05000390/LER-2008-002, Manual Reactor Trip in Response to Start of Feedwater Heater Isolation
| ML082800248 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 10/03/2008 |
| From: | Skaggs M Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 08-002-00 | |
| Download: ML082800248 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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)(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 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3902008002R00 - NRC Website | |
text
October 3, 2008 10 CFR 50.73 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Gentlemen:
In the Matter of
)
Docket No. 50-390 Tennessee Valley Authority
)
WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - LICENSEE EVENT REPORT 390/2008-002, REVISION 0 - MANUAL REACTOR TRIP IN RESPONSE TO START OF FEEDWATER HEATER ISOLATION This submittal provides LER 390/2008-002. This LER documents an event where the reactor was manually tripped during a downpower because feedwater heaters strings were beginning to isolate. The report regarding this condition is provided in accordance with 10 CFR 50.73(a)(2)(iv)(A).
There are no regulatory commitments associated with this submittal. If you have any questions concerning this matter, please contact Mike Brandon at (423) 365-1824.
Sincerely, Original signed by Mike Skaggs Site Vice President Watts Bar Nuclear Plant Enclosure cc: See Page 2
U.S. Nuclear Regulatory Commission Page 2 October 3, 2008 Enclosure cc (Enclosure):
NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 ATTN: John G. Lamb, Project Manager U.S. Nuclear Regulatory Commission Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation MS O-8 H4 Washington, DC 20555-0001 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 Institute of Nuclear Power Operations 700 Galleria Parkway, NW Atlanta, Georgia 30339-5957
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
(See reverse for required number of digits/characters for each block)
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010
, 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
- 2. DOCKET NUMBER 05000 390
- 3. PAGE 1 OF 5
- 4. TITLE Manual Reactor Trip in Response to Start of Feedwater Heater Isolation
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL NUMBER REV NO.
MONTH DAY YEAR FACILITY NAME N/A DOCKET NUMBER N/A 08 07 2008 2008 - 002 -
0 10 06 2008 FACILITY NAME N/A DOCKET NUMBER N/A
- 9. OPERATING MODE 1
- 10. POWER LEVEL 53%
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply) 20.2201(b) 20.2203(a)(3)(i) 50.73(a)(2)(i)(C) 50.73(a)(2)(vii) 20.2201(d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A) 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B) 20.2203(a)(2)(i) 50.36(c)(1)(i)(A) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A) 20.2203(a)(2)(ii) 50.36(c)(1)(ii)(A) 50.73(a)(2)(iv)(A) 50.73(a)(2)(x) 20.2203(a)(2)(iii) 50.36(c)(2) 50.73(a)(2)(v)(A) 73.71(a)(4) 20.2203(a)(2)(iv) 50.46(a)(3)(ii) 50.73(a)(2)(v)(B) 73.71(a)(5) 20.2203(a)(2)(v) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(C)
OTHER 20.2203(a)(2)(vi) 50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D)
Specify in Abstract below or in II.
DESCRIPTION OF EVENT (continued):
D.
Other Systems or Secondary Functions Affected
The plant trip was normal except that the response of the auxiliary feedwater (EIIS BA) level control valves (EIIS LCV) caused the steam generator (EIIS SG) levels to swing. When this was observed, operators placed the valves into manual control. The valves controlled level properly in the manual state. This anomaly was documented in PER 150508.
E.
Method of Discovery
Plant personnel identified, through control room indications, that the A and B string low pressure heaters were isolating. Operators discussed and established manual reactor trip criteria and once the C string began to isolate, the plant was manually tripped, as established.
F.
Operator Actions
The Operations staff (licensed personnel) responded to the event by manually tripping the reactor. The crew met all Operations standards and expectations, and no human performance issues were identified.
G.
Safety System Responses All systems performed their intended safety functions.
III.
CAUSE OF EVENT
The manual trip was the operator response to the low pressure feedwater heater string isolation caused by failure of the #7 bypass level control valve (LCV). The LCV failure was caused by failure of the signal air line.
The air line, a stainless steel bellows type flexible hose that is enclosed in a stainless steel braid, failed due to vibration induced fatigue as a result of improper installation; in that, minimum bend radius and maximum arc recommendations were not known and therefore not followed. The work order for installing these hoses did not contain vendor requirements specifying the allowed bend radius and arc allotments for installation.
IV.
ANALYSIS OF THE EVENT
Plant safety systems performed intended safety functions in response to the manual reactor trip. The plant was stabilized using Auxiliary Feedwater and the Main Steam (EIIS SB) dump valves. See Section V, Assessment of Safety Consequences, below for further discussion.
V.
ASSESSMENT OF SAFETY CONSEQUENCES
The manual reactor trip on 08/07/2008 was compared to the FSAR "Loss of Normal Feedwater Event,"
UFSAR section 15.2.8 (page 15.2-25). The reactor was manually tripped at approximately 02:28 (53%
power, U2125) due to secondary side low pressure heater string isolations. The heater string isolations occurred because the # 7 HDT bypass to condenser control valve (1-LCV-006-0190B) failed to open which caused levels in the #7 HDT and associated low pressure heaters to increase, resulting in subsequent low pressure heater string isolations (all three strings). The "A heater string isolated first, then "B" string isolated.
Moments later "C" string began to isolate, and Operators manually tripped the reactor. The manual reactor trip occurred before SG water level reached the low-low setpoint. The plant was stabilized using Auxiliary Feedwater (AFW) and the Main Steam dump valves.
V.
The secondary side steam generator (SG) atmospheric relief valves (EIIS RV) and safety valves were not challenged during the transient. The reactor coolant system (EIIS AB) responded to the initial transient as expected with no pressurizer PORV relief and no safety injection initiation.
The FSAR analysis assumes the reactor trip is caused by low-low water level in any SG and in this case was initiated by operator action before steam generator levels decreased to the low-low setpoint. In addition, no credit is taken for the steam dump system, and no credit is taken for the SG atmospheric relief valves (only steam generator safety valves are credited). The actual event had steam dumps and SG atmospheric relief valves available. The FSAR assumes that the plant is initially operating at 110.6% of the Nuclear Steam Supply System (NSSS) power level while the manual reactor trip occurred at approximately 50% power.
Failure of the Turbine Driven (TD) AFW pump is assumed in the FSAR. The TD AFW pump, along with both Motor Driven (MD) AFW pumps, started within the required 60 second response time and were available for decay heat removal.
Therefore, the 08/07/2008 trip is bounded by the FSAR safety analysis assumptions.
VI.
CORRECTIVE ACTIONS - The corrective actions for this condition are being managed within TVAs Corrective Action Program (PERs 149778 and 149790) and therefore are not considered to be regulatory
commitments
An overview of the corrective action plan is provided below:
A.
Immediate Corrective Actions
- 1.
When the A and B string low pressure heaters began isolating, monitoring for C isolation began. Once this indication was received, the immediate action was to trip the reactor.
B.
Corrective Actions to Prevent Recurrence
- 1.
The failed air line was replaced with the same type hose with the configuration modified to comply with manufacturer installation requirements.
- 2.
A walkdown of all critical air operated valves in the Turbine Building and the valve vaults was performed. Of the 81 valves inspected, three were found with discrepancies. One had an abrasion of the braid and a severe twist, and one had a sharp bend. The third had a piece of bent copper tubing at the end of a hose. The hoses and tubing were replaced according to vendor installation requirements. Extent of condition was restricted to the turbine building.
Safety related installations of flexible tubing are controlled by the modification process which provides explicit installation details.
- 3.
A walkdown of the four replaced air lines was performed by the system engineer with an attribute checklist to ensure proper installation.
- 4.
The post maintenance testing for hose installation was reviewed for adequacy.
- 5.
The maintenance procedures will be modified to clearly specify the vendor installation requirements for hoses.
VII.
ADDITIONAL INFORMATION
A.
Failed Components The main failed component associated with this event is the flexible metal hose, as discussed above.
B.
Previous LERs on Similar Events A review was performed of the previous WBN Licensee Event Reports (LERs) for any events associated with high feedwater heater level. LER 1997-015 documents a series of events where the reactor was manually tripped during plant startup due to decreasing steam generator levels resulting from isolation of feedwater heaters. The cause of the 1997 event was that additional operator attention was not provided to control the heater level during the turbine roll and main generator synchronizing. This extra oversight should have been required because the C1 heater was out of service. The issues documented in LER 1997-015 were not related to heater isolation during down power or caused by failed air lines. No other relevant WBN LERs were identified.
C.
Additional Information
None.
D.
Safety System Functional Failure This event did not involve a safety system functional failure as defined in NEI 99-02, Revision 5.
E.
Loss of Normal Heat Removal Consideration There was no loss of normal heat removal due to this condition.
VIII.
COMMITMENTS
None.