Information Notice 2001-05, Through-Wall Circumferential Cracking of Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzles at Oconee Nuclear Station Unit 3

Through-Wall Circumferential Cracking of Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzles at Oconee Nuclear Station Unit 3

ML011160588
Person / Time
Issue date:	04/30/2001
From:	Marsh L Operational Experience and Non-Power Reactors Branch
To:
References
IN-01-005
Download: ML011160588 (8)

(1)

Axial flaws are flaws that propagate along the inside or outside diameter length of the

CRDM nozzle. Below-the-weld circumferential indications are apparent flaws oriented

around the circumference of the nozzle, beneath the RPV head and below the area where

the nozzle is welded to the RPV head. A recordable indication is one that exceeds the NDE

acceptance criteria.

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D. C. 20555-0001

April 30, 2001

NRC INFORMATION NOTICE 2001-05: THROUGH-WALL CIRCUMFERENTIAL CRACKING

OF REACTOR PRESSURE VESSEL HEAD

CONTROL ROD DRIVE MECHANISM PENETRATION

Addressees

All holders of operating licenses for pressurized water nuclear power reactors except those who

have ceased operations and have certified that fuel has been permanently removed from the

reactor vessel.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert

addressees to the recent detection of through-wall circumferential cracks in two of the control

rod drive mechanism (CRDM) penetration nozzles and weldments at the Oconee Nuclear

Station, Unit 3 (ONS3).

It is expected that recipients will review the information for applicability to their facilities and

consider actions, as appropriate. However, suggestions contained in this information notice are

not NRC requirements; therefore, no specific actions or written response is required.

Description of Circumstances

On February 18, 2001, with ONS3 in Mode 5, Duke Energy Corporation (the licensee)

performed a visual examination (VT-2) of the outer surface of the units reactor pressure vessel

(RPV) head to inspect for indications of borated water leakage. This RPV head inspection was

performed as part of a normal surveillance during a planned maintenance outage. The VT-2 revealed the presence of small amounts of boric acid residue in the vicinity of nine of the

69 CRDM penetration nozzles (Figures 1 and 2). Subsequent nondestructive examinations

(NDEs) identified 47 recordable crack indications in these nine degraded CRDM penetration

nozzles. The licensee initially characterized these flaws as either axial or below-the-weld

circumferential indications, and initiated repairs of the degraded areas. NDEs of nine additional

CRDM penetration nozzles from the same heat of material were conducted for extent of

condition purposes, but did not detect recordable indications.(1)

Subsequent dye-penetrant testing (PT) revealed additional indications in two of the nine

degraded penetration nozzles. While affecting further repairs of these indications, the licensee

identified that each nozzle had significant circumferential cracks in the nozzle above the weld.

Further investigations and metallurgical examinations revealed that these cracks had initiated

from the outside diameter (OD) of the CRDM penetration nozzles. The circumferential crack in

the #56 CRDM nozzle was through-wall, and the #50 nozzle had pin hole through-wall

indications. These cracks followed the weld profile contour, and were nearly 165° in length.

The licensee stated that pre-repair ultrasonic testing (UT) examinations had identified

indications in these areas during the initial inspections, but these indications had been

misinterpreted as craze cracking with unusual characteristics. The characterization for these

two nozzle indications was revised after the initial post-repair PT examinations. The licensee

concluded that the root cause for the CRDM penetration nozzle cracking was primary water

stress corrosion cracking (PWSCC). This conclusion was based on metallurgical

examinations, crack location and orientation, and finite element analyses.

Discussion

The 69 CRDM nozzles at ONS3 are approximately 5 feet long and are J-groove welded to the

inner radius of the RPV head, with the lower end of each nozzle extending about 6 inches

below the inside of the RPV head (see Figure 2). The nozzles are constructed from 4-inch OD

Alloy 600 Inconel procured in accordance with the requirements of Specification SB-167,Section II to the 1965 Edition (including Addenda through Summer 1967) of the American

Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. During initial

construction, each nozzle was machined to final dimensions to assure a match between the

RPV head bore and the OD of the nozzle. The nozzles were shrink-fit by cooling to at least

minus 140 degrees F, inserted into the closure head penetration, and then allowed to warm to

room temperature (70 degrees F minimum). The CRDM nozzles were tack-welded and then

permanently welded to the closure head using 182-weld metal (see Figure 2). The shielded

manual metal arc welding process was used for both the tack weld and the J-groove weld.

During weld buildup, the weld was ground and PT inspected at each 9/32 inch of the weld.

The final weld surface was ground and PT inspected. The weld prep for installation of each

nozzle in the RPV head was accomplished by machining and buttering the J-groove with 182- weld metal.

Axial cracking in pressurized water reactor (PWR) CRDM nozzles has been previously

identified, evaluated, and repaired. Numerous small-bore Alloy 600 nozzles and pressurizer

heater sleeves have experienced leaks attributed to PWSCC. Generally, these components

are exposed to temperatures of 600 degrees F or higher and to primary water, as are the

ONS3 CRDM nozzles. However, circumferential cracks above the weld from the OD to the

inside diameter (ID) have not been previously identified in the U.S.

An action plan was implemented by the NRC staff in 1991 to address PWSCC of Alloy 600

vessel head penetrations (VHPs) at all U.S. PWRs. This action plan included a review of the

safety assessments by the PWR owners groups (Westinghouse Owners Group, Combustion Engineering Owners Group, and Babcock & Wilcox Owners Group) submitted for staff review

on June 16, 1993, by the Nuclear Management and Resource Council (NUMARC, now the

Nuclear Energy Institute [NEI]).

After reviewing the industry's safety assessments and examining the overseas inspection

findings, the NRC staff concluded, in a safety evaluation (SE) dated November 19, 1993, that

PWR CRDM nozzle and weld cracking was not an immediate safety concern. The bases for

this conclusion were that if PWSCC occurred (1) the cracks would be predominately axial in

orientation, (2) the axial cracks would result in detectable leakage before catastrophic failure, and (3) the leakage would be detected during visual examinations performed as part of

surveillance walkdown inspections before significant damage to the RPV head would occur.

However, the NRC staff noted concerns about potential circumferential cracking (which would

need to be addressed on a plant-specific basis), high residual stresses from initial manufacture

and from tube straightening sometimes done after welding, and the need for enhanced

leakage monitoring.

By letter dated March 5, 1996, NEI submitted a white paper entitled "Alloy 600 RPV Head

Penetration Primary Stress Corrosion Cracking," which reviewed the significance of PWSCC in

PWR VHPs, described how the PWR licensees were managing the issue. NEI assumed that

the issue was primarily an economic issue rather than a safety issue, and described an

economic decision tool to be used by PWR licensees to evaluate the probability of a VHP

developing a crack or a through-wall leak during a plant's lifetime. This information would then

be used by a PWR licensee to evaluate the need to conduct a VHP inspection at their plant.

To verify the conclusions in the industrys safety assessments, sampling inspections were

performed at three PWR units in 1994. The results of these domestic inspections were

consistent with the February 1993 analyses by the PWR owners groups, the staffs November

19, 1993, SE, and the PWSCC found in European reactors. On the basis of the results of the

first five inspections of U.S. PWRs, the PWR owners groups' analyses, and the European

experience, the NRC staff determined that it was probable that CRDM penetrations at other

plants contained similar axial cracks, but that such cracking did not pose an immediate- or

near-term safety concern. Further, the NRC staff recognized that the scope and timing of

inspections may vary for different plants, depending on their individual susceptibility to this

form of degradation. In the long term, however, the staff determined that degradation of the

CRDM and other RPV head penetrations is an important safety consideration because of the

possibility of (1) exceeding the ASME Code safety margins if the cracks are sufficiently deep

and continue to propagate during subsequent operating cycles and (2) eliminating a layer of

defense in depth for plant safety.

On April 1, 1997, NRC issued Generic Letter (GL) 97-01, Degradation of Control Rod Drive

Mechanism Nozzle and Other Vessel Closure Head Penetrations, which requested

addressees to inform the staff of their inspection activities related to VHPs. Based on the

industrys GL 97-01 response, which took credit for periodic inspections of the RPV head, the

staff agreed that the conclusions in its November 19, 1993, SE remained valid. The recent identification of significant circumferential cracking of two CRDM nozzles at ONS3 raises concerns about a potentially risk-significant generic condition affecting all domestic

PWRs. RPV head penetrations, including CRDM nozzles, provide the function of maintaining

the reactor coolant system (RCS) pressure boundary. Cracking of CRDM nozzles and welds

is a degradation of the primary RCS boundary. Industry experience has shown that Alloy 600

is susceptible to stress corrosion cracking (SCC). Further, the environment in the CRDM

housing annulus will likely be far more aggressive after any through-wall leakage, because

potentially highly concentrated borated primary water will become oxygenated, increasing

crack growth rates.

The repair activities at ONS3 were extensive. The licensee stated that all flaws would be

removed entirely from both weld material and nozzle base metal and repaired prior to plant

restart. The licensee plans to perform a thorough visual inspection of the Unit 2 RPV head

penetrations during the next outage and is investigating the eventual replacement of the RPV

heads on all three units to prevent recurrence of this event. Foreign PWRs in France and

Japan have already replaced a number of their RPV heads.

The NRC held a public meeting with the Electric Power Research Institute (EPRI) Materials

Reliability Project (MRP) personnel on April 12, 2001, to discuss CRDM nozzle circumferential

cracking issues. During the meeting, the industry representatives said that they were

developing a generic safety assessment, recommendations for revisions of near-term

inspections, and long-term inspection and flaw evaluation guidelines.

The ONS3 cracking reinforces the importance of examining the upper PWR RPV head area

(e.g., visual under-the-insulation examinations of the penetrations for evidence of borated

water leakage or volumetric examinations of the CRDM nozzles) and of using appropriate NDE

methods (e.g., UT, ET, PT, etc.) to adequately characterize cracks. Presently, licensees are

not required to remove RPV head insulation to visually inspect the head penetrations;

however, some licensees have recently performed expanded VT-2 examinations by using

cameras to inspect between the CRDM nozzles and the insulation.

The NRC has recently developed a Web page to keep the public informed of generic activities

on PWR Alloy 600 weld cracking (http://www.nrc.gov/NRC/REACTOR/MRP/index.html). The

NRC will update this Web page and assess the need for further generic action as new

information becomes available.

This information notice requires no specific action or written response. If you have any

questions about the information in this notice, please contact one of the technical contacts

listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager. Related Generic Communications

Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure

Boundary Components in PWR Plants, March 17, 1988

Generic Letter 97-01, Degradation of Control Rod Drive Mechanism Nozzle and Other

Vessel Closure Head Penetrations, April 1, 1997

Information Notice 90-10, Primary Water Stress Corrosion Cracking of INCONEL 600,

February 23, 1990

Information Notice 96-11, Ingress of Demineralizer Resins Increases Potential for

Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations, February

14, 1996

NUREG/CR-6245, Assessment of Pressurized Water Reactor Control Rod Drive

Mechanism Nozzle Cracking, October 1994

/RA/

Ledyard B. Marsh, Chief

Events Assessment, Generic Communications

and Non-Power Reactors Branch

Division of Regulatory Improvement Programs

Office of Nuclear Reactor Regulation

Technical contacts:

Ian Jung, NRR

James Medoff, NRR

301-415-1837

301-415-2715 E-mail: ixj@nrc.gov

E-mail: jxm@nrc.gov

C. E. (Gene) Carpenter, NRR

301-415-2169 E-mail: cec@nrc.gov

Attachments:

1. Figure 1: Oconee Reactor Pressure Vessel Head Map

2. Figure 2: Oconee CRDM Nozzle Penetration (Typical)

3. List of Recently Issued NRC Information Notices

ML011160588 Temple #=NRR-052

ÿ Publicly Available

ÿ Non-Publicly Available

ÿ Sensitive

ÿ Non-Sensitive

OFFICE

REXB

Tech Editor

EMCB

EMCB

REXB

C:REXB

NAME

IJung*

PKleene*

JMedoff*

CCarpenter*

JTappert/txk*1.

LMarsh

DATE

04/25/2001

04/25/2001

04/26/2001

04/25/2001

04/26/2001

04/30/2001

Figure 1: Oconee Reactor Pressure Vessel Head Map

Figure 2:Oconee CRDM Nozzle Penetration (Typical)

______________________________________________________________________________________

OL = Operating License

CP = Construction Permit

Attachment 3 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

_____________________________________________________________________________________

Information

Date of

Notice No.

Subject

Issuance

Issued to

______________________________________________________________________________________

2001-04

Neglected Fire Extinguisher

Maintenance Causes Fatality

04/11/01

All holders of licenses for nuclear

power, research, and test

reactors and fuel cycle facilities

2001-03

Incident Reporting

Requirements for Radiography

Licensees

04/06/01

All industrial radiography

licensees

2001-02 Summary of Fitness-for-Duty

Program Performance Reports

for Calendar Years 1998 and

1999

03/28/01

All holders of operating licenses

for nuclear power reactors, and

licensees authorized to possess

or use formula quantities of

strategic special nuclear material

(SSNM) or to transport formula

quantities of SSNM

2001-01

The Importance of Accurate

Inventory Controls to Prevent

the Unauthorized Possession

of Radioactive Material

03/26/01

All material licensees

2000-17, Supp. 2

Crack in Weld Area of Reactor

Coolant System Hot Leg Piping

at V.C. Summer

02/28/01

All holders of operating licenses

for nuclear power reactors except

those who has ceased operations

and have certified that fuel has

permanently removed from

reactor vessel

2000-22

Medical Misadministrations

Caused by Human Errors

Involving Gamma Stereotactic

Radiosurgery (GAMMA KNIFE)

12/18/00

All medical use licensees

authorized to conduct gamma

stereotactic radiosurgery

treatments

2000-21

Detached Check Valve Disc

not Detected by Use of

Acoustic and Magnetic

Nonintrusive Test Techniques

12/15/00

All holders of OLs for nuclear

power reactors except those who

have ceased operations and have

certified that fuel has been

permanently removed from the

reactor