Maintenance Technologies for
SCC which Support Stable
Operations of Pressurized
Water Reactor Power Plants
KOJI OKIMURA* 1NOBUYUKI HORI*1

MASAYUKI MUKAI*1

KOICHIRO MASUMOTO*1

KAZUHIKO KAMO* 2

MASAAKI KUROKAWA*2

Of the twenty-three pressurized water reactor (PWR) plants currently operating in Japan, the older ones are over
30 years old, and damage to aged plants, such as stress corrosion cracking (SCC), has been observed. This paper
describes countermeasures against SCC and introduces the latest maintenance techniques for ensuring the safe andsteady operation of PWR plants, such as inspection techniques to confirm the integrity of components, mitigation(degradation reducing) techniques to prevent the occurrence of damage, and repair and replacement techniques forany damage detected.
In order to prevent damage and leakage and to ensure the long-term steady operation of plants, we havedeveloped the inspection and maintenance techniques
for reducing the degradation shown in Fig. 1.

1. Introduction
It is important to prevent various problems arisingfrom damage and/or leakage in nuclear power plants.
This involves diagnosing the life remaining, reducing
degradation and repairing components such as pressure
vessels and piping. This is a major problems for agedplants especially.
One conspicuous problem seen in aged plants is theSCC (stress corrosion cracking) issue, which is considered to be jointly caused by the environment, thematerials and stress, as shown in Fig. 1
1. The damage
caused by SCC to high nickel based alloy (Alloy 600 below) in pressurized water reactor (PWR) plants has
become conspicuous both in Japan and abroad. Figure 2
shows the regions where Alloy 600 is used in conventional PWR plants. The stress corrosion cracking of Alloy
600 in PWR plants is called PWSCC (primary waterSCC), and it is considered that PWSCC will occur whenparts are used under high residual stress caused bymachining, welding and so on.

Having susceptibility of SCC
Material improvement(replacement, cladding, etc.)

MaterialSCC
Environment

High-temperature environment
in primary water
Reduction of environmental
temperature

StressTensile residual stress
Improved tocompressive stressFig. 1 Three factors causing stress corrosion cracking andmitigation methods

Pressurizer nozzle to safe end joints

Reactor vessel head penetration nozzle
Reactor vessel head penetration nozzle joint

Reactor vessel outlet nozzle safe end joint
Steam generator tube
Reactor vessel inlet nozzle safe end joint

Steam generator inlet
nozzle safe end joint

Reactor vessel bottom mounted instrument
Reactor vessel bottom mounted instrument joint

Steam generator outlet
nozzle safe end joint

Fig. 2 Region used Alloy 600 in primary system of PWR plants in Japan

Mitsubishi Heavy Industries, Ltd.
Technical Review Vol. 43 No. 4 (Dec. 2006)*1 Kobe Shipyard & Machinery Works*2 Takasago Research & Development Center, Technical Headquarters

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with complex shapes and materials. Surface inspection
is considered very effective for detecting PWSCC where
it is caused by the surface contacting water. Accordingly,
we developed and verified the ECT techniques with remote control and automatic search. Figure 4 shows anexample of the inspection technique for the reactor vessel inlet/outlet nozzle safe end joints (Alloy 600), whereeven extremely tiny flaws in welding can be detected byECT.
Further, by allowing the installation of multiple inspection heads, different inspection methods can be
simultaneously applied, for example, when a defect isdetected by ECT, defect quantification (confirmation offlaw depth) can also be carried out by using a UT probe
capable of measuring defect depth.
Further, improved detectability is also required for
confirming the integrity of components before applying
preventative maintenance techniques (for example,stress improvement techniques). The synergistic effectof securing integrity by inspections prior to stress improvement techniques, will suppress damage byreducing degradation, which is considered to furtherimprove reliability.

2. Inspection Technique
The inspection techniques require higher defect detection and defect quantification accuracy, in order toensure the reliability of components. In the steam generator tube (composed of about 4,000 small tube bundles
with an inner diameter of approximately 20 mm) shownin Fig. 3
3, the intelligent ECT (eddy current test), which
are densely arranged coils using a multi-induction system for higher noise reduction, allow precise inspection.
This technique was developed and applied in order toshorten the inspection time and improve the accuracy of
the inspection of shallow defects.
Further, the phased array UT (ultrasonic test) has
been developed for pipe welding. In this method, the direction and focal position of an ultrasonic-wave beam
can be freely changed by the UT probe, which is composed of multiple piezoelectric composites, which
improves the accuracy of the quantification of shallow
defects and defects in regions with complex shapes, whichwere conventionally considered difficult regions.
Alloy 600 is used for the reactor vessel nozzle safe
end joints (connecting nozzle and piping) and other joints

SCCSCC

Axial
direction

2 coils for defectdetection

Thin-film array coil

Tube

Inclined drive coil
Realization of array coil for enhanced
delectability and high-speed detection

Circumferential
direction
Tube expansion
transition

Fig. 3 The steam generator tube inspection technique of Intelligent ECT

Joint
Outer
(Alloy 600 weld metal)
surfaceNozzle(low-alloy steel)Innersurface

Reactorvesselside

Stainless steel cladding

Defectsignal

Alloy 600 weld metal
Example of ECT detection

Capable of mounting
multiple inspection
heads, enabling diverse
inspections includingECT, UT, etc.

Fig. 4 Inspection technique for Alloy 600 welding
Example of the reactor vessel outlet/inlet nozzle safe end joints

Mitsubishi Heavy Industries, Ltd.
Technical Review Vol. 43 No. 4 (Dec. 2006)2

disposed shot material.
On the other hand, when access to the inner surfaceof piping is difficult, it is necessary to relieve the tensilestress using an outer surface approach. One such technique, the outer surface irradiated laser stress
improvement process, which we call L-SIP, is the use of
the thermal expansion strain caused by the temperature
difference between the surface walls generated by thehigh speed heating of the outer surface by a high power
laser beam moving along the outer surface. Since highspeed heating is possible through laser irradiation, themethod can be used easily and simply by selecting theappropriate conditions without requiring water coolingof the inner surface. Figure 6 shows an example of theapplication of this method to the nozzle joints, where it
is confirmed that the high tensile residual stress at andaround the welded region on the inner surface has beenrelieved into compressive stress.

3. Stress Improvement Technique
As a countermeasure against high residual stress, shot
peening is used for components installed in air and waterjet peening is used for components installed in water, inorder to change the residual tensile stress into compressive stress of surfaces contacting the primary water.
Figure 5 shows an outline of these peening techniques. It
has been confirmed that the stress near the inner surface
can be made into compressive stress by the plastic strain
applied to the surface by using cavitation shock pressureduring the water jet peening and by using the collision
force of the shot material during the shot peening. For
the shot peening, we apply a new method employing ultrasonic-wave vibration. Using a high power piezo-electricelement as the drive source, we use shot material with alarger particle size. This technique will deeply relieve thetensile stress at the surface and allow easy control of theWater jet peening

Ultrasonic shot peening
Shot material drive source(piezo-electric element)

Principle and feature

Water jet with cavitations
bubbles

Shot material with larger
particle sizePlastic strain applied to thesurface by shock pressureof cavitations collapse

Plastic strain applied to the surface
by shot collision force

Material surface relieved into compressive stressdue to cavitation impact caused by water jet

Material surface relieved into compressive stress byshot material collision force

Example of the water jet peening(the reactor vessel bottom mounted instruments mock-up)

Water jet nozzle

Water jet withcavitations
cavitation bubble
bubbles

The surface and the surrounding relieved into
compressive stress at tensile residual stress field

Fig. 5 Stress improvement technique by peening

Temp

Laser oscillator
Duringheating

Stress+

Temp

Principle and feature

AftercoolingTemperature
distribution+

Stressdistribution

When the outer surface is subjected to high-speed heating,temperature distribution occurs in the sheet metal, resulting
in thermal expansion strain, causing the inner surface toundergo compressive stress during the process of cooling.
A high energy laser is used as the heat source and as it ismovable, high-speed heating as well as suppression oftemperature rise on the inner surface can be obtained at
the same time.

Temperature at inner and outer surfaces of piping during heating (analysis)
Optical head
Example of the outer surface irradiated laser stress improvement process

The inner surface can undergo compressive stresseven in a tensile residual stress field (welding residual stress).

Fig. 6 Outer surface irradiated laser stress improvement technique

Mitsubishi Heavy Industries, Ltd.
Technical Review Vol. 43 No. 4 (Dec. 2006)

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4. Improved Material and Cladding Techniques

5. Replacement with improved material

When a surface touches the primary system feed water a countermeasure employing high nickel based alloy
TT690 (Alloy 690), which has excellent PWSCC resistance,
is used. Alloy 690 has double the chromium (Cr) contentof Alloy 600. It is used for cladding, and prevents the Alloy 600 from coming in contact with the water by overlay
welding its surface (with Alloy 690) or by replacement,
which is described in the next section.
As for an example of the application of the degradation technique to the reactor vessel outlet nozzle safe end
joints (using welded Alloy 600), the cladding techniqueon the reactor vessel nozzle welded region is shown inFig. 7
7. In order to ensure a flow passage and to allow inspection, a groove is machined at surface before cladding(overlay welding). An automatic 3-layer, temper-beadwelding process, not requiring post-weld heat treatment,
has been developed and applied to actual reactors. The
cladding repair technique has also been applied to reactor vessel nozzle head penetration joints.

When components such as the reactor vessel head, thesteam generator, and the reactor internals (core internals)
are replaced, the new components are made with material with excellent SCC resistance. As a recent exampleof component replacement, the replacement of the coreinternals is shown in Fig. 88. Since a PWR's core internals
are designed to be removed, they can be removed as aunit, stored and preserved in a container, and new reactor internals installed. The replacement of the coreinternals has been carried out for the first time in theworld, being done in three PWR plants in Japan.
A partial replacement technique (replacement of nozzle
safe end joint) is also used when total replacement is difficult and to repair damage and for preventative
maintenance. An example of this is shown in Fig. 99, thereplacement of a reactor vessel outlet nozzle safe end joint.
In this method, the welding is done using Alloy 690 welding material with excellent SCC resistance when Alloy
600 joints are removed and replaced with new ones.

Outline
Inner surface (Alloy 600) is clad (overlay welding) with Alloy 690
on preventing the occurrence of SCC.
In order to secure a flow passage equivalent to the conventional
one and for inspectability, clad welding is performed at the channelregion. In order to protect the low alloy steel from the heat, 3-layer
temper-bead welding is performed.

Cladding with Alloy 690
Nozzle(low alloy steel)

Safe end(stainless steel)

Welding (Alloy 600)
Image of on-site work

View of automatic welding on
the nozzle inner surface

Fig. 7 Cladding technique: Example of the reactor vessel outlet nozzle safe end joint(3) Carrying out through theequipment hatch(1) Lifting and storage of the oldcore internals
Ceiling (Polar) craneTemporary floor

Temporary lifting
equipment
Outer shield
Containment vesselContainer for preserving
the old core internals
Reversing pedestal(2) Reversing and horizontal
removal of the old core internals

View of installing thenew core internals tothe reactor vessel

Fig. 8 component replacement technique: Example of the reactor core internals

Mitsubishi Heavy Industries, Ltd.
Technical Review Vol. 43 No. 4 (Dec. 2006)

4Outline
After removing Alloy 600 joint, new member(spool piece) is welded using Alloy 690 withexcellent SCC resistance.
Special welding process (temper-bead weldingor local heat treatment of narrow-region) isused to prevent the low-alloy steel from theheat effects of welding.
Welding step 1: Butte ring
Alloy 600 joint

Nozzle(low-alloy steel)

Welding step 2: Joint welding

Main coolant pipe
(stainless steel)Fig. 9 Partial replacement technique: Example of the reactor vessel outlet nozzle safe end joint

Removal of defect(drilled repair)

Installation of cap for new boundary(cap repairing)If size of the defect is within the permissible
range of the component structure, the bottom
mounted instrument is drilled for removing
the defect by enlarging the inner diameter.
Enlargement of
inner diameter
Original
innerdiameter

Initial(inner diameter 10mm)

The cap for new boundary is installedoutside the bottom mounted instruments
to prevent the leakage from component
by retaining the leakage from the defectinside the cap.

The bottom
mountedinstruments

AssumeddefectThe bottom
mountedinstruments

The reactor
vessel lower
headCap

After drilling(inner diameter 16mm)

Fig. 10 Repair technique: Example of the reactor vessel bottom mounted instrument

6. Repair Technique

7. Conclusion

When damage is detected and repair is required, it ispossible to remove the defective part and install a new
boundary component. Figure 10 shows an example of therepair of reactor vessel bottom mounted instruments. The
removal of defects is a technique of partial removal when
the component is structurally safe after removing thedefect. The installation of a component for the newboundary is installed outside the damaged component
to prevent leakages from the defect.

In this paper, we have explained the latest maintenance techniques including inspection, mitigation (stressimprovement and material improvement), replacement
and repair.
We consider that these techniques contribute to thesafety and reliability of nuclear power plants for the continuous economic efficiency of plant operation over a longtime period. In the future, we will continue to developand verify these techniques.

Koji Okimura

Nobuyuki Hori

Masayuki Mukai

Koichiro Masumoto

Kazuhiko Kamo

Masaaki Kurokawa

Mitsubishi Heavy Industries, Ltd.
Technical Review Vol. 43 No. 4 (Dec. 2006)5