Bulletin 79-17: Pipe Cracks in Stagnant Borated Water Systems at PWR Plants
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D. C. 20555
July 26, 1979
IE Bulletin No. 79-17
PIPE CRACKS IN STAGNANT BORATED WATER SYSTEMS AT PWR PLANTS
Description of Circumstances:
During the period of November 1974 to February 1977 a number of cracking
incidents have been experienced in safety-related stainless steel piping
systems and portions of systems which contain oxygenated, stagnant or
essentially stagnant borated water. Metallurgical investigations revealed
these cracks occurred in the weld heat affected zone of 8-inch to 10-inch
type 304 material (schedule 10 and 40), initiating on the piping I.D.
surface and propagating in either an intergranular or transgranular mode
typical of Stress Corrosion Cracking. Analysis indicated the probable
corrodents to be chloride and oxygen contamination in the affected systems.
Plants affected up to this time were Arkansas Nuclear Unit 1, R. E. Ginna,
H.B.Robinson Unit 2, Crystal River Unit 3, San Onofre Unit 1, and Surry
Units 1 and 2. The NRC issued Circular 76-06 (copy attached) in view of the
apparent generic nature of the problem.
During the refueling outage of Three Mile Island Unit 1 which began in
February of this year, visual inspections disclosed five (5) through-wall
cracks at welds in the spent fuel cooling system piping and one (1) at a
weld in the decay heat removal system. These cracks were found as a result
of local boric acid build-up and later confirmed by liquid penetrant tests.
This initial identification of cracking was reported to the NRC in a
Licensee Event Report (LER) dated May 16, 1979. A preliminary metallurgical
analysis was performed by the licensee on a section of cracked and leaking
weld joint from the spent fuel cooling system. The conclusion of this
analysis was that cracking was due to Intergranular Stress Corrosion
Cracking (IGSCC) originating on the pipe I.D. The cracking was localized to
the heat affected zone where the type 304 stainless steel is sensitized
(precipitated carbides) during welding. In addition to the main through-wall
crack, incipient cracks were observed at several locations in the weld heat
affected zone including the weld root fusion area where a minuscule lack of
fusion had occurred. The stresses responsible for cracking are believed to
be primarily residual welding stresses in as much as the calculated applied
stresses were found to be less than code design limits. There is no
conclusive evidence at this time to identify those aggressive chemical
species which promoted this IGSCC attack. Further analytical efforts in this
area and on other system welds are being pursued.
IE Bulletin No. 79-17 July 26, 1979
Page 2 of 4
Based on the above analysis and visual leaks, the licensee initiated a broad
based ultrasonic examination of potentially affected systems utilizing
special techniques. The systems examined included the spent fuel, decay heat
removal, makeup and purification, and reactor building spray systems which
contain stagnant or intermittently stagnant oxygenated boric acid
environments. These systems range from 2 1/2-inch (HPCI) to 24-inch (borated
water storage tank suction), are type 304 stainless steel, schedule 160 to
schedule 40 thickness respectively. Results of these examinations were
reported to the NRC on June 30, 1979 as an update to the May 16, 1979 LER.
The ultrasonic inspection as of July 10, 1979 has identified 206 welds out
of 946 inspected having UT indications characteristic of cracking randomly
distributed throughout the aforementioned sizes (24"-14"-12"-10"-8"-2" etc.)
of the above systems. It is important to note that six of the crack
indications were found in 2 1/2-inch diameter pipe of the high pressure
injection lines inside containment. These lines are attached to the main
coolant pipe and are nonisolable from the main coolant system except for
check valves. All of the six cracks were found in two high pressure
injection lines containing stagnated borated water. No cracks were found in
the high pressure injection lines which were occasionally flushed during
makeup operations. The ultrasonic examination is continuing in order to
delineate the extent of the problem.
The above information was previously provided in Information Notice 79-19.
For All Pressurized Water Reactor Facilities with an Operating License:
1. Conduct a review of safety related stainless steel piping
systems,within 30 days of the date of this Bulletin to identify systems
and portions of systems which contain stagnant oxygenated borated
water. These systems typically include ECCS, decay/residual heat
removal, spent fuel pool cooling, containment spray and borated water
storage tank (BWST-RWST) piping.
(a) Provide the extent and dates of the hydrotests, visual and
volumetric examinations performed per 10 CFR 50.55a(g) (Re: IE
Circular 76-06 enclosed) of identified systems. Include a
description of the nondestructive examination procedures,
procedure qualifications and acceptance criteria, the sampling
plan, results of the examinations and any related corrective
(b) Provide a description of water chemistry controls, summary of
chemistry data, any design changes and/or actions taken, such as
periodic flushing of recirculation procedures to maintain required
water chemistry with respect to pH, B, CL-, F-, 02.
IE Bulletin No. 79-17 July 26, 1979
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(c) Describe the preservice NDE performed on the weld joints of
identified systems. The description is to include the applicable
ASME Code sections and supplements (addenda) that were followed,
and the acceptance criterion.
(d) Facilities having previously experienced cracking in identified
systems, Item 1, are requested to identify (list) the new
materials utilized in repair or replacement on a system-by-system
basis. If a report of this information and that requested above
has been previously submitted to the NRC, please reference the
specific report(s) in response to this Bulletin.
2. Facilities at which ISI examinations have not been performed (i.e.,
visual and volumetric UT) on stagnant portions of systems identified in
Item 1 above, shall complete the following actions at the earliest
practical date but not later than 90 days after the date of the
(a) Perform ASME Section XI visual examination (IWA 2210) of normally
accessible* welds of all engineered safety systems at service
pressure to verify system integrity.
(b) Conduct ultrasonic examination and liquid penetrant surface
examination or a representative number of circumferential welds in
normally accessible* portions of systems identified by above. It
is intended that the sample number of welds include all pipe
diameters in the 2-1/2 inch to 24-inch range with no less than a
10 percent sample by system and pipe wall thickness. It is also
intended that the U.T. examination cover the weld fusion zone and
a minimum of 1/2-inch on each side of the weld at the pipe I.D.
The examination shall be in accordance with the provisions of ASME
Code Section XI-Appendix III and Supplements of the 1975 Winter
Addenda except all signal responses shall be evaluated as to the
nature of the indications. These code methods or alternative
examination methods, combination of methods, or newly developed
techniques may be used provided the procedures yield a
demonstrated effectiveness in detecting stress corrosion cracking
in austenitic stainless steel piping.
(c) If cracking is identified during Item (a) and (b) examinations,
all welds of safety-related piping systems and associated
subsystems where dynamic flow conditions do not exist during
normal operations (Item 1) shall be subject to volumetric
examination and repair including piping in areas which are
* Normally accessible refers to those areas of the plant which can be
entered during reactor operation.
IE Bulletin No. 79-17 July 26, 1979
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3. Identification of cracking in one unit of a multi-unit facility which
causes safety-related systems to be inoperable shall require immediate
examination of accessible portions of other similar units which have
not been inspected under the ISI provisions of 10 CFR 50.55a(g) unless
justification for continued operation is provided.
4. Any cracking identified shall be reported to the Director of the
appropriate NRC Regional Office within 24 hours of identification
followed by a 14 day written report.
5. Provide a written report to the Director of the appropriate NRC
Regional Office within 30 days of the date of this Bulletin addressing
the results of your review required by Item 1.
6. Complete the examination required by Item 2 within 90 days of the date
of this Bulletin and provide a written report to the Director of the
appropriate NRC Regional Office within 120 days of the date of this
Bulletin describing the results of the inspections required by Item 2
and any corrective measures taken.
7. Copies of the reports required by Items 4, 5 and 6 above shall also be
provided to the Director, Division of Operating Reactors, Office of
Inspection and Enforcement, Washington, D.C. 20555.
Approved by GAO, B180225 (R0072), clearance expires 7/31/80. Approval was
given under a blanket clearance specifically for identified generic
1. IE Circular 76-06
2. List of IE Bulletins
Issued in 1979
Page Last Reviewed/Updated Friday, May 22, 2015