Industry’s Efforts toward Technology Development Related to Aging Management of PWR Plants

The term, “aging plants”, which refers to nuclear power plants operating more than 30 years, has become popular among local residents around nuclear power plants as well as among the media. The term “aging” was first used in the title of the document issued by Agency of Natural Resources and Energy in 1996, “The Basic Concept of Aging Management”.

In addressing aging degradation, it was a matter of question which measure of time (date, month or year) should be used as a unit to consider an events. Although the aging plant was defined as a plant operating more than 30 years, the mode and timing of aging degradation depend on the environment and service conditions under which a component has been operating. Previous efforts made by related parties have contributed to the prevention of potential aging events which may challenge the reactor safety. Owing to such efforts, causes of some events attributable to defective design have been clarified and preventive maintenance measures, including the replacement and mitigation, have been taken accordingly. As a result, the number of events resulting from such aging phenomena has been reduced. On the other hand, the events caused by aging phenomena attributable to manufacturing processes (welding, surface treatment, etc.), which hardly emerge, are slightly increasing. There have been many cases that a shorter weld line due to lack of penetration in the narrow gap has been led to fatigue cracking. More recently, incidents of stress corrosion cracking due to work hardening and local tensile stresses on the surface have been observed.

It should be noted that an effort to analyze not only the phenomena but also the mechanisms of the above events to clarify the root causes can improve the quality of preventive maintenance by means of rolling-out the analytical results to relevant plants.
This paper introduces valuable experiences with the application of the results of technology development regarding to aging degradation of operating plants from the viewpoints of R&D and reactor operators.

2. Electric power companies' efforts to pursue aging-related R&D and apply the results to operating plants

2.1 History of aging-related R&D

2.1.1 History of Japan’s PWR (Fig.1)

2.1.2 Past Achievements and Future Prospects for Ageing-related Research Activities (Fig.2)

2.1.3 Outline of Next Generation LWR Development

2.1.4 Industrial Efforts to Promote PLM-related R&D Activities (Fig.3)

2.1.5 Studies Performed by 10 Sub Groups

2.1.6 The History of PLM/AM at Japanese NPPs (Fig.4)

2.1.7 Replacement of Major Components at Mihama#3 (Fig.5)

Several examples of difficulties we have faced in applying the R&D results to operating plants are introduces in this article.

Following incidents of SCC in low pressure turbine rotors and discs at nuclear power plants in the USA and in Japan, partial repair and replacement of concerned sections had been conducted. Then, considering the results of R&D, it was determined to adopt the integrated rotor and disc designs to improve the SCC resistance. Similar to the surveillance tests for the reactor vessel structural materials, accelerated aging tests were conducted by placing test pieces made of same material as actual turbine parts into the casing of a steam turbine. From the test results, the residual life of turbine parts was evaluated.

The next subject of aging was irradiation assisted stress corrosion cracking (IASCC) in baffle former bolts. The research project on this phenomenon was initiated before any incident attributable to this phenomenon happened. Since the discovery of IASCC in the baffle former bolts at Bugey, France 15 years ago, research projects have been promoted and an international network has been established.

The last one is primary water stress corrosion cracking (PWSCC). Since the discovery of PWSCC in the rolled expansion zone of steam generator tube at a PWR plant, which was called “Achilles’ Heel”, incidents of PWSCC have been observed in the reactor vessel penetration nozzles at Bugey-3, in France where the inside diameter was expanded by machining, and then at Ohi-3 in the RV head nozzle welds, which had been subject to grinder polishing. These PWSCC were supposed to be caused by residual stresses on the component surface. Recently, PWSCC has been found in pipe nozzles of replaced steam generators. All these cracks occurred in nickel-based alloy components. However, it should be noted that in some cases, cracking in stainless steel material has propagated to an extended length. There is growing expectation for new findings to be obtained from the future R&D.

2.2.1 Application of Developed Technologies

2.2.2 Inspection of Steam turbines (Fig.6)

2.2.3 Case of deterioration : Baffle former bolts (Fig.8)

Technical evaluation : Degradation modes of Core Internal(PWR) (Fig.9)

Technical Evaluation of IASCC for Ohi#1(1/2) (Fig.10)

Technical Evaluation of IASCC for Ohi#1 (2/2) (Fig.11)

Integrity Evaluation of BFB - Multiple Ageing Phenomena - (Fig.12)

2.2.4 Main Components Using Alloy 600 in J-PWR (Fig13)

The History of Trouble and PM of RV-Head (Fig.14)

Case of deterioration : RV-haed penetration (1)(Fig.15)

Case of deterioration : RV-head penetration (2)(Fig.16)

Experiences in Japan

1. Discovery of Cracking

2. Results of Inspections at Other NPPs

Plant	Operating hours	Loops	Indications by ECT	UT results (length shows ECT results)
Kansai Mihama - 2*	About 93,000	2	A-inlet nozzle weld: 13	Max. length : about 17mm Max. depth : about 13mm
JAPC Tsuruga - 2**	About 151,000	4	A-inlet nozzle weld: 1	Detected (the depth and length sizing was impossible although a significant indication was confirmed.)
			B-inlet nozzle weld: 5	Max. length : about 21mm, Max. depth : about 12mm (an another crack having a similar size was also confirmed)
			C-inlet nozzle weld: 23	Max. length : about 14mm, Max. depth : about 13mm (cracks were also found at other 6 locations)
Kansai Takahama-2	About 100,000	3	A-inlet nozzle weld:3	Not detected
			B-inlet nozzle weld:2	Max. length: about 7mm, Max. depth: about 6mm
			C-inlet nozzle weld:4	Max. length: about 14mm, Max. depth: about 8mm
Kansai Takahama-3	About 175,000	3	A-inlet nozzle weld:7	Max. length : about 28mm, Max. depth : about 9mm
			B-inlet nozzle weld:16	Max. length : about 38mm, Max. depth : about 15mm
			C-inlet nozzle weld:9	Max. length : about 12mm, Max. depth : about 9mm
Kyushu Genkai-1	About 96,000	2	A-inlet nozzle weld:3	Not detected
Hokkaido Tomari-2	About 131,000	2	A-inlet nozzle weld:3	Max. length : about 13mm, Max. depth : about 7mm
			B-inlet nozzle weld:10	Max. length : about 10mm, Max. depth : about 5mm
Detailed investigation was performed at the representative plants * Investigation by cutting the pipe section ** Tests using boat samples

3. Detailed Investigation (Fig.18)

4. Suspected Causes (Fig.19)

5. Corrective Actions to Addresss PWSCC

Mihama-2

Tsuruga-2, Takahama-2, Takahama-3, Tomari-2

Genkai-1

2.2.5 Non-Destructive Testing Technique
- Stream Generator Nozzle Safe End Weld - (Fig.21)

2.2.6 Preventive Maintenace (Water Jet Peening) (Fig.22)

2.2.7 Material Improvement - Alloy 690 cladding

2.2.8 PWSCC in Stainless Steel Components
- Fukui Prefecture Ageing Management Investigation Committee in FY 2006

Background and Achievements

(Background)
At BWRs, the initiation and development of SCC have been reported even in stainless steel components, such as the core shroud and recirculation pipe, which are subjected to hard cold work (machining in room temperature). On the other hand, under the PWR environment, no SCC in stainless steel components has been observed. (Fig.24)

A verification program regarding SCC under the PWR environment should be included in the ageing management activities.

(Achievements obtained so far)
- Laboratory experiments revealed that cracks caused by SCC would develop even under the PWR environment if cold work was added.
- The crack growth rate in PWR increases as the yield stress (degree of cold work) increases like BWR.

Issues to be addressed
1. Clarification of SCC mechanism in cold worked stainless steel parts
2. Identification of conditions simulating an actual unit and data sampling for tests and evaluation
3. Investigation of SCC prevention measures

3.1 Application of Lessons Learned from Past Events/Failures (Fig.25)

3.2 Results of Ageing Management Study and Total Management System (Fig.26)

3.3 Preparedness for the maintenance activities (viewpoint of information) (Fig.27)

3.4 Conclusion

1) It is necessary to fully understand the actual status in the field if insights cannot be obtained from laboratory tests.

2) Confirmation of the consistency between the current knowledge and actual status versus time is the key to clarifying the cause of an event.
If cause investigation has not be fully conducted, a similar event occurs in the field.
It is necessary to clarify the degradation mechanism in order to complete the cause investigation.

3) For the purpose of the Plant Life Management, maintenance activities are conducted based on an assumed operation period (timing of rebuilding the plant). Therefore, it is likely to presume that future trouble can be prevented by replacing the components which are expected to experience trouble in the near future considering the operation records achieved in a half of the assumed plant life.
However, the idea of adopting the unconditional replacement may be hazardous because the replacement work conditions may affect on degradation. Therefore, thorough consideration is needed before determining the actions.