United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 86-108, Supplement 2: Degradation of Reactor Coolant System Pressure Boundary Resulting from Boric Acid Corrosion

                                                  IN 86-108, Supplement 2 

                                UNITED STATES
                        NUCLEAR REGULATORY COMMISSION
                    OFFICE OF NUCLEAR REACTOR REGULATION
                           WASHINGTON, D.C. 20555

                              November 19, 1987

Information Notice No. 86-108, SUPPLEMENT 2:  DEGRADATION OF REACTOR 
                                                  COOLANT SYSTEM  PRESSURE 
                                                  BOUNDARY  RESULTING FROM 
                                                  BORIC  ACID CORROSION 
Addressees: 

All holders of operating licenses or construction permits for nuclear power 
reactors. 

Purpose: 

This supplement to Information Notice (IN) 86-108 is intended to provide 
addressees with additional information concerning potential problems result-
ing from the boric acid-induced corrosion of ferritic steel components of 
systems important to safety. It is expected that recipients will review the 
information for applicability to their facilities and consider actions, as 
appropriate, to avoid similar problems. However, suggestions contained in 
this information notice do not constitute NRC requirements; therefore, no 
specific action or written response is required. 

Description of Circumstances: 

On August 7, 1987, after an unplanned shutdown, Salem Unit 2 was brought to 
a cold shutdown condition. 

Inspection teams entered the containment building to look for reactor 
coolant leaks that would account for the increased radioactivity in 
containment air that was noted before the shutdown. The team assigned to the 
reactor head area found boric acid crystals on a seam in the ventilation 
cowling surrounding the reactor head area. The licensee then removed some of 
the cowling and insulation and discovered a mound of boric acid residue at 
one edge of the reactor vessel head. A pile of rust-colored boric acid 
crystals 3 feet by 5 feet by 1 foot high had accumulated on the head, and a 
thin white film of boric acid crystals had coated several areas of the head 
and extended 1 to 2 feet up the control rod mechanism housings. The source 
of the boric acid was reactor coolant leakage through three pinholes in the 
seal weld at the base of the threaded connection (conoseal) for thermocouple 
instrumentation. During the previous operating period, reactor coolant 
leakage had not exceeded 0.4 gallon per minute (gpm). 

8711130008 

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                                                  IN 86-108, Supplement 2 
                                                  November 19, 1987 
                                                  Page 2 of 3 

Corrosion damage to the reactor vessel head was caused by borated water that
had dripped from the ventilation supports onto the head. The licensee found 
nine corrosion pits in the ferritic steel vessel head. The pits were 1 to 3 
inches in diameter and 0.4 to 0.36 inch deep. In the corroded area, the 
minimum thickness of 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 still met 
ASME Code requirements. 

Another incident of boric acid corrosion, which occurred at San Onofre Unit 
2, was reported on August 31, 1987. With the plant shut down and the reactor
coolant temperature at 125F, the control room operator was attempting 
to change valve positions in the shutdown cooling system, when he found that
an isolation valve in a 10-inch pipe was stuck closed. Personnel were sent 
into the containment to manually open the valve with a pipe wrench. During 
an attempt to open the valve, the valve packing follow plate was dislodged 
when the carbon steel holddown bolts, corroded by previous boric acid leak-
age, failed. The reactor coolant system pressure, which was 350 psig, caused
the valve packing to extrude. A leak of 60 to 100 gpm developed and 18,000 
gallons of reactor coolant spilled into the containment and was subsequently
pumped to the liquid radwaste system. Five workers were contaminated. The 
concentration of radioactive gases at the site boundary reached 17 percent 
of the permissible concentration for noble gases. 

Discussion: 

As a consequence of the accelerated rate of the boric acid corrosion 
observed at the Salem plant and the extensive corrosion previously reported 
at Turkey Point Unit 4 (discussed in Information Notice 86-108, Supplement 
1), Westinghouse issued letters to its customers, on or about October 15, 
1987, which addressed the potential for degradation of the reactor coolant 
system pressure boundary resulting from boric acid corrosion and enclosed a 
report entitled "Corrosion Effects of Boric Acid Leakage on Steel Under 
Plant Operating Conditions - A Review of Available Data." The following are 
excerpts from that report. 

     As a result of the recent boric acid leakage at reactor vessel head 
     penetrations at the Turkey Point 4 and Salem 2 stations, Westinghouse 
     has reviewed available literature and has conducted certain experiments
     regarding the corrosion effects of such leakage on the reactor vessel 
     steels and stud materials. 

     The primary effect of boric acid leakage that can concentrate is 
     'wastage' (or general dissolution corrosion) of both carbon steel and 
     stainless steel. Pitting, stress corrosion cracking (SCC), inter-
     granular attack, and other forms of corrosion are not generally of 
     concern in concentrated boric acid solutions at elevated temperatures. 
     It should be recognized, however, that the general corrosion rate 
     (wastage) of carbon steel can be unacceptably high under conditions 
     that can prevail when primary coolant leaks onto surfaces and 
     concentrates at the temperatures that pertain to reactor external 
     surfaces. In 
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                                                  IN 86-108, Supplement 2 
                                                  November 19, 1987 
                                                  Page 3 of 3 

     one series of tests performed by Westinghouse, aerated 25 percent boric
     acid solutions were shown to corrode carbon steel at about 400 mils/ 
     month in a 200 degrees F environment. Deoxygenating the test solution 
     reduced the corrosion rate to 250 mils/month. Similar corrosion rates 
     (358-418 mils/month) were obtained by dripping 15 percent boric acid at
     200 degrees F onto carbon steel surfaces at 210 degrees F in air. Both 
     types of experiments demonstrate that aqueous solutions of boric acid, 
     when allowed to concentrate, are highly corrosive to carbon steel sur-
     faces that are at approximately 200 degrees F. 

     In one series of Westinghouse tests relating to leakage of boric acid, 
     a mock-up of the Inconel control rod drive mechanism (CRDM) head weld 
     with a typical crevice geometry, was exposed to dripping 15 percent 
     boric acid at 210 degrees F. Extensive general corrosion of the steel 
     occurred (to approximately 400 mils/month), but there was no 
     preferential attack in the crevice or on the Inconel. 

The information provided by Westinghouse confirmed and supplemented the 
evidence that recently observed boric acid corrosion rates are greater than 
those that were either previously known or estimated. A review of existing 
inspection programs may be warranted to ensure that adequate monitoring pro-
cedures are in place to detect boric acid leakage and corrosion before it 
could result in significant degradation of the reactor coolant pressure 
boundary. The information herein is being provided as an early notification 
of a potentially significant matter that is still under consideration by the
NRC staff. If NRC evaluation so indicates, specific licensee actions may be 
requested. 

No specific action or written response is required by this information 
notice. If you have any questions about this matter, please contact the 
technical contact listed below or the Regional Administrator of the 
appropriate regional office. 


                              Charles E. Rossi, Director
                              Division of Operational Events Assessment 
                              Office of Nuclear Reactor Regulation 

Technical Contact:  Sam MacKay, NRR 
                    (301) 492-8394 

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