermu~, D.A. J. Nucl. Mater. J u l y (11) 1988, 1 5 5 - 1 b - / A , 8 2 - 9 4 The design and fabrication of suitable plasma-facing components for fusion devices is a difficult task, with often conflicting requirements. Most operating devices utilize stainless steel as the first wall material and graphite for the main lirniters. There are potential problems with the extrapolation of graphite to power producing reactors and, hence, the evaluation of metallic alternatives is important. Prime candidate metals are beryllium and tungsten for plasma interactive components. There are several potential advantages to metallic compo nents over graphite or ceramics. In general, metals have superior fracture tougllness, are inert chemically to the hydrogen atmosphere, have a better developed fabrication and joining technology, and better resistance to neutron damage. Metals are limited, however,
206
in their ability to withstand disruption damage and. with the exception of Be, require a very low edge temperature to avoid sputtering problems. The current status of research in the areas of erosion, tritium permeation and inventory, thermal stress and fatigue, offnormal events, neutron irradiation an fabrication of these and other metals (Ni, Fe, Mo) is reviewed and compared to the prime non-metal choice of graphite. 90 refs
Joining Leak-before-break: an integrated approach for high energy piping. Chexal, VK., Norris, D.M. alxt Ser~r, W.L. Int. J. Pressure Vessels Pip/rig 1988, 34, ( 1 - 5 ) , 2 3 7 - 2 5 4 An integrated approach that invoives system design, thermal hydraulics, materials, and fracture mechanics analyses to assure that pipe failure is highly unlikely is described. This approach is based on a leak-before-break (LBB) promise and includes through-wall flaw detectable leakage, through-wall flaw stability, and part-through-wall flaw fatigue crack propagation catculations. A successful application of LBB can reduce the amount of excessive pipe rupture restraint hardware, Assuring LBB not only reduces initial construction. future maintenance, and radiation exposure costs, but the overatl safety and integrity of the plant are improved. This last benefit comes about from gaining additional insight into the piping systems and their capabilities. Details of the LBB methodology are presented, with specific examples of two pressurized water reactor lines (one inside containment fabricated of stainless steel 304, 316, and the other outside containment made from ferritic steel SA106B), The application of this approach at Beaver Valley Power Station (Unit 2) indictes that pipe rupture hardware is not necessary for stainless steel lines inside containment as small as 152 mm nominal pipe size that have passed screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel lines as small as 76 mm diameter (outside containment) can qualify for pipe rupture hardware elimination. 19 refs,
Fatigue strength of tungsten-copper duplex structures for divertor plates. Seki, M., Horie, T. and Tone, T. J. NucL Mater. J u l y (11) 1988, 1 5 5 - 1 5 7 A , 3 9 2 - 3 9 7 A W-Cu duplex structure is specified in a conceptual design of the Japan fusion experimental reactor (FER). The evaluation of the fatigue and creep life of the interface region between tungsten and Cu is essential for design of the divertor plate. Fatigue crack initiation life and crack propagation behaviour at room temperature and 200"C were measured for fully-annealed OFHC Cu and For W-OFHC Cu joints brazed with amorphous Ni-base filler metal. The debonding fatigue strength for the brazed joints was relatively high, but less than that of the Cu. Fatigue crack growth rates in the braze layer were approximately similar to those of the Cu. Fatigue lives were estimated for the divertor plate with small defects, and a method for analysing the apparent K-values of interface cracks was presented. 20 refs,
The influence of radiation on the properties of welds and joints. Tavassoli, A. A. J. Nucl. Mater. J u l y (11) 1988, 1r:8-1b'TA, 1 0 5 - 1 1 2 The effect of radiation on mechanical properties of candidate structural materials for the first wall and breeder blanket of fusion reactors is reviewed. The emphasis is placed on austenitic stainless steel type 316L and its weld metals; design parameters considered are similar to those currently specified for the next European torus, irradiation doses ~<15 dpa, temperatures ~< 400°C., number of pulse cycles approx 105 and hold times ~<15 min The effect of irradiation on other materials, including austenitic stainless steel Type 304L, weld metal Type 308L and ferritic/martensitic steels (~12% CFMo), and other service conditions such as temperatures as high as 550"C are also briefly discussed. The data collected and presented are those usually measured before and after irradiation, through tensile, impact toughness, fracture toughness fatigue, creep-fatigue and fatigue crack propagation in each case, the influence of irradiation parameters on the observed changes are discussed and relative conclusions are drawn The most important observation made is the lack of medium dose irradiation data on the weld metal and, in particular, on the electron beam welded joints 66 refs
Welding consumables and equipment for Honshu-Shikoku bridge project. Ueda, K. Kobelco Technol. Rev. 2 7 - 3 0 Aug. 1 988, (4) Some of the high grade welding consumables and equipment used for the Honshu-Shikoru bridges are introduced, Ultra low hydrogen covered electrodes (UL Series) with extremely low H content and low moisture pick-up have been developed in an oblique Y-groove weld cracking test, good results have been obtained. The 'LBF-62' covered electrode has remarkably improved fatigue strength in fillet welds. Applications of flux- and metalcored wires have enabled good welding performance and efficiency. Mechanized welding equipment for fillet welding (MIGTRAIN) has been extensively used for inner welds of box column, etc 1 ref
Damage to an isotope container. Stange, J. andZaminer, C. Prakt. Matallogr. Aug. 1988, 25, (8), 4 0 2 - 4 0 8 An isotope container used for the storage of an (r source for radio-therapeutic purposes was damaged by the fracture of the isotope positioning wire (made of X12CrNi17 7 steel). Positioned in the upper part of the container is the angular shaped holder in conjunction with a plug and an auxiliary plate onto which the positioning wire is welded, the tr source being mounted onto the opposite end of the positioning wire, The fracture of the wire occurred at its point of exit from the auxiliary plate. It was the aim of the investigation to clarify the causes of the wire fracture. Of special interest was the question, if material or working failures were pre-existing or if only the design-dependent sharp notch at the point of exit from the auxiliary plate and the operational conditions caused the fracture
Use of radioactive tracers for detecting cracks in steel structural welds i n s e a w a t e r . Jones, J.E., Natalie, C.A. and Burns, M. Weld. J. Sept. 1988, 67, (9), 4 7 - 5 3 The feasibili W of a system is summarized where, during fabrication, a radioactive tracer is included in the interior passes of a multiple pass weldment, but not in the outer cap and root passes. The feasibility has been examined using only chemical tracers, When a stress corrosion or corrosion fatigue crack penetrates through these outer passes and
Int J Fatigue May 1989