05000293/LER-2013-004
| Pilgrim Nuclear Power Station (Pnps) | |
| Event date: | 4-14-2013 |
|---|---|
| Report date: | 6-18-2013 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 2932013004R00 - NRC Website | |
BACKGROUND
Steam Seal System Discussion:
The shaft sealing system seals the turbine against two kinds of leakage: air in-leakage to the turbine casing and, steam leakage out of the turbine casing. Each kind of leakage could occur at different operating conditions if the shaft glands were not sealed.
The source of steam pressure for the shaft steam sealing system is main steam line "A". The system components supply sealing steam, provide steam/air-leakage boundaries, condense leaking steam, and exhaust the non-condensable. Besides sealing the turbine stages, the system also seals the control valves, stop valves, bypass valves, and combined intermediate intercept and stop valves.
The steam seal regulator automatically maintains the seal header pressure at 3 to 4.5 psig. A motor operated valve (MO-S-2) bypasses the regulator for start-up or backup operation. In the event the steam seal regulator malfunctions during power operation and excessive steam seal header pressure results, pressure may be maintained by opening unloading valve MO-B. Also, relief valves PSV-3195/6/7/8 protect the system from over pressurization. The lift pressure for PSV-3195 and 3196 is 20 psig as sensed on the steam seal header. PSV- 3197 and 3198, located immediately downstream of the steam seal header bypass valve, MO-S-2 are set to lift at 120 psig. The steam seal header pressurization source (either the regulator or the bypass line) determines the relief valve(s) that have the potential to lift in response to the steam line high pressure. PSV-3197 and 3198 are located on an 8" line downstream of MO-S-2 that is separated from the steam seal loads by 30' long 4" diameter piping that acts like a restricting orifice to limit the pressure sensed by Steam Seal Pressure Indicator PI-3160 and supplied to the steam seal loads when MO-S-2 is opened. Thus, the potential exists to lift PSV- 3197 and 3198 without lifting PSV-3195 and 3196.
EVENT DESCRIPTION
PNPS was in the process of inserting control rods with the intention of completing a "soft" shutdown for RFO 19.
At 2200 hours0.0255 days <br />0.611 hours <br />0.00364 weeks <br />8.371e-4 months <br />, the reactor was subcritical on IRM range 2 with reactor pressure at approximately 800 psig and slowly lowering. Reactor pressure and temperatures were being monitored through the use of PNPS 2.1.7, "Vessel Heatup and Cooldown". In accordance with station procedure 2.1.5, Controlled Shutdown from Power, Section F, Controlled Shutdown Without Manual Scram step [32], the RO bypassed the turbine steam seal regulator by opening the gland seal regulator bypass valve (MO-S-2) and closing the gland seal regulator supply valve (MO-S-1). When the RO initially jogged open MO-S-2, he noticed that the steam seal pressure indicator (PI-3160) quickly jumped to 9 psig then quickly settled to within its normal manual operating band of 3 to 8 psig.
Subsequent review of the event revealed that coincident with the MO-S-2 valve being opened, TE-3611 indicated a temperature increase that rose to above 307 deg F. This temperature rise indicates a steam flow path from MO-S-2 through PSV-3197/PSV-3198 to the main condenser.
At 2206, the RO completed placing MO-S-2 into service with steam seal header pressure indicating 6 psig and reactor pressure indicating 756 psig. The RO then continued to perform the actions of PNPS 2.1.7. At 2208, reactor pressure was 688 psig and lowering and the control room received the Reactor Water Level Low Alarm (+25"). At 2211 the Reactor Water Level Low Alarm (+25") clears. (The control room had been experiencing fluctuations in controlling letdown through RWCU (Reactor Water Clean Up) earlier in the shift, and the focus was placed on the operation of CV-1239, RWCU System Reject Flow Control Valve as to the cause of the alarm). At 2212, the RO recognized the reactor pressure decrease was excessive. Coincident with the identification of the excessive reactor pressure decrease, the control room received the Group I MSIV Scram Bypass Alarm (576 psig). The Control Room Supervisor (CRS) updated the crew that Reactor pressure was lowering quicker than expected. The operation's crew then began evaluating the excessive reactor pressure decrease by investigating main turbine mechanical hydraulic control system indications, main steam safety relief valve tailpipe temperature indication, and steam supply lineups to the offgas system, steam jet air ejectors, and augmented offgas system jet compressors. Main steam safety relief valve SR-203-3B was initially suspected as a potential cause because of the known pilot degradation and the presence of a Kaye alarm indicating a first stage pilot differential temperature (ref CR-PNP-2013-0378). SRV 3B was ruled out after acoustic monitors and SRV tailpipe temps indicated the relief valve was closed. The steam seal regulator was ruled out as a potential cause due to PI-3160, Steam Seal Header Pressure, indicating 6 psig which is within the acceptable band of 3-8 psig per procedure 2.1.5. The CRS briefed the control room to insert a manual reactor scram, place HPCI in pressure control, and close the MSIVs. At 2218 with reactor pressure at 418 psig, the CRS directed the initiation of a manual scram. The operation's crew performed the actions of 2.1.6, REACTOR SCRAM. At approximately 2223 with a reactor pressure of 345 psig, PI point TE-3611 began to lower from its peak temperature. At the direction of the CRS, HPCI was placed in pressure control mode in preparation for the closure of the MSIVs. At 2227 and reactor pressure at 293 psig, the CRS directed closure of the MSIVs which isolated the reactor from the steam seal system thereby mitigating the excessive reactor pressure decrease event.
CAUSE
The Direct Cause of this event was operation of MO-S-2, Steam Seal Bypass Valve, at a steam line pressure above the design operating limit (250 psig) of the steam seal bypass. This operation allowed for exceeding the lift pressure of inline relief valves PSV-3197/3198.
The Root Cause of the event was that procedure PNPS 2.1.5 did not limit operation of MO-S-2, Steam Seal Bypass Valve, to below the steam line pressure design operating limit (250 psig) of the steam seal bypass.
CORRECTIVE ACTIONS
The following corrective action has been completed.
- Revised Procedure 2.1.5 "Controlled Shutdown from Power" to limit operation of the Gland Seal Regulator Bypass Valve MO-S-2 to reactor pressures below 250 psig (Corrective Action to Preclude Recurrence) The following procedure improvement is planned:
- Revise Procedure 2.2.93 "Main Condenser Vacuum System" to ensure MO-S-2 is jogged closed and closed as Reactor Pressure rises to 250 psig and provide a caution or note that pressure indicated on PI- 3160 is not reflective of pressure downstream of MO-S-2.
The corrective actions above and additional corrective actions are addressed in the Corrective Action Program.
SAFETY CONSEQUENCES
The event occurred during a planned reactor shutdown with the reactor mode switch in the "Startup/Hot Standby" position.
This event consisted of a reactor depressurization through the steam seal system relief valves followed by manual operator action to isolate the reactor. The cooldown rate remained below the TS limit of 100°F/hr.
Power and steam generation rates were low and well within the capacity of available pressure control systems.
Reactor water level and pressure were manually controlled within acceptable bands to avoid any adverse consequences. With respect to analyzed transients evaluated in the FSAR, this is a non-limiting combination of a low power reactor depressurization and low water level incident.
This event is analyzed in the FSAR and did not challenge safety limits or fission product barriers. All other Engineered Safeguard System functions were operable during this event.
REPORTABILITY
This report is submitted in accordance with 10 CFR 50.73(a)(2)(iv)(A) — Any event or condition that resulted in manual or automatic actuation of any system listed in paragraph 10 CFR 50.73 (a)(2)(iv)(B). The Reactor Protection System (RPS) and Containment Isolation System are included in 10 CFR 50.73 (a)(2)(iv)(B).
PREVIOUS EVENTS
A review for similarity was conducted of Pilgrim Station Licensee Event Reports (LERs) submitted to the NRC.
The review focused on LERs involving a similar event or cause involving depressurization of the reactor. The review identified no similar event.
ENERGY INDUSTRY IDENTIFICATION SYSTEM (EllS) CODES:
SYSTEMS CODES
Plant Protection System JC Containment Isolation Control System JM Reactor Building NG Turbine Steam Seal System TC Reactor Water Cleanup System CE
REFERENCES
Condition Report, CR-PNP-2013-2275 — Unexpected Lowering of Reactor Pressure During Shutdown.