Headquarters Daily report MAY 22, 1995 *************************************************************************** REPORT NEGATIVE NO INPUT ATTACHED INPUT RECEIVED RECEIVED HEADQUARTERS X REGION I X REGION II X REGION III X REGION IV X *************************************************************************** PRIORITY ATTENTION REQUIRED MORNING REPORT - HEADQUARTERS MAY 22, 1995 MR Number: H-95-0101 NRR DAILY REPORT ITEM GENERIC COMMUNICATIONS BRANCH/EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT OFFICE OF NUCLEAR REACTOR REGULATION Subject: N/A NRC Generic Letter 92-01, Rev. 1, Supp. 1, "Reactor Vessel Structural Integrity," issued May 19, 1995. The NRC issued this supplement to Generic Letter (GL) 92-01, Revision 1, to require that all addressees identify, collect and report any new data pertinent to analysis of structural integrity of their reactor pressure vessels (RPVs) and to assess the impact of that data on their RPV integrity analyses relative to the requirements of Section 50.60 of Title 10 of the Code of Federal Regulations (10 CFR 50.60), 10 CFR 50.61, Appendices G and H to 10 CFR Part 50, (which encompass pressurized thermal shock (PTS) and upper shelf energy (USE) evaluations) and any potential impact on low temperature overpressure (LTOP) limits or pressure-temperature (P-T) limits. Technical contacts: Edwin M. Hackett (301) 415-2751 Keith R. Wichman (301) 415-2757 PRIORITY ATTENTION REQUIRED MORNING REPORT - HEADQUARTERS MAY 22, 1995 MR Number: H-95-0102 NRR DAILY REPORT ITEM GENERIC COMMUNICATIONS BRANCH/EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT OFFICE OF NUCLEAR REACTOR REGULATION Subject: N/A NRC Generic Letter 92-01, Rev. 1, Supp. 1, "Reactor Vessel Structural Integrity," issued May 19, 1995. The NRC issued this supplement to Generic Letter (GL) 92-01, Revision 1, to require that all addressees identify, collect and report any new data pertinent to analysis of structural integrity of their reactor pressure vessels (RPVs) and to assess the impact of that data on their RPV integrity analyses relative to the requirements of Section 50.60 of Title 10 of the Code of Federal Regulations (10 CFR 50.60), 10 CFR 50.61, Appendices G and H to 10 CFR Part 50, (which encompass pressurized thermal shock (PTS) and upper shelf energy (USE) evaluations) and any potential impact on low temperature overpressure (LTOP) limits or pressure-temperature (P-T) limits. Technical contacts: Edwin M. Hackett (301) 415-2751 Keith R. Wichman (301) 415-2757 NRC Information Notice 94-61, Supp. 1, "Corrosion of William Powell Gate Valve Disc Holders," will be issued May 25, 1995. The NRC issued this supplement to Information Notice (IN) 94-61, to alert addressees to problems caused by the corrosion of carbon steel disc holders in William Powell (the vendor) gate valves and to correct an error in IN 94-61, "Corrosion of William Powell Gate Valve Disc Holders." Technical contacts: Michael J. Morgan, RII (334) 899-3386 or 3387 Geoffrey P. Hornseth, NRR (301) 415-2756 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION III MAY 22, 1995 Licensee/Facility: Notification: Illinois Power Co. MR Number: 3-95-0090 Clinton 1 Date: 05/21/95 Clinton,Illinois Dockets: 50-461 BWR/GE-6 Subject: DELAYED RESTART FOLLOWING REACTOR TRIP Reportable Event Number: 28807 Discussion: On May 16, 1995, during startup following the reactor trip described in event notification 28807, a control rod was inadvertently mispositioned. All control rods were then returned to the full in position. The licensee implemented a plant standdown and conducted training sessions on the importance of attention to detail. On May 19 the reactor was taken critical and on May 20 the generator was synchronized to the grid. There were no attention to detail problems during startup and power ascension. The event notification 28807 stated that the licensee was investigating the cause of the reactor recirculation (RR) pump downshift. The licensee determined that a power supply bus in the RR flow control valve instrument rack had shorted. This generated a variety of erroneous signals that led to the downshift. Regional Action: Information only Contact: B. CLAYTON (708)829-9602 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION III MAY 22, 1995 Licensee/Facility: Notification: Commonwealth Edison Co. MR Number: 3-95-0091 Dresden 1 2 Date: 05/22/95 Morris,Illinois Dockets: 50-010,50-237 BWR/GE-1,BWR/GE-3 Subject: TECHNICAL SPECIFICATION REQUIRED SHUTDOWN Reportable Event Number: N/A Discussion: The licensee entered Technical Specification 3.0.3 and declared an Unusual Event for both Units 1 and 2 at approximately 10:25 a.m. (EDT) on May 19, 1995. Reactor power was reduced from 100 to approximately 98 percent on both units prior to the licensee exiting TS 3.0.3 and terminating the Unusual Event at 1:10 p.m. On May 16, 1995, the licensee identified that ultrasonic testing required by ASME Code Section XI had not been properly performed on large bore branch connections in the reactor coolant system (RCS) for both units. The transducer was not in a position to examine the applicable welds. These branch connections include all hot and cold leg ECCS injection lines. Per ASME Section XI, 3 welds out of a population of 12 welds per unit were required to be ultrasonically tested every 10 years. After entering TS 3.0.3, the licensee contacted Region III and NRR for the purpose of requesting relief from the Code and terminating the plant shutdown. The licensee was informed that, per Generic Letter 91-18, if operability of the affected systems could be justified, the plant could continue to operate while pursuing Code relief. Following this discussion, the licensee exited TS 3.0.3 based on the following; all welds had acceptable radiographic and penetrant exams during construction, the systems had successful hydro-static tests, walkdowns for evidence of boric acid during startups and shutdowns showed no signs of leakage, and RCS leak rate monitoring was within acceptable limits. Regional Action: Region III had a conference call with the licensee on May 19, 1995, to discuss the details of the event and will review the licensee's planned corrective actions. Contact: WAYNE KROPP (708)829-9633