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Category:Licensee Event Report (LER)
MONTHYEAR05000423/LER-2024-002, Door Latch Failure Resulted in Loss of Safety Function for Secondary Containment Boundary2024-11-26026 November 2024 Door Latch Failure Resulted in Loss of Safety Function for Secondary Containment Boundary 05000336/LER-2024-002, Two Main Steam Safety Valves Failed to Lift within the Acceptance Criteria Resulting in a Condition Prohibited by Technical Specifications2024-11-26026 November 2024 Two Main Steam Safety Valves Failed to Lift within the Acceptance Criteria Resulting in a Condition Prohibited by Technical Specifications 05000423/LER-2024-001, Loss of Safety Function and Condition Prohibited by Technical Specifications for Loss of Secondary Containment Boundary2024-10-14014 October 2024 Loss of Safety Function and Condition Prohibited by Technical Specifications for Loss of Secondary Containment Boundary 05000423/LER-2023-006-02, Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-09-26026 September 2024 Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications 05000336/LER-2024-001, Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications2024-06-10010 June 2024 Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications 05000423/LER-2023-006-01, Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-05-20020 May 2024 Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications 05000423/LER-2023-006, Pressurizer Power Operated Relief Valve Failed to Stroke Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-05-0202 May 2024 Pressurizer Power Operated Relief Valve Failed to Stroke Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications 05000423/LER-2023-005, Oil Leakage from C RSS Pump Motor Challenged Meeting Its Mission Time Resulting in a Condition Prohibited by Technical Specifications2024-02-0808 February 2024 Oil Leakage from C RSS Pump Motor Challenged Meeting Its Mission Time Resulting in a Condition Prohibited by Technical Specifications 05000423/LER-2023-004, For Millstone Power Station, Unit 3 Regarding Reactor Coolant System Pressure Isolation Valves Operational Leakage Exceeded the Acceptance Criteria Resulting in a Condition Prohibited by Technical Specifications2023-12-19019 December 2023 For Millstone Power Station, Unit 3 Regarding Reactor Coolant System Pressure Isolation Valves Operational Leakage Exceeded the Acceptance Criteria Resulting in a Condition Prohibited by Technical Specifications 05000423/LER-2023-003, RCS Temperature Detector Exceeded Time Response Acceptance Criteria Resulting in a Condition Prohibited by Technical Specifications2023-12-0808 December 2023 RCS Temperature Detector Exceeded Time Response Acceptance Criteria Resulting in a Condition Prohibited by Technical Specifications 05000423/LER-2023-002, Auxiliary Feedwater Control Valve Failure Resulting in a Condition Prohibited by Technical Specifications2023-11-30030 November 2023 Auxiliary Feedwater Control Valve Failure Resulting in a Condition Prohibited by Technical Specifications 05000336/LER-2023-002, Failed Check Valve Resulted in an Unanalyzed Condition2023-08-31031 August 2023 Failed Check Valve Resulted in an Unanalyzed Condition 05000423/LER-2023-001, Automatic Reactor Trip Due to Main Generator Output Breaker Ground Fault2023-07-27027 July 2023 Automatic Reactor Trip Due to Main Generator Output Breaker Ground Fault 05000336/LER-2023-001, For Millstone, Unit 2, Structural Integrity of a Train Service Water Header Piping Could Not Be Demonstrated Causing the Unit to Operate in a Condition Prohibited by Technical Specifications2023-07-0707 July 2023 For Millstone, Unit 2, Structural Integrity of a Train Service Water Header Piping Could Not Be Demonstrated Causing the Unit to Operate in a Condition Prohibited by Technical Specifications 05000423/LER-2022-002, Two Main Steam Safety Valves Installed in Wring Location Resulting in Failure to Lift within the Technical Specification Acceptance Criteria2022-05-31031 May 2022 Two Main Steam Safety Valves Installed in Wring Location Resulting in Failure to Lift within the Technical Specification Acceptance Criteria 05000423/LER-2022-001, Emergency Core Cooling and Reactor Plant Component Cooling Water Systems Inoperable for Time Greater than Allowed by Technical Specifications2022-03-24024 March 2022 Emergency Core Cooling and Reactor Plant Component Cooling Water Systems Inoperable for Time Greater than Allowed by Technical Specifications 05000336/LER-2022-001, Structural Integrity of Reactor Building Component Cooling Water Cracked Threaded Fitting Could Not Be Established Resulting in System Inoperable Longer than Allowed by Technical Specifications2022-03-18018 March 2022 Structural Integrity of Reactor Building Component Cooling Water Cracked Threaded Fitting Could Not Be Established Resulting in System Inoperable Longer than Allowed by Technical Specifications 05000336/LER-2021-002, Failed Check Valve Resulting in Unnalyzed and Operation Prohibited by Technical Specifications2022-01-0505 January 2022 Failed Check Valve Resulting in Unnalyzed and Operation Prohibited by Technical Specifications 05000336/LER-2021-001, Incorrectly Placed Spent Fuel Assemblies in Unit 2 Spent Fuel Pool2021-09-23023 September 2021 Incorrectly Placed Spent Fuel Assemblies in Unit 2 Spent Fuel Pool 05000423/LER-2020-005, Loss of Safety Function - Secondary Containment2020-11-19019 November 2020 Loss of Safety Function - Secondary Containment 05000423/LER-2020-002, 3, Automatic Reactor Trip Due to Main Generator Ground Fault2020-05-28028 May 2020 3, Automatic Reactor Trip Due to Main Generator Ground Fault 05000423/LER-2020-001, Technical Specification Allowed Outage Time Exceeded by Accident Monitoring Instrument2020-03-0505 March 2020 Technical Specification Allowed Outage Time Exceeded by Accident Monitoring Instrument 05000336/LER-2019-001, Manual Reactor Trip Due to Loss of the a Steam Generator Feed Pump2020-02-19019 February 2020 Manual Reactor Trip Due to Loss of the a Steam Generator Feed Pump 05000423/LER-2019-001, Regarding Plant Shutdown Required by Technical Specifications, Emergency Diesel Generator Exceeded Allowed Outage Time2020-02-13013 February 2020 Regarding Plant Shutdown Required by Technical Specifications, Emergency Diesel Generator Exceeded Allowed Outage Time 05000336/LER-2018-001, Regarding Loss of Both Trains of Control Room Emergency Ventilation Resulting in the Loss of Safety Function2018-12-19019 December 2018 Regarding Loss of Both Trains of Control Room Emergency Ventilation Resulting in the Loss of Safety Function 05000423/LER-2017-001, Regarding Loss of Safety Function - Secondary Containment2017-03-20020 March 2017 Regarding Loss of Safety Function - Secondary Containment 05000336/LER-2016-002, Regarding Manual Reactor Trip Due Loss of Two Circulating Water Pumps2016-09-20020 September 2016 Regarding Manual Reactor Trip Due Loss of Two Circulating Water Pumps 05000423/LER-2016-005, Technical Specification Required Shutdown and Manual Reactor Trip Due to Steam Generator Level Oscillation2016-08-0909 August 2016 Technical Specification Required Shutdown and Manual Reactor Trip Due to Steam Generator Level Oscillation 05000423/LER-2016-004, Manual Reactor Trip Due to Low Hydrogen Gas Pressure in Main Generator2016-07-13013 July 2016 Manual Reactor Trip Due to Low Hydrogen Gas Pressure in Main Generator 05000336/LER-2016-001, Turbine Driven Auxiliary Feedwater Pump Room HELB Door Left Open2016-06-27027 June 2016 Turbine Driven Auxiliary Feedwater Pump Room HELB Door Left Open 05000423/LER-2016-003, Regarding Loss of Safety Function-Supplementary Leak Collection and Release System2016-06-0808 June 2016 Regarding Loss of Safety Function-Supplementary Leak Collection and Release System 05000423/LER-2016-002, Regarding Feedwater Isolation Signal Defeated Due to Wiring Error2016-03-23023 March 2016 Regarding Feedwater Isolation Signal Defeated Due to Wiring Error 05000423/LER-2016-001, Regarding Automatic Reactor Trip on Reactor Coolant System Low Flow Due to Loss of B Reactor Coolant Pump2016-03-16016 March 2016 Regarding Automatic Reactor Trip on Reactor Coolant System Low Flow Due to Loss of B Reactor Coolant Pump 05000336/LER-2015-003, Regarding Valid Actuation of the Reactor Protection System2016-01-0707 January 2016 Regarding Valid Actuation of the Reactor Protection System 05000336/LER-2015-002, Regarding Degraded Emergency Core Cooling System Check Valve2015-08-10010 August 2015 Regarding Degraded Emergency Core Cooling System Check Valve 05000336/LER-2015-001, Regarding Historical Oil Leakage from C RBCCW Pump Bearings Challenged Meeting Mission Time2015-06-16016 June 2015 Regarding Historical Oil Leakage from C RBCCW Pump Bearings Challenged Meeting Mission Time 05000423/LER-2015-001, Regarding Unlatched Dual Train HELB Door Results in Potential Loss of Safety Function2015-04-20020 April 2015 Regarding Unlatched Dual Train HELB Door Results in Potential Loss of Safety Function 05000423/LER-2014-004, Regarding Unlatched Dual Train HELB Door Results in Potential Loss of Safety Function2015-02-0909 February 2015 Regarding Unlatched Dual Train HELB Door Results in Potential Loss of Safety Function 05000336/LER-2014-007, Regarding Completion of Plant Shutdown Required by Technical Specifications2014-09-24024 September 2014 Regarding Completion of Plant Shutdown Required by Technical Specifications 05000336/LER-2014-006, Regarding Millstone Power Station Dual Unit Reactor Trip on Loss of Offsite Power2014-07-24024 July 2014 Regarding Millstone Power Station Dual Unit Reactor Trip on Loss of Offsite Power 05000336/LER-2014-005, Regarding Train a Containment Spray Inoperable Due to Gas Voids2014-07-16016 July 2014 Regarding Train a Containment Spray Inoperable Due to Gas Voids 05000423/LER-2014-003, Regarding Turbine Driven Auxiliary Feedwater Pump Operability Impacted by Incorrect Bearing2014-06-30030 June 2014 Regarding Turbine Driven Auxiliary Feedwater Pump Operability Impacted by Incorrect Bearing 05000336/LER-2014-004, Regarding Foreign Material Found in a Motor Lead Rendered a Motor Driven Auxiliary Feedwater Pump Inoperable2014-06-0909 June 2014 Regarding Foreign Material Found in a Motor Lead Rendered a Motor Driven Auxiliary Feedwater Pump Inoperable 05000336/LER-2014-003, Regarding Loss of Safety Function Due to Inoperable Enclosure Building2014-06-0202 June 2014 Regarding Loss of Safety Function Due to Inoperable Enclosure Building 05000336/LER-2014-002, 2 Regarding DC Circuit Hot Shorts2014-05-0505 May 2014 2 Regarding DC Circuit Hot Shorts 05000423/LER-2014-002, Regarding DC Circuit Hot Shorts2014-05-0505 May 2014 Regarding DC Circuit Hot Shorts 05000336/LER-2014-001, Regarding Completion of Plant Shutdown Required by Technical Specifications2014-03-31031 March 2014 Regarding Completion of Plant Shutdown Required by Technical Specifications 05000423/LER-2014-001, Limiting Condition for Operation Exceeded Upon Approval of Enforcement Discretion2014-03-17017 March 2014 Limiting Condition for Operation Exceeded Upon Approval of Enforcement Discretion 05000423/LER-2013-009, Regarding Secondary Containment Boundary Breach Could Have Prevented Safety Function2014-01-15015 January 2014 Regarding Secondary Containment Boundary Breach Could Have Prevented Safety Function 05000336/LER-2013-004, Unit 2 Regarding Reactor Trip While Backwashing D Waterbox2013-12-19019 December 2013 Unit 2 Regarding Reactor Trip While Backwashing D Waterbox 2024-09-26
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text
Dominion Nuclear Connecticut, Inc.
Dominion-Millstone Power Station I
Rope Ferry Road Waterford, CT 06385 APR I 4 2003 Docket No. 50-336 B18865 RE: 10 CFR 50.73(a)(2)(i)(B)
U.S. Nuclear Regulatory Cornmission Attention: Document Control Desk Washington, DC 20555 Millstone Power Station, Unit No. 2 Licensee Event Report 2003-001-00 Pressurizer Water Volume Periodically Exceeded The Technical Specifications Limit This letter forwards Licensee Event'Report (LER) 2003-001-00, which documents an event at Millstone Power Station, Unit No. 2, on February 17, 2003. This LER is being submitted pursuant to 10 CFR 50.73(a)(2)(i)(B) as any operation or condition prohibited by the plant's Technical Specifications.
There are no regulatory commitments contained within this letter.
Should you have any questions regarding this submittal, please contact Mr. Paul Willoughby at (860) 447-1791, extension 3655.
Very truly yours, DOMINION NUCLEAR CONNECTICUT, INC.
Stephen P/arver, Director Nuclear Station Operations and Maintenance Attachment (1):
LER 2003-001 -00 cc:
H. J. Miller, Region I Administrator R. B. Ennis, NRC Senior Project Manager, Millstone Unit No. 2 Millstone Senior Resident Inspector
I NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 7-31-2004
, the NRC may not conduct or sponsor, (See reverse for reouired number of dioitslcharacters for each block) and a person Is not required to respond to, the information collection.
FACILITY NAME (1)
DOCKET NUMBER (2)
PAGE (3)
Millstone Power Station - Unit No. 2 05000336 1 OF 3
TITLE (4)
Pressurizer Water Volume Periodically Exceeded The Technical Specifications Limit EVENT DATE (5)
LER NUMBER (6)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (8)
MO DAY YEAR YEAR SEQUENTIAL REV MO DAY YEAR FACILITY NAME DOCKET NUMBER Z jNUMBER NO.
I l
05000 02 17 2003 2003 - 001 - 00 04 14 2003 FACILITY NAME l DOCKET NUMBER 1
05000 OPERATING 1
THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) (11)
MODE (9) 20.2201 (b) 202203(a)(3)C(i) 50.73(a)(2)(il)(B) 50.73(a)(2)(ix)(A)
POWER 100 20.2201(d) 20.2203(a)(4) 50.73(a)(2)(iii) 50.73(a)(2)(x)
LEVEL (10)
= 20 2203(a)(1)
= 50 36(c)(116)(Al
=
50.73(a)(2)(iv)(A) 73.71 (a)(4) 20.2203(a)(2)(i) 50 36(c)(1)(ii)(A) 50.73(a)(2)(v)(A)
_ 73.71 (a)(5) 20.2203(a)(2)Q(i) 50 36(c)(2) 50.73(a)(2)(v)(B)
_ OTHER
- - q_
20.2203(a)(2)(iii) 150.46(a)(3)(ii) 50.73(a)(2)(v)(C)
Specify in Abstract below or
>^A; 5
20 2203(a)(2)(v) 50.73(a)(2)(1)(A) 50.73(a)(2)(v)(D)
In (If more space Is required, use addibonal copies of NRC Form 366A) (17) relief valves following any analyzed moderate frequency event in FSAR Chapter 14. Liquid discharge out the pressurizer relief valves could lead to a more serious plant condition due to the potential for the loss of Reactor Coolant System (RCS) [AB] fission product barrier integrity.
The limiting moderate frequency events with respect to pressurizer overfill are the loss of external load event presented in Final Safety Analysis Report (FSAR) Section 14.2.1 and the loss of normal feedwater flow event presented in FSAR Section 14.2.7. The FSAR Section 14.2.1 loss of external load event transient results in approximately a 10 percent increase in pressurizer level. The FSAR section 14.2.7 loss of normal feedwater transient results in approximately 5 percent increase in pressurizer level. Operating the plant due to the condition described above, with an actual pressurizer level approximately 3 percent higher than the maximum value allowed by the Technical Specifications, will not result in pressurizer overfill or the potential for the loss of the RCS fission product barrier integrity for either of these events.
The initial pressurizer level does not have a significant impact on the calculated peak RCS pressure for the limiting moderate frequency event with respect to RCS pressurization, the loss of external load event. The maximum calculated RCS pressure for the loss of external load event is 2717 psia, and is limited by the opening set pressure and relief capacity of the pressurizer safety valves. Operating the plant due to the condition described above, with an actual pressurizer level approximately 3 percent higher than the maximum value allowed by the Technical Specifications, will not result in a significant increase in the peak RCS pressure for this event.
Additionally, operating the plant due to the condition described above, with the actual pressurizer level approximately 3 percent higher than the maximum value allowed by the Technical Specifications, would result in a slight increase in the maximum liquid mass inventory contained within the RCS. The impact of this increased RCS liquid mass will not result in a significant increase in the calculated containment pressure consequences of the main steam line break or loss of coolant accident.
Based upon the above discussion this event has a low safety significance.
4. Corrective Action
The Engineering documents were revised to reflect the correct differential pressure calibration values for the pressurizer level transmitters. The results were subsequently incorporated into the surveillance procedure and on February 24, 2003 the two level transmitters were recalibrated to the correct values.
5. Previous Occurrences
No previous similar events were identified.
Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
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05000336/LER-2003-001, Re Pressurizer Water Volume Periodically Exceeded the Technical Specifications Limit | Re Pressurizer Water Volume Periodically Exceeded the Technical Specifications Limit | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 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)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(1) | 05000336/LER-2003-002, From Millstone, Unit 2 Regarding Reactor Trip While Performing RPS Matrix Testing | From Millstone, Unit 2 Regarding Reactor Trip While Performing RPS Matrix Testing | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2) | 05000423/LER-2003-002, From Millstone, Unit 3 Regarding Fire Safe Shutdown Strategy May Not Be Adequate for Fire Scenarios Which Assume Loss of All AC Power | From Millstone, Unit 3 Regarding Fire Safe Shutdown Strategy May Not Be Adequate for Fire Scenarios Which Assume Loss of All AC Power | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2) | 05000336/LER-2003-003, For Millstone Unit 2 the Charging System Did Not Perform Its Design Function in Response to Falling Pressurizer Level | For Millstone Unit 2 the Charging System Did Not Perform Its Design Function in Response to Falling Pressurizer Level | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 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)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000336/LER-2003-004, Reactor Coolant System Pressure Boundary Leakage Event | Reactor Coolant System Pressure Boundary Leakage Event | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 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)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2) | 05000336/LER-2003-005, Regarding Loss of Shutdown Cooling During Refueling Outage | Regarding Loss of Shutdown Cooling During Refueling Outage | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(x) | 05000336/LER-2003-006, Regarding Two Manual Reactor Due to Turbine Trips Caused by Turbine Vibrations | Regarding Two Manual Reactor Due to Turbine Trips Caused by Turbine Vibrations | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2) |
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