05000346/LER-2016-002, Regarding Unanticipated Steam and Feedwater Rupture Control System Actuation
| ML16091A115 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  (NPF-003) |
| Issue date: | 03/29/2016 |
| From: | Boles B FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-16057 LER 16-002-00 | |
| Download: ML16091A115 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3462016002R00 - NRC Website | |
text
FE NOC' RrstEnergy Nuclear Operating Company Brian D. Boles Vice President, Nuclear March 29, 2016 L-16-057 ATTN: Document Control Desk United States Nuclear Regulatory Commission Washington, D.C. 20555-0001
Subject:
Davis-Besse Nuclear Power Station, Unit 1 Docket Number 50-346, License Number NPF-3 Licensee Event Report 2016-002 5501 North State Route 2 Oak Harbor, Ohio 43449 10 CFR 50.73 419-321-7676 Fax: 419-321-7582 Enclosed is Licensee Event Report (LER) 2016-002-00, "Unanticipated Steam and Feedwater Rupture Control System Actuation." This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A).
There are no regulatory commitments contained in this letter or its enclosure. The actions described represent intended or planned actions and are described for information only. If there are any questions or if additional information is required, please contact Mr. Patrick J. McCloskey, Manager-Site Regulatory Compliance, at (419) 321-7274.
Sincerely,
~J~
Brian D. Boles vaw Enclosure: LER 2016-002 cc: NRC Region Ill Administrator NRC Resident Inspector NRR Project Manager Utility Radiological Safety Board
NRCFORM366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2018 (11-2015)
, the NRC may not conduct or sponsor, and a nerson is not reouired to resnond to the information collection.
- 3. LER NUMBER YEAR SEQUENTIA REV NUMBER NO.
Davis-Besse Nuclear Power Station Unit 1 05000-346 2016
- - 002 DESCRIPTION OF EVENT (Continued) on High Reverse DifferentiaJ 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. 00 (11-2015)
U.S. NUCLEAR REGULATORY COMMISSION' APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2018 LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
, the NRG may not conduct or sponsor, and a oerson is not reauired to resoond to the information collection.
- 3. LER NUMBER YEAR SEQUENTIA REV NUMBER NO.
Davis-Besse Nuclear Power Station Unit 1 05000-346 2016
- 002
CORRECTIVE ACTIONS
Procedure, DB-OP-0691 o
.. Trip Recovery, will be revised to add the requirement to verify that th~
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.
/ 00