05000249/LER-2002-005
| Dresden Nuclear Power Station Unit 3 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2492002005R01 - NRC Website | |
LICENSEE EVENT REPORT (LER) and to the Paperwork Reduction Project (3150-0104), Office Of Management display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Dresden Nuclear Power Station Unit 3 05000249 NUMBER NUMBER (If more space is required, use additional copies of NRC Form 366A)(17)
A. Plant Conditions Prior to Event:
Unit: 03 � Event Date: 07-05-2001 Reactor Mode: 1 � Mode Name: Run Reactor Coolant System Pressure: 1000 psig Event Time: 1006 CDT Power Level: 100 percent
B. Description of Event:
On July 19, 2001, it was identified that two out of four base plate anchors for support M-1187D-80 were partially pulled out of the concrete on the underside of the 517'-6" slab, above the torus. Engineering performed an immediate extent of condition walk-down of the HPCI [BJ] pump discharge piping that could be accessed from the torus catwalk area. The engineers inspected the adjacent supports for similar damage. No other visible damage to accessible supports was identified. A corrective action program condition report (CR) was written and a prompt operability determination was completed.
On July 24, 2001, the final Operability Determination 01-031 was completed in accordance with the station's operability determination process and NES-MS-03.2, "Evaluation of Discrepant Piping and Support Systems." It concluded that the system was operable for design basis loads with the existing support damage. The conclusion of the operability determination was based on the past experience and practice that systems could be considered operable even with the failure of one support due to conservatism in design.
A work request (WR) was written to repair the support and was identified as part of the operability determination.
The WR completion was scheduled for November 1, 2001 based on the normal work control scheduling process.
Subsequently, the WR Screening Committee assigned the work request a lower priority, and the repair was rescheduled for a later date.
On July 26, 2001, a management review committee screened the CR and an apparent cause evaluation (ACE) was assigned to engineering to investigate the cause of the failed support. On August 24, 2001, the ACE was completed and approved. The ACE stated, "It is likely that the anchor deficiencies are a result of a transient (water hammer) possibly associated with the Unit 3 scram, or could be the result of another event." The ACE also stated the following corrective action: "The Dresden water hammer subject matter expert (SME) should perform a water hammer evaluation to determine if any further actions are required." This action was tracked by an action tracking item (ATI) with a due date of December 7, 2001. The completed ACE, which lacked a quality supervisory review, did not present a basis for the stated apparent cause and did not assess the impact of its conclusion on the open operability determination.
On September 26, 2001, the NRC Regional Inspector expressed concerns regarding the adequacy of the operability determination. The concerns were: (a) the failed support was not repaired, (b) there were no corrective actions in place to prevent the piping loads necessary to fail the support from recurring, (c) the system had not been vented, and (d) the Operability Determination did not consider any transient that might have failed the support.
Following additional walkdowns, on September 29, 2001, Operability Determination 01-031 was revised to remove credit for the two intact anchors on the damaged support The supports were repaired on September 30, 2001, and Unit 3 HPCI pump discharge piping was vented at the high point vent per the venting procedure. Air was LICENSEE EVENT REPORT (LER) and to the Paperwork Reduction Project (3150-0104), Office Of Management display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Dresden Nuclear Power Station Unit 3 05000249 NUMBER NUMBER (If more space is required, use additional copies of NRC Form 366A)(17) found, but no steam or hot water was noted. As a result of finding air, a review for past reportabillty and was initiated. The Unit 2 HPCI pump discharge piping was vented on October 1, 2001, and air was found. It was determined that the amount of air found in the Unit 2 HPCI system would not cause a damaging transient.
During this time period the Transient Analysis Display System (TADS) was reviewed for the July 5, 2001, scram.
Engineering personnel concluded the HPCI system had experienced a hydrodynamic transient / water hammer on July 5, 2001.
October 17, 2001, a CR was issued to document the July 5, 2001, HPCI pressure and flowrate transient data and its impact on Operability Determination 01-031. From the period between October 2001 and October 2002, numerous communications between Exelon and the NRC were conducted. Although the support had been repaired and the discharge piping vented, engineering performed detailed analyses assuming the pipe support was in a degraded condition and the system not properly vented in order to determine past reportability of the condition.
Subsequently, DNPS has been unable to demonstrate through analysis that the Unit 3 HPCI piping and supports would have met operability evaluation acceptance criteria following an additional automatic initiation. Thus, it has been concluded that the Unit 3 HPCI system was inoperable for the period following the hydrodynamic transient / water hammer on July 5, 2001, until September 30, 2001.
C. Cause of Event:
Two root causes were identified as a result of this event. One root cause addressed the cause of the damage to the support and the other addressed the station's management of the issue. The root cause of the HPCI support failure was a hydrodynamic transient / water hammer during the system actuation on July 5, 2001, due to air pockets and steam voids in the HPCI pump discharge piping. The cause of the air pockets was due to inadequate venting of the system. The root cause of the management issues was the failure of Design Engineering to evaluate the HPCI operability issue from the proper safety perspective because the focus was on demonstrating operability and not on recognizing the extent of the degraded condition, which resulted in untimely corrective actions. Corrective actions to prevent recurrence are: (1) to implement modifications and revise appropriate procedures as required, to ensure that air/steam voids do not recur and (2), to develop and implement a Design Engineering Excellence Plan to address the management issues and knowledge deficiencies.
D. Safety Analysis:
This event has been determined to be of low to moderate safety significance. This result is based on an assumption that the HPCI discharge piping stresses would exceed operability limits if HPCI injected during the period of approximately 80 days during which the support was degraded and the potential for an additional hydraulic transient existed. The Automatic Depressurization, Isolation Condenser [BL], Low Pressure Coolant Injection [BO], and Core Spray [BM] systems were operable and available, except for short periods during maintenance (within Technical Specification compliance), to provide reactor pressure / inventory control under postulated design basis accident conditions.
LICENSEE EVENT REPORT (LER) and to the Paperwork Reduction Project (3150-0104), Office Of Management display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Dresden Nuclear Power Station Unit 3 05000249 NUMBER NUMBER (If more space is required, use additional copies of NRC Form 366A)(17)
E. Corrective Actions:
The Unit 3 HPCI supports M-1187D-80 and M-1187D-83 were repaired The Unit 2 and 3 HPCI pump discharge piping in the X-Area, torus Area and HPCI pump room was vented.
Appropriate station procedures were revised to include all high point vents when venting and require venting of the HPCI pump discharge piping when lined up to the CCST.
A modification was performed on Unit 3 to eliminate potential steam voids on HPCI initiation by adding a pressure permissive on the injection valves. A similar modification will be performed on Unit 2.
Developed and implemented a Design Engineering Plan to address the knowledge and management deficiencies in the areas of problem solving, performance of operability evaluations, walk-downs, recognizing safety significance of issues without being prompted by outside agencies, identification and resolution of issues, communication, and issue management.
NES-MS-03.2 "Evaluation of Discrepant Piping and Support Systems," was revised to recommend expediting repair when the cause of the degradation is unknown and to include guidance on when to consider cause as an integral element of operability determinations.
Open operability determinations were reviewed for extent of condition.
Performance managed individuals involved on challenging the quality of ACE's, Operability Determinations and being aware of compensatory actions. Reinforced expectations regarding ACE quality and operability determinations.
Developed guidance on how to properly evaluate and respond to NRC concerns and presented as a discussion topic at the Engineering Support Personnel Training session.
A post-transient system walk-down process was developed with a standard walk-down checklist and appropriate review by engineering management.
F. Previous Occurrences:
HPCI Water Hammer Events Licensee Event Report 1989-029-05, "Elevated High Pressure Coolant Injection (HPCI) Discharge Piping operation at 100 percent power, an evaluation of the High Pressure Coolant Injection (HPCI) system discharge piping revealed discharge piping temperatures as high as 275 degrees Fahrenheit. The Unit 2 HPCI system was declared inoperable due to the possibility of a steam void existing within the discharge piping. A seven day Limiting Condition for Operation (LCO) was entered in accordance with Technical Specification 3.5.C.
Subsequently, on October 31, 1989, during normal Unit 3 operation at 100 percent power, the Unit 3 HPCI system was also declared inoperable due to the discharge piping temperatures being as high as 256 degrees Fahrenheit 133), U.S Nuclear Regulatory Commission, Washington, DC 20555-0001, LICENSEE EVENT REPORT (LER) and to the Paperwork Reduction Project (3150-0104), Office Of Management display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Dresden Nuclear Power Station Unit 3 05000249 NUMBER NUMBER (if more space is required, use additional copies of NRC Form 366A)(17) the HPCI temperature increase was system back leakage into the HPCI discharge piping on both Units 2 and 3 from the HPCI pump discharge valve MOV 2(3)-2301-8 and HPCI pump discharge check valve 2(3)-2301-7. The corrective actions associated with this event would not have prevented the event described in this LER.
Leakage." On Unit 3, piping supports were discovered damaged on the Gland Seal Condenser (GSC) subsystem following an April 8, 1992 event. The investigation concluded that the degradation may have occurred during HPCI overspeed testing, when the only source of cooling water supply to the GSC was secured in order to maintain the proper oil temperatures. Without cooling water flow, steam voids would form in the GSC tubes, and a water hammer would occur once the cooling water pump was restarted.
Water Hammer." On April 11, 1992, at 1540 hours0.0178 days <br />0.428 hours <br />0.00255 weeks <br />5.8597e-4 months <br />, several Unit 2 supports on its HPCI GSC subsystem were determined to be outside of Final Safety Analysis Report (FSAR) allowable stress limits. These damaged supports were found subsequent to the Unit 3 April 8, 1992 event described above.
The corrective actions associated with both GSC incidents would not have prevented the event described in this
LER
Isolation Condenser Water Hammer Event On January 8, 2002, while performing DOS 0010-16, "Unit 2(3) Isolation Condenser Safe Shutdown Valve Operability," Operations noted a loud noise while manually opening the Isolation Condenser condensate return inboard isolation valve, 3-1301-3. During the next eight days, extensive investigation and analyses were performed to determine the cause and effects of the loud noise (water hammer). The cause was determined to be flashing of the hot water trapped between the Isolation Condenser condensate return isolation valves, 3-1301-3 and 3-1301-4, serving as a pressure source to drive fluid into the two 12 inch Isolation Condenser steam headers The water hammer occurred when the two fluid fronts collided in the 12 inch horizontal steam headers. The effects of the water hammer were damage to one support, degradation and shearing of pass plate bolts internal to the isolation condenser, and bowing of the pass plates. The corrective actions to this event included repairs to the degraded components (support and pass plates) and revisions to procedures. Repairs to all degraded components were completed, a thermal performance test for the Isolation Condenser was satisfactorily completed, and the Isolation Condenser was declared Operable on January 17, 2002. Two causes were identified as a result of this event. One cause was inadequate design, in that the current design does not provide instrument indication (pressure or temperature) for the volume between condensate return isolation valves 3-1301-3 and 3-1301-4. The second cause was inadequate procedures. The procedures did not provide adequate instructions to assure proper pressure equalization across the Isolation Condenser condensate return isolation valve 3-1301-3, prior to valve opening. The corrective actions associated with this event would not have prevented the event described in this LER.
G. � Component Failure Data:
N/A