Headquarters Daily Report SEPTEMBER 22, 1997 *************************************************************************** REPORT NEGATIVE NO INPUT ATTACHED INPUT RECEIVED RECEIVED HEADQUARTERS û REGION I û REGION II û REGION III û REGION IV û PRIORITY ATTENTION REQUIRED MORNING REPORT - HEADQUARTERS SEP. 22, 1997 MR Number: H-97-0114 NRR DAILY REPORT ITEM GENERIC COMMUNICATIONS NRC Information Notice 97-72, "Potential for Failure of the Omega Series Sprinkler Heads," dated September 22, 1997. The NRC is issuing this information notice to alert addressees to a potential for failure of the Omega Series sprinkler heads manufactured by Central Sprinkler Company of Lansdale, Pennsylvania. Technical contacts: E. A. Connell, NRR 301-415-2838 P. W. Lain, NMSS 301-415-6317 NRC Information Notice 97-73, "Fire Hazard in the Use of a Leak Sealant," dated September 23, 1997. The NRC is issuing this information notice to alert addressees to a potential fire hazard in the use of a leak sealant. Technical contacts: Geoffrey P. Hornseth, NRR 301-415-2756 Joseph I. Tapia, RIV (817) 860-8243 _ REGION I MORNING REPORT PAGE 2 SEPTEMBER 22, 1997 Licensee/Facility: Notification: Public Service Electric & Gas Co. MR Number: 1-97-0051 Hope Creek 1 Date: 09/22/97 Hancocks Bridge,New Jersey SRI PC Dockets: 50-354 BWR/GE-4 Subject: PINHOLE LEAKS IN ASME CLASS I CORE SPRAY INJECTION LINE WELD Reportable Event Number: 32962 Discussion: On September 19, 1997, with the Hope Creek plant shutdown in Operational Condition 5 for a refueling outage, the licensee identified unisolable through wall leakage on a 12 inch diameter ASME Code Class I core spray system pipe in the drywell. Subsequent ultrasonic non-destructive examination (NDE) determined that the leakage was from three separate pinholes through the weld material at the A core spray nozzle safe-end weld, fabricated by General Electric. Based on initial evaluation, the licensee does not believe the leaks are the result of intergranular stress corrosion cracking. Total leakage was approximately 1 drop/second with a ten foot static head of water. The leak location is between the reactor pressure vessel wall and the bioshield. The nozzle is located ten feet above the top of active fuel. The licensee stated that this weld was examined during the last refueling outage with no indications noted. Additionally, similar examinations conducted following the leak discovery were unsuccessful at identifying the flaws. Employment of more sophisticated NDE than required by the Hope Creek inservice inspection (ISI) program was necessary to identify and validate the pinhole leaks. Repair options are being considered and root cause analysis is underway. A General Electric evaluation of the weld flaws indicates that the structural condition of the core spray line is acceptable for reactor cavity flood up and subsequent core alterations. A full core offload is scheduled to begin today. This offload was previously planned in order to conduct detailed core shroud and other invessel inspections mandated before completion of the ten year ISI interval. Regional Action: Resident inspector followup. Additionally, a previously planned regional initiative ISI inspection is scheduled to begin September 29, 1997. Contact: Scott A. Morris (609)935-3850 James Linville (610)337-5129 _ REGION II MORNING REPORT PAGE 3 SEPTEMBER 22, 1997 Licensee/Facility: Notification: Georgia Power Co. MR Number: 2-97-0073 Hatch 1 2 Date: 09/22/97 Baxley,Georgia Dockets: 50-321,50-366 BWR/GE-4,BWR/GE-4 Subject: FUEL SPACER FABRICATION NOT TO DRAWING SPECIFICATIONS Discussion: On September 9, 1997, General Electric (GE) personnel identified and informed the licensee that some Hatch fuel spacers on GE fuel fabricated at the GE facility in Wilmington, NC, did not meet GE's drawing specification tolerance. The fuel spacer tab angle for some Hatch fuel was fabricated at an incorrect degree. GE determined that the affected fuel includes some fuel spacer tabs for fuel already installed in Hatch Unit 2 and new fuel fabricated for the upcoming Hatch Unit 1 refueling outage. GE is reviewing the fabrication issue to determine the root cause and whether or not a 10 CFR Part 21 was warranted. GE conducted an evaluation of the fuel spacer tab angle error for Hatch Units 1 and 2 and determined that there was no impact, (i.e., Safety Limit Minimum Critical Power Ratio, Operating Limit Minimum Critical Power Ratio or fuel assembly pressure drop). Regional Action: For information only. Contact: PIERCE H. SKINNER (404)562-4520 _ REGION III MORNING REPORT PAGE 4 SEPTEMBER 22, 1997 Licensee/Facility: Notification: Consumers Power Co. MR Number: 3-97-0098 Big Rock Point 1 Date: 09/22/97 Charlevoix,Michigan INSPECTION Dockets: 50-155 BWR/GE-1 Subject: STUCK FUEL BUNDLES RECONSTRUCTED AND CORE DEFUELED Discussion: The two stuck fuel bundles were reconstructed on September 19-20, 1997, and moved to the spent fuel pool (SFP). At 5:56 p.m. EDT, on September 20, 1997, all of the fuel had been removed from the reactor core. One fuel bundle was reconstructed at 6:47, September 19, 1997, and the other fuel bundle was reconstructed at 3:14, September 20, 1997. The force to lift the reconstructed bundles was about 430 pounds, the weight of each fuel assembly. These activities completed emptying the core of fuel and marked the beginning of decommissioning activities at Big Rock Point. The force to lift the empty cages was about 1000 pounds. The cages were also stored in the SFP. Regional Action: The senior resident inspector monitored the reconstruction and core unloading. Contact: BRUCE BURGESS (630)829-9629 _ REGION IV MORNING REPORT PAGE 5 SEPTEMBER 22, 1997 Licensee/Facility: Notification: Entergy Operations, Inc. MR Number: 4-97-0072 River Bend 1 Date: 09/22/97 St Francisville,Louisiana UPDATE TO MORNING REPORT 4-97-0071 Dockets: 50-458 BWR/GE-6 Subject: UNEXPECTED INCREASE IN REACTOR COOLANT TEMPERATURE (UPDATE) Discussion: On September 13, 1997, the licensee inadvertently re-entered Mode 3 from Mode 4 while shutdown cooling was secured. Morning Report 4-97-0071 contains details of the occurrence. In response, the licensee initiated a Significant Event Response Team (SERT) to determine the cause of the event, evaluate the impact on plant conditions, and determine safety significance. The SERT preliminarily determined that the average RCS temperature exceeded 200 degrees Fahrenheit for approximately 10 minutes. During this period, plant staff calculated that the reactor coolant reached a maximum saturation temperature of approximately 244 degrees Fahrenheit in the core. The SERT determined the root cause of this problem was related to inadequate management of changes to equipment and procedures. Decisions were made based on confidence that the normal methods of monitoring temperatures at the shutdown cooling heat exchanger inlet or the Reactor Water Cleanup (RWCU) system were accurate and representative of RCS average temperature. Consideration was not adequately given to the limited circulation provided by the Reactor Water Cleanup System, and the resultant stratification in the reactor vessel. Plant staff did not consult available "time to boil" information when determining the allowable time that shutdown cooling could be secured while placing Alternate Decay Heat Removal (ADHR) in-service. The inspectors noted that the licensee secured shutdown cooling to start the ADHR alignment only 36 hours after plant shutdown. The inspectors noted that the licensee focused on the two-hour outage time for shutdown cooling allowed by Technical Specifications, rather than the shorter time to boil due to decay heat conditions. In conjunction with the time to boil, the amount of time required to place ADHR in-service caused the inadvertent mode change. The licensee issued an operational event notice to the industry which discussed this event. They are developing actions to adjust training and procedures related to the maintenance of shutdown cooling based on lessons-learned. They are also evaluating their post-modification testing process for weaknesses in procedure validation. Regional Action: A Special Inspection of the causes and circumstances of this event has been initiated. Contact: E. E. Collins (817)860-8291 W. F. Smith (504)635-3193 _