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Event Notification Report for July 13, 2017

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
07/12/2017 - 07/13/2017

** EVENT NUMBERS **


52692 52850

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Part 21 Event Number: 52692
Rep Org: GE HITACHI NUCLEAR ENERGY
Licensee: GE HITACHI NUCLEAR ENERGY
Region: 1
City: WILMINGTON State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: LISA SCHICHLEIN
HQ OPS Officer: BETHANY CECERE
Notification Date: 04/19/2017
Notification Time: 10:29 [ET]
Event Date: 02/27/2017
Event Time: [EDT]
Last Update Date: 07/12/2017
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(a)(2) - INTERIM EVAL OF DEVIATION
Person (Organization):
PHILIP McKENNA (R2DO)
LAURA KOZAK (R3DO)
RAY AZUA (R4DO)
PART 21/50.55 REACTO (EMAI)

Event Text

PART 21 REPORT - CONTROL ROD DRIVE MECHANISMS CONTAMINATED WITH CHLORIDES

Dale E. Porter
GE-Hitachi Nuclear Energy
Safety Evaluation Program Manager
3901 Castle Hayne Rd., Wilmington, NC 28401
(910) 819-4491
Dale.Porter@GE.Com

"The inappropriate addition of chlorinated water from container box desiccants into the CRDMs [Control Rod Drive Mechanisms] during leak testing after rebuild could potentially initiate Intergranular Stress Corrosion Cracking (IGSCC) or Transgranular Stress Corrosion Cracking (TGSCC). These two types of SCC could cause a separation of the stop piston or separation of the index tube contained within the CRDM. The stop piston separation could cause a slower scram speed and damage the drive so it could not be withdrawn. An index tube separation could result in a similar type of rod uncoupling event that would have the potential to result in a rod drop accident (RDA). The piston tube located within the CRDM is a reactor coolant pressure boundary (RCPB) and is an ASME component. There is a possibility of cracking causing a RCPB leak. SCC initiation on the Cylinder Tube and Flange (CTF) area of the CRDM could result in a separation that could prevent a scram or normal insertion of a CRDM.

"Reports have been issued to River Bend, LaSalle Unit 2, and Hatch Unit 2 providing the results of an evaluation that concludes that the condition will not create a substantial safety hazard or potentially cause a Technical Specification Safety Limit violation for a minimum of one operating cycle. The Browns Ferry Unit 2 drives were shipped but were not installed prior to recall, thus a short-term evaluation for Browns Ferry has not been completed.

"River Bend, Entergy, Shipped Date: 2017, Quantity Shipped: 15, Customer PO Number: 10478763
LaSalle Unit 2, Exelon, Shipped Date: 2017, Quantity Shipped: 24, Customer PO Number: 00414787-66
Hatch Unit 2, Southern Nuclear, Shipped Date: 2017, Quantity Shipped: 15, Customer PO Numbers: SNG50295-0001 & SNG50295-0002
Browns Ferry Unit 2, TVA, Shipped Date: 2017, Quantity Shipped: 32, Customer PO Number: 2424171"

* * * UPDATE ON 7/12/17 AT 1035 EDT FROM LISA SCHICHLEIN TO BETHANY CECERE * * *

"Pursuant to 10 CFR 21.21(d)(4), GEH is providing the final report with the conclusion that, in limited cases, the chloride contamination could create a substantial safety hazard. Attachment 1 identifies the potentially impacted plants and Attachment 2 contains the final report information. The enclosure provides additional details of the evaluation.

Updated Attachment 1 notes:
"A portion of the CRDMs at the Hatch Plant are Reportable while the remainder are not.

Summary of updates to Attachment 2:
"The inappropriate addition of chlorinated water from container box desiccants into the CRDMs during leak testing after rebuild could potentially initiate Intergranular Stress Corrosion Cracking (IGSCC) or Transgranular Stress Corrosion Cracking (TGSCC). These two types of SCC were initially considered for the potential to cause a separation of the stop piston or separation of the index tube contained within the CRDMs that are constructed of 304 Stainless steel. The completed evaluation indicates that a stop piston separation could cause a slower scram speed and damage the drive so it could not be withdrawn. The potential exists for the control rod to drift out. The piston tube located within the CRDM is a reactor coolant pressure boundary (RCPB) and is an ASME component. The possibility of cracking causing RCPB leakage was eliminated by the evaluation. An index tube separation was eliminated as a potential failure mode. Likewise, the potential for SCC initiation on the Cylinder Tube and Flange (CTF) area of the CRDM resulting in a separation that could prevent a scram or normal insertion/withdrawal of a CRD was eliminated.

"The long-term evaluation concluded there is no concern for TGSCC. This evaluation also determined the stop piston to piston tube separation is the only failure mechanism that could occur, and only if those components were manufactured from 304 SS.

"GEH initiated a Root Cause Evaluation (RCA) to determine why this event occurred and has implemented process changes to ensure that the condition does not reoccur. Actions to prevent recurrence, such as eliminating the desiccant material and flushing the closed loop water system, have been completed.

"For Hatch Unit 2, the 12 CRDMs that have the 304 SS piston tubes should be replaced prior to those CRDMs exceeding 10 years in service. See table below:
CRDM S/N, Piston Tube
SE0474, 304 SS
A8737, 304 SS
A9423, 304 SS
A6791, 304 SS
3095, 304 SS
A8729, 304 SS
7253, 304 SS
A6786, 304 SS
SE0368, 304 SS
A5409, 304 SS
7080, 304 SS
A9484, 304 SS

"Interim Reports were issued to River Bend, LaSalle Unit 2, and Hatch Unit 2 providing the results of an evaluation that concluded the condition would not create a substantial safety hazard or potentially cause a Technical Specification Safety Limit violation for a minimum of one operating cycle. The Browns Ferry Unit 2 drives were shipped but were not installed prior to recall, thus a short-term evaluation for Browns Ferry was not provided.

"GEH has completed the long-term CRDM evaluation with the following results:

"A Safety Information Communication is being issued to River Bend, LaSalle Unit 2, and Browns Ferry Unit 2 stating that the CRDMs exposed to the chloride intrusion will not cause a substantial safety hazard, or cause a Technical Specification Safety Limit violation and is therefore not reportable (see enclosure 1 for details).

"For Hatch Unit 2, the introduction of chlorides could cause a substantial safety hazard for the 12 CRDMs that were manufactured from 304 SS material, and is therefore a reportable condition per 10CFR 21.21(d); however, the 3 CRDMs manufactured from XM-19 material would be considered not reportable.

"The results of the short-term evaluations for all three plants where CRDMs were installed remains valid."

Notified the R2DO (Bonser), R3DO (Peterson), R4DO (Proulx), and Part 21/50.55 Reactors Group by email.

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Power Reactor Event Number: 52850
Facility: WATTS BAR
Region: 2 State: TN
Unit: [1] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: DAVID ALLEN
HQ OPS Officer: DONG HWA PARK
Notification Date: 07/12/2017
Notification Time: 16:05 [ET]
Event Date: 07/12/2017
Event Time: 12:38 [EDT]
Last Update Date: 07/12/2017
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
Person (Organization):
BRIAN BONSER (R2DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 100 Power Operation 100 Power Operation
2 N N 0 Cold Shutdown 0 Cold Shutdown

Event Text

UNANALYZED CONDITION RELATED TO GDC-5 [GENERAL DESIGN CRITERION-5] FOR DUAL UNIT OPERATION

"On July 12, 2017, at 1238 Eastern Daylight Time (EDT), Watts Bar Nuclear Plant (WBN) determined that a preliminary analysis shows adequate Essential Raw Cooling Water (ERCW) flow may not be in place during dual unit limiting design basis conditions of one unit in Hot Shutdown on Residual Heat Removal (RHR) cooling when the other unit experiences a Loss of Coolant Accident (LOCA). Based on preliminary analysis, during a Unit 1 LOCA, Unit 1 receives adequate flow when following existing procedural guidance; however, Unit 2 may not receive adequate flow to meet cool-down requirements with design basis maximum temperatures. During a Unit 2 LOCA, however, current procedural guidance is not adequate to ensure the proper system alignment to establish correct ERCW Component Cooling Water (CCS) Heat Exchanger A and B flow rates for either unit's cool down requirements.

"Unit 2 has been shutdown for an extended period of time such that the flow delivered by ERCW is adequate to serve both Unit 1 in a LOCA and Unit 2 in less than Mode 3.

"The NRC Resident Inspector has been notified."

Page Last Reviewed/Updated Thursday, July 13, 2017
Thursday, July 13, 2017