05000293/LER-2019-001, Reactor Core Isolation Cooling System Declared Inoperable During Surveillance Testing
| ML19064A593 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 02/28/2019 |
| From: | Miner P Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 2019-001-00 | |
| Download: ML19064A593 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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)(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)(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)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| 2932019001R00 - NRC Website | |
text
2.19.006 February 28, 2019 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Peter J. Miner Manager, Regulatory Assurance 10 CFH 50.73 SUBJECT: Licensee Event Report 2019-001-00, Reactor Core Isolation Cooling System Declared Inoperable During Surveillance Testing Pilgrim Nuclear Power Station NRG Docket No. 50-293 Renewed Facility Operating License No. DPR-35
Dear Sir or Madam:
The enclosed Licensee Event Report 2019-001-00, Reactor Core Isolation Cooling System Declared Inoperable During Surveillance Testing, is submitted in accordance with Title 1 O Code of Federal Regulations 50.73.
There are no regulatory commitments contained in this letter.
If you have any questions or require additional information, please contact me at 508-830-7127.
Sincerely, PJM/rjm : Licensee Event Report 2019-001-00, Reactor Core Isolation Cooling System Declared Inoperable During Surveillance Testing
Letter No. 2.19.006 Page 2 of 2 cc:
NRG Region I Regional Administrator NRG NRR Project Manager - Pilgrim NRG Senior Resident Inspector - Pilgrim Letter Number 2.19.006 Licensee Event Report 2019-001-00, Reactor Core Isolation Cooling System Declared Inoperable During Surveillance Testing (3 Pages)
SNRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2017) ht!Q://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3L) the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
.PAGE Pilgrim Nuclear Power Station 05000-293
- 1 OF3
- 4. TITLE Reactor Core Isolation Cooling System Declared Inoperable During Surveillance Testing
- 5. EVENT DATE
- 6. LEA NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR REV MONTH DAY YEAR N/A NUMBER NO.
N/A 01 08 2019 2019
- - 001
- - 00 02 28 2019 FACILITY NAME DOCKET NUMBER N/A N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
N D 20.2201 (b)
D 20.2203(a)(3)(i)
D 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
D 20.2201 (d)
D 20.2203(a)(3)(ii)
D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)(1)
D 20.2203(a)(4)
D so.73(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(i)
D 50.36(c)(1 )(i)(A)
D 50.73(a)(2)(iv)(A)
D 50.73(a)(2)(x)
- 10. POWER LEVEL D 20.2203(a)(2)(ii)
D 50.36(c)(1 )(ii)(A)
D 50.73(a)(2)(v)(A)
D 73.71 (a)(4)
D 20.2203(a)(2)(iii)
D so.36(c)(2)
D 50.73(a)(2)(v)(B)
D 73.71(a)(5) 100 D 20.2203(a)(2)(iv)
D so.46(a)(3)(ii)
D 50.73(a)(2)(v)(C)
D 73.77(a)(1)
D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A) 18l 50.73(a)(2)(v)(D)
D 73.77(a)(2)(i)
D 20.2203(a)(2)(vi)
D 50.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 73.77(a)(2)(ii)
D 50.73(a)(2)(i)(C) 0 OTHER Specify in Abstract below or in NRG Form 366A
- 12. LICENSEE CONTACT FOR THIS LEA LICENSEE CONTACT IELEPHONE NUMBER (Include Area Code)
!Mr. Peter J. Miner - Regulatory Assurance Manager
~08-830-7127 CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TOEPIX FACTURER TOEPIX X
BN FIC G080 y
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR 0 YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
~ NO SUBMISSION DATE
!ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On January 8, 2019, the Reactor Core Isolation Cooling (RCIC) system was declared inoperable. This action was taken because the RCIC system turbine-pump did not achieve the acceptance criteria of flow rate, discharge pressure, and speed during the performance of a surveillance test.
Due to the unexpected response a Failure Modes Analysis was initiated to determine the cause. No deficient conditions were identified during troubleshooting. However, based on the investigation, the team determined that the direct cause was output signal loss from the RCIC flow controller. A spare calibrated controller was installed, the system was tested, and successfully returned to service.
The High Pressure Coolant Injection system was operable and capable of providing high pressure core cooling similar to the RCIC system. This event had no impact on the health and/or safety of the public.
This report is submitted in accordance with Title 1 O Code of Federal Regulations 50.73(a)(2)(v)(D).
BACKGROUND SEQUENTIAL NUMBER
- - 001 REV NO.
- - 00 The Pilgrim Station Updated Final Safety Analysis Report (UFSAR) states that the Reactor Core Isolation Cooling (RCIC) system is designed to provide makeup water to the reactor vessel following reactor isolation in order to prevent the release of radioactive materials to the environment as a result of inadequate reactor core cooling. The system consists of a steam driven turbine-pump and associated valves and piping capable of delivering makeup water to the reactor vessel over a range of reactor pressures. The system can be operated automatically or manually, and is one of the systems credited in the UFSAR for a design basis Control Rod Drop Accident (CRDA). The RCIC system is sufficient to maintain reactor vessel water level at an acceptable limit during this event.
EVENT DESCRIPTION
On January 8, 2019, the RCIC turbine was started in automatic from the main control room per a quarterly
~urveillance. During the test run, the turbine did not reach rated conditions. Operators identified that flow controller, FIC-1340-01, output meter was indicating zero, which was unexpected. Operators attempted to change demand by varying the flow controller setpoint from 375 gallons per minute (gpm) to 425 gpm, but no change in output or flow occurred. Operators then manually stopped the RCIC turbine from the main control room and declared the RCIC system inoperable.
rrhe NRG Operations Center was notified of the event in accordance with Title 1 O Code of Federal Regulations (CFR) 50. 72 at 1545 hours0.0179 days <br />0.429 hours <br />0.00255 weeks <br />5.878725e-4 months <br /> on January 8, 2019.
The event occurred during power operation while at 100 percent reactor power. The reactor mode selector switch was in the RUN position.
CAUSE OF THE EVENT
Due to the unexpected response a Failure Modes Analysis was initiated to investigate the cause. No deficient conditions were identified during troubleshooting. However, based on the investigation, the team determined that the direct cause was output signal loss from RCIC flow controller FIC-1340-01.
CORRECTIVE ACTIONS
rrhe installed flow controller was removed and replaced with a spare calibrated controller. An operability test run was completed and the system was successfully returned to service.
~ny further corrective actions will be documented in the corrective action program.
Page 2 of 3 (04-2017)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 3/31/2020 LICENSEE EVENT REPORT {LER)
CONTINUATION SHEET (See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc.gov/readinq-rm/doc-collections/nuregs/staff/sr1022/r3/)
, the NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. LER NUMBER YEAR Pilgrim Nuclear Power Station 05000- 293 2019
SAFETY CONSEQUENCES
SEQUENTIAL NUMBER
- - 001 REV NO.
- - 00 The RCIC system function is to provide high pressure makeup water to the reactor vessel after isolation of the vessel. The RCIC and High Pressure Coolant Injection (HPCI) systems provide Single-Failure Proof response to high pressure reactor vessel injection following a CRDA. The HPCI system was operable while the RCIC system was inoperable. If the HPCI system were to become inoperable and core cooling was necessary, the automatic depressurization system would depressurize the reactor in order to allow low pressure core cooling by the residual heat removal system (low pressure coolant injection mode), and/or core spray system.
There were no actual consequences to safety of the general public, nuclear safety, industrial safety, or radiological safety for this event based on the availability of appropriate high pressure core cooling.
REPORT ABILITY The condition is reportable under 10 CFR 50.72 and 50.73 based on NUREG-1022 guidance as a condition that could have prevented fulfillment of a safety function because the RCIC system was inoperable as defined in the Technical Specifications. However, the condition is not a safety system functional failure because an engineering evaluation concluded that the safety function would have been fulfilled by HPCI, if high pressure core cooling were needed. HPCI was operable at the time that RCIC failed to meet it surveillance requirements and was declared inoperable.
PREVIOUS EVENTS LER 2004-004-00, RCIC System Declared Inoperable During Surveillance Testing due to Flow Controller Potentiometer Oxidation REFERENCES CR-PNP-2019-00145 Page 3 of 3