United States Nuclear Regulatory Commission - Protecting People and the Environment

Bulletin 79-17: Revision 1, Pipe Cracks in Stagnant Borated Water Systems at PWR Plants

                                                            SSINS No.: 6820 
                                                            Accession No.:  
                                                            7908220137  

                                UNITED STATES
                        NUCLEAR REGULATORY COMMISSION
                    OFFICE OF INSPECTION AND ENFORCEMENT
                           WASHINGTON, D.C. 20555

                              October 29, 1979 

                                                  IE Bulletin No. 79-17 
                                                  Revision 1

PIPE CRACKS IN STAGNANT BORATED WATER SYSTEMS AT PWR PLANTS  

Description of Circumstances:

IE Bulletin No. 79-17, issued July 26, 1979, provided information on     R1
the cracking experienced to date in safety-related stainless steel       R1
piping systems at PWR plants. Certain actions were required of all PWR   R1
facilities with an operating license within a specified 90-day time      R1
frame.                                                                   R1

After several discussions with licensee owner group representatives and  R1
inspection  agencies it has been determined that the requirements of     R1
Item 2, particularly  the ultrasonic examination, may be impractical     R1
because of unavailability of  qualified personnel in certain cases to    R1
complete the inspections within the time specified by the Bulletin. To   R1
alleviate this situation and allow licensees the resources of improved   R1
ultrasonic inspection capabilities, a time extension and clarifications  R1
to the bulletin have been made. These are referenced to the affected     R1
items of the original bulletin.                                          R1

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 No. 76-06 (copy enclosed) in view of the apparent generic
nature of the problem.

During the refueling outage of Three Mile Island Unit l 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 buildup 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.

R1 - Identifies those additions or revision to IE Bulletin No. 79-17   
.

IE Bulletin No. 79-17                                  October 29, 1979 
Revision 1                                             Page 2 of 5 

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
miniscule 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.

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 If -10 11 811 2"
etc.) of the above systems. It is important to note that six of the crack
indications were reportedly found in 2 1/2-inch diameter pipe of the     R1
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 crack indications were found in  R1
two high pressure injection lines containing stagnated borated water. No R1
crack indications were found in high pressure injection lines which were R1
utilized for makeup operations.                                          R1

Recent data reported from Three Mile Island Unit 1 indicates that the    R1
extent of IGSCC experienced in stainless steel piping at that facility   R1
may be more limited than originally stated above. Of the 1902 total      R1
welds originally inspected 350 contained U.T. indications which required R1
further evaluation. These 350 welds have been reinspected with a second  R1
U.T. procedure which pur- portedly provides better discrimination        R1
between actual cracks and geometrical reflectors. Hence, the Licensee    R1
now estimates that approximately 38 of the 350 welds contain IGSCC and   R1
the remaining welds, including those in high pressure injection and      R1
decay heat lines, contain only geometrical reflectors. Further           R1
metallurgical analysis of these welds is required to verify the adequacy R1
of the U.T. procedures and to determine the nature of the cracking.      R1
.

IE Bulletin No. 79-17                                       October 29, 1979
Revision 1                                                  Page 3 of 5 

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 (July 26, 1979) to identify    R1
     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.

     For this review, the term "stagnant, oxygenated borated water       R1
     systems" refers to those systems serving as engineered safeguards   R1
     having no normal operating functions and contain essentially air    R1
     saturated borated water where dynamic flow conditions do not exist  R1
     on a continuous basis. However, these systems must be maintained    R1
     ready for actuation during normal power operations. Where your      R1
     definition for stagnant differed from the one given above please    R1
     supplement your previous response within 30 days of this Bulletin   R1
     revision.                                                           R1

     (a)  Provide the extent and dates of the hydrotests, visual and
          volumetric examinations performed per 10 CFR 50.55a(g) (Re: IE
          Circular No. 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 actions taken.

     (b)  Provide a description of water chemistry controls, summary of
          chemistry data, any design changes and/or actions taken, such as
          periodic flushing or recirculation procedures to maintain required
          water chemistry with respect to pH, B, C1-, F-, 02.

     (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.   All operating PWR facilities shall complete the following inspectionR1
     on the stagnant piping systems identified in Item 1 at the          R1
     earliest practical date but not later than twelve months from the   R1
     date of this bulletin revision. Facilities which have been          R1
     inspected in accordance with the original Bulletin, Sections 2(a)   R1
     and 2(b) satisfy the requirements of this Revision.                 R1
.

IE Bulletin No. 79-17                                  October 29, 1979 
Revision 1                                             Page 4 of 5 

     (a)  Until the examination required by 2(b) is completed a visual   R1
          examination shall be made of all normally accessible welds of  R1
          the engineered safety systems at least monthly to verify       R1
          continued systems integrity.  Similarly, the normally          R1
          inaccessible welds, shall be visually examined during each     R1
          cold shutdown.                                                 R1

          The relevant provisions of Article IWA 2000 of ASME Code       R1
          Section XI and Article 9 of Section V are considered           R1
          appropriate and an acceptable basis for this examination. For  R1
          insulated piping, the examination may be conducted without the R1
          removal of insulation. During the examination particular       R1
          attention shall be given to both insulated and noninsulated    R1
          piping for evidence of leakage and/or boric acid residues      R1
          which may have accumulated during the service period preceding R1
          the examination.  Where evidence of leakage and/or boric acid  R1
          residues are detected at locations, other than those normally  R1
          expected, (such as valve stems, pump seals, etc.) the piping   R1
          shall be cleaned (including insulation removal) to the extent  R1
          necessary to permit further evaluation of the piping condition.R1
          In cases where piping conditions observed are not sufficiently R1
          definitive, additional inspections (i.e., surface and/or       R1
          volumetric) shall be conducted in accordance with Item 2.(b).  R1

     (b)  An ultrasonic examination shall be performed on a              R1
          representative sample of circumferential welds in normally     R1
          accessible* portions of systems identified by 1 above. It is   R1
          intended that the sample number of welds selected for          R1
          examination include all pipe diameters within the 2 1/2- inch  R1
          to 24-inch range with no less than a 10 percent sampling being R1
          taken. The approach to selection of the sample shall be based  R1
          on the following criteria:                                     R1

          (1)  Pipe Material Chemistry - As a first consideration, those R1
               welds in austenitic stainless steel piping (Types 304 and R1
               316 ss) having 0.05 to 0.08 wt. % carbon content based on R1
               available material certification reports.                 R1

          (2)  Pipe Size and Thickness - An unbiased mixture of pipe     R1
               diameters and actual wall thickness distributed among     R1
               both horizontal and vertical piping runs shall be includedR1
               in the sample.                                            R1

          (3)  System Importance - The sample welds shall focus the      R1
               examination primarily on those systems required to        R1
               function in the emergency core cooling mode and secondly, R1
               on the containment spray system.                          R1


          The U.T. examination sample may be focused on noninsulated     R1
          piping runs. The evaluation shall cover the weld root fusion   R1
          zone and a minimum of 112 inch on the pipe I.D. (counterbore   R1
          area) on each side of the weld. The procedure(s) for this      R1
          examination shall be essentially                               R1

*Normal accessible refers to those areas of the plant which can be       R1
entered during reactor operation.                                        R1
.

IE Bulletin No. 79-17                                  October 29, 1979 
Revision 1                                             Page 5 of 5 

          in accordance with ASME Code Section XI, Appendix III and      R1
          Supplements of the 1975 Winter Addenda, except all signal      R1
          responses shall be evaluated as to the nature of the           R1
          reflectors. Other alternative examination methods, combination R1
          of methods, or newly developed techniques may be used provided R1
          the procedure(s) have a proven capability of detecting stress  R1
          corrosion cracking in austenitic stainless steel piping.       R1

          For welds of systems included in the sample having pipe wall   R1
          thickness of 0.250 inches and below, visual and liquid         R1
          penetrant surface examination may be used in lieu of           R1
          ultrasonic examination.                                        R1

     (c)  If cracking is identified during Item 2(a) and 2(b)            R1
          examinations, all welds in the affected system, shall be       R1
          subject to examination and repair considerations. In addition, R1
          the sample welds to be examined on the remaining normally      R1
          accessible noninsulated piping shall be increased to 25        R1
          percent using the criteria outlined in paragraph 2(b). In the  R1
          event that cracking is identified in other systems at this     R1
          sampling level,  all accessible and inaccessible welds of the  R1
          systems identified in  item 1 shall be subject to examination. R1
 
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     R1
     Regional Office within 30 days of the date of this bulletin         R1
     revision addressing the results of your review if required by Item  R1
     1. Provide a schedule of your inspection plans in response to Item  R1
     2(b) in those cases in which the inspections have not been          R1
     completed.                                                          R1

6.   Provide a written report to the Director of the appropriate NRC     R1
     Regional Office within 30 days of the date of completion of the     R1
     examinations required by Items 2(a) , 2(b), or 2(c) describing the  R1
     inspection results and any corrective actions taken.                R1

7.   Copies of the reports required by Items above shall also be provided to
     the Director, Division of Operating Reactors, Office of Inspection and
     Enforcement, Washington, D.C. 20555.

Approved by GAO, 8180225 (R0072), clearance expires 7/31/80. Approval was 
given under a blanket clearance specifically for identified generic problems.

Enclosures:
1. IE Circular No. 76-06
2. List of IE Bulletins Issued 
     in the Last Six Months  
Page Last Reviewed/Updated Tuesday, July 23, 2013