05000395/LER-2012-001, Regarding Core Exit Thermocouples Inoperable Due to an Inadequate Maintenance Procedure
| ML12079A286 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 03/16/2012 |
| From: | Gatlin T South Carolina Electric & Gas Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 12-01-00 | |
| Download: ML12079A286 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 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)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3952012001R00 - NRC Website | |
text
Thomas D. Gatlin Vice President, Nuclear Operations 803.345.4342 A SCANA COMPANY March 16, 2012 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555
Dear Sir / Madam:
Subject:
VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LICENSEE EVENT REPORT (LER 2012-001-00)
CORE EXIT THERMOCOUPLES INOPERABLE DUE TO AN INADEQUATE MAINTENANCE PROCEDURE Attached is Licensee Event Report (LER) No. 2012-001-00 for the Virgil C. Summer Nuclear Station Unit 1. This report describes a condition where Core Exit Thermocouples would not be operable for accident monitoring. This report is submitted in accordance with 10 CFR 50.73(a)(2)(i)(B).
This letter and attached LER contain no new commitments and no revisions to existing
commitments
Should you have any questions, please call Bruce Thompson at (803) 931-5042.
Very truly yours, Thomas D. Gatlin TSITDG/wm Attachment c:
K. B. Marsh S. A. Byrne J. B. Archie N. S. Carns J. H. Hamilton R. J. White W. M. Cherry V. M. McCree R. E. Martin NRC Resident Inspector M. N. Browne P. Ledbetter J. C. Mellette EPIX Coordinator K. M. Sutton INPO Records Center Marsh USA, Inc.
R. J. Schwartz NSRC RTS (CR-11-01807)
File (818.07)
PRSF (RC-12-0037)
. yJfL Virgil C. Summer Station - Post Office Box 88
- Jenkinsville, SC. 29065. F (803) 345-5209
NRC FORM'366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010)
, the NRC may sfor each block) not conduct or sponsor, and a person is not required to respond to, the digits/characters finformation collection.
- 3. PAGE Virgil C. Summer Nuclear Station Unit 1 05000 395 1 OF 5
- 4. TITLE Core Exit Thermocouples Inoperable due to an Inadequate Maintenance Procedure
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NO.
05000 FACILITY NAME DOCKET NUMBER 01 17 2012 2012 1
0 03 16 2012 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1 El 20.2201(b)
[J 20.2203(a)(3)(i)
E] 50.73(a)(2)(i)(C) 5 50.73(a)(2)(vii) 5 20.2201(d)
E] 20.2203(a)(3)(ii)
E] 50.73(a)(2)(ii)(A) 5 50.73(a)(2)(viii)(A)
O 20.2203(a)(1)
[J 20.2203(a)(4) 5] 50.73(a)(2)(ii)(B) 5 50.73(a)(2)(viii)(B) 5 20.2203(a)(2)(i) 5 50.36(c)(1)(i)(A) 5 50.73(a)(2)(iii) 5 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL E] 20.2203(a)(2)(ii) 5 50.36(c)(1)(ii)(A) 5 50.73(a)(2)(iv)(A) 5 50.73(a)(2)(x)
[]
20.2203(a)(2)(iii) 5 50.36(c)(2)
E] 50.73(a)(2)(v)(A) 5 73.71(a)(4) 100%
E] 20.2203(a)(2)(iv)
E] 50.46(a)(3)(ii) 5 50.73(a)(2)(v)(B) 5 73.71(a)(5)
O 20.2203(a)(2)(v) 5 50.73(a)(2)(i)(A)
[J 50.73(a)(2)(v)(C) 5 OTHER El 20.2203(a)(2)(vi)
[] 50.73(a)(2)(i)(B)
[] 50.73(a)(2)(v)(D)
Specify in Abstract below or in =
PLANT IDENTIFICATION Westinghouse - Pressurized Water Reactor EQUIPMENT IDENTIFICATION Control Rod Drive Mechanism (CRDM) Cable Bridge Core Exit Thermocouples (TI)
Channel A ITEs 2, 4, 9, 12, 13, 15, 19, 21, 22, 23, 24, 25, 26, 27, 28, 29, 31, 32, 33, 35, 39, 41, 42, 45, 46, and 47 Channel B ITEs 1, 3, 5, 6, 7, 8, 10, 11, 14, 16, 17, 18, 20, 30, 34, 36, 37, 38, 40, 43, 44, 48, 49, 50, and 51 IDENTIFICATION OF EVENT At the beginning of Refuel (RF) 19, Virgil C. Summer Nuclear Station (VCSNS) contract employees attempted to raise the CRDM Cable Bridge using VCSNS Maintenance Procedure GMIP-100.007, "Maintenance Support for Refueling." When preparing to lift the Cable Bridge, the crew noted that none of the twenty-four (24) hold-down bolts that were supposed to be removed, per the procedure, were installed. The discovery was entered into the VCSNS Corrective Action Program under CR-11-01807, and the subsequent investigation determined that the hold-down bolts were not installed at the end of RF-18. The evaluation for past operability (completed on January 17, 2012), determined that, had there been a Loss of Coolant Accident (LOCA), the CRDM Cable Bridge could have moved enough to damage the Core Exit Thermocouple (CETC) cables and resulted in a loss of core exit temperature monitoring capability. This condition was a violation of Technical Specification 3.3.3.6, "Accident Monitoring Instrumentation," since the required number of CETCs per core quadrant per channel would not have been available during a postulated LOCA.
EVENT DATE January 17, 2012 REPORT DATE March 16, 2012
CONDITIONS PRIOR TO EVENT
Mode 1, 100% Power
DESCRIPTION OF EVENT
Each refueling outage, 24 hold-down bolts are removed when the CRDM Cable Bridge is raised to support Reactor Vessel Head removal. On April 18, 2011, a VCSNS contract crew attempted to raise the CRDM Cable Bridge using Maintenance Procedure GMP-100.007. When the crew was performing Step 7.2.4.A, "REMOVE bolts which hold bridges to support," it was noted that none of the 24 hold-down bolts were installed. The discovery was documented in the VCSNS Corrective Action Program under CR-11-01807. An Apparent Cause Evaluation (ACE) was performed to determine why the hold-down bolts had not been installed at the end of RF-18. The ACE determined the cause was an inadequate procedure because it did not require verification or documentation of bolt installation during CRDM Cable Bridge reassembly.
On January 17, 2012, VCSNS personnel completed the past operability analysis that determined the CETCs required for accident monitoring were inoperable. The analysis determined the CETCs would have been inoperable if a LOCA had occurred during station operation during the period the CRDM Cable Bridge hold-down bolts were not installed because the CRDM Cable Bridge would have been free to pivot upward, damaging cable connections at the plug boards.
CAUSE OF THE EVENT
The ACE was conducted to identify why the CRDM Cable Bridge hold-down bolts were not reinstalled during RF-18. The ACE identified that Maintenance Procedure GMIP-100.007 was inadequate because Step 7.4.20(G), "Bolt Cable Bridge sections to floor supports," did not require verification or documentation of the installation of the support bolts.
VCSNS performed an extent of condition evaluation that included review of similar refueling activities which require unbolting of components. The applicable procedures were reviewed to ensure they require appropriate verification of removal and reinstallation of bolting.
No additional instances of procedure inadequacy were identified.
ANALYSIS OF EVENT
Design Calculation DC0316F-002, "Reactor Building CRDM Missile Shield Design," provides an analysis of the movable CRDM Cable Bridge loading for LOCA pressure transient and Operating Basis Earthquake (OBE)/Safe Shutdown Earthquake (SSE) events. The calculation evaluates the potential movement of the CRDM Cable Bridge should one of these events occur. The movement of the Cable Bridge determines the affect on the cables running through the bridge or any other plant equipment. No other accidents were deemed to have a credible impact on the CRDM Cable Bridge loading.
Normal Operation During normal operation, the absence of the hold-down bolts would not impact the ability of the CRDM Cable Bridge to perform its function. No external forces are exerted on the bridge; therefore, the bridge and cables running inside it would remain undamaged.
Seismic Event The seismic load used in DC0316F-002 is 0.5g vertical based on the station's design basis earthquake (DBE) values. The acceleration required to lift the CRDM Cable Bridge is Ig; therefore, the 0.5g vertical acceleration calculated for a DBE would not be sufficient to lift the bridge. The bridge rests on a lateral support beam located beneath the bridge. The bolts are not necessary for any forces encountered in the downward direction.
Based on this analysis, the absence of the hold-down bolts would not have a negative effect on the CRDM Cable Bridge during a design basis seismic event.
LOCA The calculated upward force on the CRDM Cable Bridge at the bolt locations due to a LOCA is 49.9 kips. This force is generated by assuming the reactor coolant flashes to steam at the worst-case location of the leak. Without the hold-down bolts installed, the CRDM Cable Bridge would be free to pivot up towards the cables plugged into the fixed plug boards. The exact amount of upward movement was not determined; however, it was assumed to be significant since the calculated force from the LOCA is much more than the downward force from the weight of the bridge. Based on this analysis, it is assumed that the CRDM Cable Bridge would move sufficiently to damage the cable connections at the plug boards. An engineering evaluation is underway to determine the potential impacts on other Systems, Structures, or Components (SSCs) if the CRDM Cable Bridge were to become disconnected from the wall during a LOCA. If required, VCSNS will submit a revision to this LER based on the engineering analysis results.
Overall Conclusion During a seismic event, the absence of the hold-down bolts does not impact the Reactor Vessel, Reactor Coolant System integrity, or any reactor trip/accident mitigation equipment because the CRDM Cable Bridge would remain affixed to the vertical concrete wall by the pivot shafts and would therefore be capable of fulfilling anti-falldown requirements. During a LOCA scenario, the uplift force on the CRDM Cable Bridge is assumed to result in a loss of the cable connections. The loss of CRDM power and rod position indication during a LOCA is not significant as CRDM power and rod position indication are not required for safe shutdown. However, the CETCs are required for Regulatory Guide 1.97, "Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants," accident monitoring and for compliance with Technical Specification 3.3.3.6. Per Technical Specification Table 3.3-10, "Accident Monitoring Instrumentation," four CETCs per Core Quadrant per Channel are required. Because the CETC requirements would not have been met during a postulated LOCA, the associated Post Accident Monitoring System qualification was not met, and this event is reportable.
The impacts of the as-found configuration documented in this LER (loss of CRDM power, Control Rod position indication, and CETC indication during LOCAs) is not risk significant. Losses of CRDM power and Control Rod position indication are not important because the resulting insertion of all control rods is the desired response following a LOCA. Loss of the CETC indication is not risk significant because the CETCs do not provide control functions and because alternate means of indication (RCS Hot Leg temperature or Reactor Vessel level indication) are provided as options in the Emergency Operating Procedures. For these reasons, the CETCs are not considered important contributors to risk and are, therefore, not modeled in the VCSNS Probabilistic Risk Assessment model either explicitly as failures, or implicitly as part of the Human Error Probability calculations for operator action failure rates. Because the alternate indication is not affected by this condition, there is no increase in Core Damage Frequency (CDF). Additionally, there is no impact of this event on Large Early Release Frequency (LERF) because the unavailability of the CETCs does not affect the ability to mitigate the containment bypass events calculated in the VCSNS model and because lack of CETC availability does not impact the assumptions in the LERF model regarding containment failure. For these reasons, this condition is not risk significant.
CORRECTIVE ACTIONS
The CRDM Cable Bridge hold-down bolts were re-installed prior to startup from RF-19. To prevent a recurrence of this event, Maintenance Procedure GMP-100.007 was revised to add a verification step for installing the hold-down bolts. The completion of this step must be documented and verified on page 2 of 2 of Attachment 11, "DATA SHEET FOR REFUELING AND RECOVERY FROM REFUELING."
PRIOR OCCURRENCES A search was conducted within the station's Corrective Action Program using search criteria for "CRDM" and "missing bolts." No prior events were identified related to the CRDM Cable Bridge.