05000346/LER-2016-002

LER-2016-002, Unanticipated Steam and Feedwater Rupture Control System Actuation

Davis-Besse Nuclear Power Station, Unit 1
Event date:	01-30-2016
Report date:	03-29-2016
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation
Initial Reporting
ENS 51702	10 CFR 50.72(b)(3)(iv)(A), System Actuation
3462016002R00 - NRC Website

LER 16-002-00 for Davis-Besse Nuclear Power Station, Unit 1, Regarding Unanticipated Steam and Feedwater Rupture Control System Actuation

ML16091A115
Person / Time
Site:	Davis Besse
Issue date:	03/29/2016
From:	Boles B D FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-16057 LER 16-002-00
Download: ML16091A115 (5)

Reported lessons learned are incorporated into the licensing process and fed back to industry.

Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by intemet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not 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.

Energy Industry Identification System (EIIS) codes are identified in the text as [XX].

System Description:

The Davis-Besse Nuclear Power Station (DBNPS) Steam and Feedwater Line Rupture Control System (SFRCS) [JB] is a protection system that initiates the Auxiliary Feedwater System (AFW) [BA] and isolates the affected Steam Generator (SG) [AB-SG] on a steam or feedwater line rupture. The SFRCS is required to ensure an adequate feedwater supply to the steam generators to, remove reactor decay heat during periods when the normal feedwater supply and/or the electric power supply to essential auxiliaries has been lost. The design of the SFRCS is to mitigate release of high energy steam, to automatically start the AFW System in the event of a main steam line or Main Feedwater (MFW) line rupture, or on the loss of both main feed pumps [SJ-P] or the loss of all four Reactor Coolant Pumps (RCPs) [AB-P], and to prevent steam generator overfill and subsequent spillover into the main steam lines. The SFRCS also provides a trip signal to the Anticipatory Reactor Trip System (ARTS). In the event of a main steam line rupture, the SFRCS will close both Main Steam Isolation Valves (MSIVs) [SB-ISV] and all MFW control [SJ-LCV] and stop valves [SJ-ISV] and trip the main turbine [TA-TRB].

A loss of normal feedwater is sensed by monitoring the differential pressure across the main feedwater check valves. On a high reverse differential pressure, the SFRCS opens the steam supply from either steam generator to either auxiliary feedwater pump turbine and aligns the auxiliary feedwater pump to its associated steam generator. SFRCS also isolates MFW, main steam, and trips the reactor and the main turbine. On a reduced steam generator(s) inventory, the SFRCS responds similarly, with the exception that MFW and main steam will not be isolated.

DESCRIPTION OF EVENT:

On January 29, 2016, at 1322 hours0.0153 days <br />0.367 hours <br />0.00219 weeks <br />5.03021e-4 months <br />, the DBNPS experienced a reactor trip and subsequent SFRCS actuation on SG 1 High Level. This resulted in various automatic actions, including the isolation of MFW to the SGs and starting both trains of AFW. The Plant was stable in Mode 3 with the AFW System providing AFW to the SGs, as designed. Refer to LER 2016-001 for more information on the January 29, 2016 events.

On January 30, 2016, with the plant remaining in Mode 3, Operations personnel were proceeding with Post Trip recovery actions using procedure DB-OP-06910, Trip Recovery, to transition from AFW to MFW supplying feedwater to the SGs. The Motor Driven Feedwater Pump (MDFP) was started and operated on minimum recirculation flow back to the Deaerator. The MDFP discharge flowpath was then aligned to the MFW system to refill and pressurize the downstream MFW header through the individual Minimum Bypass Throttle Valves (FW139, FW44) up to the closed MFW Isolation Valves (FW612, FW601). The MDFP continued to deliver feedwater to the MFW header establishing a stable FW pressure upstream of the Minimum Bypass Throttle Valves (FW139, FW44). The MDFP flow remained steady at approximately 12-15 gallons per minute for approximately 5 minutes. In accordance with procedure DB-OP-06910, Operations personnel verified MFW pressure upstream of the Minimum Bypass Throttle Valves (FW139 and FW44) was greater than SG pressure. MFW to SG 2 Isolation Valve (FW601) was opened and in approximately 1 minute, MFW to SG 1 Isolation Valve (FW612) was opened. This resulted in a reverse delta pressure indication across the MFW check valve at 0123 hours0.00142 days <br />0.0342 hours <br />2.03373e-4 weeks <br />4.68015e-5 months <br /> as MFW to SG 1 Isolation Valve (FW612) was opened, resulting in an automatic actuation of the SFRCS Reported lessons learned are incorporated into the licensing process and fed back to industry.

Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB,10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not 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.

on High Reverse Differential Pressure and realignment of several valves in the Main Steam Feedwater System as designed. AFW continued to operate throughout the event and the Plant was verified to be stable.

CAUSE OF EVENT

The direct cause of the event was insufficient pressure in the piping upstream of the MFW to SG 1 Check valve (FW147), resulting in the SFRCS Actuation. When MFW to SG 1 Isolation Valve (FW612) was opening, the downstream SG pressure of approximately 870 pounds per square inch gauge (psig) was immediately applied to the downstream side of MFW to SG 1 Check Valve (FW147), developing greater than 125 pounds per square inch differential (psid) reverse DP, thus causing the SFRCS SG 1 High Reverse DP trip. The unexpected SFRCS SG 1 High Reverse D/P trip resulted in the reclosure of affected valves, including the MFW Isolation Valves, as designed. The AFW Pumps remained in operation, as expected. The MDFP remained operating aligned to the MFW Header and continued to pressurize the piping between the FW Heaters up to the MFW Isolation Valves.

The apparent cause of the event is that less than adequate procedural guidance is contained in procedure DB-OP-06910, Trip Recovery. The guidance in this procedure was inadequate to ensure that the MFW piping segment between MFW Bypass Throttle Valve (FW139) and MFW to SG 1 Isolation Valve (FW612) was sufficiently pressurized prior to opening MFW to SG 1 Isolation Valve (FW612)..

ANALYSIS OF EVENT

The SFRCS is a protection system required to actuate AFW to the SG's to remove reactor decay heat during periods when normal feedwater supply has been lost and/or upon loss of power to the RCP motors. Crossover piping exists that may be used to direct feedwater from either AFW source to either SG. The SFRCS also functions to isolate steam and main feedwater lines to mitigate overcooling events caused by steam depressurization:

At the time of the event, the AFW System was operating and feeding the Steam Generators. The unit remained in Mode 3 and stable. This actuation did not have any negative impact to the AFW system and the ability to feed the steam generators.

A bounding quantitative evaluation of risk impact was performed, using best available information, which estimated the delta Core Damage Frequency (CDF) to be 4.9E-07/yr. Based on this analysis, this event is considered to have a very low safety significance.

Reportability Discussion:

The Auxiliary Feedwater System by the SFRCS on a valid reverse differential pressure is reportable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the event in accordance with 10 CFR 50.72(b)(3)(iv)(A). This notification was completed at 07:32 on January 30, 2016 (Event Number 51702). The receipt of an actuation signal of the Auxiliary Feedwater System and the actuation/closure of several main steam system valves is reportable as a Licensee Event Report per 10 CFR 50.73(a)(2)(iv)(A). All safety systems performed as required in response to the event, and no loss of safety function occurred.

Reported lessons learned are incorporated into the licensing process and fed back to industry.

Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not 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.

3. LER NUMBER

• 05000-346 2016

SEQUENTIA

- 002 00

CORRECTIVE ACTIONS

Procedure, DB-OP-06910, Trip Recovery, will be revised to add the requirement to verify that the, pressure in SG 1 (2) Feedwater piping segment between valves MFW to SG 1 Check Valves (FW139 and FW147) and MFW to SG 2 Check Valves (FW44 and FW156) exceeds SG 1 (2) steam pressure prior to opening MFW to SG 2 (SG 1) Isolation Valve FW612 (FW601). This verification will require the installation of a temporary pressure indicator(s).

PREVIOUS SIMILAR EVENTS

Licensee Event Report (LER) 2015-002 documents the manual actuation of SFRCS in May 2015 to isolate a steam leak in the turbine building. LER 2016-001 documents the automatic actuation of SFRCS on January 29, 2016, in response to high SG levels following a reactor trip. There have been no LERs at the DBNPS involving an SFRCS reverse differential pressure trip in the past three years.