text

LER-2015-006, Regarding Reactor Scram Due to Group 1 Isolation from Foreign Material in the Main Steam Flow Instrument Line
Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation; 10 CFR 50.73(a)(2)(iv)(B), System Actuation
LER closed by
	IR 05000263/2016004 (13 February 2017)
	2632015006R00 - NRC Website
	v • d • e

Regarding Reactor Scram Due to Group 1 Isolation from Foreign Material in the Main Steam Flow Instrument Line
	ML16022A106
Person / Time
Site:	Monticello
Issue date:	01/21/2016
From:	Gardner P; Northern States Power Co, Xcel Energy
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-MT-16-005 LER 15-006-00
	Download: ML16022A106 (5)
	v • d • e

(l Xcel Energy January 21, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket 50-263 Renewed Facility Operating License No. DPR-22 Monticello Nuclear Generating Plant 2807 W County Road 75 Monticello, MN 55362 L -MT 005 10 CFR 50.73 LER 2015-006-00 "Reactor Scram due to Group 1 Isolation from Foreign Material in the Main Steam Flow Instrument Line" Enclosed is the Monticello Nuclear Generating Plant (MNGP) Licensee Event Report (LER) 2015-006-00, Reactor Scram due to Group 1 Isolation from Foreign Material in the Main Steam Flow Instrument Line. This condition is reportable to the NRC in accordance with 10 CFR 50.73(a)(2)(iv)(A), as an event that resulted in automatic actuation of a reactor protection system.

Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.

~~

Peter A. Gardner

~

Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company-Minnesota Enclosure cc:

Regional Administrator, Region Ill, USNRC Project Manager, MNGP, USNRC Resident Inspector, MNGP, USNRC

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150*0104 EXPIRES: 01/31/2017 (02-2014)

~'l<l.taltt

, the NRC may not conduct or sponsor, and a person is not required to respond to, the infonnation collection.

YEAR 2015

6. LER NUMBER I

SEQUENTIAL I NUMBER 006 REV NO.

00 2

3. PAGE OF 4

On November 23, 2015, at approximately 10:40, with the plant in Mode 1 and 100 percent power, a trip of the# 11 Reactor Recirculation Pump [AD] occurred. Following the # 11 Reactor Recirculation Pump trip, at approximately 10:41, a large spike in differential pressure on the 'C' main steam line flow instrumentation line occurred. Due to this differential pressure spike, all four of the Main Steam Line 'C' High Flow Isolation Pressure Differential Indicating Switches [PDIS] actuated, causing a Group 1 isolation~ The Group 1 isolation caused the main steam isolation valves (MSIV) [ISV] to close, resulting in a reactor scram due to the MSIVs being less than 90% open at approximately 10:41. The 'D' outboard MSIV did not close until approximately 130 seconds after the Group 1 signal was sent. The other seven MSIVs closed as expected. All control rods [AA] inserted, and all other safety systems responded as expected. Division II secondary containment isolation instrumentation, secondary containment isolation valves, and the Standby Gas Treatment System were inoperable for planned maintenance prior to the event.

Also prior to the event, the# 4 Turbine Stop Valve 90% open limit switch, which inputs to the reactor protection system (RPS), was inoperable and the channels had previously been placed in trip to meet Technical Specification Required Actions 3.3.1.1.A and 3.3.1.1.8 and remained in trip throughout the event. The inoperable equipment did not impact the scram response.

With the MSIVs closed, normal Reactor Pressure Vessel (RPV) [RPV] pressure control [PC] was not available. The Low-Low Set System operated as expected to control reactor pressure by cycling the 'H' Safety Relief Valve (SRV) [RV]. The 'H' SRV first opened at approximately 10:49. At approximately 10:58, the High Pressure Coolant Injection (HPCI) [BJ] System was placed in service in the pressure control mode. The 'H' SRV cycled five (5) times prior to placing HPCI in service and cycled one additional time after HPCI was placed in pressure control mode.

Following the scram, RPV water level immediately decreased to +3 inches, causing a Group 2 Primary Containment Isolation signal when RPV level decreased below +9 inches. The Group 2 signal remained in effect until it was reset at approximately 13:05. In response to the RPV water level reduction, the Feedwater Regulating Valves (FRVs) opened to restore RPV level. With the MSIVs closed, the only steam exiting the RPV was through the 'H' SRV and the steam supplied to operate the HPCI system, however pressure was below the 'H' SRV Low-Low Set setpoint and HPCI had not yet been placed in service. Under these conditions, RPV water level rose. Operations responded by closing the FRVs.

This action lessened the RPV l~el increase, but RPV water level continued to rise, tripping the Reactor Feedwater Pumps at approximately 10:42.

At approximately 11 :02, with HPCI in the pressure control mode and RPV water level at approximately

+22 inches the final actuation of the 'H' SRV occurred. Indicated RPV level immediately increased above the high RPV level setpoint of +48 inches, causing a trip of the HPCI System.* HPCI could not be placed in service until the 'H' SRV closed and RPV water level was below +48 inches. At approximately 11 :03, the 'H' SRV closed and indicated RPV water level lowered to +6 inches. HPCI was once again placed in service in the pressure control mode and the# 11 Reactor Feedwater Pump (RFP) was started.

The HPCI System was able to control RPV pressure without additional actuations of the 'H' SRV and the Condensate and Feedwater System was able to control RPV water level for the remainder of the event.

Following plant stabilization, Mode 4 was achieved on November 24, 2015, at approximately 14:28.

EVENT ANALYSIS

Pursuant to 10 CFR 50.72 paragraph (b)(2)(iv)(B) for the RPS actuation and paragraph (b)(3)(iv)(A) as an emergency safety feature (ESF) actuation, a four hour event notification was made to the NRC. The event was determined to be reportable in accordance with 10 CFR 50.73 (a)(2)(iv)(A), "Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph 10 CFR 50.73 (a)(2)(iv)(B)".

This event is not classified as a safety system functional failure.

SAFETY SIGNIFICANCE

The occurrence of the Group 1 isolation signal caused by the 'C' main steam line flow instrument did not compromise safety systems or components required to mitigate the consequences of an accident or transient, with the exception of temporarily resulting in losing the Main Condenser heat sink as an alternative means for decay heat removal. The excess flow check valve (EFCV) on the upstream 'C' main steam line flow instrument was degraded by the pressure wave that caused the Group 1 isolation.

The EFCV is designed to isolate the downstream instrument line in the event of an instrument line break. The integrity of the line was not compromised during the event. Therefore, there were no consequences resulting from the failure of this valve. The Group 1 isolation and subsequent reactor scram is within the analyzed and anticipated frequency of a Group 1 initiation and is bounded by the Main Steam Line Break accident analysis documented in the USAR.

The 'D' outboard MSIV was delayed in closing following the Group 1 isolation signal. However, the 'D' inboard MSIV closed within the required time frame, isolating the 'D' main steam line. Furthermore, subsequent evaluation has determined a condition requiring a Group 1 isolation did not exist. All other safety systems and components responded as required to mitigate the transient and ensure adequate core cooling. Operators performed as expected and in accordance with procedures.

In conclusion, this transient that occurred on November 23, 2015 did not impact the health and safety of the public.

CAUSE

The Group 1 isolation and subsequent reactor scram was caused by legacy foreign material in the common high side instrument line tap of the 'C' main steam flow instrument line which obstructed proper steam line flow instrument operation. This foreign material resulted in a high sensed flow condition by all four flow switches on the 'C' main steam flow instrument line during the down-power event (loss of a recirculation pump).

The foreign material in the instrument line is considered a legacy issue (believed to be from initial construction). Foreign Material Exclusion (FME) programs and controls have been improved to preclude similar conditions since that time. Therefore, this legacy issue is not indicative of current plant programs, processes, or behaviors.

The cause of the #11 Reactor Recirculation Pump trip that precipitated the down power was less than adequate maintenance strategy that did not provide sufficient process controls to ensure the bench testing of high critical components identified and then corrected equipment deficiencies prior to installation. As a result, a loose wire on the #11 Recirculation Pump Motor-Generator set voltage regulator was not detected and corrected prior to the event. This loose wire led to a trip of the #11 Reactor Recirculation Pump, a core flow reduction, and a reactor power reduction.

CORRECTIVE ACTION

Immediate corrective actions included removing the foreign material (nut) from the instrumentation line and properly attaching the loose wire. The EFCV and the MSIV issues were fixed prior to plant start up.

Long term corrective actions are to revise a fleet procedure to require verification of proper torque on accessible electrical connections for critical components which are bench tested, and also to ensure that accessible soldered and crimped electrical terminations are inspected for signs of degradation during bench testing.

Since the original plant construction, the FME programs and controls have been improved to preclude similar conditions. Therefore, no additional FME corrective actions are required.

PREVIOUS SIMILAR EVENTS

There were no previous similar licensee event reports in the past three years.

ADDITIONAL INFORMATION

The Institute of Electrical and Electronics Engineer codes for equipment are denoted by [XX].

EVENT DESCRIPTION