No Findings at Special Inspection for Brunswick Event, but Great Observations by NRC.

by Bob Meyer

Read the NRC assessment of Operator actions. Good lessons learned. For instance, the Operators did not recognizing trends for several hours.

On November 16, 2011, at 2:12 a.m., operators at Brunswick Nuclear Plant Unit 2 calculated a drywell floor drain leak rate of 5.88 gpm following several hours of gradually rising floor drain leakage during a plant startup. Technical Specification 3.4.4 A was entered requiring floor drain leakage to be restored below 5 gpm within 8 hours. At 2:53 a.m., the calculated leak rate was 10.11 gpm. At 3:01 a.m., a Notice of Unusual Event (NOUE) was declared for unidentified leakage exceeding 10 gpm. At 3:09 a.m., the licensee initiated a manual reactor scram from approximately 7 percent power. Following the scram, reactor pressure was decreased and the unidentified leak rate dropped below 10 gpm within 1 hour and was less than 5 gpm within 2 hours. The leak rate at 6:14 a.m. was 3.82 gpm with reactor pressure at 228 psig.

The NOUE was exited at 8:15 a.m. on November 16, 2011, when leakage could be maintained below 10 gpm due to decreasing pressure. The unit was cooled down and reached cold shutdown at 2:38 p.m. on November 16, 2011.

On November 17, 2011, the licensee determined that the reactor head flange leakage was due to inadequate reactor vessel head stud tensioning.

Event Follow-up - Special Inspection (93812).

Develop a timeline associated with this event.

On November 16, 2011, with Unit 2 operating in mode 2 at 7 percent power, operators responded to an increase in unidentified leakage inside of primary containment (drywell) by inserting a reactor scram and declaring an Unusual Event. Events occurred as indicated by the following timeline.

11/4/11 9:07 p.m. The Unit 2 reactor was shut down for a mid-cycle maintenance outage. The licensee shutdown to address indications of a reactor fuel leak, which requires that the reactor be disassembled and the fuel bundles individually inspected for leakage.

11/6/11 7:27 a.m.: Unit 2 entered mode 5 and reactor disassembly commenced.

11/12/11 5:00 p.m.: The Unit 2 reactor head was set during reactor reassembly after completion of the fuel inspections and bundle replacements.

11/13/11 3:13 a.m.: Initial reactor head stud measurements were completed in preparation for reactor head stud tensioning. Tensioning of the studs began. During the evolution, over the next three hours, operators of the stud tensioning pump failed to apply proper pressure to the stud tensioning devices due to a lack of understanding of the pump pressure indication. Approximately 1,300 psi was applied to the tensioning devices vice the required 13,000 psi.

11/13/11 6:39 a.m.: The reactor head stud tensioning procedure was completed, including final stud elongation measurements. During review of the elongation data, maintenance personnel mistakenly interpreted the readings to conclude that sufficient stud elongation was obtained. However, the data actually showed that the studs were not sufficiently elongated.

11/13/11 7:35 a.m.: The licensee declared the Unit 2 reactor to be in Mode 4.

11/15/11 2:00 a.m.: The Unit 2 reactor Mode switch was placed in STARTUP/HOT STANDBY and the licensee declared the reactor to be in mode 2.

11/15/11 2:58 a.m.: Unit 2 reactor control rod withdrawal commenced.

11/15/11, approximately 8:00 p.m.: Based on the frequency and amount of the drywell sump pumping, the operating crew determined that the drywell floor drain (DWFD) leakage was abnormal. The calculated 4-hour average leak rate from 4:00 p.m. to 8:00 p.m. was 0.1 gpm, indicating that significant leakage started late in the 4 hour period between 4:00 p.m. and 8:00 p.m. The Outage Control Center (OCC) was notified of the abnormal leakage. The operating crew attributed the increased DWFD leakage to the 2-B32-F031B valve, which was noted to have packing leakage during the previous startup. The OCC prepared to send personnel into the drywell to backseat the valve.

11/16/11 12:00 a.m.: Backseating of 2-B32-F031B was complete. The midnight 4-hour leakage rate was calculated to be 3.99 gpm. Startup activities were paused to determine the effect of backseating 2-B32-F031B and to further investigate the cause of the increased DWFD leakage.

11/16/11 12:54 a.m.: OCC and Operations management started preparations for a second drywell entry to assess the leakage. A chemistry sample of the DWFD sump was also planned to assess the origin of the leak.

11/16/11 2:12 a.m.: DWFD leakage rate was calculated to be 5.88 gpm based on the cycling of the DWFD sump pumps. The crew declared the TS LCO for unidentified leakage inside of the primary containment greater than 5 gpm to be in effect.

11/16/11, approximately 2:30 a.m.: Personnel entered the drywell, noted water dripping from equipment and the drywell walls, and reported to the OCC that the drywell atmosphere was “very humid.”

11/16/11 2:35 a.m.: The 92’ elevation of the drywell (highest elevation) temperature was observed to be 240 degrees F by the operations crew. The normal temperature of the 92’ elevation is approximately 200 degrees F. Operators recognized that this was an indication of a leak in the upper portion of the drywell, and not a leak from the 2-B32-F031B valve. Operators started making preparations to shut down the reactor.

11/16/11 2:53 a.m.: The DWFD leak rate was calculated to be 10.11 gpm based on the cycling of the DWFD sump pumps. The Operations Shift Manager ordered the evacuation of personnel from the drywell.

11/16/11 3:01 a.m.: Operators declared an Unusual Event due to unidentified leakage exceeding 10 gpm.

11/16/11 3:09 a.m.: After a short operations crew brief, the Unit 2 reactor was scrammed, and plant cooldown and depressurization commenced.

11/16/11 3:09 a.m. – 2:38 p.m.: The Unit 2 reactor was cooled down and depressurized. DWFD leakage lowered as reactor pressure was reduced. DWFD leakage was calculated to be 5.27 gpm at 4:30 a.m. and 3.82 gpm at 6:14 a.m..

11/16/11 2:38 p.m.: Unit 2 entered mode 4. The DWFD leakage rate in mode 4 was approximately 0.13 gpm.

11/17/11, approximately 1:00 p.m.: During trouble-shooting activities on the refueling floor, maintenance personnel were able to turn several of the reactor head retaining nuts by hand, indicating improper reactor head stud tensioning. The licensee declared the Unit 2 reactor to be in mode 5.

.2 Assess the ability of the reactor vessel to meet its design basis functions with the asfound condition.

The inspectors interviewed licensee training personnel and found that a formal refueling training qualification for reactor vessel disassembly and reassembly had not been conducted since 2000 per lesson plan ME501B. Maintenance began conducting preoutage information training; however this was not tracked as a formal qualification. Nine of the twelve personnel who performed reactor vessel reassembly on November 13, 2011, did not have the formal qualification to perform the work. Subsequent investigation by the licensee revealed that this qualification training for refueling personnel is still being conducted at other Progress Energy plants and that the qualification needs to remain active. Procedure TRN-NGCC-1000, Conduct of Training, required training be conducted per the Biennial Period Training Matrix. The training matrix required that refueling floor personnel receive initial qualification (MB81) for reactor vessel reassembly per Lesson Plan ME501B. However, the qualification for ME501B has not been provided since 2000.

The inspectors noted that the licensee conducted informal just-in-time training prior to the Fall 2011 Unit 2 outage. This training covered reactor vessel disassembly and reassembly but did not specifically focus on stud tensioning. A separate SEMS III table top training session was conducted prior to the Spring 2011 Unit 2 refueling outage to describe operation of the SEMS III device for measuring stud elongation, but was not conducted for the Fall 2011 outage. Seven of the mechanics who performed the reactor vessel reassembly during the Fall 2011 Unit 2 outage received all the just-in-time training and two received the SEMS III table-top training. The remainder did not. The use of the SEMS III measuring device was introduced and successfully used during the Spring 2011 Unit 2 refueling outage. However, the use and interpretation of the tool was not understood during the Fall 2011 outage which resulted in measurement errors that led to the failure of the maintenance crew to recognize that the reactor vessel studs were not properly tensioned. The inspectors also determined that the QC inspector for the job did not receive any specific training related to the use of the SEMS III measuring device.

In conclusion, the licensee did not conduct a specific qualification training course as required by procedure TRN-NGCC-1000, Conduct of Training. Personnel received a combination of general mechanical training and some just-in-time training prior to the Fall 2011 Unit 2 outage, however, this informal training conducted by the licensee did not prevent the refueling team from inadequately tensioning the reactor vessel studs and incorrectly performing stud elongation measurements. The licensee’s failure to follow TRN-NGCC-1000, Conduct of Training is documented in section 4OA5.1 of this report.

Assessment: Operators appropriately declared an Unusual Event when DWFD leakage exceeded 10 gpm.

Overall Operator Assessment: No procedural, TS, or regulatory requirements obligated the operators to shutdown the reactor. Plant procedures and guidance required the operators to identify the cause of the increased leakage. TS 3.4.4 would have required a shutdown eventually if leakage had remained above 5 gpm. Operators made the decision to shut down the reactor when it became apparent that leakage was rapidly increasing.

Operators were actively pursuing identification of the leak throughout the night. Plant Management, including the Operations Manager, was present and actively engaged in decision making. At approximately 7:00 p.m., an automatic start of the DWFD sump pump made operators aware of increased leakage. Since the 8:00 p.m. 4-hour leak rate calculation was 0.1 gpm, a significant increase in leakage occurred between 1600 and 7:00 p.m. on November 15, 2011. Prior to 4:00 p.m., there were no signs of significant DWFD leakage. The small amount of DWFD sump ingress prior to 4:00 p.m. was attributed to condensed atmospheric humidity due to the drywell being ventilated with outside air, and a known packing leak on the B recirculation loop discharge valve.

These were reasonable assumptions. Operators were aware that the plan to backseat the valve was being executed by maintenance and the OCC. This valve was backseated at approximately 11:30 p.m.. The 3.99 gpm calculation for DWFD leakage at 12:00 a.m. was a 4-hour average and included 3.5 hours when the recirculation discharge valve was not backseated. The 1:00 a.m. leakage calculation of approximately 4.2 gpm was for the hour from 12:00 a.m. to 1:00 a.m.. This signaled the operators that the leakage was not due to the recirculation pump discharge valve. At 2:12 a.m., the leakage rate was calculated to be 5.88 gpm. At 2:35 a.m., DW 92’ elevation temperatures were noted to be above normal at 240 degrees F and the operators realized that steam was leaking on the 92’ elevation of the drywell. During the time between 1:00 a.m. and 2:35 a.m., operations directed a second DW entry to identify the leak location and to obtain a sump sample for chemical analysis.

Overall operator performance in the identification of the leak was not optimal. For example, slowly rising trends of drywell airborne radioactivity and 92’ elevation temperatures were not recognized for several hours. Several operators stated that because the leak from the recirculation pump discharge valve was identified previously, and because unidentified DWFD leakage during a reactor startup was not uncommon at Brunswick, they weren’t rigorously pursuing other potential causes. However, inspectors noted that operators were attentive to the magnitude and progression of the leak, and took sufficient actions to ensure reactor safety.

Licensee personnel:

M. Annacone, Site Vice President

A. Brittain, Security Manager

J. Burke, Engineering Director

P. Dubrouillet, Training Manager

C. Dunsmore, Shift Operations

K. Gerald, Maintenance Manager

S. Gordy, Operations Manager

L. Grzeck, Lead Engineer - Technical Support

K. Hill, Control Room Supervisor

R. Ivey, Nuclear Oversight Services Manager

J. Johnson, Environmental and Radiological Controls Manager

P. Mentel, Support Services Manager

J. Miller, Operation Shift Manager

A. Pope, Licensing and Regulatory Affairs Supervisor

T. Sherrill, Technical Support

NRC personnel:

Randall A. Musser, Chief, Reactor Projects Branch 4, Division of Reactor Projects Region II