Information Notice No. 79-19, Pipe Cracks in Stagnant Borated Water Systems at PWR Plants


                               July 17, 1979 

MEMORANDUM FOR:     B. H. Grier, Director, Region I 
                    J. P. O'Reilly, Director, Region II 
                    J. G. Keppler, Director, Region III 
                    K. V. Seyfrit, Director, Region IV 
                    R. H. Engelken, Director, Region V 

FROM:               Norman C. Moseley, Director, ROI:IE 

SUBJECT:            Information Notice No. 79-19, PIPE CRACKS IN STAGNANT

The subject Information Notice is transmitted for issuance on July 17, 1979.
The Information Notice should be issued to all holders of Reactor Operating 
Licenses and Construction Permits. Although a Bulletin action section was 
circulated for expedited comment on July 13, 1979 issuance of the Bulletin 
is delayed pending accumulation of information by NRR on previous 
inspections. Also enclosed is a draft copy of the transmittal letter for 
this purpose. 

                                        Norman C. Moseley, Director 
                                        Division of Reactor Operations 
                                        Office of Inspection and Enforcement

1.   Information Notice No.   
       No. 79-19 
2.   Draft Transmittal Letter 

CONTACT:  W. J. Collins 

(Draft letter to all power reactors with an operating license or 
construction permit) 

                                            Information Notice No. 79-19 

This Information Notice is provided as an early notification of a possibly 
significant matter. It is expected that recipients will review the 
information for possible applicability to their facilities. No response is 
requested at this time however licensees should be aware that the NRC is 
evaluating the issuance of a Bulletin to operating PWR's requesting 
information on previous inservice inspections of stagnant borated water 
systems and requesting inspection of systems which have not been inspected 
recently. If you have questions or comments regarding this matter, please 
contact the Director of the appropriate NRC Regional Office. 


                                        (Regional Director) 

IE Information Notice
  No. 79-19

                              UNITED STATES 
                       NUCLEAR REGULATORY COMISSION 
                          WASHINGTON D. C. 20555 
                               July 17, 1979 

                                            Information Notice No. 79-19 


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 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. 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 is being pursued. 

Information Notice No. 79-19                              July 17, 1979 
                                                             Page 2 of 2  

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. 

1.   IE Circular 76-06 
2.   List of Information 
       Notices Issued in 1979 

                                                    November 26, 1976    
                                                    IE Circular No. 76-06 



During the period November 7, 1974 to November 1, 1975, several incidents of
through-wall cracking have occurred in the 10-inch, schedule 10 type 304 
stainless steel piping of the Reactor Building Spray and Decay Heat Removal 
Systems at Arkansas Nuclear Plant No. 1. 

On October 7, 1976, Virginia Electric and Power also reported throughwall 
cracking in the 10-inch schedule 40 type 304 stainless discharge piping of 
the "A" recirculation spray heat exchanger at Surry Unit No. 2. A recent 
inspection of Unit 1 Containment Recirculation Spray Piping revealed 
cracking similar to Unit 2. 

On October 8, 1976, another incident of similar cracking in 8-inch schedule 
10 type 304 stainless piping of the Society Injection Pump Suction Line at 
the Ginna facility was reported by the licensee. 

Information received on the metallurgical analysis conducted to date 
indicates that the failures were the result of intergranular stress 
corrosion cracking that initiated on the inside of the piping. A commonality 
of factors observed associated with the corrosion mechanism were: 

1.   The cracks were adjacent to and propagated along weld zones of the 
     thin-walled low pressure piping, not part of the reactor coolant 

2.   Cracking occurred in piping containing relatively stagnant boric acid 
     solution not required for normal operating conditions. 

3.   Analysis of surface products at this time indicate a chloride ion 
     interaction with oxide formation in the relatively stagnant boric acid 
     solution as the probable corrodant, with the state of stress probably 
     due to welding and/or fabrication. 

The source of the chloride ion is not definitely known. However, at ANO-1 
the chlorides and sulfide level observed in the surface tarnish film near 
welds is believed to have been introduced into the piping during testing of 
the sodium thiosulfate discharge valves, or valve leakage. Similarly, at 
Ginna the chlorides and potential oxygen 

IE Circular No. 76-06            - 2 -                   November 26, 1976 

availability were assumed to have been present since original construction 
of the borated water storage tank which is vented to atmosphere. Corrosion 
attack at Surry is attributed to in-leakage of chlorides through 
recirculation spray heat exchange tubing, allowing buildup of contaminated 
water in an otherwise normally dry spray piping. 


1.   Provide a description of your program for assuring continued integrity 
     of those safety-related piping systems which are not frequently 
     flushed, or which contain nonflowing liquids. This program should 
     include consideration of hydrostatic testing in accordance with ASME 
     Code Section XI rules (1974 Edition) for all active systems required 
     for safety injection and containment spray, including their 
     recirculation modes, from source of water supply up to the second 
     isolation valve of the primary system. Similar tests should be 
     considered for other safety-related piping systems. 
2.   Your program should also consider volumetric examination of a 
     representative number of circumferential pipe welds by non-destructive 
     examination techniques. Such examinations should be performed generally
     in accordance with Appendix I of Section XI of the ASME Code, except 
     that the examined area should cover a distance of approximately six (6)
     times the pipe wall thickness (but not less than 2 inches and need not 
     exceed 8 inches) on each side of the weld. Supplementary examination 
     techniques, such as radiography, should be used where necessary for 
     evaluation or confirmation of ultrasonic indications resulting from 
     such examination. 

3.   A report describing your program and schedule for these inspections 
     should be submitted within 30 days after receipt of this Circular. 

4.   The NRC Regional Office should be informed within "24 hours, of any 
     adverse findings resulting during nondestructive evaluation of the 
     accessible piping welds identified above. 

5.   A summary report of the examinations and evaluation of results should 
     be submitted within 60 days from the date of completion of proposed 
     testing and examinations. 

IE Circular No. 76-06              -3-                   November 26, 1976 

     This summary report should also include a brief description of plant 
     conditions, operating procedures or other activities which provide 
     assurance that the effluent chemistry will maintain low levels of 
     potential corrodants in such relatively stagnant regions within the 

Your responses should be submitted to the Director of this office, with a 
copy to the NRC Office of Inspection and Enforcement, Division of Reactor 
Inspection Programs, Washington, D.C. 20555. 

Approval of NRC requirements for reports concerning possible generic 
problems has been obtained under 44 U.S.C 3152 from the U.S. General 
Accounting Office. (GAO Approval B-180255 (R0062), expires 7/31/77) 


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