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Weldability of irradiated stainless steel (SS) has been studied to develop the technical guideline regarding the repair-welding of reactor internals. Type 304 and 316L SSs were irradiated at ambient temperature in the US Advanced Test Reactor. The multi-pass bead-on-plate TIG (GTA) and YAG laser welding with heat input levels less than 1 MJ/m were performed on specimens containing helium up to 18 appm. In this paper, results of cross-sectional micrograph observations of the heat affected zone were considered in light of helium bubble properties. The tendency for weld crack formation of irradiated Type 316L SS was compared with that of irradiated Type 304 SS.

Helium bubble structure was examined on a helium-implanted stainless steel after applying thermal and stress cycle using a weld thermal and stress cycle simulator. Helium ions were implanted on Type 304 stainless steels up to 200 appm uniformly to a depth of 3.5 μm. The specimens were heated at various temperatures between 1073 and 1473 K for 2 s in Ar gas atmosphere. Tensile stresses from 0.5 to 8 MPa were applied during the thermal cycle. TEM observations revealed that size of the bubbles at grain boundaries was larger for the specimens with a higher tensile stress and with a higher annealing temperature. Densities of bubbles increased with increasing helium content. A theoretical model calculation showed a good agreement with the experimental results.

Helium bubble structure was examined in helium-implanted stainless steels after applying thermal cycles and tensile stresses using a weld thermal and stress cycle simulator. SUS304 specimens implanted with helium ions to 5 appm were heated at 1473 K for 2 s in an argon gas atmosphere. The heat-up rate and cooling rate were controlled to be 90 and 130 K/s, respectively. Tensile stresses ranging from 0 to 4 MPa were applied immediately after reaching a temperature of 1473 K. TEM observation revealed that bubble formation occurred even after short annealing times and that the size of the helium bubbles was strongly dependent on the tensile stress during heating.

Degradation of weldability caused by helium, which is generated by nuclear transmutation in neutron 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 × 1024 to 1.4 × 1026 n/m2 (E>1 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 actual reactor components.

Bead-on-plate welding experiments using a 400 W YAG laser were conducted on SUS304 stainless steels implanted with helium ions of 0.5, 5 and 50 appm uniformly to a depth of 0.25 mm. High heat input welding at 20 kJ/cm caused surface grain boundary cracking in the heat-affected zone at 50 appm He. Cross-sectional observations after etching in oxalic acid solution revealed that bubble growth at grain boundaries in the heat-affected zone was enhanced at higher heat input and at higher helium concentrations. Bubble growth was negligible for the laser welding condition of 1 kJ/cm even at 50 appm He. The results suggest that YAG laser welding is a promising welding technique for stainless steels containing high amounts of helium.

A simulation model of helium bubble behavior and crack initiation at grain boundaries in the heat-affected zone (HAZ) during welding is proposed, and the effects of welding conditions and helium concentration on bubble evolution and cracking are evaluated. The model was based on the following assumptions; homogeneous bubble nucleation, bubble coalescence by random motion, bubble growth by vacancy absorption and ductile fracture. The result of calculations for different welding heat inputs reveal that the final bubble size increases with increasing the weld heat input and with decreasing bubble spacing at the grain boundary. The calculated critical heat input for cracking as a function of helium concentration is in good agreement with the results of welding experiments on neutron-irradiated stainless steels and helium-doped stainless steels.

Stress corrosion cracking (SCC) is one of the major reasons to reduce the reliability of aged reactor components. Toshiba has been developing underwater laser welding onto surface of the aged components as maintenance and repair techniques. Because most of the reactor internal components to apply this underwater laser welding technique have 3-dimensional shape, effect of welding positions and welded shapes are examined and presented in this report.Copyright © 2006 by ASME

A technology to prevent the occurrence of Intergranular Stress Corrosion Cracking (IGSCC) by irradiating a high power Nd:YAG laser beam was developed. Laser Desensitization Treatment (LDT) process was realized by irradiating a laser beam onto the sensitized Heat Affected Zone (HAZ) surface of SUS304 stainless steel. LDT was formed by both a molten layer of approximately 0.2mm depth and a solution heat treated layer. The results of a Creviced Bent Beam (CBB) test showed that no cracks had appeared on the surface of LDT. After LDT was applied in the vicinity of welding joints on the inside surface of pipes, tensile residual stress was measured there. On the other hand, the tensile stress of outside surface of the pipes was decreased. From these results, LDT processing on the inside surface of a pipe can be expected to prevent the occurrence of IGSCC owing to the effect of both metallurgical improvement and decrease of the residual stress on the outside surface of the pipe. We developed the LDT processing system and successfully applied on the pipes of some actual nuclear power plants.A technology to prevent the occurrence of Intergranular Stress Corrosion Cracking (IGSCC) by irradiating a high power Nd:YAG laser beam was developed. Laser Desensitization Treatment (LDT) process was realized by irradiating a laser beam onto the sensitized Heat Affected Zone (HAZ) surface of SUS304 stainless steel. LDT was formed by both a molten layer of approximately 0.2mm depth and a solution heat treated layer. The results of a Creviced Bent Beam (CBB) test showed that no cracks had appeared on the surface of LDT. After LDT was applied in the vicinity of welding joints on the inside surface of pipes, tensile residual stress was measured there. On the other hand, the tensile stress of outside surface of the pipes was decreased. From these results, LDT processing on the inside surface of a pipe can be expected to prevent the occurrence of IGSCC owing to the effect of both metallurgical improvement and decrease of the residual stress on the outside surface of the pipe. We developed the LDT processing sy...