Information Notice 1979-19, Pipe Cracks in Stagnant Borated Water Systems at PWR Plants

Pipe Cracks in Stagnant Borated Water Systems at PWR Plants

ML031180155
Person / Time
Site:	05000000
Issue date:	07/17/1979
From:	NRC/IE
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References
IN-79-019, NUDOCS 7907230164
Download: ML031180155 (7)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, D. C. 20555 July 17, 1979 IE Information Notice No. 79-19 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 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.

7907230164

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IE Information Notice No. 79-19 July 17, 1979 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 environ- ments. 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.

Enclosures:

1. IE Circular 76-06

2. List of Information

Notices Issued in 1979

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November 26, 1976 IE Circular No. 76-06 STRESS CORROSION CRACKS IN STAGNANT, LOW PRESSURE STAINLESS

PIPING CONTAINING BORIC ACID SOLUTION AT FWR' 6

DESCRIPTION OF CIRCUMSTANCES

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 Bulldlng Spray and Decay Heat

• Removal Systems at Arkansas Nuclear Plant No. 1.

On October 7, 1976, Virginia Electric and Power also reported through- wall cracking in the 1.0-inch schedule 40 type 304 stainless discharge

piping of the "A" recirculation spray lhcat exchanger at Surry Unit

No. 2. A recent inspection of Unit 1.Containnent 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 Sefety Injection Pump

Suction Line at the Ginna facilitv was rpnnrrtd hv rh. llrnnA..

Jnfnfiat1nn rorpivoct nn thp tnpt:4ll.fg4rzra l mannlyd4.Q ectrnnAvr- te sto

indicates that the failures were the result of intergranular stress

corrosion cracking that initiated on tie inside of the piping. A

commonality of factors observed associated with the corrosion mechanism

were:

1. flu CLAIV. were adjacent to and propagated along weld zones of tne

thin-walled low pressure piping, not part of .the reactor coolant

system.

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

_ 2 - November 26, 1976 lE Circular No. 76-06 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.

ACTION TO BE TAKEN BY LICENSEE:

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

modsb at. aLut_% %j[ WiLeL

oprOy. X1IcLUU2L1 tI:1LC zcL:+/-LLt;u+/-4L.1or 9 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 shoutl also consider volumetfic 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 inspec- tions should be submitted within 30 days after receipt of this

Circular.

4. The NRC Regional Office should be informed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, of any adverse findings resulting during noude!;tructiva

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.

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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 piping.

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iour responses atnUUa ve tUiUtvt.LLLc LtU LLIW "JLL-:LVL UC

with a copy to the NRC Office of Inspection anad Enforcement, Division

of Reactor Inspection Programs, Washington, D.C. 20555.

Approval of NRC requirements for reports concerning possible generic

problems has been obtained tinder 44 U.S.C 3152 from the U.S. General

Accounting Office. (GAO Approval B-180255 (R0062), expires 7/31/77.)

IE Information Notice No. 79-19 Enclosure

July 17, 1979 LISTING OF IE INFORMATION NOTICES

ISSUED IN 1979 Information Subject Date Issued To

Notice No. Issued

79-01 Bergen-Paterson Hydraulic 2/2/79 All power reactor

Shock and Sway Arrestor facilities with an

OL or a CP

79-02 Attempted Extortion - 2/2/79 All Fuel Facilities

Low Enriched Uranium

79-03 Limitorque Valve Geared 2/9/79 All power reactor

Limit Switch Lubricant facilities with an

OL or a CP

79-04 Degradation of 2/16/79 All power reactor

Engineered facilities with an

Safety Features OL or a CP

79-05 Use of Improper Materials 3/21/79 All power reactor

in Safety-Related Components facilities with an

OL or CP

79-06 Stress Analysis of 3/23/79 All Holders of

Safety-Related Piping Reactor OL or CP

79-07 Rupture of Radwaste Tanks 3/26/79 All power reactor

facilities with an

OL or CP

79-08 Interconnection of 3/28/79 All power reactor

Contaminated Systems with facilities with an

Service Air Systems Used OL and Pu Processing

As the Source of Breathing fuel facilities

Alr

79-09 Spill of Radioactively 3/30/79 All power reactor

Contaminated Resin facilities with an

OL

79-10 Nonconforming Pipe 4/16/79 All power reactor

Support Struts facilities with a

CP

79-11 Lower Reactor Vessel Head 5/7/79 All holders of Reactor

Insulation Support Problem OLs and CPs

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IE Information Notice No. 79-19 Enclosure

July 17, 1979 LISTING OF IE INFORMATION NOTICES

ISSUED IN 1979 Information Subject Date Issued To

Notice No. Issued

79-12 Attempted Damage to New 5/11/79 All Fuel Facilities

Fuel Assemblies Research Reactors, and Power Reactors

with an OL or CP

79-13 Indication of Low Water 5/29/79 All Holders of Reactor

Level in the Oyster Creek OLs and CPs

Reactor

79-14 NRC Position of Electrical 6/11/79 All Power Reactor

Cable Support Systems Facilities with a

CP

79-15 Deficient Procedures 6/7/79 All Holders of Reactor

OLs and CPs

79-16 Nuclear Incident at Three 6/22/79 All Research Reactors

Mile Island and Test Reactors

with OLs

79-17 Source Holder Assembly Damage 6/20/79 All Holders of Reactor

Damage From Misfit Between OLs and CPs

Assembly and Reactor Upper

Grid Plate

79-18 Skylab Reentry 7/5/79 All Holders of Reactor

OLs