ML18039A898

LER 99-009-00:on 990915,manual Reactor Scram Was Noted Due to EHC Leak.Caused by Failure of Stainles Steel Tubing Connection.Removed Damaged Tubing & Connection Plug

ML18039A898
Person / Time
Site:	Browns Ferry
Issue date:	10/14/1999
From:	Rogers A TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18039A897	List: ML18039A896 ML18039A898
References
LER-99-009, NUDOCS 9910200198
Download: ML18039A898 (10)

Text

NRC FORM 366 U.S. NUCLEAR'REGULATORY COMMISSION APPROVED BY OMB NO. 3160-0104 EKPIREs (6-1998) 0 8/30/2001 Estimated burden per response to comply with th)s mandatory hformation coection request: 50 hrs. Reported lessons learned are ncorporated into LICENSEE EVENT REPORT (LER) the'Ecensing process and fed back to Industry. Forward comments regarding burden estimate to the Records Management Branch (TA F33), U.S.

Nuclear Regulatory Commission. Washington, OC 205554001. and to the (See reverse for required number of Paperwork ReducUon Project (31500104). Offee of Management and digits/characters for each block) Budget, Washington. OC 20503. If sn information coaection does not dispLIy a currently vaEd OMB control number. the NRC may not conduct or sponsor. and a person Is not required to respond to. the intormation collection.

FACIUTY NAME l1) DOCKET NUMBER I2) PAGE IS)

Browns Ferry Nuclear Plant Unit 2 05000260 1of5 TITLE I4)

Manual Reactor Scram due to an EHC leak EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) 0 HER FACILITIES INVOL ED IB)

MONTH DAY YEAR SEQVENTIAL REVISION IU NAM DOCKET NUMBER NUMBER NUMBER NA DOCKETNUMBER 09 15 99 1999 009 000 10 99 NA OPERATING THIS REPORT IS SUBMITTED PURSUANT To THE REQUIREMENTS OF 10 CFR rn (Check one or more) (11)

MODE (9) 20.2201(b) 20.2203(a) (2) (v) 50.73(a)(2)(i)(B) 50.73(a) (2)(viii)

POWER 20.2203(a)(l) 20. 2203(a) (3) (i) 50.73 (a) (2) (ii) 50.73(a)(2)(x)

LEVEL (10) 100 20.2203(a) (2)(i) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.71 20.2203(a)(2) lii) 20.2203(a) (4) 50.73(a)(2)(iv) OTHER 20.2203(a) (2)(iii) 50.36(c) (1) 50.73(a)(2)(v) Specify In Abstract below or in NRC Form 366A 20.2203(a)(2) (iv) 50.36(c)(2) 50.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER'(12)

NAME TELEPHONE NUMBER ilndude Ares Code)

Anthony T. Rogers, Senior Licensing Project Manager (256) 729-2977 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE NPRDS TO NPRDS TG TBG G080 eels

'%,: rd x:.'.:...z.':< .:%;

SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR YES No SUBMISSION X DATE (15)

(lf yes, complete EXPECTED SUBMISSION DATE).

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On September 15, 1999, at 1825 CDT, Unit 2 operators manually scrammed the reactor from 54 percent power due to an Electro-Hydraulic Control (EHC) tTG] leak that could not be isolated. The reactor had been at 100 percent power prior to the leak. As expected, the reactor scram caused reactor water level to go below the low level setpoint (level 3) which generated a redundant scram signal and initiated Primary Containment Isolation, Standby Gas Treatment, and Control Room Emergency Ventilation Systems. All systems responded as expected and all control rods fully inserted.

The cause of the leak was failure of a stainless steel tubing connection that was installed for the power uprate modification package to measure pressure perturbations in the EHC system. The damaged tubing was removed and the connection plugged.

TVA is reporting this event in accordance with 10 CFR 50.73 (a)(2)(iv) as an event that resulted in a manual actuation of an engineered safety feature, including the reactor protection system.

99i0200i98 99iOi4 PDR ADOCl('5000260 S PDR NRC FORM 366B (6-1996)

II h'

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-I 998I LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 1 DOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL REVISION NUMBER 2 of 5 Browns Ferry Nuclear Plant - Unit 2 05000260 1999, 009 000 TEXT (Ifmore space is required, use additional copies of NRC Form 366A/ I17)

I. PLANT CONDITIONS Prior to the initiation of the event, Unit 2 and Unit 3 were at 100 percent power. Unit 1 was shutdown and defueled.

II.,DESCRIPTION OF EVENT A. Event:

On September 15, 1999, Unit 2 developed an Electro-Hydraulic Control (EHC) [TG] leak that could not be isolated, upon recognition of the problem, the operators took appropriate action by reducing reactor power and manually scramming the reactor prior to an automatic scram from a turbine trip. The reactor was initially at 100 percent power prior to the leak and was scrammed from 54 percent power after initial operator action. As expected, the reactor scram caused reactor water level to go below the low level setpoint (level 3) which generated a redundant scram signal and initiated a Primary Containment Isolation, Standby Gas Treatment, and Control Room Emergency Ventilation Systems. All systems responded as expected and all control rods fully inserted.

The cause of the leak was failure of a stainless steel tubing connection that was installed for the power uprate modification package to measure pressure perturbations in the EHC system.

The scram resulted in the expected automatic actuation or isolation of the following PCIS [JE] systems and components:

~ PCIS group 2, Shutdown cooling mode of Residual Heat Removal (RHR) [BO] system; drywell floor drain isolation valves; drywell equipment drain isolation valves [WP].

~ PCIS group 3, Reactor Water Cleanup (RWCU) system [CE].

~ PCIS group 6, primaIy containment purge and ventilation [JM], Unit 2 reactor zone ventilation [VB];

refuel zone ventilation [VA]; Standby Gas Treatment system [BH]; Control Room Emergency Ventilation system [VI].

~ PCIS group 8, Traversing Incore Probe (TIP) [IG].

This event is reportable in accordance with 10 CFR 50.73 (a)(2)(iv), as an event that resulted in a manual actuation of an engineered safety feature, including the reactor protection system.

B. Ino erable Structures Com onents orS stems that Contributedtothe Event:

None.

C. Dates and A roximate Times of Ma or Occurrences:

September 15, 1999, at 1758 hours0.0203 days <br />0.488 hours <br />0.00291 weeks <br />6.68919e-4 months <br /> CDT An EHC Reservoir Level Low alarm was received in the control room and personnel were dispatched to investigate.

NRC FORM 366 {6-199BI

41 Qi NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION I6.1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 1 DOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL REVISION NUMBER 3 of 5 Browns Ferry Nuclear Plant - Unit 2 05000260 1999 009 000 TEXT llfmore spece is required, use eddi tionel copies of lVRC Form 366AJ I 17)

C. Dates and A roximate Times of Ma or Occurrences continued:

September 15, 1999, at 1804 hours0.0209 days <br />0.501 hours <br />0.00298 weeks <br />6.86422e-4 months <br /> CDT Reactor power was lowered to 50-60% core flow upon receipt of report that the EHC reservoir level was reported low.

September 15, 1999, at 1825 hours0.0211 days <br />0.507 hours <br />0.00302 weeks <br />6.944125e-4 months <br /> CDT Reactor manually scrammed. Expected PCIS signals and actuations occurred when reactor water reached level 3 following the scram.

September 15, 1999, at 1844 hours0.0213 days <br />0.512 hours <br />0.00305 weeks <br />7.01642e-4 months <br /> CDT A four-hour non-emergency report is made to the NRC pursuant to 10 CFR 50.72 (b) (2) (ii).

D. Other S stems or Seconda Functions Affected:

None.

E. Method of Discove Operators received alarms indicating an EHC leak had occurred.

F. 0 erator Actions:

Operations personnel responded to the event in accordance with applicable plant procedures.

G. Safet S stem Res onse:

All required safety systems operated as designed.

III. CAUSE OF THE EVENT A. 'Immediate Cause:

The immediate cause of this event was failure of a stainless steel tubing connection in the heat affected zone of the weld.

B. Root Cause:

The root cause of the failure was poor fabrication and work practices used to install the stainless steel tubing.

C. Contributin Factors:

None.

NRc FQRM 366 I6-1998)

i' NRC FORM 366A u.s. NucLEAR REGULATORY coMMissioN (6. I 998 i LiCENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 1 DOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL REVISION NUMBER 4 of 5 Browns'Ferry Nuclear Plant - Unit 2 05000260 1999 009 000 TEXT Iffmore space is required, use additional copies of PIRC Form 366A/ {17)

IV. ANALYSIS OF THE EVENT As part of the installation for five percent power uprate during the Browns Ferry Unit 2 Cycle 10 refueling outage, a design change installed four EHC accumulator packages on the main turbine control valves to dampen EHC pressure peiturbations. Part of this installation package included a 3/8 inch nominal outer diameter (0.035 inch nominal wall thickness) tubing connection which consisted of socket weld glands and standard nuts to connect the accumulator to a pressure transmitter on the number four main turbine control valve. This stainless steel tubing connection completely fractured at the toe of the weld and resulted in the necessity to initiate a manual scram.

The subject tubing failure was evaluated in order to determine the failure mechanism and root cause for the failure. Plant personnel that initially discovered the failed EHC tubing failure indicated that the broken segments of the tubing were off-set at least 2-3 inches. This amount of offset would result in excessive cold springing for tubing of this diameter and.length. A visual examination performed by Site Engineering on the inside of the tubing at the failure location showed evidence of weld melt-through at the root of the joint for almost the entire circumference of the tubing. The weld melt-through is the result of excessive heat input from welding during fabrication of the failed socket weld joint. Examination of the tubing fracture surfaces using a stereo microscope revealed a relatively flat fracture surface at the toe of the weld, ratchet marks and the absence of gross deformation. Scanning electron microscopy, which was performed at TVA's Central Laboratories, on the tubing side of the fracture surface revealed striations.

The features (i.e., relatively flat fracture surface, ratchet marks, striations and the absence of gross deformation) revealed by stereo and scanning electron microscopy on the fracture surface of the stainless steel tubing failure are indicative of a high cycle fatigue failure. This connection was exposed to constant vibration during plant operation. The excessive cold springing and weld melt-through resulted in additional residual stresses which attributed to this failure. Therefore, the root cause of this tubing failure is poor fabrication and.installation practices.

No other similar installations were identified on Unit 2 or Unit 3.

V. ASSESSMENT OF SAFETY, CONSEQUENCES The evaluation of plant system and component responses to the event concluded that responses were as designed and within the time-frames expected. The normal heat removal path was not lost during this event since the condenser was used for decay heat removal and no main steam relief valves opened. Personnel performance was also evaluated and found to be timely, appropriate, and met expectations for performance during an event of this type.

There were no equipment failures during or following the scram that complicated recovery. In addition, there were no radioactive material released and no actual or potential safety consequences as a result of this event. Therefore, this event did not adversely affect the safety of plant personnel or the public.

NRC FORM 366 I6-1998)

~I NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION I6-1998)

LICENSEE EVENT REPORT (LER).

TEXT CONTINUATION FACILITY NAME 1 DOCKET LER NUMBER 6 PAGE 3 YEAR, SEQUENTIAL REVISION NUMBER 5 of 5 Browns Ferry Nuclear Plant - Unit 2 05000260 1999 009 000 TEXT (If more space is required, use additional copies of NRC Form 366A/ t17)

VI. CORRECTIVE ACTIONS A. Immediate Corrective Actions:

The Operations crew stabilized the reactor following the scram using the appropriate operating instructions.

The failed tubing was removed and the connections were plugged.

An inspection of the area affected by the EHC fluid was performed and cleanup activities were completed prior to restart.

B. Corrective Action to Prevent Recurrence:

General Electric will evaluate the cause of failure and provide recommendations to TVA to prevent recurrence.'VA will evaluate the design of the tubing and accumulator arrangement to determine the long term desired configuration.'he cabling contacted by the EHC fluid will be inspected during the next refueling outage to determine if any deterioration is evident.'II.

ADDITIONALINFORMATION A. Failed Com onents:

None.

B. Previous Similar Events:

None.

C. Additional Information:

This event did not result in loss of the normal heat removal path as described in draft NEI 99-02, Rev. C, since the. condenser was used for decay heat removal.

D. Safet S stem Functional Failure:

I This event did not result in a safety system functional failure in accordance with draft NEI 99-02, Rev. C.

VIII. COMMITMENTS None.

'TVA does not consider this corrective action a regulatory commitment. The completion of this item will be tracked in TVA's Corrective Action Program.

NRC FORM 366 t6-1998)

0 I