Information Notice 2002-11, Recent Experience with Degradation of Reactor Pressure Vessel Head

Recent Experience with Degradation of Reactor Pressure Vessel Head

ML020700556
Person / Time
Site:	Davis Besse
Issue date:	03/12/2002
From:	Beckner W NRC/NRR/DRIP/RORP
To:
Dozier J, NRR/RLSB 415-1014
References
IEB-01-001 IN-02-011
Download: ML020700556 (11)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, DC 20555-0001 March 12, 2002 NRC INFORMATION NOTICE 2002-11: RECENT EXPERIENCE WITH DEGRADATION

OF REACTOR PRESSURE VESSEL HEAD

Addressees

All holders of operating licenses for pressurized-water reactors (PWRs), except those who have

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

from the reactor.

Purpose

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

addressees about findings from recent inspections and examinations of the reactor pressure

vessel (RPV) head at Davis-Besse Nuclear Power Station. It is expected that recipients will

review the information for applicability to their facilities and consider actions, as appropriate, to

avoid similar problems. However, suggestions contained in this information notice are not

NRC requirements; therefore, no specific action or written response is required.

Description of Circumstances

On February 16, 2002, the Davis-Besse facility began a refueling outage that included

inspection of the vessel head penetration (VHP) nozzles, which focused on the inspection of

control rod drive mechanism (CRDM) nozzles, in accordance with the licensees commitments

to NRC Bulletin 2001-01, ?Circumferential Cracking of Reactor Pressure Vessel Head

Penetration Nozzles, which was issued on August 3, 2001. These inspections identified axial

indications in three CRDM nozzles, which had resulted in pressure boundary leakage.

Specifically, these indications were identified in CRDM nozzles 1, 2, and 3, which are located

near the center of the RPV head. These findings were reported to the NRC on February 27,

2002, and supplemented on March 5 and March 9, 2002. The licensee decided to repair these

three nozzles, as well as two other nozzles that had indications but had not resulted in pressure

boundary leakage.

The repair process for these nozzles included roll expanding the CRDM nozzle material into the

surrounding RPV head material, followed by machining along the axis of the CRDM nozzle to

an elevation above the indications in the nozzle material. On March 6, 2002, the machining

process on CRDM nozzle 3 was prematurely terminated and the machining apparatus was

removed from the nozzle. During the removal process, nozzle 3 was mechanically agitated and

subsequently displaced in the downhill direction (i.e., tipped away from the top of the RPV

head) until its flange contacted the flange of the adjacent CRDM nozzle.

To identify the cause of the CRDM nozzle displacement, the licensee began an investigation

into the condition of the RPV head surrounding CRDM nozzle 3. This investigation included

removing the CRDM nozzle from the RPV head, removing boric acid deposits from the top of

the RPV head, and performing ultrasonic thickness measurements of the RPV head in the

vicinity of CRDM nozzles 1, 2, and 3. Upon completing the boric acid removal on March 7,

2002, the licensee conducted a visual examination of the area, which identified a large cavity in

the RPV head on the downhill side of CRDM nozzle 3. Followup characterization by ultrasonic

testing indicated wastage of the low alloy steel RPV head material adjacent to the nozzle.

The wastage area was found to extend approximately 5 inches downhill on the RPV head from

the penetration for CRDM nozzle 3, with a width of approximately 4 to 5 inches at its widest

part. The minimum remaining thickness of the RPV head in the wastage area was found to be

approximately d inch. This thickness was attributed to the thickness of the stainless steel

cladding on the inside surface of the RPV head, which is nominally d inch thick.

Background

The Davis-Besse Nuclear Power Station has an RPV head that is constructed from low alloy

steel, fabricated in accordance with the American Society of Mechanical Engineers (ASME)

specification SA-533, Grade B, Class1, and clad on the inside surface with stainless steel. Of

those 69 VHP nozzles, 61 are used for CRDMs, 7 are spare (empty) nozzles, and 1 is used for

the RPV head vent piping. Each of the 69 nozzles is approximately 4 inches in outside

diameter, with a wall-thickness of approximately e inch. Each is constructed of Alloy 600 and

is attached to the RPV head by a partial-penetration, J-groove weld using Alloy 82 and 182.

The distance from the center of one nozzle to the center of the next is approximately 12 inches.

The vessel head is insulated with metal reflective insulation, which is located on a horizontal

plane slightly above the RPV head (i.e., it is not in direct contact with the head). The minimum

distance between the RPV head and the insulation is approximately 2 inches at the center (top)

of the head. The CRDM nozzles pass from the RPV head through the insulation and terminate

at flanges to which the CRDM housings are attached.

The limited gap between the insulation and the RPV head does not impede the performance

of a visual inspection of the CRDM nozzles, as described in Bulletin 2001-01. This is because

the top of the RPV head is surrounded by a service structure that has 18 openings (referred to

as weep holes) near the bottom of the structure, through which small cameras can be inserted

to facilitate visual inspections of the RPV head.

During refueling outages in 1998 and 2000, the licensee performed visual inspections of the

RPV head surface that was accessible through the service structure weep holes. The scope of

these visual inspections covered the bare metal of the RPV head to identify the presence of

boric acid deposits, which would be indicative of primary coolant leakage. These inspections

also included checking for leakage from any of the CRDM flanges, located above the insulation, in response to Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure

Boundary Components, which the NRC issued on March 17, 1988.

The visual inspections in 1998 showed an uneven layer of boric acid deposits scattered over

the RPV head (including deposits near CRDM nozzle 3). The outside diameter of the CRDM

nozzles had white streaks, which indicated to the licensee that the boric acid evident on the

head flowed downward from leakage in the CRDM flanges. During the refueling outage in 2000, the licensee also performed visual inspections of the

CRDM flanges and nozzles. Above the RPV head insulation, those inspections revealed five

CRDM flanges with evidence of leakage, including one flange that was the principal leakage

point. Boric acid deposits on the vertical faces of three of these five flanges and the associated

nozzles confirmed leakage from the flanges. Similarly, one of the other two leaking CRDM

flanges had boric acid deposits between the flange and the insulation, which indicated leakage

from the flange. All of these leaking flanges were repaired by replacing their gaskets. The

faces of the flange that was the principal leakage point were also machined to ensure a better

seal.

Visual inspections performed below the RPV head insulation during the 2000 refueling outage

indicated some accumulation of boric acid deposits on the RPV head. These deposits were

located beneath the leaking flanges, with clear evidence of downward flow from the flange area.

No visible evidence of CRDM nozzle leakage (i.e., leakage from the gap between the nozzle

and the RPV head) was detected. The licensee described that the RPV head area was cleaned

with demineralized water to the greatest extent possible, while trying to maintain the dose

as low as reasonably achievable (ALARA). Subsequent video inspection of the partially

cleaned RPV head and nozzles was performed for future reference.

A subsequent review of the 1998 and 2000 inspection videotapes in 2001 confirmed that there

was no evidence of leakage from the RPV head nozzles, although many areas of the RPV head

were not accessible because of persistent boric acid deposits that the licensee did not clean

because of ALARA issues (including the region around nozzle 3).

The inspections in 2002 did not reveal any visual evidence of flange leakage from above the

RPV head. However, as discussed above, three CRDM nozzles had indications of cracking

(identified by ultrasonic testing of the nozzles), which could result in leakage from the RPV to

the top of the RPV head.

Discussion

The following documents describe reactor operating experience with boric acid corrosion

of ferritic steel reactor coolant pressure boundary components in PWR plants:

• Information Notice 86-108, Degradation of Reactor Coolant System Pressure Boundary

Resulting from Boric Acid Corrosion, issued December 29, 1986

• Information Notice 86-108, Supplement 1, Degradation of Reactor Coolant System

Pressure Boundary Resulting from Boric Acid Corrosion, issued April 20, 1987

• Information Notice 86-108, Supplement 2, Degradation of Reactor Coolant System

Pressure Boundary Resulting from Boric Acid Corrosion, issued November 19, 1987

• Information Notice 86-108, Supplement 3, Degradation of Reactor Coolant System

Pressure Boundary Resulting from Boric Acid Corrosion, issued January 5, 1995

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

Components in PWR Plants, issued March 17, 1988 Several instances of boric acid corrosion discussed in these generic communications are

associated with corrosion of the RPV head. NRC Information Notice 86-108, Supplement 1, for example, described an instance in which boric acid had severely corroded three of the RPV

flange bolts, the control rod drive shroud support, and an instrument tube seal clamp. Similarly, NRC Information Notice 86-108, Supplement 2, described an instance in which boric acid

resulted in nine pits in the surface of the RPV head, ranging in depth from 0.9 to 1 cm

[approximately 0.4 inch] and ranging in diameter from 2.5 to 7.5 cm [1 to 3 inches].

As discussed in Information Notice 86-108, Supplement 2, the primary effect of boric acid

leakage onto the ferritic steel RPV head is wastage or general dissolution of the material.

Pitting, stress corrosion cracking (SCC), intergranular attack, and other forms of corrosion are

not generally of concern in concentrated boric acid solutions at elevated temperatures such as

those that may occur on the surface of the RPV head. The rate of general corrosion (wastage)

of ferritic steel from boric acid varies and depends on several conditions, including whether the

boric acid is dry or in solution. If the boric acid is dry (i.e., boric acid crystals), the corrosion rate

is less severe; however, boric acid crystals are not completely benign to carbon steel. During

operation, the temperature of the RPV head is sufficiently high that any leaking primary coolant

would be expected to flash to steam, leaving behind dry boric acid crystals.

Given the wide range of conditions around reactor primary coolant leakage sites and the wide

variation in boric acid corrosion rates, the deleterious effects of boric acid on ferritic steel

components indicate the importance of minimizing boric acid leakage, detecting and correcting

leaks in a timely manner, and promptly cleaning any boric acid residue.

The investigation of the causative conditions surrounding the degradation of the RPV head at

Davis-Besse is continuing. Boric acid or other contaminants could be contributing factors.

As discussed above, factors contributing to the degradation might also include the environment

of the head during both operating and shutdown conditions (e.g., wet/dry), the duration for

which the RPV head is exposed to boric acid, and the source of the boric acid (e.g., leakage

from the CRDM nozzle or from sources above the RPV head such as CRDM flanges).

Related Generic Communications

Bulletin 2001-01, Circumferential Cracking of Reactor Pressure Vessel Head Penetration

Nozzles, August 3, 2001.

Bulletin 82-02, Degradation of Threaded Fasteners in the Reactor Coolant Pressure Boundary

of PWR Plants, June 2, 1982.

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 Nozzles and Other Vessel

Closure Head Penetrations, April 1, 1997.

Information Notice 80-27, Degradation of Reactor Coolant Pump Studs, June 11, 1980.

Information Notice 82-06, Failure of Steam Generator Primary Side Manway Closure Studs, March 12, 1982. Information Notice 86-108, Degradation of Reactor Coolant System Pressure Boundary

Resulting from Boric Acid Corrosion, December 29, 1986.

Information Notice 86-108, Supplement 1, Degradation of Reactor Coolant System Pressure

Boundary Resulting from Boric Acid Corrosion, April 20, 1987.

Information Notice 86-108, Supplement 2, Degradation of Reactor Coolant System Pressure

Boundary Resulting from Boric Acid Corrosion, November 19, 1987.

Information Notice 86-108, Supplement 3, Degradation of Reactor Coolant System Pressure

Boundary Resulting from Boric Acid Corrosion, January 5, 1995.

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

February 23, 1990.

Information Notice 94-63, Boric Acid Corrosion of Charging Pump Casing Caused by Cladding

Cracks, August 30, 1994.

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

Corrosion Cracking of Control Rod Drive Mechanism Penetrations, February 14, 1996.

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, April 30, 2001.

This information notice does not require any 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 project manager in the NRCs Office of Nuclear Reactor

Regulation (NRR).

/RA/

William D. Beckner, Program Director

Operating Reactor Improvements Program

Division of Regulatory Improvement Programs

Office of Nuclear Reactor Regulation

Technical contacts: Allen Hiser, NRR Ken Karwoski, NRR

(301) 415-1034 (301) 415-2752 E-mail: alh1@nrc.gov E-mail: kjk1@nrc.gov

Jerry Dozier, NRR

(301) 415-1014 E-mail: jxd@nrc.gov

Attachment: List of Recently Issued NRC Information Notices Information Notice 86-108, ?Degradation of Reactor Coolant System Pressure Boundary

Resulting from Boric Acid Corrosion, December 29, 1986.

Information Notice 86-108, Supplement 1, Degradation of Reactor Coolant System Pressure

Boundary Resulting from Boric Acid Corrosion, April 20, 1987.

Information Notice 86-108, Supplement 2, Degradation of Reactor Coolant System Pressure

Boundary Resulting from Boric Acid Corrosion, November 19, 1987.

Information Notice 86-108, Supplement 3, Degradation of Reactor Coolant System Pressure

Boundary Resulting from Boric Acid Corrosion, January 5, 1995.

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

February 23, 1990.

Information Notice 94-63, ?Boric Acid Corrosion of Charging Pump Casing Caused by Cladding

Cracks, August 30, 1994.

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

Corrosion Cracking of Control Rod Drive Mechanism Penetrations, February 14, 1996.

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, April 30, 2001.

This information notice does not require any 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 project manager in the NRCs Office of Nuclear Reactor

Regulation (NRR).

/RA/

William D. Beckner, Program Director

Operating Reactor Improvements Program

Division of Regulatory Improvement Programs

Office of Nuclear Reactor Regulation

Technical contacts: Allen Hiser, NRR Ken Karwoski, NRR

(301) 415-1034 (301) 415-2752 E-mail: alh1@nrc.gov E-mail: kjk1@nrc.gov

Jerry Dozier, NRR

(301) 415-1014 E-mail: jxd@nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DISTRIBUTION:

PUBLIC IN Reading File

ADAMS ACCESSION NO.: ML020700556 Template: NRR-052

• See previous concurrence

OFFICE RSE:RORP:DRIP RSE:EMCB:DE BC:EMCB:DE (A)SC:RORP:DRIP PD:RORP:DRIP

NAME IJDozier* KJKarwoski* WHBateman* TKoshy* WDBeckner

DATE 03/11/2002 03/11/2002 03/11/2002 03/11/2002 03/12/2002 OFFICIAL RECORD COPY

Attachment 1 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

_____________________________________________________________________________________

Information Date of

Notice No. Subject Issuance Issued to

_____________________________________________________________________________________

2002-10 Nonconservative Water Level 03/07/2002 All holders of operating licenses

Setpoints on Steam for nuclear power reactors, Generators except those who have

permanently ceased operations

and have certified that fuel has

been permanently removed from

the reactor.

2002-09 Potential for Top Nozzle 02/13/2002 All holders of operating licenses

Separation and Dropping of for nuclear power reactors, and

Certain Type of Westinghouse non-power reactors and holders

Fuel Assembly of licenses for permanently

shutdown facilities with fuel

onsite.

2002-08 Pump Shaft Damage Due to 01/30/2002 All holders of operating licenses

Excessive Hardness of Shaft for nuclear power reactors, Sleeve except those who have

permanently ceased operations

and have certified that fuel has

been permanently removed from

the reactor.

2002-07 Use of Sodium Hypochlorite for 01/28/2002 All holders of operating licenses

Cleaning Diesel Fuel Oil for nuclear power except those

Supply Tanks who have ceased operations and

have certified that fuel has been

permanently removed from the

reactor vessel.

2002-06 Design Vulnerability in BWR 01/18/2002 All holders of operating licenses

Reactor Vessel Level or construction permits for boiling

Instrumentation Backfill water reactors (BWRs).

Modification

2002-05 Foreign Material in Standby 01/17/2002 All holders of licenses for nuclear

Liquid Control Storage Tanks power reactors.

2002-04 Wire Degradation at Breaker 01/10/2002 All holders of operating licenses

Cubicle Door Hinges for nuclear power reactors.

______________________________________________________________________________________

OL = Operating License

CP = Construction Permit