05000247/LER-1917-003, Regarding Technical Specification Violation of Section 3.3.1 RPS Instrumentation
| ML17241A047 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 08/23/2017 |
| From: | Vitale A Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-17-111 LER 17-003-00 | |
| Download: ML17241A047 (5) | |
| 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) |
| 2471917003R00 - NRC Website | |
text
- =w=* Entergx NL-17-111 August 23, 2017 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop O-P1-17 Washington, D.C. 20555-0001 Indian Point Energy Center 450 Broadway, GSB P.O. Box249 Buchanan, N.Y. 10511-0249 Tel (914) 254-6700 Anthony J. Vitale Site Vice President
Subject:
Licensee Event Report# 2017-003-00, "Technical Specification Violation of Section 3.3.1 RPS Instrumentation" Indian Point Unit No. 2 Docket No. 50-247 DPR-26
Dear Sir or Madam:
Pursuant to 10 CFR 50. 73(a)(1 ), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2017-003-00. The attached LER identifies an event where there was a Technical Specification Violation of Section 3.3.1 Reactor Protection System Instrumentation, which is reportable under 10 CFR 50. 73(a)(2)(i)(B).
This condition was recorded in the Entergy Corrective Action Program as Condition*
Report CR-IP2-2017-02193.
There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, Regulatory Assuranc~, at Indian Point Energy Center at (914) 254-6710.
Sincerely, tr ~"~J v.~
AJV/trj b/z_g/'/f cc:
Mr. Daniel H. Dorman, Regional Administrator, NRC Region I NRC Resident Inspector's Office Ms. Bridget Frymire, New York State Public Service Commission
NRCFORM366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2017) httg://www.nrc.gov/reading-rm/doc-collections/nur§9s/staff/sr1022/Qll the NRC may not conduct or sponsor, and a person is not required to respond.to, the information collection.
- 3. PAGE Indian Point Unit 2 05000247 1 OF4
- 4. TITLE Technical Specification Violation of Section 3.3.1 RPS Instrumentation
- 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.
05000 06 27 2017 2017
- - 003
- - 00 08 23 2017 FACILITY NAME DOCKET NUMBER N/A 5000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: Check all that anoly) 2 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 20.22o~(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.3s(c)(2)
D 50.73(a)(2)(v)(B)
D 73.71(a)(5) 0%
D 20.2203(a)(2)(iv)
D so.4s(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)
D 50.73(a)(2)(v)(D)
D 73.77(a)(2)(i)
D 20.2203(a)(2)(vi) 18150.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 73.77(a)(2)(ii)
D 50.73(a)(2)(i)(C) 00THER Specify in Abstract below or in NRG Form 366A
- 12. LICENSEE CONTACT FOR THIS LEA ICENSEE CONT ACT rELEPHONE NUMBER {Include Area Code) lcharles Bristol
~142546665 CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TOEPIX FACTURER TOEPIX A
N/A N/A N/A N/A
- :. *~:.1
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED' *
.. *~:*,, ' -
MONTH DAY YEAR
\\*
0 YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
[8J NO
. SUBMISSION DATE
~BSTRACT (Umit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On June 27, 2017, during reactor startup, power was raised from Mode 3 to Mode 2 and above the P-6 (Intermediate Range Neutron Flux) interlock with the P-6 Switches in the wrong position. With P-6 inoperable this was a violation of the requirement of Technical Specification (TS) Limiting Condition of Operation (LCO) 3.3.1 and resulted in a 60 Day Licensee Event Report (LER). LCO 3.3.1, Table 3.3.1-1, Item 17 states that the Intermediate Range Neutron Flux, P-6 shall be operable.
On June 27, 2017, during performance of 2-PT-V63A, Reactor Protection System (RPS) Logic Train 'A' Partial Functional Test, Instrumentation and Control (l&C) Technicians left two Intermediate Range P-6.
switches in the wrong-position, which resulted in an unplanned entry intoLCO 3.3.1, due to inoperable RPS instrumentation. This inoperable RPS instrumentation resulted in a TS LCO 3.3.1 violation when reactor power was raised from Mode 3 to Mode 2.
I NR"c FORM 366 (04-2017)
SEQUENTIAL NUMBER
- - 003 REV NO.
- - 00 On June 27, 2017, during reactor startup, power was raised from Mode 3 to Mode 2 and above the P-6 (Intermediate Ranger Neutron Flux) interlock. With P-6 inoperable this was a violated of the requirement of Technical Specification (TS) Limiting Condition of Operation (LCO) 3.3.1 and resulted in a 60 Day Licensee Event Report (LER).
Unit 2 reactor was in Mode 3 in preparation for start-up following a forced outage to repair the 22 Main Boiler ~eed-Pump control system. As required by plant technical specifications, forced outage surveillance test 2-PT-V63A, Reactor Protection Logic Train 'A' Partial Functional Test and 2-PT-V63B, Reactor Protection Logic Train 'B' Partial Functional Test were scheduled to be performed prior to entering Mode 2. These tests perform an actuation logic test, channel operation test, and trip actuating device operational test (T ADOT) on portions of the reactor protection system (RPS) logic circuit that are not able to be tested at full power conditions.
Some of the switch manipulations, including the P-6 permissive switches, involve multiple actions requiring a team of individuals to coordinate rotation of switches to the right, and then pushing and holding switches while verifying test panel trip lamps and trip bus volt lamps are illuminated.
The test is performed in th~ Unit 2 Control Room, and all RPS test switches are located in adjacent racks designated as Panel 'RLTRA' (Train 'A') and 'RLTRB' (Train 'B').
Three I&C technicians were assigned to perform the two surveillance tests sequentially. The 'A' train test was perfortned first. The technicians and their first-line supervisor conducted the pre-job briefing for the tests in accordance with EN-HU-102 Human Performance Traps &Tools requirements. The test details were discussed, along with roles and responsibilities and stop work criteria. During the work preparations, the technicians identified tec;hnical errors in the prerequisites sections of both surveillance tests, requiring DRNs to be issued prior to the start of work. Also identified was that three of the sections in the tests were not required to be performed because they had been performed during the 2R22 refueling outage and were within the required surveillance frequency. The test was not considered a high risk acti".'ity and consequences for leaving switches in the wrong position were not identified during the pre-job briefing. Had the importance of the switch configuration to reactor safety been understood and discussed during work preparations, additional measures would have been taken to provide for and ensure high quality work. Page 2 of 4 (04-2017)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: 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/reading-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.
ri Indian Point Unit 2 05000-247 YEAR 2017
CAUSE OF EVENT
- 3. LER NUMBER SEQUENTIAL NUMBER
- - 003 REV NO.
- - 00 I&C technicians did not restore test permissive intermediate range switches 1/N35D and 1/N36D to their proper position during surveillance testing as required by procedure.
Poor Work Practices Workers did not apply essential maintenance fundamentals to ensure proper switch configuration was preserved during the surveillance test:
Risk Recognition and Mitigation Proficiency Supervisory I Management Methods The responsible supervisor and maintenance leadership did not provide sufficient oversight or guidance to promote error-free performance.
Pre-Job Briefing I Work Preparation Performing Monitoring - Human Performance (HU) elements Written Communications-l&C Procedures The test procedure 2-PT-V063A is not aligned with the writer's guide best practices to reduce the potential for human error:
Operations Interface - Configuration Control Operations did not apply adequate methods to validate system configuration and overly relied on maintenance processes and procedures
CORRECTIVE ACTIONS
A site-wide focused stand down was held on July 10, 2017 to review recent IPEC human performance event The l&C technicians involved had their qualifications administratively suspended pending the completion of the human performance culpability reviews A Performance Analysis was completed as part of the AC~ report to identify knowledge and skill gaps for determining training needs Revise IP2 and IP3 SI Logic and Reactor Protection System surveillance tests to include steps to
.perform independent verification of as-left switch lineups Revise 2-PT-V053 and 3-PT-V053 mode change checkoff list(s) to validate critical Reactor Protection System switch lineups following testing Page 3 of 4 (04-201,7)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 3/31/2020
~
\\:.* ~.Y LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET (See NUREG-1022,-R.3 for instruction and guidance for completing this form http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/@)
, the NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. LEA NUMBER Indian Point Unit 2 05000-247 YEAR SEQUENTIAL NUMBER REV NO.
2017
- - 003
- - 00
EVENT ANALYSIS
This LER is being submitted pursuant to Title 10 Code of Federal Regulations 50.73(a)(2)(i)(B) "Any operation or condition which was prohibited by the plant's Technical Specifications" PAST SIMILAR EVENTS In July 2016, the IP2 reactor tripped due to the inadvertent operation of the 'B' RPS bypass key out of sequence during Reactor Protection logic testing (CR-IP2-2016-04320. In March 2013, an inadvertent Safety Injection (SI) occurred during the 3Rl 7 outage while conducting the reactor protection system functional monthly test (CR-IP3-2013-02115). These events were not discussed prior to the task because the specific conditions were not relevant to the performance of 2-PT-V63A.
SAFETY SIGNIFICANCE
The plant operation impact was that the Source Range flux high level trip was not able to be manually bypassed to continue plant startup. The inoperability of P-6 did not impact the ability to shut down the reactor or maintain it in a safe shutdown condition (both the Source Range and Intermediate Range reactor trip functions were fully operable) or to mitigate the consequences of an accident as described in theFSAR.
This event is reportable and a 60 day LER is required due to violation of the requirements of LCO 3.3.1 During reactor startup, power was raised from Mode 3 to Mode 2 and above the P-6 interlock. P-6 was inoperable at the time, resulting in LCO 3.3.1, Table 3.3.1-1, Item 17 not being met.
This event was classified as a Level 2 Consequential (Major) Component Mispositioned Event because it was an unintentional or unexpected component manipulation that resulted in a major impact to operation of the plant (unplanned shutdown and reportable entry into an LCO). Page 4 of 4