05000306/LER-2017-001

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LER-2017-001, 23 Containment Fan Coil Unit Operability
Prairie Island Nuclear Generating Plant Unit 2
Event date: 5-2-2016
Report date: 11-29-2017
Reporting criterion: 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications

10 CFR 50.73(a)(2)(i)(b)

10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident

10 CFR 50.73(a)(2)(ii)
3062017001R00 - NRC Website
LER 17-001-00 for Prairie Island, Unit 2, Regarding 23 Containment Fan Coil Unit Operability
ML17334A040
Person / Time
Site: Prairie Island Xcel Energy icon.png
Issue date: 11/29/2017
From: Northard S
Northern States Power Company, Minnesota, Xcel Energy
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-17-048
Download: ML17334A040 (5)


comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to 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.

Prairie Island Nuclear Generating Plant 05000-306 NUMBER NO.

Unit 2 2017 - 001 - 00

DESCRIPTION OF EVENT

During an extent of condition review for an issue with correct application of Technical Specifications (Tech Specs) LCO 3.0.6 that occurred August of 2017, a similar condition was discovered. While 122 Control Room Chiller (CRC) was out-of- service (OOS) due to chiller oil temperature outside operability limit per Tech Specs 3.7.11 Condition A, from May 2, 2016 to May 6, 2016, 23 Containment Fan Coil Unit (FCU1) was OOS due to a problem with the discharge damper. According to guidance in C18.1, "Engineered Safeguards Equipment Support Systems", Bus 16 load sequencer and Bus 121 are inoperable when 122 CRC is OOS (safeguards room cooling is provided by CRC). Supported systems including 21 Safeguards Screenhouse Roof Exhaust Fan (powered from Bus 121), supported B Train Diesel Driven Cooling Water Pump, and B Train containment cooling (22/24 FCUs) were also OOS.

Having both trains of containment FCUs OOS at the same time for approximately 35.

6 hours
6.944444e-5 days
0.00167 hours
9.920635e-6 weeks
2.283e-6 months

would require entry into LCO 3.0.3 for Unit 2, actions required to be in MODE 3 within

7 hours
8.101852e-5 days
0.00194 hours
1.157407e-5 weeks
2.6635e-6 months

; MODE 4 within

13 hours
1.50463e-4 days
0.00361 hours
2.149471e-5 weeks
4.9465e-6 months

; and MODE 5 within

37 hours
4.282407e-4 days
0.0103 hours
6.117725e-5 weeks
1.40785e-5 months

.

This did not occur. The Senior Reactor Operator failed to correctly assess the Technical Specifications (Tech Specs) impact to Unit 2 when applying Tech Specs 3.0.6. This event is reportable under 10 CFR 50.73(a)(2)(i)(B), Operation or Condition Prohibited by Tech Specs.

EVENT ANALYSIS

The event is reportable under 10 CFR 50.73(a)(2)(i)(B). The licensee shall report any operation or condition which was prohibited by the plant's Tech Specs. This condition meets the reporting criteria because both trains of containment FCUs were OOS at the same time for approximately 35.

6 hours
6.944444e-5 days
0.00167 hours
9.920635e-6 weeks
2.283e-6 months

, this required entry into LCO 3.0.3, putting Unit 2 in MODE 3 within

7 hours
8.101852e-5 days
0.00194 hours
1.157407e-5 weeks
2.6635e-6 months

. Tech Specs 3.0.3 for Limiting Condition for Operation was not entered and the required actions were not initiated within

1 hour
1.157407e-5 days
2.777778e-4 hours
1.653439e-6 weeks
3.805e-7 months

to place the unit, as applicable, in MODE 3 in

7 hours
8.101852e-5 days
0.00194 hours
1.157407e-5 weeks
2.6635e-6 months

; MODE 4 within

13 hours
1.50463e-4 days
0.00361 hours
2.149471e-5 weeks
4.9465e-6 months

; and MODE 5 within

37 hours
4.282407e-4 days
0.0103 hours
6.117725e-5 weeks
1.40785e-5 months

.

The Containment Air Cooling System consists of four fan coil units, a duct distribution system, and the associated instrumentation and controls. During normal operation the fans may be run at high or low speed and during post-accident conditions the fans run at low speed. The Containment Air Cooling System is designed to recirculate and cool the containment atmosphere in the event of a loss-of-coolant or main steam line break accident and thereby ensure that the containment pressure cannot exceed its design value of 46 psig at 268 degrees F (100% relative humidity).

Two of the four containment cooling units and one containment spray pump provide sufficient heat removal capability to maintain the post-accident containment pressure and temperature below the design value, assuming that the core residual heat is released to the containment as steam. Analysis has shown that the operation of one containment spray pump during the injection phase and the heat removal capability equivalent to a single fan coil unit at maximum fouling conditions is sufficient to maintain containment pressure less than design. While B Train FCU's 22 and 24 were OOS and A Train FCU 23 was OOS, A Train FCU 21 was operable. Because analysis has shown that single FCU with Containment Spray is sufficient to maintain containment pressure and 21 FCU was operable the event is not reportable under 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident (safety system functional failure). Even though 23, 22 and 24 FCU's were inoperable per Tech Specs, 21 FCU was operable and the ability to mitigate postulated accidents was not lost and the system was not in an unanalyzed condition as described 10 CFR 50.73(a)(2)(ii).

IEEE Component Code — FCU comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to 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 1. FACILITY NAME Prairie Island Nuclear Generating Plant 05000-306 - 001 2017 - 00

SAFETY SIGNIFICANCE

Safety function was not lost, because with both B Train FCU's and one A Train FCU inoperable per Tech Specs, the ability to mitigate postulated accidents was not lost and the system was not in an unanalyzed condition. Analysis has shown that the operation of one containment spray pump during the injection phase and the heat removal capability equivalent to a single fan coil unit at maximum fouling conditions is sufficient to maintain containment pressure less than design.

There were no radiological, environmental, or industrial impacts associated with this event, and PINGP did not adversely affect the health and safety of the public. This event report does not identify any safety system functional failures.

CAUSE

Cause evaluation determined that the Senior Reactor operators failed to utilize Human Performance Tools (Verification/Validation and Procedure Use/Adherence) when assessing the Technical Specification impact to Unit 2 for applying LCO 3.0.6 when 122 CRC was taken 00S.

CORRECTIVE ACTION(s) 1. Revise operations work instructions SWI 0-200.3, TECHNICAL SPECIFICATION ENTRY & EXIT to require independent assessment of shared system LCO's for each unit. This action is complete.

2. Revise the LCO database to Limit the use of "Unit 0" to ISFSI Technical Specifications. This action is complete.

3. Establish the standard that LCO 3.0.6 log entries, carried over a shift, are in Narrative Logs using the Open Item option.

This action is complete 4. Revise site work instruction 5AWI 3.15.8, SAFETY FUNCTION DETERMINATION PROGRAM to be more user friendly.

Including graphical explanations. This action is expected to be completed by the end of the year.

PREVIOUS SIMILAR EVENTS

A review of the Corrective Action Program (CAP) and Licensee Event Reports (LERs) for PINGP revealed one similar event over the last three years.

On September 11, 2015, it was identified that 122 Control Room Chiller was removed from service and control valve CV-31837 (121/122 Control Room Chiller Outlet) and CV-31838 (121/122 Control Room Chiller Inlet) were closed. This isolated Train B Safeguards Chilled Water and rendered Bus 16 Unit Cooler non-functional, which will result in unacceptable temperatures in the associated bus room during a postulated High Energy Line Break (HELB). Bus 16 would not have performed its safety function and was inoperable for greater than the time allowed by Tech Specs. Tech Specs 3.8.9 for Distribution Systems-Operating was not entered and the required actions were not taken to restore to an operable status within

8 hours
9.259259e-5 days
0.00222 hours
1.322751e-5 weeks
3.044e-6 months

or to enter MODE 3 in

6 hours
6.944444e-5 days
0.00167 hours
9.920635e-6 weeks
2.283e-6 months

and MODE 5 in

36 hours
4.166667e-4 days
0.01 hours
5.952381e-5 weeks
1.3698e-5 months

. This is a reportable event under 10 CFR 50.73(a)(2)(i)(b), Operation or Condition Prohibited by Tech Specs.