visual inspection of reactor vessel head penetration nozzles

International Conference

Nuclear Energy for New Europe 2003 Portorož, Slovenia, September 8-11, 2003

http://www.drustvo-js.si/port2003

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Visual Inspection of Reactor Vessel Head Penetration Nozzles

Simon Farley, Thomas Weiss Westinghouse Reaktor GmbH

Dudenstrasse 44, 68167 Mannheim, Germany [email protected]

ABSTRACT

Since the detection of a through wall axial crack in 1991 in a reactor vessel head penetration (VHP) in a French nuclear power plant, inspections of control rod drive mechanism (CRDM) and thermocouple nozzles have been performed at plants throughout the world to early detect and reliably size such defects.

One function of VHP nozzles is to maintain the reactor coolant system pressure boundary. Cracking of VHP nozzles and welds represents a degradation of the reactor coolant system boundary, and hence, is potentially safety significant.

Past inspection was initially restricted to the inner surface of VHP nozzles and the occurrence of axial cracking was deemed to be of limited safety concern by the United States Nuclear Regulatory Commission (NRC). However, recent discoveries of cracked and leaking Alloy 600 VHP nozzles in the USA, which include circumferential cracking and cracking of the J-groove weld, has raised concerns about the potential safety implications and prevalence of cracking in VHP nozzles and has lead to more extensive inspection of the nozzles and the surrounding area.

In 1992 Westinghouse developed a complete system to non-destructively inspect the inner surface of VHP nozzles using the eddy current testing (ET) and the ultrasonic testing (UT) techniques. Since then our inspection methods have continually improved and expanded to include the visual inspection (VT) of the nozzle inner and outer surfaces and the ET and UT of the J-groove weld.

The latest success in our product range has been the optimisation of our VT-gap probes, which, like the ET and UT probes, enter into the tight gap (approx. 3,25 mm) between thermal sleeve and the inner surface of the VHP nozzle. With these probes it is now possible to visually inspect cracks of down to 6 micrometers.

This presentation will give a general overview of VHP nozzle inspection, concentrating on the latest development of VT-gap capabilities.

1 INTRODUCTION

Control rod drive shafts pass through reactor vessel head penetration nozzles, which sit at the top of a reactor pressure vessel (RPV) head. CRDM’s control the movement of control rods in and out of a reactor core. In Figure 1 a typical configuration is shown as used in most of the Pressurized Water Reactors (PWR’s) worldwide. The Alloy 600 VHP nozzle is fitted into the ferritic steel vessel head. The join between RPV head and VHP nozzle for mechanical fixture and sealing of the pressure boundary is achieved by the so called J-Groove weld with the Alloy 600 equivalent filler material Alloy 82 or Alloy 182.

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Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 8-11, 2003

Figure 1: Sleeved CRDM Penetration

Due to its optimal thermal expansion behaviour, good corrosion resistance and good welding ability on stainless, nickel based and carbon steels, Alloy 600 became the material of choice for the VHP nozzles during plant design. However, industrial experience has shown that Alloy 600 is susceptible to primary water stress corrosion cracking (PWSCC).

In 1991 a through wall crack, resulting in a leak to the top of the reactor vessel, was detected by an acoustic leak monitoring system during the 10th year hydro pressure test at the French nuclear power plant (NPP) Bugey 3.

Until May 1992 no remote inspection equipment was available with sufficient sensitivity to detect cracks in VHP nozzles. This was due to the specific design of the component. As can be seen in Figure 1 thermal sleeves limit the access to the inside surface of the VHP nozzle. The radial gap between the VHP nozzle and the thermal sleeve is usually only about 3 mm. Due to the impossibility of entering this gap with a suitable inspection probe the thermal sleeves had to be removed to access the potentially cracked area.

In February 1992 Westinghouse Electric Germany began developing a remote operated manipulator and inspection techniques that allowed cracks in all VHP nozzles to be detected and sized without the necessity of dismantling the CRDM’s and thermal sleeves. Thus avoiding the accompanying plant shut down time and exposure of personnel to excessive radiation. The first remote inspection with the system was performed early May 1992. Since that time the system has been continually improved to increase the inspection capability and minimize the inspection time and radiation dose of the inspection personnel.

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More than 6000 penetrations were inspected world wide (during more than 130 inspections) between 1991 and 2000, during which no additional leaks were found. Approximately four percent of the inspected penetrations showed crack indications.

Until certain incidents in the USA the occurrence of axial cracking was deemed to be of limited safety concern and did not warrant inspection by the United States Nuclear Regulatory Commission (NRC) and past inspection at non U.S. sites were initially restricted to the inner surface of VHP nozzles.

However, in early 2001 four new leaks were identified in the U.S. plants Oconee 1, 2, 3 and ANO 1 as a result of cracking in the J-groove weld and VHP nozzle. Thereby, the circumferential cracking for the first time found at Oconee 3 led to safety concerns by the NRC. These findings resulted in an additional request by the NRC for inspections during the coming outages [1]. The youngest and most severe event was seen at the NPP Davis-Besse in March 2002. Significant wastage of the vessel head material resulted in a serious degradation of its structural integrity and the potential for a Loss of Reactor Coolant Accident (LOCA) [2].

As a result of these incidents an extensive inspection programme was initiated to inspect the inner surface of VHP nozzles throughout the United States and a tool was developed to inspect the J-groove weld.

2 SAFETY ASSESSMENT

The recent identification of cracking indicate that circumferential cracks above the J-groove weld can occur, in contrast to an earlier conclusion that the cracks would be predominantly axial in orientation and that cracking of the J-groove weld metal can precede cracking of the base metal. Therefore a revised susceptibility model taking into account the above mentioned items had to be established.

In addition, the presence of circumferential cracking where only a small amount of boric acid residue indicated a problem, raised questions on the adequacy of current visual examinations. For boric acid deposits from VHP nozzle cracks to be detectable at the outer surface of the RPV head, sufficient reactor coolant has to leak through the primary pressure boundary into the annulus between the VHP nozzle and the RPV head base metal, propagate up the annulus and finally emerge onto the outer surface of the RPV head. Since PWSCC cracks in Alloy 600 and Alloy 182 welds are very tight, leakage from these is expected to be small.

The EPRI Materials Reliability Program (MRP) offered an approach by using an assessment of the relative susceptibility of each PWR to OD-initiated or weld PWSCC based on the operating time and temperature of the penetrations. Based on this simplified model, each plant was ranked by the MRP according to the operating time in Effective Full Power Years (EFPY) required for the plant to reach an effective time-at-temperature equivalent to Oconee 3 at the time the above-weld circumferential cracks were identified in early 2001. Based on the experiences at Oconee it was recommended that plants ranked within 10 EFPY should perform a visual inspection of the RPV top head capable of detecting small amounts of leakage.

Although the industry susceptibility ranking model has limitations, such as large uncertainties and no predictive capability, the model does according to NRC provide a starting point for assessing the potential for vessel head penetration nozzle cracking in PWR plants.

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3 INSPECTION CAPABILITIES AND TECHNIQUES

Since the first inspections in 1992 several non-destructive examination techniques have been developed to inspect VHP nozzles and J-Groove welds. These encompass ET, UT, VT and dye penetrant testing (PT). This chapter gives a brief summary of the main methods of inspection, applied by Westinghouse.

3.1 Top of the Head Visual Inspection

Boric acid deposits on the top of the reactor vessel head may be indicative of primary coolant leakage through the penetration wall or the J-groove weld from the underside of the reactor vessel. Figure 2 shows a typical example for boric acid deposits.

Figure 2: Boron Accumulation on Vessel Head

Westinghouse has remote tooling and inspection technology to perform “under the insulation” visual inspections to identify evidence of leakage in the form of boric acid deposits. The system consists of a remotely controlled delivery vehicle, high resolution cameras, video monitors and screen writing capability to perform and document these visual inspections. Figure 3 shows the BTRIS manipulator on a reactor vessel head mock-up.

Figure 3: BTRIS Delivery Vehicle on a Reactor Vessel Head

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3.2 Inspection of the VHP Inner Surface

A range of slim sword like probes (gap probes) has been developed to inspect the inner surface of VHP nozzles that are equipped with thermal sleeves. These include a differential ET probe with two pancake coils used for crack detection and a TOFD (Time of Flight Defraction) UT probe for subsequent crack depth sizing. If deposits cover the surface, the surface can be cleaned before starting the inspection using specially developed sword type cleaning probes. In case the roughness or geometry of the surface (for example, if the penetration has been repaired) does not allow an ET inspection, Westinghouse has developed and qualified a visual gap probe, which will be described in chapter 4.

The Gap Scanner inspection tool, shown in Figure 4, is specifically designed to guide eddy current, ultrasonic and visual inspection probes into the annulus between the inner surface of the VHP nozzle and the thermal sleeve and to manipulate the probes so as to scan the surface area to be inspected.

Figure 4: Gap scanner for Inspections of Penetrations with Thermal Sleeves

For penetrations without thermal sleeves, or open housings, the Gap scanner is replaced with the open housing scanner inspection tool. The open housing scanner delivers an eddy current probe and an array of ultrasonic probes to provide simultaneous inspections of the inside diameter surface of the VHP nozzle. The scanner can be used to provide circumferential or axial scanning motion for specific applications.

Figure 5 shows a typical inspection scenario. If any detected cracks exceed defined criteria an engineering justification is conducted to justify continuous operation or the penetration is repaired. Westinghouse has developed a range of repair capabilities, which are not covered by this paper.

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VT InspectionECT Inspection

Penetration Cleaning(if needed)

PT Inspection

UT Sizing Future Inspections

Cracks No Cracks

Cracks exceeding Criteria Small Cracks

Future Inspections

Engineering Justification for

Continuous Operation

Repair

Figure 5: VHP Nozzle Inner Surface Inspection Scenario

3.3 Inspection of the J-Groove Weld

A probe manipulation tool, called Groove Man, was specifically developed to perform eddy current examination of the Alloy 182 J-Groove weld and/or the outside surface of the Alloy 600 VHP nozzle. The purpose for performing this inspection is to determine if surface flaws exist in the J-Groove welds or nozzle OD and to characterize the indications as axial or circumferential. The tool is designed to conform to the geometry of the J-Groove weld to allow the eddy current probe to follow the contour of the assembly. The Groove Man is shown in Figure 6.

Figure 6: Groove Man Inspection Tool In addition to ET the J-groove weld is also inspected using PT and VT inspection

techniques.

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4 VISUAL INSPECTION OF THE VHP NOZZLE SURFACE

In 1995 a probe was developed and qualified to visually inspect the inner surface of VHP nozzles, equipped with a thermal sleeve. Until recently these probes have been used to inspect the penetrations in which a known defect had been removed by machining (turning) the inner surface. The subsequent abrupt change in diameter of the inner surface made it impossible to inspect the transition area using the eddy current detection method because of probe “lift off”, caused when passing from the smaller to the larger diameter of the inner surface.

Recent years have seen the replacement of most reactor vessel heads in France and the decision was taken by EDF to qualify an upgraded VT-probe with a higher resolution as its predecessor. The aim is to use the VT-probe to verify ambiguous ET signals so as to distinguish between scratches and cracks and hence to avoid unnecessary UT sizing of indications that aren’t cracks. As with the ET and UT probes the Gap Scanner is used to manipulate the optical probe.

As shown in Figure 7 the probe is equipped with a bundle of 30 000 optic fibres capped at the end with a 90° prism to transmit the optical image to a CCD camera, mounted on the Gap scanner. The surface of the penetration is illuminated by 2 separate plastic fibres that transmit light from the light source, located outside the vessel head, to the head of the probe.

Figure 7: VT Inspection Probe

Both the optic fibre bundle and the light transmitting fibres are packed into a metal

ribbon (band) that delivers the probe head to the object of inspection. Like the ET & UT probes, the VT probe has been designed to penetrate the tight annulus between the thermal sleeve and the inner surface of the VHP nozzle.

Laboratory tests have shown that cracks of down to 6 micrometers can be inspected using the VT-probe.

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5 VESSEL HEAD INSPECTION MANIPULATOR

RPV head penetration inspection tools are remotely delivered by the Westinghouse DERI 700 manipulator. The DERI 700 is a multi-axis, remotely operated robot that can access all nozzles without repositioning. More importantly, the DERI 700 manipulator provides a common delivery system for not only inspection equipment but also for repair tooling.

Figure 8: DERI 700 Manipulator under RVH Mockup

The manipulator is comprised of a rotating central leg and vertically moving arm, with elbow and wrist. Inspection tools are mounted on a tilt unit allowing them to be remotely passed through the man way of the RPV head stand. This allows inspection probe and tool changes to be made outside the RPV head, resulting in lower dose rates. Cameras underneath the head allow the operators to verify tool position and monitor all probe delivery and inspection activities. When used for inspection purposes the DERI 700 manipulator has the added advantage that it can be dismantled and retracted from the inspection stand without requiring the reactor vessel head to be moved.

REFERENCES

[1] NRC Bulletin 2001-01, Circumferential cracking of reactor pressure vessel head penetration nozzles, United States Nuclear Regulatory Commission, Office of Nuclear Regulation, Washington, D. C. 20555-0001, August 3, 2001.

[2] NRC Bulletin 2002-01, Reactor pressure vessel head degradation and reactor coolant pressure boundary integrity”, United States Nuclear Regulatory Commission, Office of Nuclear Regulation, Washington, D. C., March 18, 2002.

visual inspection of reactor vessel head penetration nozzles

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