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

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
(StriderTol Bot change)
 
Line 14: Line 14:
| page count = 9
| page count = 9
}}
}}
{{#Wiki_filter:UNITED STATES
{{#Wiki_filter:ML041260136


===UNITED STATES===
NUCLEAR REGULATORY COMMISSION
NUCLEAR REGULATORY COMMISSION


OFFICE OF NUCLEAR REACTOR REGULATION
===OFFICE OF NUCLEAR REACTOR REGULATION===
 
WASHINGTON, D.C. 20555 May 6, 2004 NRC INFORMATION NOTICE 2004-11:
WASHINGTON, D.C. 20555 May 6, 2004 NRC INFORMATION NOTICE 2004-11:               CRACKING IN PRESSURIZER SAFETY AND


===CRACKING IN PRESSURIZER SAFETY AND===
RELIEF NOZZLES AND IN SURGE LINE NOZZLE
RELIEF NOZZLES AND IN SURGE LINE NOZZLE


Line 36: Line 37:
addressees to cracking and leakage indications found on pressurizer safety and relief nozzles
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
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
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
avoid similar problems. However, suggestions contained in this information notice are not NRC


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


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


Line 54: Line 54:
Tsuruga 2, which started commercial operation in February 1987, was designed and fabricated
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
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
average primary coolant temperature is 289 C (552 F) in the cold leg and 322 C (612 F) in


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


==Description of Circumstances==
==Description of Circumstances==
Tsuruga 2 Tsuruga 2 was in its 13th annual inspection, which started in September of 2003. During a
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
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
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
on the pressurizer safety and relief nozzles revealed linear indications in the nozzle-to-safe-end
Line 75: Line 75:
weld metal on the relief nozzle and on safety nozzle A, one of the three safety nozzles
weld metal on the relief nozzle and on safety nozzle A, one of the three safety nozzles


(Figure 1).
(Figure 1). Each nozzle consists of a safe end made of Type 316 stainless steel and a nozzle end made of
 
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.
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
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
(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.
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
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.
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
Plant personnel found that a repair was made to the nozzle-to-safe-end weld on safety
Line 95: Line 93:
nozzle A.
nozzle A.


Figure 1.       Top View of Reactor Pressurizer (Courtesy of Japan Atomic Power)
Figure 1.
        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
 
Top View of Reactor Pressurizer (Courtesy of Japan Atomic Power)
Figure 2.


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


the centerline axis of the pressurizer cylinder, marked by the spray line nozzle in Figure 1. On
extend into the base metal.  The 0 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
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
35 - 45 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
(1.4 inches) were found at the 90 location, and one indication with a maximum length of


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


The samples removed for destructive examinations contained the entire weld and a portion of
The samples removed for destructive examinations contained the entire weld and a portion of
Line 114: Line 116:
the base metal on each side of the weld.
the base metal on each side of the weld.


Radiography was performed on the severed pieces, confirming the linear flaws. Metallurgical
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
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
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
in nature. A through-wall crack was confirmed at the 90 location in the weld on the relief


nozzle. The conclusion of the metallurgical analysis was that the nozzle failures were caused
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.
by primary water stress corrosion cracking (PWSCC) in the nozzle weld.
Line 128: Line 130:
Personnel at the plant stated that visual inspections, with insulation removed, were performed
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
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
(normal beam method using 0 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
Figure 3.


late 1999. Plant personnel stated that no indications were detected during the previous
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.
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
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
steel product with an Alloy 82/182 filler metal butter and welded with Alloy 82/182 filler metal to
Line 146: Line 150:
to the steam generator A hot leg of the primary coolant loop and normally is operating at
to the steam generator A hot leg of the primary coolant loop and normally is operating at


317 EC (602 EF).
317 C (602 F).


TMI-1 was performing planned manual ultrasonic testing (UT) of the surge line nozzle-to-safe
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
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
examinations, the licensee characterized the indication as spanning the width of the weld on the
Line 158: Line 162:
The indication was confined in the Alloy 82/182 weld material and stopped at the base metal
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
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.
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
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
installed using an machine tungsten arc welding, temper bead process and Alloy 52 filler
Line 170: Line 174:
Discussion
Discussion


It is well known that Alloy 600/82/182 materials are susceptible to PWSCC. A pressurizer relief
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
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
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
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
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
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
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
cracks caused by PWSCC were also detected in hot-leg nozzle welds at Ringhals Unit 3 in
Line 190: Line 194:
Based on currently available information, the NRC believes that the degradation that occurred
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
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
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
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
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
questions about the information in this notice, please contact one of the technical contacts
Line 203: Line 207:


/RA/
/RA/
                                              William D. Beckner, Chief


===William D. Beckner, Chief===
Reactor Operations Branch
Reactor Operations Branch


Division of Inspection Program Management
===Division of Inspection Program Management===
 
Office of Nuclear Reactor Regulation
Office of Nuclear Reactor Regulation


Technical Contacts:   Bart Fu, NRR
Technical Contacts:


===Bart Fu, NRR===
(301) 415-2467 E-mail: zbf@nrc.gov
(301) 415-2467 E-mail: zbf@nrc.gov


Attachment: List of Recently Issued NRC Information Notices
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


ML041260136 OFFICE OES:IROB              Tech Editor                  EMCB              EMCB
CPJackson


NAME CVHodge                PKleene                      ZBFu              EJSullivan
WDBeckner


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


NAME BBateman                CPJackson                    WDBeckner
04/26/2004
05/05/2004
05/06/2004


DATE    04/26/2004          05/05/2004                    05/06/2004
______________________________________________________________________________________
OL = Operating License


Attachment 1 LIST OF RECENTLY ISSUED
CP = Construction Permit


===Attachment 1 LIST OF RECENTLY ISSUED===
NRC INFORMATION NOTICES
NRC INFORMATION NOTICES


_____________________________________________________________________________________
_____________________________________________________________________________________
Information                                             Date of
Information
 
Date of
 
Notice No.


Notice No.              Subject                         Issuance         Issued to
Subject
 
Issuance
 
Issued to


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


Generators                                             for pressurized-water reactors
===Loose Parts in Steam===
Generators
 
05/04/2004
 
===All holders of operating licenses===
for pressurized-water reactors


(PWRs), except those who have
(PWRs), except those who have
Line 251: Line 306:
the reactor.
the reactor.


2004-09           Corrosion of Steel                   04/27/2004       All holders of operating licenses
2004-09
 
===Corrosion of Steel===
Containment and Containment
 
Liner
 
04/27/2004


Containment and Containment                            for nuclear power reactors except
===All holders of operating licenses===
for nuclear power reactors except


Liner                                                  those who have permanently
those who have permanently


ceased operation and have
ceased operation and have
Line 265: Line 328:
reactor vessel.
reactor vessel.


2004-08           Reactor Coolant Pressure             04/22/2004       All holders of operating licensees
2004-08
 
===Reactor Coolant Pressure===
Boundary Leakage Attributable
 
to Propagation of Cracking in
 
===Reactor Vessel Nozzle Welds===
04/22/2004


Boundary Leakage Attributable                          for nuclear power boiling-water
===All holders of operating licensees===
for nuclear power boiling-water


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


Reactor Vessel Nozzle Welds                            who have permanently ceased
who have permanently ceased


operations and have certified that
operations and have certified that
Line 279: Line 351:
removed from the reactor vessel.
removed from the reactor vessel.


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


Pump Lubrication Oil Coolers                           or construction permits for
===Plugging of Safety Injection===
Pump Lubrication Oil Coolers


with Lakeweed                                         nuclear power reactors, except
with Lakeweed
 
04/07/2004
 
===All holders of operating licenses===
or construction permits for
 
nuclear power reactors, except


those who have permanently
those who have permanently
Line 295: Line 375:
reactor vessel.
reactor vessel.


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


issued by subscribing to the NRC listserver as follows:
issued by subscribing to the NRC listserver as follows:
                To subscribe send an e-mail to <listproc@nrc.gov >, no subject, and the following
To subscribe send an e-mail to <listproc@nrc.gov >, no subject, and the following


command in the message portion:
command in the message portion:
                                  subscribe gc-nrr firstname lastname
subscribe gc-nrr firstname lastname}}
 
______________________________________________________________________________________
OL = Operating License
 
CP = Construction Permit}}


{{Information notice-Nav}}
{{Information notice-Nav}}

Latest revision as of 03:11, 16 January 2025

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)
v • d • e


ML041260136

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 C (552 F) in the cold leg and 322 C (612 F) 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 0 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

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

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

34 mm (1.3 inches) was found at the 315 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 90 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 0 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 C (602 F).

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

______________________________________________________________________________________

OL = Operating License

CP = Construction Permit

Attachment 1 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

_____________________________________________________________________________________

Information

Date of

Notice No.

Subject

Issuance

Issued to

_____________________________________________________________________________________

2004-10

Loose Parts in Steam

Generators

05/04/2004

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.

2004-09

Corrosion of Steel

Containment and Containment

Liner

04/27/2004

All holders of operating licenses

for nuclear power reactors except

those who have permanently

ceased operation and have

certified that fuel has been

permanently removed from the

reactor vessel.

2004-08

Reactor Coolant Pressure

Boundary Leakage Attributable

to Propagation of Cracking in

Reactor Vessel Nozzle Welds

04/22/2004

All holders of operating licensees

for nuclear power boiling-water

reactors (BWRs), except those

who have permanently ceased

operations and have certified that

fuel has been permanently

removed from the reactor vessel.

2004-07

Plugging of Safety Injection

Pump Lubrication Oil Coolers

with Lakeweed

04/07/2004

All holders of operating licenses

or construction permits for

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


Retrieved from "https://kanterella.com/w/index.php?title=Information_Notice_2004-11,_Cracking_in_Pressurizer_Safety_and_Relief_Nozzles_and_in_Surge_Line_Nozzle&oldid=5402032"