05000390/LER-1917-012, Regarding Error in Plant Emergency Procedures Leads to a Condition Prohibited by the Technical Specifications
| ML17296A447 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 10/23/2017 |
| From: | Simmons P Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 17-012-00 | |
| Download: ML17296A447 (8) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2) |
| 3901917012R00 - NRC Website | |
text
Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381 October 23,2017 10 cFR 50.73 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Subject:
Watts Bar Nuclear Plant, Units 1 and2 Facility Operating License Nos. NPF-90 and NPF-96 NRC Docket Nos. 50-390 and 50-391 Licensee Event Report 39012017-012-00, Error in Plant Emergency Procedures Leads to a Condition Prohibited by the Technical Specifications This submittal provides Licensee Event Report (LER) 39012017-012-00. This LER provides details concerning a condition where a procedure error related to use of the Essential Raw Cooling Water crosstie resulted in a condition prohibited by the Technical Specifications (TS). This condition is being reported as a condition prohibited by TS in accordance with 10 CFR 50.73(a)(2)(iXB).
There are no regulatory commitments contained in this letter. Please direct any questions concerning this matter to Kim Hulvey, WBN Licensing Manager, at (423) 365-7720.
Respectfully, ak:--
Paul Simmons Site Vice President Watts Bar Nuclear Plant Enclosure cc: See Page 2
U.S. Nuclear Regulatory Commission Page 2 October 23,2017 cc (Enclosure):
NRC Regional Administrator - Region ll NRC Senior Resident Inspector - Watts Bar Nuclear Plant
NRC FORM 366 (o4-2o17)
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iffi U.S. NUCLEAR REGULATORY COMMISSION LTGENSEE EVENT REPORT (LER)
APPROVED BY OMB: NO.3150-0104 EXPIRESi 0313112020
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection,
- 1. FACILITY NAME Watts Bar Nuclear Plant, Unit 1
- 2. DOCKET NUMBER 05000390
- 3. PAGE 10F6
- 4. TITLE Error in Plant Emergency Procedures Leads to a Condition Prohibited by the Technical Specifications
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTHI DAY I YEAR YEAR I tt-ir"?ff' REV NO.
MONTH DAY YEAR FACILIry NAME I
DOCKET NUMBER Watts Bar Nuclear Ptant, Unit 2 IOSOOOSS1 08 23 I 2017 2017 - 012 00 10 23 2017 FACILITY NAME I
OOCrcET NUMBER
- 9. OPERATING MODE
'II.THISREPORTISSUBMITTEDPURSUANTTOTHEREQUIREMENTSOF'10CFR$:
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X 50.73(aX2XiXB) tr 50 73(aX2Xvii) n B7r(a)(2xii) tl 50 73(a)(2XiXC) tr OTHER Specify in Abstract below or in Date Event 8123117 CR 1331422 generated to document concern that step to cross tie ERCW headers had been removed without evaluating impact on past operability.
1016117 Past Operability Evaluation (POE) for CR 1331422 determines that opening ERCW cross tie would have resulted in inoperability of Train A for 18.9 days between March 19,2017 and April25,2017.
D. Manufacturer and Model Number of Components that Failed During the Event There were no failed components that contributed to this event.
E. Other Systems or Secondary Functions Affected
No other systems or secondary functions were affected.
F. Method of discovery of each Component or System Failure or Procedural Error
The issue was identified as concern subsequent to it being corrected. A past operability evaluation performed after the condition was identified determined the condition to be reportable.
G. Failure Mode and Effect of Each Failed Component No actual equipment failures occurred during this event.
H. Operator Actions
No actual event was ongoing related to this report.
l. Automatically and Manually lnitiated Safe$ System Responses Not applicable.
III. CAUSE OF THE EVENT
A. The cause of each component or system failure or personnel error, if known.
The cause of this procedural error is under investigation and will be provided in a supplement to this report.
B. The cause(s) and circumstances for each human performance related root cause.
The cause of this procedural error is under investigation and will be provided in a supplement to this report.
IV. ANALYSIS OF THE EVENT
NRC FORM 365A (04-2017)
The WBN plant is provided with a common ERCW system serving both units (see Figure). Eight ERCW pumps are provided, with two pumps aligned to each 6.9 kV Shutdown Board (SDBD) and aligned to two ERCW trains A and B. The ERCW system supplies cooling water to the Component Cooling Water System (CCS) heat exchangers, the containment spray heat exchangers, the emergency diesel generators, containment coolers, and various other system loads.
The CCS is an intermediate cooling loop cooled by the ERCW system. Three CCS heat exchangers serve both units aligned into two trains. CCS heat exchangers A and B serve Train A for Units I and2, respectively, and CCS heat exchanger C serves Train B for both units. The CCS heat exchangers provide heat removal for the Residual Heat Removal (RHR) heat exchangers, the Spent Fuel Pool heat exchangers, the Chemicaland Volume Control System (CVCS), various pump coolers, and various other loads. ERCW flow to the CCS heat exchangers is controlled using the discharge flow control valves, each heat exchanger with a pair of valves, one normally throttled which is used during normal conditions and the other normally closed and used during accident conditions which can be positioned to one of three (A, B, Open) preset opening positions.
The limiting design for both the ERCW and CCS systems is a normal shutdown on one unit with a design basis Loss of Coolant Accident (LOCA) on the other unit. The operation of the RHR system on the unit in shutdown and the transition to cold leg recirculation heat load on the RHR heat exchange(s) for the LOCA unit represent the maximum heat load that is applied to the CCS system.
During normal operation, ERCW flow is modulated to the CCS heat exchangers to allow the CCS temperature to be maintained in an optimaloperating band. Because heat loads during normaloperation are low, ERCW flow would be maintained at flows lower than during an accident to allow for optimal CCS system operation. Two separate, but related issues are identified in this event.
ln LER 390-2017-009, WBN reported several issues associated with the ERCW, the most significant being that for an accident on Unit 2, no provisions existed in the Emergency Operating lnstructions (EOls) for adjusting ERCW flows to the Train A CCS heat exchangers by opening the ERCW to the CCS heat exchanger discharge control valves to one of the desired pre-set opening positions to support accident operation (minimum analysis required ERCW flow of 3500 gpm). These valve position adjustments should have been performed in procedure 2-E-0, Reactor Trip or Safety lnjection. Without repositioning these valves in advance of switchover to recirculation, unacceptably high temperature in the CCS Train A system for both units may occur following transfer to recirculation post LOCA. The required valve repositioning steps were also incorrect in 1-E-0 for Unit 1, but supported an alignment yielding adequate flow for Unit 1 and Unit 2.
A separate issue was identified in Condition Report (CR) 1331422that is the subject of this report where incorrect change to procedure 1-E-1 and 2-E-1, Loss of Reactor or Secondary Coolant, were made prior to Unit 2 initial operation. Steps were added to both procedures where for a LOCA concurrent with a loss of offsite power and a train failure while the other unit was on RHR cooling, an ERCW supply side cross connect between the A and B trains would be open. This erroneous change had the potential to reduce ERCW flow to the functioning train.
A detailed engineering review of the positions of the ERCW discharge valves from the CCS heat exchangers was performed from the time period of the introduction of the 1-E-1 procedure error until correction of this issue via EOI revisions as part of the evaluation of LER 390-2017-009. This review was
used as a basis to document the impact of considering a LOCA with a train failure to address the impact of opening the ERCW supply side cross connect of system safety function.
The review performed for LER 390-2017-009 determined that inadequate ERCW flow would have been present for approximalely 0.252 years for a large break LOCA on Unit 2. The review of the open cross tie issue identified a total of 18.9 days of inoperability on Unit 2Train A. Of the 18.9 days, 7.8 days were already deemed inoperable by the review performed for LER 390-2017-009.
V. ASSESSMENT OF SAFETY CONSEQUENCES
As described in the previous section, the ERCW system Train A was not able to perform its safety function for a design basis accident on Unit 2 for about 18.9 days associated with the ERCW cross tie valve being opened in the event of a LOCA with a train failure. A probabilistic risk assessment (PRA) performed determined that the increase in core damage frequency (CDF) for this issue, when coupled with the issue described in LER 390-2017-009, to be less than 1E-7.
A. Availability of systems or components that could have performed the same function as the components and systems that failed during the event Train B ERCW was able to perform its safety function for a design basis accident on Unit 2 related to this issue for the periods where Train A was inoperable.
B. For events that occurred when the reactor was shut down, availability of systems or components needed to shutdown the reactor and maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident Not applicable.
C. For failure that rendered a train of a safety system inoperable, an estimate of the elapsed time from the discovery of the failure until the train was returned to service This issue was corrected during the resolution of the ERCW design issue reported in LER 390-2017-009.
VI. CORRECTIVE ACTIONS
This event was entered into the Tennessee Valley Authority (TVA) Corrective Action Program and is being tracked under Condition Report (CR) 1331422.
A. lmmediate Corrective Actions The issue identified was already corrected when it was identified.
B. Corrective Actions to Prevent Recurrence or to Reduce Probability of Similar Events Occurring in the Future Corrective actions will be provided in a supplement to this report.
VII
PREVIOUS SIMILAR EVENTS AT THE SAME SITE
vilt ERCW ERCW Train 14 Train 2A CCS Train A IX LER 390/2017-009 reported an unanalyzed condition that showed adequate ERCW flow may not be available during dual unit limiting design basis conditions of one unit in Hot Shutdown on RHR cooling when the other unit experiences a LOCA during specific single failure events due to a design error. Due to procedural errors, the analysis showed during a Unit 1 LOCA, Unit 1 received adequate flow when following existing procedural guidance. During a Unit 2 LOCA, however, procedural guidance was not adequate to ensure the proper system alignment to establish correct ERCW to CCS Heat Exchanger A and B flow rates for either unit's cool down requirements, due to procedure errors. The cause of this event was human performance related to both a knowledge of personnel associated with the design and a failure by procedure preparers to properly capture configuration changes specified in a plant design change when WBN went to dual unit operation.
LER 390/2016-004 reported an unanalyzed condition where the Unit 1 Fire Protection Report (FPR) analysis for 10 CFR 50, Appendix R contained a non-conservative time for isolation of the Volume Control Tank (VCT) following a postulated fire in room 737.0-A1A. Multiple fire-induced failures were postulated to result in a loss of suction to the Centrifugal Charging Pumps (CCPs), which could cause Reactor Coolant Pump seal damage and loss of Reactor Coolant System inventory. Fire modeling subsequently determined that for any credible combination of failures or equipment spurious operation that the CCPs would remain operable. This issue was determined to be the result of a latent engineering error associated with the original Appendix R analysis performed for Unit 1.
ADDITIONAL I NFORMATION None.
COMMITMENTS
None.
Other Loads Simplified ERCW FigurePage 6 of 6