United States Nuclear Regulatory Commission - Protecting People and the Environment

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

                               UNITED STATES  
                       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 
          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 of recirculation procedures to maintain required
          water chemistry with respect to pH, B, CL-, F-, 02. 
.

IE Bulletin No. 79-17                                       July 26, 1979 
                                                            Page 3 of 4 

     (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 
     Bulletin. 

     (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 inaccessible. 

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

IE Bulletin No. 79-17                                       July 26, 1979 
                                                            Page 4 of 4 

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 
problems. 

Enclosures: 
1. IE Circular 76-06 
2. List of IE Bulletins 
     Issued in 1979 
Page Last Reviewed/Updated Tuesday, July 23, 2013