05000382/LER-2009-005
| Docket Number | |
| Event date: | 10-19-2009 |
|---|---|
| Report date: | 05-13-2010 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3822009005R01 - NRC Website | |
REPORTABLE OCCURRENCE
On October 19, 2009 at 09:44, Waterford 3 manually tripped the reactor due to lowering condenser hotwell level caused by a stuck open reheat system [SB] moisture separator reheater (MSR) relief valve [RV] (RS 203B). Subsequently, Emergency Feedwater Actuation Signals (EFAS-1 and EFAS-2) were received due to low Steam Generator (SG) levels actuating the Emergency Feedwater (EFW) System [BA]. The condition was reported to the NRC Operations Center within four hours. The event was reportable within four hours under criteria 10CFR50.72(b)(2)(iv)(B) for a manual reactor trip of the plant to preclude receiving an automatic Reactor Protection System (RPS) [JC] trip while the reactor was critical. Additionally, the event was reportable within eight hours under criteria 10CFR50.72(b)(3)(iv) for the automatic actuation of EFAS upon low Steam Generator levels. The manual reactor trip is reportable in writing (Licensee Event Report) within,60 days in accordance with 10CFR50.73(a)(2)(iv)(A) due to the manual actuation of the RPS and due to the automatic actuation of the Emergency Feedwater system.
INITIAL CONDITIONS
Just prior to the initiating events, the plant was operating in Mode 1 at 100% power. There were no procedures being implemented specific to this condition. There were no Technical Specification Limiting Conditions of Operation specific to this condition in effect.
BACKGROUND
Waterford 3 has two MSRs. The purpose of the MSRs are to remove the moisture and to reheat the steam from the high pressure turbine [TRB] exhaust to the low pressure turbine inlet. Each MSR is provided with overpressure protection by three pilot-operated relief valves mounted on top of the shell. These relief valves discharge to atmosphere at approximately 250 psig to prevent damage to the MSR shell. The design pressure of the shell is 265 psig.
With the stuck open moisture separator relief valve (RS-203B) open, the calculated flow rate based on hotwell level loss was 2500 GPM. The maximum makeup rate to the condenser hotwell through the makeup line is approximately 1250 gpm.
EVENT DESCRIPTION
On October 19, 2009 at 09:15, while operating at 100% power moisture separator reheater relief valve, RS 203B, spuriously opened causing reactor power to increase from 100% to 100.27% Rated Thermal Power (RTP). At 09:16, operations promptly reduced main turbine generator load to restore reactor power to less than 100% RTP. At 09:17 hotwell emergency makeup valve (CMU-715) opened. At approximately 09:42, operations commenced a rapid plant shutdown. At 09:44, operations manually tripped the reactor due to lowering condenser hotwell level and entered OP-902-000 (Standard Post Trip Actions). Emergency Feedwater actuation signals EFAS-1 and EFAS-2 automatically initiated due to low Steam Generator levels, an anticipated response to the reactor trip with the plant at or near full power. Steam was isolated to the MSRs. The inventory loss through the open relief valve stopped and hotwell level recovered. The plant was maintained in Mode 3 with both Steam Generators being fed from the non-safety main feedwater [SJ] system with steam generator levels in the normal operational band for Mode 3. The EFAS actuation signals were reset. Following the event, a post trip review was performed.
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CAUSAL FACTORS
Results of condition investigations and failure modes analysis conclude that the cause of the condition was a broken pilot valve spring in RS-203B. When the pilot valve spring fractured, the pressure load on the main valve disc was lost and the main valve spring opened RS-203B.
Based on metallurgical analysis performed, the root cause for the fractured spring was that spring material had cracks introduced during the manufacturing process. The springs were made of ASTM A401 spring wire that achieves it strength by being heated to the austenitizing temperature and then quenched.
Because this steel is brittle in the as-quenched condition, it is subsequently "tempered.
ASTM A401 wire for springs is drawn to size by pulling through a die. If during the drawing process there is an interruption of the drawing lubricant at the die, the surface of the wire will be heated to the austenizing temperature, and as the wire exits the carbide "nib" or throat of the die, the body of the wire will quench this locally hot surface and will become brittle., When combined with the tensile stresses from pulling the wire through the die, the wire will sometimes develop "cross cracking" on the surface of the wire. The "cross cracking" provides initiation points for propagation of stress cracks.
In the replacement springs, the substitution of A313, type 316 stainless steel wire in place of A401 carbon steel wire eliminates the mechanism that was the root cause of the failure of the original springs that were in these valves.
CORRECTIVE ACTIONS
The carbon steel pilot springs in all six MSR relief valves were replaced with stainless steel springs during refueling outage 16. The wire that was used in the replacement springs was ASTM type 316 stainless steel. This is an austenitic stainless steel that does not harden when heated and quenched. Type 316 stainless steel hardens due to the work hardening from drawing the wire to successively smaller diameters. Because type 316 wire cannot be hardened by heating and quenching, it is immune to embrittlement that leads to "cross cracking.
SAFETY SIGNIFICANCE
The plant remained within safety limits throughout the event. The condition did not prevent the fulfillment of any safety function and did not result in a safety system functional failure (i.e. ability to shut down the reactor and maintain it in a safe shutdown condition, ability to remove residual heat, ability to control the release of radioactive material, and ability to mitigate the consequences of an accident) as defined by 10CFR50.73(a)(2)(v). The only engineered safety feature actuations were emergency feedwater actuation signals EFAS-1 and EFAS-2, which automatically initiated due to low steam generator levels. This is an anticipated response to the reactor trip with the plant at or near full power. Main feedwater maintained SG water level above the setpoint at which EFW control valves open; therefore, no injection of EFW occurred during the event.
Operations manually tripped the reactor which caused a turbine trip and steam to be isolated to the MSRs.
There were no structures, systems, or components that were inoperable at the time of the event that contributed to this condition. Since engineered safety features actuated as required, and considering that primary system parameters were maintained within acceptable limits, the safety significance of this event is considered minimal.
SIMILAR EVENTS
A search was performed for other similar reported events at Waterford 3. No similar events were identified.
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ADDITIONAL INFORMATION
Energy industry identification system (El IS) codes are identified in the text within brackets [].
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