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- 2) 0 1987 I N 86-108, Supplement 2 q p EDlSON ED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D. C.
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November 19, 1987 NRC INFORMATION NOTICE NO.86-108, SUPPLEMENT 2:
DEGRADATION OF REACTOR COOLANT SYSTEM PRESSURE BOUNDARY RESULTING FROM BORIC ACID CORROSION Addressees:
A l l holders o f operating licenses o r construction permits f o r nuclear power reactors.
Purpose :
This supplement t o Information Notice (IN)86-108 i s intended t o provide ad-dressees with additional information concerning potential problems resulting from the boric acid-induced corrosion o f f e r r i t i c steel components o f systems important t o safety. It i s expected t h a t recipients w i l l review the informa-t i o n f o r applicability t o t h e i r f a c i l i t i e s and consider actions, as appropriate, t o avoid similar problems.
notice do not constitute NRC requirements; therefore, no specific action o r w r i t t e n response i s required.
Description o f Circumstances:
However, suggestions contained i n t h i s information On August 7, 1987, a f t e r an unplanned shutdown, Salem Unit 2 was brought t o a cold shutdown condition.
Inspection teams entered the containment building t o look. f o r reactor coolant leaks t h a t would account f o r the increased radioactivity i n containment a i r t h a t was noted before the shutdown.
found boric acid crystals on a seam i n the ventilation cowling surrounding the reactor head area.
t i o n and discovered a mound o f boric acid residue a t one edge o f the reactor vessel head.
A p i l e o f rust-colored boric acid crystals 3 feet by 5 feet by 1 foot high had accumulated on the head, and a t h i n white f i l m o f boric acid crystals had coated several areas o f the head and extended 1 t o 2 feet up the control rod mechanism housings.
coolant leakage through three pinholes i n the seal weld a t the base o f the threaded connection (conoseal) for thermocouple instrumentation.
previous operating period, reactor coolant leakage had not exceeded 0.4 gallon per minute (gpm).
The team assigned t o the reactor head area The licensee then removed some o f the cowling and insula-The source of the boric acid was reactor During the 8711130008
IN 86-108, Supplement 2 November 19, 1987 I
I Page 2 o f 3 Corrosion damage t o the reactor vessel head was caused by borated water that had dripped from the ventilation supports onto the head.
The licensee found nine corrosion p i t s i n the f e r r i t i c steel vessel head.
The p i t s were 1 t o 3 inches i n diameter and 0.4 t o 0.36 inch deep.
I n the corroded area, the minimum thickness o f the head as specified by design could have been 7 inches, while the actual wall thickness was 8 inches.
Calculations performed by the licensee and Westinghouse confirmed that the affected areas s t i l l met ASME Code requirements.
Another incident o f boric acid corrosion, which occurred a t San Onofre Unit 2, was reported on August 31, 1987.
coolant temperature a t 125OF, the control room operator was attempting t o change valve positions i n the shutdown cooling system, when he found that an isolation valve i n a 10-inch pipe was stuck closed.
Personnel were sent i n t o the containment t o manually open the valve with a pipe wrench.
.During an attempt t o open the valve, the valve packing follow plate was dislodged when the carbon steel holddown bolts, corroded by previous boric acid leakage, failed.
valve packing t o extrude.
A leak o f 60 t o 100 gpm developed and 18,000 gallons o f reactor coolant spilled i n t o the containment and was subsequently pumped t o the l i q u i d radwaste system.
Five workers were contaminated.
The concentration o f radioactive gases a t the s i t e boundary reached 17 percent o f the permissible concentration f o r noble gases.
With the plant shut down and the reactor The reactor coolant system pressure, which was 350 psig, caused the Discussion:
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As a consequence o f the accelerated rate o f the boric acid corrosion observed a t the Salem plant and the extensive corrosion previously reported a t Turkey Point Unit 4 (discussed i n Information Notice 86-108, Supplement l),
Westinghouse issued l e t t e r s t o i t s customers, on o r about October'l5, 1987, which addressed the potential f o r degradation o f the reactor coolant system pressure boundary resulting from boric acid corrosion and enclosed a report e n t i t l e d "Corrosion Effects o f Boric Acid Leakage on Steel Under Plant Operat-ing Conditions - A Review o f Available Data."
The following are excerpts from t h a t report.
As a result o f the recent boric acid leakage a t reactor vessel head penetrations a t the Turkey Point 4 and Salem 2 stations, Westinghouse has reviewed available l i t e r a t u r e and has conducted certain experiments regarding the corrosion effects o f such leakage on the reactor vessel steels and stud materials.
The primary effect o f boric acid leakage that can concentrate i s 'wastage' (or general dissolution corrosion) o f both carbon steel and stainless steel.
Pitting, stress corrosion cracking (SCC),
intergranular attack, and other forms o f corrosion are not generally o f concern i n concentrated boric acid solutions a t elevated temperatures.
It should be recognized, however, t h a t the general corrosion rate (wastage) o f carbon steel can be unacceptably high under conditions that can prevail when primary coolant leaks onto surfaces and concentrates a t the temperatures that pertain t o reactor external surfaces.
I n
0 0 Q 0 0 4 0 3 4 c q i.j I N 86-108, Supplement 2 November 19, 1987 Page 3 o f 3 one series o f tests performed by Westinghouse, aerated 25 percent boric acid solutions were shown t o corrode carbon steel a t about 400 milshonth i n a 200 degrees F environment.
sol ution reduced the corrosion rate t o 250 m i 1 shonth.
Similar corrosion rates (358-418 m i 1 shonth) were obtained by dripping 15 percent boric acid a t 200 degrees F onto carbon steel surfaces a t 210 degrees F i n air.
aqueous solutions o f boric acid, when allowed t o concentrate, are highly corrosive t o carbon steel sucfaces t h a t are a t approximately 200 degrees F.
I n one series o f Westinghouse tests r e l a t i n g t o leakage o f boric acid, a mock-up o f the Inconel control rod drive mechanism (CRDM) head weld with a typical crevice geometry, was exposed t o dripping 15 percent boric acid a t 210 degrees F.
of the steel occurred (to approximately 400 mils/month),
but there was no preferential attack i n the crevice o r on the Inconel.
Deoxygenating the t e s t Both types o f experiments demonstrate that Extensive general corrosion The information provided by Westinghouse confirmed and supplemented the evi-dence that recently observed boric acid corrosion rates are greater than those t h a t were either previously known o r estimated.
t i o n programs may be warranted t o ensure t h a t adequate monitoring procedures are i n place t o detect boric acid leakage and corrosion before it could result i n significant degradation o f the reactor coolant pressure boundary.
information herein i s being provided as an early notification o f a potentially significant matter t h a t i s s t i l l under consideration by the NRC staff.
I f NRC evaluation so indicates, specific 1 icensee actions may be requested.
No specific action o r written response i s required by t h i s information notice.
I f you have any questions about t h i s matter, please contact the technical contact l i s t e d below o r the Regional Administrator o f the appropriate regional office.
A review of existing inspec-The Technic 1 Contact
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Jh-Charles E. Rossi, Director Division o f Operational Events Assessment Office o f Nuclear Reactor Regulation Sam MacKay, NRR (301) 492-8394
Attachment:
L i s t o f Recently Issued NRC Information Notices