Weld residual stress is one of the most important factors in order to assess the structural integrity of safety-related components such as reactor pressure vessel (RPV) in long-term operation of nuclear power plant since the residual stress significantly affects crack initiation and growth behaviors. The inner surface of the RPV made of low alloy steel is protected against corrosion by weld-overlay cladding of austenitic stainless steel. At the J-weld of the vessel head penetrations, Ni-based alloys are used for weld material. The residual stresses generated within the cladding, J-weld and base material were measured as-welded and after PWHT conditions using the deep-hole-drilling method. Thermal-elastic-plastic-creep analyses considering the phase transformation were also performed to evaluate the weld residual stress. By comparing analytical results with the measured ones, it was shown that there was roughly a good agreement of residual stress distribution within the cladding, J-weld and base metal. It was also suggested that taking the phase transformation during welding and PWHT into account was important to improve the accuracy of weld residual stress analysis. Using the residual stress distributions, fracture mechanics analyses for primary water stress corrosion cracking (PWSCC) have been performed using probabilistic fracture mechanics analysis code. Effects of the weld residual stress and scatter of PWSCC growth rate on the crack penetration were evaluated through some case studies.

Environmental mitigation is being looked forward to as one of the promising preventive measures for better plant maintenance. Especially, it would be much more effective, when environmental measures are introduced in combination with mechanical measures, such as stress improvement, material replacement, and so on.

Dissolved hydrogen (DH) concentration in the primary cooling system has been controlled at the level of 25~35 ml STP/kg H₂O in all PWR plants in Japan, maintaining a sufficient margin to the utilities’ self-regulated band 15~50 ml STP/kg H₂O. The lower limit of this self-regulated band was determined to suppress reactor coolant radiolysis based on experiments using a test reactor over 50 years ago, though the test was performed at much lower than typical operational temperature and is considered over-conservative at present.

On the other hand, primary water stress corrosion cracking (PWSCC) events have occurred on Dissimilar Metal Welds (DMW) at vessel nozzles or at vessel head penetration in some PWR plants in Japan in recent years.  PWSCC growth test data have been showing that the current management band of DH content is around the peak of crack growth rate (CGR) at about 340ºC, corresponding to the operating temperature of the pressurizer. To avoid the higher CGR levels, the DH content band should be shifted to the higher side, or to the lower side. The authors suggest much lower DH content than that at present as an alternative control, since some experimental data on PWSCC initiation show monotonically decreasing dependency of initiation time with increasing DH content.

In this paper, we show the effects of DH content on PWSCC initiation property. Reverse-u-bend (RUB) tests and three-point-bend (3PB) tests were conducted for Alloy 600 base metal and Alloy 182, respectively, to examine the dependency of the lower DH content than currently controlled band on PWSCC initiation time.  Test results showed the advantage of lowering DH content down to around 5 ml STP/kg H₂O compared to 15~25 ml STP/kg H₂O. Analyses results of oxide and deposits on the surface of the test specimens were also shown to try to clarify how hydrogen interacts with nickel-base alloys and influences the propensity to PWSCC.  Based on the analyses results of surface oxide and deposits, an outline of the interface region was drawn at low and intermediate to high DH content.

Current status of the total studies on DH optimization and challenges left were summarized.

Inspection and Evaluation Guidelines for reactor internals has been taken into the Rules on Fitness-for–Service for Nuclear Power Plants of The Japan Society of Mechanical Engineers. It is a base of the maintenance plan of each Nuclear Power Plant. A plant maintenance methodology will have more importance to maintain the plant safety and stable plant operation. This paper introduces the systematization of the maintenance such as repair, replacement, preventive maintenance in these guidelines. Maintenance methodologies are classified follows.

• Repair : methodology to reinforce degraded parts by some methods or prevent progress of degradation of without replacement of the existing structure when the degradation of structure is actualized.

• Replacement : methodology to replace the existing structure with new one when the degradation of structure is actualized.

• Preventive maintenance : methodology to mitigate the damaged condition.

When the maintenance methodologies are implemented in the actual plant, we have to consider the feedback of the inspection program and plant life management.