text

LER-2015-007, Regarding Loss of Residual Heat Removal Capability
Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(vii), Common Cause Inoperability; 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded; 10 CFR 50.73(a)(2)(viii)(A); 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition; 10 CFR 50.73(a)(2)(viii)(B); 10 CFR 50.73(a)(2)(iii); 10 CFR 50.73(a)(2)(ix)(A); 10 CFR 50.73(a)(2)(iv)(A), System Actuation; 10 CFR 50.73(a)(2)(x); 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor; 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown; 10 CFR 50.73(a)(2)(v), Loss of Safety Function; 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications; 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat
LER closed by
	IR 05000263/2016004 (13 February 2017)
	2632015007R00 - NRC Website
	v • d • e

Regarding Loss of Residual Heat Removal Capability
	ML16022A223
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-16-006 LER 15-007-00
	Download: ML16022A223 (4)
	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 LER 2015-007, "Loss of Residual Heat Removal Capability" Monticello Nuclear Generating Plant 2807 W County Road 75 Monticello, MN 55362 L-MT 006 10 CFR 50.73 Enclosed, is the Monticello Nuclear Generating Plant (MNGP) Licensee Event Report (LER) 2015-007 regarding a loss of Residual Heat Removal capability. This condition is reportable to the NRC in accordance with 10 CFR 50. 73(a)(2)(v)(B), as an Event or Condition that Could have Prevented the Fulfillment of the Safety Function of Structures or Systems that are Needed to Remove Residual Heat.

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: 0113112017 (02-2014) 1)>-***e.a..,

Estimated burden per response to comply v.ith this mandatory collection request: 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months .

(~) LICENSEE EVENT REPORT (LER)

Reported lessons learned are Incorporated into the licensing process and fed back to indusby.

Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T*5 F53), U.S. Nudear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to lnfocollects.Resource@nrc.gov, and to the Desk OffiCer, Office of Information and (See Page 2 for required number of Regulatory Affairs, NEOB-10202, (31500104), OffiCe of Management and Budget, Washington, DC digits/characters for each block) 20503. If a means used to impose an Information collection does not display a currenlly valid OMB control number, the NRC may not conducl or sponsor, and a person is not required to respond to, the information collection.

3. PAGE Monticello Nuclear Generating Plant 05000-263 1 of 3

4. TITLE Loss of Residual Heat Removal Capability

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED YEAR I SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO.

MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 11 24 2015 2015 -

007

- 00 01 21 2016 05000

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: {Check all that apply)

D 20.2201 (b)

D 20.2203(a)(3)(i)

D

50. 73(a)(2)(i)(C)

D 50.73(a)(2)(vii) 3 D

20.2201 (d)

D 20.2203(a)(3)(ii)

D 50.73(a)(2)(ii)(A)

D 50.73(a)(2)(viii)(A)

D 20.2203(a)(1)

D 20.2203(a)(4)

D 50.73(a)(2)(ii)(B)

D 50.73(a)(2)(viii)(B)

D 20.2203(a)(2)(i)

D 50.36(c)(1)(i)(A)

D 50.73(a)(2)(iii)

D 50.73(a)(2)(ix)(A)

10. POWER LEVEL D

20.2203(a)(2)(ii)

D 50.36(c)(1 )(ii)(A)

D 50.73(a)(2)(iv)(A)

D 50.73(a)(2)(x)

D 20.2203(a)(2)(iii)

D 50.36(c)(2)

D 50.73(a)(2)(v)(A)

D 73.71(a)(4) 0%

D 20.2203(a)(2)(iv)

D 50.46(a)(3)(ii)

[gj 50. 73(a)(2)(v)(B)

D 73.71(a)(5)

D 20.2203(a)(2)(v)

D 50.73(a)(2)(i)(A)

D 50.73(a)(2)(v)(C)

D OTHER D

20.2203(a)(2)(vi)

D 50.73(a)(2)(i)(B)

D 50.73(a)(2)(v)(D)

Specify in Abstract below or In NRC Fonn 366A

12. LICENSEE CONTACT FOR THIS LER LICENSEE CONTACT

~~ELEPHONE NUMBER (Include Area Code}

Andrew Kouba, Licensing Engineer (763) 271-7251 CAUSE SYSTEM COMPONENT MANU-REPORTABLE

CAUSE

SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TOEPIX N/A N/A N/A N/A N/A I

N/A N/A N/A N/A N/A

14. SUPPLEMENTAL REPORT EXPECTED

15. EXPECTED MONTH DAY YEAR DYES (If yes, complete 15. EXPECTED SUBMISSION DATE)

[gj NO SUBMISSION DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)

On November 24, 2015 at 0534 hours0.00618 days 0.148 hours 8.829365e-4 weeks 2.03187e-4 months , the Monticello Nuclear Generating Plant was at 0% power in Mode 3 (Hot Shutdown) for a forced outage. While initially placing Shutdown Cooling (SOC) in service, the 12 Residual Heat Removal (RHR) pump tripped approximately 8-10 seconds after start due to the closure of the RHR SOC suction isolation valves. When placing SOC in service, flow rapidly increased after opening the RHR Division 2 Low Pressure Coolant Injection (LPCI) outboard injection valve causing a localized pressure transient in the reactor recirculation pump suction piping that resulted in an isolation of the SOC suction line. Reactor pressure vessel (RPV) pressure remained stable at approximately 30 psig.

Prior to attempting to place '8' SOC in service, the Condensate system and the 'F' Safety Relieve Valve were in service providing decay heat removal. Immediate actions were taken to restore 'B' RHR SOC to operable status, thus an alternative method of decay heat removal was already established by the Condensate system and 'F' Safety Relief Valve.

NRC FORM 366 (02-2014)

EVENT DESCRIPTION

Estimated burden per response to comply with this mandatory collection request 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months .

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 lnfocollects.Resourre@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 OMS con~ol number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

YEAR 2015

6. LER NUMBER l

SEQUENTIAL I NUMBER 007 REV NO.

00 2

3. PAGE OF 3

On November 24, 2015 at 0534 hours0.00618 days 0.148 hours 8.829365e-4 weeks 2.03187e-4 months , the Monticello Nuclear Generating Plant was at 0% power in Mode 3 (Hot Shutdown) for a forced outage. While initially placing Shutdown Cooling (SOC) [KE] in service, the 12 Residual Heat Removal (RHR) [BO] pump [P] tripped approximately 8-10 seconds after start due to the closure of the RHR SOC suction isolation valves [ISV].

When placing SOC in service, flow was rapidly increased after opening the RHR Division 2 Low Pressure Coolant Injection (LPCI) outboard injection valve [INV], causing a localized pressure transient in the reactor recirculation pump suction piping that resulted in an isolation of the SOC suction line. The RHR High Reactor Pressure annunciator [PA] was received and immediately cleared as the pressure switch [PS] upstream of 12 Recirculation Pump [AD] Suction valve in the 'B' Recirculation Loop actuated causing a Group 2 containment isolation signal. However, this was not expected as Reactor Pressure Vessel (RPV) [RPV] steam dome pressure remained stable at approximately 30 psig.

At 0535 hours0.00619 days 0.149 hours 8.845899e-4 weeks 2.035675e-4 months , immediate actions were taken to restore 'B' RHR SOC to operable status.

At 0545 hours0.00631 days 0.151 hours 9.011243e-4 weeks 2.073725e-4 months , an alternative method of decay heat removal was established by utilizing the Condensate [SO] system and 'F' Safety Relief valve [RV].

At 1354 hours0.0157 days 0.376 hours 0.00224 weeks 5.15197e-4 months , the 12 RHR pump and 12 RHR Service Water pump were successfully placed in service on SOC and the plant reached Mode 4 (Cold Shutdown) at 1428 hours0.0165 days 0.397 hours 0.00236 weeks 5.43354e-4 months .

EVENT ANALYSIS

The event was determined to be reportable in accordance with 10 CFR 50.73(a)(2)(v)(B) as an Event or Condition that Could have Prevented the Fulfillment of the Safety Function of Structures or Systems that are Needed to Remove Residual Heat. This event is considered a Safety System Functional Failure per NEI 99-02, Revision 7.

SAFETY SIGNIFICANCE

Prior to attempting to place 'B' SOC in service, the Condensate system and the 'F' Safety Relieve Valve were in service providing decay heat removal. These systems remained in service and, as demonstrated by steadily lowering RPV pressure and temperature, provided adequate decay heat removal until SOC was placed in service. Additionally, the Reactor Water Cleanup System [CE] was available for decay heat reject if needed. After the closure of the SOC suction valves and subsequent trip of the 12 RHR pump, immediate actions were taken to restore SOC to operable status. Since the reactor remained adequately cooled, there were no actual consequences as a result of the initial failed attempt to place SOC in service. There was no impact to the health and safety of the public.

CAUSE

U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET

2. DOCKET

6. LER NUMBER

3. PAGE OF YEAR I SEQUENTIAL I REV NUMBER NO.

05000-263 3

3 2015 007 00 Both reactor high pressure SOC isolation pressure switches are located on the 'B' Recirculation Suction Piping. When initially placing SOC in service the LPCI outboard injection valve was opened and flow into the 'B' Recirculation system increased to approximately 4000 gpm in several seconds. This rapid flow increase caused a localized pressure transient in the 'B' Recirculation pump piping that resulted in the isolation of the SOC suction valves. Closure of the SOC suction valves subsequently caused a trip of the 12 RHR pump due to loss of pump suction. Written documentation in the operations manual did not adequately address the sensitivity of the pressure switches while placing 'B' SOC in service.

CORRECTIVE ACTION

Since the Condensate system and the 'F' Safety Relieve Valve were already in service providing decay heat removal, an alternate method of decay removal did not need to be established. Immediate actions were taken to restore 'B' SOC to operable status.

The Operations Manual used to place 'B' SOC in service was re-performed in its entirety to verify proper valve alignment, ensure the piping was full of water, and verify acceptable temperatures existed prior to attempting to place the system in service. This included venting the RHR suction and discharge lines prior to placing 'B' SOC in service. Existing procedural guidance allowed the associated LPCI injection valve to be slowly throttled open to achieve required RHR pump flow without introducing a pressure transient that would challenge the reactor high pressure SOC isolation setpoint.

The 12 RHR pump was successfully started and placed in SOC mode to cool down the plant to MODE

4. The Operations Manual has been updated to provide additional guidance for placing SOC in service including the pressure switch sensitivity to injection flow rate changes when changing the position of the LPCI outboard injection valve.

PREVIOUS SIMILAR EVENTS

There were no Licensee Event Reports with similar causes of loss SOC within the 3 last years.

ADDITIONAL INFORMATION

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

EVENT DESCRIPTION

Estimated burden per response to comply with this mandatory collection request 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months .

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 lnfocollects.Resourre@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 OMS con~ol number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

YEAR 2015

6. LER NUMBER l

SEQUENTIAL I NUMBER 007 REV NO.

00 2

3. PAGE OF 3

On November 24, 2015 at 0534 hours0.00618 days 0.148 hours 8.829365e-4 weeks 2.03187e-4 months , the Monticello Nuclear Generating Plant was at 0% power in Mode 3 (Hot Shutdown) for a forced outage. While initially placing Shutdown Cooling (SOC) [KE] in service, the 12 Residual Heat Removal (RHR) [BO] pump [P] tripped approximately 8-10 seconds after start due to the closure of the RHR SOC suction isolation valves [ISV].

When placing SOC in service, flow was rapidly increased after opening the RHR Division 2 Low Pressure Coolant Injection (LPCI) outboard injection valve [INV], causing a localized pressure transient in the reactor recirculation pump suction piping that resulted in an isolation of the SOC suction line. The RHR High Reactor Pressure annunciator [PA] was received and immediately cleared as the pressure switch [PS] upstream of 12 Recirculation Pump [AD] Suction valve in the 'B' Recirculation Loop actuated causing a Group 2 containment isolation signal. However, this was not expected as Reactor Pressure Vessel (RPV) [RPV] steam dome pressure remained stable at approximately 30 psig.

At 0535 hours0.00619 days 0.149 hours 8.845899e-4 weeks 2.035675e-4 months , immediate actions were taken to restore 'B' RHR SOC to operable status.

At 0545 hours0.00631 days 0.151 hours 9.011243e-4 weeks 2.073725e-4 months , an alternative method of decay heat removal was established by utilizing the Condensate [SO] system and 'F' Safety Relief valve [RV].

At 1354 hours0.0157 days 0.376 hours 0.00224 weeks 5.15197e-4 months , the 12 RHR pump and 12 RHR Service Water pump were successfully placed in service on SOC and the plant reached Mode 4 (Cold Shutdown) at 1428 hours0.0165 days 0.397 hours 0.00236 weeks 5.43354e-4 months .

EVENT ANALYSIS

The event was determined to be reportable in accordance with 10 CFR 50.73(a)(2)(v)(B) as an Event or Condition that Could have Prevented the Fulfillment of the Safety Function of Structures or Systems that are Needed to Remove Residual Heat. This event is considered a Safety System Functional Failure per NEI 99-02, Revision 7.

SAFETY SIGNIFICANCE

Prior to attempting to place 'B' SOC in service, the Condensate system and the 'F' Safety Relieve Valve were in service providing decay heat removal. These systems remained in service and, as demonstrated by steadily lowering RPV pressure and temperature, provided adequate decay heat removal until SOC was placed in service. Additionally, the Reactor Water Cleanup System [CE] was available for decay heat reject if needed. After the closure of the SOC suction valves and subsequent trip of the 12 RHR pump, immediate actions were taken to restore SOC to operable status. Since the reactor remained adequately cooled, there were no actual consequences as a result of the initial failed attempt to place SOC in service. There was no impact to the health and safety of the public.

CAUSE

U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET

2. DOCKET

6. LER NUMBER

3. PAGE OF YEAR I SEQUENTIAL I REV NUMBER NO.

05000-263 3

3 2015 007 00 Both reactor high pressure SOC isolation pressure switches are located on the 'B' Recirculation Suction Piping. When initially placing SOC in service the LPCI outboard injection valve was opened and flow into the 'B' Recirculation system increased to approximately 4000 gpm in several seconds. This rapid flow increase caused a localized pressure transient in the 'B' Recirculation pump piping that resulted in the isolation of the SOC suction valves. Closure of the SOC suction valves subsequently caused a trip of the 12 RHR pump due to loss of pump suction. Written documentation in the operations manual did not adequately address the sensitivity of the pressure switches while placing 'B' SOC in service.

CORRECTIVE ACTION

Since the Condensate system and the 'F' Safety Relieve Valve were already in service providing decay heat removal, an alternate method of decay removal did not need to be established. Immediate actions were taken to restore 'B' SOC to operable status.

The Operations Manual used to place 'B' SOC in service was re-performed in its entirety to verify proper valve alignment, ensure the piping was full of water, and verify acceptable temperatures existed prior to attempting to place the system in service. This included venting the RHR suction and discharge lines prior to placing 'B' SOC in service. Existing procedural guidance allowed the associated LPCI injection valve to be slowly throttled open to achieve required RHR pump flow without introducing a pressure transient that would challenge the reactor high pressure SOC isolation setpoint.

The 12 RHR pump was successfully started and placed in SOC mode to cool down the plant to MODE

4. The Operations Manual has been updated to provide additional guidance for placing SOC in service including the pressure switch sensitivity to injection flow rate changes when changing the position of the LPCI outboard injection valve.

PREVIOUS SIMILAR EVENTS

There were no Licensee Event Reports with similar causes of loss SOC within the 3 last years.

ADDITIONAL INFORMATION

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

05000263/LER-2015-007, Regarding Loss of Residual Heat Removal Capability

Contents

text

CAUSE

EVENT DESCRIPTION

EVENT ANALYSIS

SAFETY SIGNIFICANCE

CAUSE

CORRECTIVE ACTION

PREVIOUS SIMILAR EVENTS

ADDITIONAL INFORMATION

EVENT DESCRIPTION

EVENT ANALYSIS

SAFETY SIGNIFICANCE

CAUSE

CORRECTIVE ACTION

PREVIOUS SIMILAR EVENTS

ADDITIONAL INFORMATION

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05000263/LER-2015-007, Regarding Loss of Residual Heat Removal Capability

text

CAUSE

EVENT DESCRIPTION

EVENT ANALYSIS

SAFETY SIGNIFICANCE

CAUSE

CORRECTIVE ACTION

PREVIOUS SIMILAR EVENTS

ADDITIONAL INFORMATION

EVENT DESCRIPTION

EVENT ANALYSIS

SAFETY SIGNIFICANCE

CAUSE

CORRECTIVE ACTION

PREVIOUS SIMILAR EVENTS

ADDITIONAL INFORMATION

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