05000458/LER-1917-003, Regarding Manual Reactor Scram Initiated in Response to Increase in Steam Pressure During Steam Leak Troubleshooting
| ML17136A274 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 05/09/2017 |
| From: | Maguire W Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RBF1-17-0053, RBG-47754 LER 17-003-00 | |
| Download: ML17136A274 (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)(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)(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)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| 4581917003R00 - NRC Website | |
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RBG-47754 May 9, 2017 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Subject:
RBF1-17-0053 Licensee Event Report 50-458 I 2017-003-00 River Bend Station - Unit 1 Docket No. 50-458 License No. NPF-47
Dear Sir or Madam:
Entergy Operations, Inc.
River Berid Station 5485 U.S. Highway 61 N St. Francisville, LA 70775 Tel 225-381-4157 William F. Maguire Site Vice President In accordance with 10 CFR 50.73, enclosed is the subject Licensee Event Report. This document contains no commitments. If you have any questions, please contact Mr. Tim Schenk at 225-381-4177.
Sincerely, WFM/dhw Enclosure cc:
I U.S. Nuclear Regulatory Commission Region IV 1600 East Lamar Blvd.
Arlington, TX 76011-4511 NRC Sr. Resident Inspector P. 0. Box 1050
- I Licensee Event Report 50-458 I 2017-003-00 May 9, 2017 RBG-47754 Page 2 of 2 Central Records Clerk Public Utility Commission of Texas 1701 N. Congress Ave.
Austin, TX 78711-3326 Department of Environmental Quality Office of Environmental Compliance Radiological Emergency Planning and Response Section Ji Young Wiley P.O. Box 4312 Baton Rouge, LA 70821-4312
(
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020
. ~
(04-2017) htto://www.nrc.gov/reading-rm/doc-collections/nuqilgs/staff/sr1022/r30 the NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 13. PAGE River Bend Station - Unit 1 05000-458 1 OF3
- 4. TITLE Manual Reactor Scram Initiated in Resoonse to Increase in Steam Pressure During Steam Leak Troubleshootina
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR REV MONTH DAY YEAR NUMBER NO.
05000 03 10 2017 2017 003 00 05 09 2017 FACILITY NAME DOCKET NUMBER 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: Check all that apply) 1 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 50.73(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(i)
D 50.36(c)(1)(i)(A)
[8J 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 50.36(c)(2)
D 50.73(a)(2)(v)(B)
D 73.71(a)(5) 15 D 20.2203(a)(2)(iv)
D 50.46(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)
D 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 LER LICENSEE CONTACT llLEPHONE NUMBER (Include Area Code) lrim Schenk, Manager - Regulatory Assurance
~25-381-4177 CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TOEPIX FACTURER TOEPIX (see text)
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR D YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
~NO SUBMISSION DATE
~BSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On March 10, 2017, at approximately7:14 a.m. CST, the reactor operator manually actuated a reactor scram in response to an.
abnormal increase in steam pressure. Reactor power was approximately 15 percent at the time. The turbine generator had been synchronized to the grid at 5:13 a.m. on March 10, and was being closely monitored by engineers and operators since a major modification to the turbine electro-hydraulic control (EHC) system had been installed during the recent refueling outage.
Approximately 45 minutes prior to the manual scram, a main control room alarm actuated indicating a problem with the EHC system.
A few minutes later, it was reported from the turbine building that there was a steam leak in the area of the EHC steam pressure transmitters. Shortly thereafter, reactor pressure began to increase with no demand signal present, at which time the reactor operator initiated the scram. The main feedwater system remained in service, and reactor water level control performed normally as designed. No reactor safety-relief valves actuated. The main turbine bypass valves did not open following the shutdown, and engineering review determined this condition was consistent with the response to the abnormal configuration of the EHC system pressure transmitters created by efforts to isolate the leak locally. Approximately five minutes after the scram, the outboard main steam isolation valves were manually closed to limit the reactor cooldown rate. This event resulted from the incorrect installation of a new compression fitting in the steam pressure instrumentation tubing for the main turbine control system. This event is being reported in accordance with 10 CFR 50. 73(a)(2)(iv)(A) as a manual actuation of the reactor protection system.
NRC FORM 366 (04-2017)
REPORTED CONDITION SEQUENTIAL NUMBER 003 REV NO.
00 On March 10, 2017, at approximately 7:14 a.m. CST, the reactor operator manually actuated a reactor scram in response to an abnormal increase in steam pressure. Reactor power was approximately 15 percent at the time, and the turbine generator was on* line. The reactor had been taken critical at 4:39 p.m. on March 8 following a refueling outage, and power ascent was in progress. The turbine generator had been synchronized to the grid at 5:13 a.m. on March 10, and was being closely monitored by engineers and operators since a major modification to the turbine electro-hydraulic control (EHC) system [JI] had been installed during the outage.
Approximately 45 minutes prior to the manual scram, a main control room alarm actuated indicating a problem with the EHC system. A few minutes later, it was reported from the turbine building that there was a steam leak in the area of the steam pressure transmitters. The operations shift manager held a briefing with the operators on potential effects of the field observations, the single-point vulnerability of the transmitter configuration, and the possibility of a main turbine trip. Shortly thereafter, reactor pressure began to increase with no demand signal present. This response likely resulted from efforts to isolate the steam leak.
The main feedwater system remained in service, and reactor water level control was performed normally. No reactor safety-relief valves actuated. The main turbine bypass valves did not open following the shutdown, and engineering review determined this condition was consistent with the response to the abnormal configuration of the EHC system pressure transmitters created by efforts to isolate the leak locally. Approximately five minutes after the scram, the outboard main steam isolation valves were manually closed to limit the reactor cooldown rate.
Other than scheduled testing on the Division 1 diesel generator, no safety-related systems were out of service at the time of the scram.
This event is being reported in accordance with 10 CFR 50. 73(a)(2)(iv)(A) as a manual *actuation of the reactor protection system INVESTIGATION During the recent refueling outage, a digital control system had been installed on the main steam turbine bypass I pressure regulation system. Part of that modification involved the installation of a new main turbine steam throttle pressure transmitter (**PT**) near the high pressure turbine. The transmitter was to be installed adjacent to two existing steam pressure transmitters by adding a tee fitting into an existing run of tubing (**TBG**). One of the newly-installed tubing compression fittings* separated during efforts to isolate the leak. Examination of the components concluded that the ferrule in the fitting was not fully inserted, and did not compress adequately to engage the surface of the tubing when the nut was tightened. Page 2 of 3
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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/nureos/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 River Bend Station - Unit 1 05000-458 YEAR 2017 SEQUENTIAL NUMBER 003 REV NO.
00
!The tee fitting was installed by two qualified pipefitters who were contracted for the refueling outage, with oversight provided by a contract pipefitter foreman. The field work to complete the fit-up of the tee connection was specified as Quality Control (QC) Hold Point. The contract foreman stated that he was present during the fit-up of the tee connection and that the technicians performed the work as required, fitting each of the three connections and tightening them one at a time starting with the top compression fitting and ending. with the bottom connection. When the fittings were tight, he observed the technicians use a gap tool to verify proper gap and engagement of the compression nut on the tee connection body.
CAUSAL ANALYSIS
- !This event resulted directly from the incorrect installation of the tee compression fitting for the new steam pressure transmitter. A contributing cause was the lack a standard process on how to properly verify compression tubing and fitting engagement is maintained during the tightening process.
(
CORRECTIVE ACTION TO PREVENT RECURRENCE A maintenance procedure will be developed to address the proper installation of compression fittings. This action will be tracked in the corrective action program.
PREVIOUS OCCURRENCE EVALUATION RBS has reported no similar events in the last three years.
SAFETY SIGNIFICANCE
h"he plant responded as designed to the transient. The response of the main turbine bypass valves resulted from the efforts to isolate the steam leak, and was, by itself, of no consequence to the operators' response to the event. The steam leak was isolated by closure of an instrument valve. The outboard main steam isolation valves were manually closed in accordance with procedures to manage reactor cooldown rate. There were no injuries as a result of the steam leak. This event was, thus, of minimal significance to the health*.and safety of the public.
(NOTE: Energy Industry Identification System component function identifier and system name of each component or system referred to in the LER are annotated as (**XX**) and [XX], respectively.) Page 3 of 3