Bulletin 82-03: Revision 1: Stress Corrosion Cracking in Thick-Wall, Large-Diameter, Stainless Steel, Recirculation System Piping at BWR Plants
SSINS NO.: 6820
IEB 82-03 Rev. 1
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555
October 28, 1982
IE BULLETIN NO. 82-03, REVISION 1: STRESS CORROSION CRACKING IN THICK-WALL,
LARGE-DIAMETER, STAINLESS STEEL,
RECIRCULATION SYSTEM PIPING AT BWR PLANTS
Addressees:
Those licensees of operating boiling water reactors (BWRs) identified in
Table 1 for action. All other licensees and holders of construction permits
(CPs) for information only.
Purpose:
This bulletin is to notify all licensees and CP holders about a matter that
may have a high degree of safety significance, and to require specific
actions as set forth below for those licensees listed in Table 1.
Specifically, this matter involves the degradation in the recirculation
system piping in the reactor coolant pressure boundary . (RCPB) that was
found at the Nine Mile Point Unit 1 Nuclear Generating Station. This
information was described in considerable detail in Information Notice
82-39, dated September 21, 1982. Action by the affected licensees identified
in Table 1 is required to (1) provide a reasonable level of assurance that
inspections which are currently being performed or scheduled are sufficient
to detect cracking in BWR thick-wall recirculation piping welds* and (2) R1
to assist the NRC in determining the generic significance of the piping
degradation found at Nine Mile Point. The affected licensees are those
owners whose plants are currently in or scheduled to be in a refueling mode
or extended outage through January 31, 1983.
This bulletin is provided to all other licensees and holders of construction
permits for information only at this time. Licensees not listed in Table 1
will be notified by January 15, 1983 as to the scope and extent of any
required actions.
Description of Circumstances:
During a primary system hydrotest in March 1982 at Nine Mile Point Unit 1
(NMP-1), leakage was visually detected at two of the ten
*Large bore piping that is not designated as "Service-Sensitive" in R1
accordance with NUREG-0313, Rev. 1. It should be noted that NUREG-0313, R1
Rev. 1 designates the recirculation riser lines as "Service-Sensitive. R1
8208190240
. IEB 82-03, Rev. 1
October 28, 1982
Page 2 of 5
Table 1
Plants Currently in or Scheduled to Be in
a Refueling Mode or Extended Outage Through January 31, 1983
LICENSEE PLANT
Northern States Power Company Monticello Nuclear Generating
Station
Tennessee Valley Authority Browns Ferry Unit 2 Nuclear
Generating Station
Commonwealth Edison Company Quad Cities Unit 1 Nuclear
Generating Station
Dresden Unit 2 Nuclear Generating
Station
Northeast Utilities Millstone Unit 1 Nuclear Generating
Station
Georgia Power Company Hatch Unit 1 Nuclear Generating
Station
Carolina Power & Light Company Brunswick Unit 1 Nuclear Generating
Station*
Jersey Central Power & Light Company Oyster Creek Nuclear Generating
Station
Iowa Electric Light & Power Company Duane Arnold Nuclear Generating
Station
*To be performed during the November 1982 refueling outage, not the current
outage.
.
IEB 82-03, Rev. 1
October 28, 1982
Page 3 of 5
furnace-sensitized, recirculation system safe-ends. Further visual
inspection revealed three pinhole indications and a single 1/2-inch-long
axial indication, all of which were located in the heat-affected zone of the
welds where the safe-end joined the pipe. About nine months before the leak,
these safe-ends were ultrasonically (UT) inspected; at that time, the
inspection did not disclose any reportable indications. Subsequent to the
leak, the UT procedure was modified; UT examination of the two affected
safe-ends and one other safe-end confirmed the presence of indications of
intermittent cracking around the pipe's inside diameter (ID). Additional
examinations revealed cracking in heat-affected zones of recirculation pump
discharge welds. Dye penetrant examination confirmed these crack
indications. The UT examinations werg extended to other welds in the five
loops of the recirculation system. The results of these examinations
disclosed ID cracking in a large number of the welds examined.
Two boat samples removed from the area of the through-wall cracks in one
safe-end were sent for evaluation--one to General Electric Co. and the other
to Battelle Laboratories. In addition, a boat sample from the crack region
of the elbow weld was evaluated by Sylvester Associates, consultants to the
licensee. The results of these metallurgical evaluations concluded that the
degradation resulted from intergranular stress corrosion cracking (IGSCC) in
the sensitized region of the weld's heat-affected zones.
Based on the fact that NMP-1 has furnace-sensitized safe-ends, the licensee
decided to replace all 10 recirculation system safe-ends without further
investigation beyond that described above. Based on recirculation system
findings, the licensee decided to also replace all recirculation system
piping while the facility was shut down for safe-end replacement.
On September 16, 1982, a meeting was held between General Electric, BWR
licensees, and NRC staff to review past IGSCC experiences and the general
implications of NMP-1 IGSCC degradation in main recirculation piping welds.
The staff had the benefit of the metallurgical evaluation of the NMP-1 event
and an update of the general IGSCC experiences relative to all operating BWR
plants.
On September 27, 1982, a meeting was held between BWR licensees and the NRC
staff to discuss the extent and results of examining welds in the
recirculation system for all BWR licensees with plants currently in or
scheduled to be in a refueling mode or extended outage through January 31,
1983. As a result of this meeting, the NRC staff has determined that
additional information is needed to assess the effectiveness of the UT
methods employed or planned to be used and to determine whether such piping
should be designated "service- sensitive" in accordance with NUREG-0313,
Rev. 1, issued by NRC letter dated February 26, 1981.
To provide a reasonable level of assurance that inspections which are
currently being performed or scheduled are sufficient to detect cracking in
thick-wall, recirculation system piping welds and to assist the NRC in
further evaluating this issue, the affected licensees (identified in Table
1) are requested to take the following actions.
.
IEB 82- 03, Rev. 1
October 28, 1982
Page 4 of 5
Actions To Be Taken by Licensees of BWR Facilities Identified in Table 1:
1. Before resuming power operations following the current refueling or
extended outage, the licensee is to demonstrate the effectiveness of
the detection capability of the ultrasonic methodology used or planned
to be used to examine welds in recirculation system piping. This
demonstration shall be made on representative service-induced cracked
pipe samples. Arrangements should be made to allow NRC to witness this
demonstration. This demonstration shall employ those procedures and
standards, the same type of equipment (same transducer size,
frequencies and calibration-standards), and representative UT personnel
from the inservice inspection (ISI) organization utilized or to be
utilized in the examinations at the plant site.*
2. Before resuming power operations following the current refueling or
extended outage, the licensee is to provide a listing of results of
recirculation system piping inspections.
3. Before resuming power operations following the current refueling or
extended outage, the licensee (if the inspections indicate the presence
of cracks) is to describe the corrective actions taken and report these
in accordance with the appropriate regulations.
4. To assist NRC's further evaluation of this issue, the following shall
be submitted by December 1, 1982:
a. A description of the sampling plan used or to be used during this
outage for UT examinations of recirculation system piping welds
and the bases for the plan. The description should:
(1) Provide an isometric drawing of the recirculation system
piping showing all the welds, and the number of welds and
their location that have been examined or will be examined.
(2) Identify criteria for weld sample selection (e.g., stress
rule index, carbon content, high stress location, and their
values for each weld examined).
(3) Describe piping material(s), including material type,
diameter, and wall thickness.
(4) Estimate the occupational radiation exposure incurred or
expected and briefly summarize measures taken to maintain
individual and collective exposures as low as reasonably
achievable.
*We understand that Electric Power Research Institute (EPRI) has arranged to
have samples from the Nine Mile Point Unit 1 plant available for industry
demonstrations of UT methodology. The samples have been taken to Battelle
Memorial Institute in Columbus, Ohio for characterization and subsequent
use.
.
IEB 82-03, Rev. 1
October 28, 1982
Page 5 of 5
b. A summary description of the UT procedures and calibration
standards used or to be employed in the examination at the
licensee's plant site. This description should include the
scanning sensitivity, the evaluation sensitivity and the recording
criteria.
c. A summary of the results of any previous inspection of the
recirculation system piping welds which used the validated
examination methodology as discussed in Action Item 1 above.
d. An evaluation of the crack-detection capability of ultrasonic
methodology used or planned to be used to examine recirculation
system piping welds. This evaluation should result from conducting
the demonstration required in Action Item 1 above, and should
include a comparison of the service-induced pipe crack sample to
those welds actually examined in the licensee's plant in terms of
pipe wall thickness and diameter, weld geometry, and materials.
5. The written reports required by Items 2, 3, and 4 shall be submitted to
the appropriate Regional Administrator under oath or affirmation under
provisions of Section 182a, Atomic Energy Act of 1954, as amended. The
original copy of the cover letters and a copy of the reports shall be
transmitted to the U.S. Nuclear Regulatory Commission, Document Control
Desk, Washington, D.C. 20555 for reproduction and distribution.
This request for information does not require Office of Management and
Budget approval since the number of plants asked to provide the information
is limited to nine reactor plants.
Although no specific request or requirement is intended, the following
information would help the NRC evaluate the cost of implementing this
bulletin:
o Staff time to perform requested demonstration
o Staff time to prepare written responses
If you have any questions regarding this matter, please contact the Regional
Administrator of the NRC Regional Office or one of the technical contacts
listed below.
Richard C. DeYoung, Director
Office of Inspection and Enforcement
Technical Contact: William J. Collins, IE
492-7275
Warren Hazelton, NRR
492-8075
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