Information Notice 2004-11, Cracking in Pressurizer Safety and Relief Nozzles and in Surge Line Nozzle

Cracking in Pressurizer Safety and Relief Nozzles and in Surge Line Nozzle

ML041260136
Person / Time
Issue date:	05/06/2004
From:	Beckner W NRC/NRR/DIPM/IROB
To:
Fu Z, NRR/DE/EMCB, 415-2467
References
IN-04-011
Download: ML041260136 (9)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555 May 6, 2004 NRC INFORMATION NOTICE 2004-11: CRACKING IN PRESSURIZER SAFETY AND

RELIEF NOZZLES AND IN SURGE LINE NOZZLE

Addressees

All holders of operating licenses or construction permits for nuclear power reactors, except

those that 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 (IN) to alert

addressees to cracking and leakage indications found on pressurizer safety and relief nozzles

and in a surge line nozzle-to-safe end weld. It is expected that the recipients of this notice 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.

Background

During an annual inspection in September of 2003, cracking and leakage were discovered on

pressurizer safety and relief nozzles in Tsuruga Power Plant, Unit 2 (Tsuruga 2), in Japan.

Tsuruga 2 is a four-loop pressurized water reactor (PWR) unit (similar to the PWRs in the U.S).

Tsuruga 2, which started commercial operation in February 1987, was designed and fabricated

by Mitsubishi Heavy Industries. Full power for Tsuruga 2 is 1160 MWe. At 100% power, the

average primary coolant temperature is 289 EC (552 EF) in the cold leg and 322 EC (612 EF) in

the hot leg.

During a refueling outage in October 2003, an indication was detected in a surge line nozzle-to- safe end dissimilar metal weld at Three Mile Island, Unit 1 (TMI-1). TMI-1 is a Babcock and

Wilcox pressurized water reactor which started commercial operation in September 1974.

Description of Circumstances

Tsuruga 2 Tsuruga 2 was in its 13th annual inspection, which started in September of 2003. During a

visual inspection of the pressurizer safety and relief nozzles with insulation removed, boric acid

deposits were found on the pressurizer relief nozzle. Subsequent ultrasonic testing performed

on the pressurizer safety and relief nozzles revealed linear indications in the nozzle-to-safe-end

weld metal on the relief nozzle and on safety nozzle A, one of the three safety nozzles

(Figure 1).

Each nozzle consists of a safe end made of Type 316 stainless steel and a nozzle end made of

a low-alloy steel, similar to Type 508 ferritic steel typically used in the same nozzles for U. S.

plants (Figures 2 and 3). The safety and relief nozzles are the same size, about 190 mm

(7.5 inches) outside diameter and about 130 mm (5.1 inches) inside diameter. Weld butter with

Alloy 132 (which has properties similar to Alloy 182) was initially applied to each of the nozzles.

The component was stress-relieved. Then a safe end was welded to each nozzle with

Alloy 132. The weld is approximately 40 mm (1.6 inches) in width.

Plant personnel found that a repair was made to the nozzle-to-safe-end weld on safety

nozzle A.

Figure 1. Top View of Reactor Pressurizer (Courtesy of Japan Atomic Power)

Figure 2. Nozzle Configuration (Courtesy of Japan Atomic Power) All of the flaws found were axially oriented and located in the welds, that is, the flaws did not

extend into the base metal. The 0E location of each nozzle is the point of the nozzle closest to

the centerline axis of the pressurizer cylinder, marked by the spray line nozzle in Figure 1. On

safety nozzle A, two indications with a maximum length of 24 mm (0.9 inches) were found at the

35E - 45E location. On the relief nozzle, two indications with a maximum length of 35 mm

(1.4 inches) were found at the 90E location, and one indication with a maximum length of

34 mm (1.3 inches) was found at the 315E location.

The samples removed for destructive examinations contained the entire weld and a portion of

the base metal on each side of the weld.

Radiography was performed on the severed pieces, confirming the linear flaws. Metallurgical

failure analysis was performed on these samples. The results showed that the cracks initiated

from the inside diameter surface, were axially oriented and were intergranular or interdendritic

in nature. A through-wall crack was confirmed at the 90E location in the weld on the relief

nozzle. The conclusion of the metallurgical analysis was that the nozzle failures were caused

by primary water stress corrosion cracking (PWSCC) in the nozzle weld.

Personnel at the plant stated that visual inspections, with insulation removed, were performed

on the pressurizer nozzles during the 1st, 2nd, 9th, and 10th annual inspections. Ultrasonic testing

(normal beam method using 0E angle wave, straight beam) and dye penetrant testing were

Figure 3. Nozzle Materials (Courtesy of Japan Atomic Power) performed during the 9th refueling outage in early 1998, and during the 10th refueling outage in

late 1999. Plant personnel stated that no indications were detected during the previous

inspections.

TMI-1 During refueling outage 15 in October 2003, an indication was detected in a surge line nozzle- to-safe end dissimilar metal weld at TMI-1. The nozzle is a 25.4 cm (10-inch) diameter, schedule 140, American Society for Testing and Materials (ASTM) A-105, Grade 2, carbon

steel product with an Alloy 82/182 filler metal butter and welded with Alloy 82/182 filler metal to

an ASTM A-336 Class F8M forged stainless steel safe end. The surge line nozzle is connected

to the steam generator A hot leg of the primary coolant loop and normally is operating at

317 EC (602 EF).

TMI-1 was performing planned manual ultrasonic testing (UT) of the surge line nozzle-to-safe

end weld and found an axial indication in the weld material. During subsequent UT

examinations, the licensee characterized the indication as spanning the width of the weld on the

inside surface and extending 12 mm (0.48 inches) into the weld.

The indication was confined in the Alloy 82/182 weld material and stopped at the base metal

interface on either side of the weld. The indication was in a region that was repaired during

original fabrication. Based on the location, acoustic response, and operating temperature, TMI-1 concluded that the indication was due to PWSCC.

TMI-1 performed a full structural weld overlay repair to maintain weld integrity. The overlay was

installed using an machine tungsten arc welding, temper bead process and Alloy 52 filler

material.

Discussion

It is well known that Alloy 600/82/182 materials are susceptible to PWSCC. A pressurizer relief

nozzle leak was detected at Palisades in 1993. The leak was attributed to PWSCC in the Alloy

82/182 nozzle weld. PWSCC in nozzles of the same materials in the reactor coolant

environment was also reported in recent years. For example, cracking and leakage were

discovered in the reactor vessel hot-leg nozzle weld at V. C. Summer during a refueling outage

in October 2000. Metallurgical examinations revealed both axial and circumferential cracks in

the nozzle weld. The root cause was attributed to PWSCC of the Alloy 82/182 weld. Axial

cracks caused by PWSCC were also detected in hot-leg nozzle welds at Ringhals Unit 3 in

1999 and at Ringhals Unit 4 in the fall of 2000 during outage inspections.

Based on currently available information, the NRC believes that the degradation that occurred

at Tsuruga 2 and TMI-1 is relevant to PWR facilities. The NRC has issued a number of generic

communications and an order over the past 2 years related to PWSCC in the reactor coolant

system of PWRs. The NRC staff continues to evaluate the adequacy of inspections to assure

that reactor coolant pressure boundary integrity is maintained at each facility. 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.

/RA/

William D. Beckner, Chief

Reactor Operations Branch

Division of Inspection Program Management

Office of Nuclear Reactor Regulation

Technical Contacts: Bart Fu, NRR

(301) 415-2467 E-mail: zbf@nrc.gov

Attachment: List of Recently Issued NRC Information Notices

ML041260136 OFFICE OES:IROB Tech Editor EMCB EMCB

NAME CVHodge PKleene ZBFu EJSullivan

DATE 04/22/2004 03/08/2004 04/22/2004 04/23/2004 OFFICE EMCB (A)SC:OES:IROB:DIPM C:IROB:DIPM

NAME BBateman CPJackson WDBeckner

DATE 04/26/2004 05/05/2004 05/06/2004

Attachment 1 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

_____________________________________________________________________________________

Information Date of

Notice No. Subject Issuance Issued to

_____________________________________________________________________________________

2004-10 Loose Parts in Steam 05/04/2004 All holders of operating licenses

Generators for pressurized-water reactors

(PWRs), except those who have

permanently ceased operations

and have certified that fuel has

been permanently removed from

the reactor.

2004-09 Corrosion of Steel 04/27/2004 All holders of operating licenses

Containment and Containment for nuclear power reactors except

Liner those who have permanently

ceased operation and have

certified that fuel has been

permanently removed from the

reactor vessel.

2004-08 Reactor Coolant Pressure 04/22/2004 All holders of operating licensees

Boundary Leakage Attributable for nuclear power boiling-water

to Propagation of Cracking in reactors (BWRs), except those

Reactor Vessel Nozzle Welds who have permanently ceased

operations and have certified that

fuel has been permanently

removed from the reactor vessel.

2004-07 Plugging of Safety Injection 04/07/2004 All holders of operating licenses

Pump Lubrication Oil Coolers or construction permits for

with Lakeweed nuclear power reactors, except

those who have permanently

ceased operations and have

certified that fuel has been

permanently removed from the

reactor vessel.

Note: NRC generic communications may be received in electronic format shortly after they are

issued by subscribing to the NRC listserver as follows:

To subscribe send an e-mail to <listproc@nrc.gov >, no subject, and the following

command in the message portion:

subscribe gc-nrr firstname lastname

______________________________________________________________________________________

OL = Operating License

CP = Construction Permit