Information Notice 2004-11, 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