05000280/LER-2005-001, Regarding Manual Reactor Trip Initiated Due to Misaligned Control Rod
| ML051040406 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 04/04/2005 |
| From: | Jernigan D Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 05-162 LER 05-001-00 | |
| Download: ML051040406 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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) 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 |
| 2802005001R00 - NRC Website | |
text
Ir 1 OCFR50.73 Virginia Electric and Power Company Surry Power Station 5570 Hog Island Road Surry, Virginia 23883 April 4, 2005 U. S. Nuclear Regulatory Commission Serial No.:
05-162 Attention: Document Control Desk SPS: JSA Washington, D. C. 20555-0001 Docket No.: 50-280 License No.: DPR-32
Dear Sirs:
Pursuant to 10CFR50.73, Virginia Electric and Power Company hereby submits the following Licensee Event Report applicable to Surry Power Station Unit 1.
Report No. 50-280/2005-001-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be forwarded to the Management Safety Revie Committee for its review.
V ryt ly yours,
- ernigan, Site Vice President Surry Power Station Enclosure Commitment contained in this letter:
- 1. Implement control rod exercising prior to any unit startup to mitigate the effects of particulate in CRDM internals.
,</s
cc:
United States Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23 T85 Atlanta, Georgia 30303-8931 Mr. N. P. Garrett NRC Senior Resident Inspector Surry Power Station
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY tPM: NO. 3150-0104 EXPIRES06i3C2007 (6-2004)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the digits/characters for each block) information collection.
- 3. PAGE Surry Power Station, Unit 1 05000 - 280 1 OF 4
- 4. TITLE Manual Reactor Trip Initiated Due to Misaligned Control Rod
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.
MONTH DAY YEAR 05000 OPRTN OE 1.HSRPR FACILITY NAME DOCKET NUMBER 02 07 2005 2005 -
001 -
00 04 04 2005 05000
- 9. OPERATING 11.THIS SUBMITTED PURSUANT TO THE REQUIREMENTS OF10 CFR §: (Check allthatapply) a 20.2201(b)
[
20.2203(a)(3)(i)
M 50.73(a)(2)(i)(C)
[
50.73(a)(2)(vii)
N EJ 20.2201(d)
[
20.2203(a)(3)(ii)
- - 1 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
M 20.2203(a)(1)
D 20.2203(a)(4) a 50.73(a)(2)(ii)(B)
J 50.73(a)(2)(viii)(B) 0_ _ 20-2203(a)(2)(i) 0 50.36(c)(1)(i)(A)
J 50.73(a)(2)(iii) 50.73(a)(2)(i.)(A)
- 10. POWER LEVEL 0
20.2203(a)(2)(ii) 50.36(c)(1)(ii)(A)
EJ 50.73(a)(2)(iv)(A) aJ 50.73(a)(2)(x) 0 20.2203(a)(2)(iii) 0J 50.36(c)(2) 0 50.73(a)(2)(v)(A)
[
73.71(a)(4) 0%
0 20.2203(a)(2)(iv) 01 50.46(a)(3)(ii) 0] 50.73(a)(2)(v)(B)
[
73.71(a)(5)
D 20.2203(a)(2)(v) 50.73(a)(2)(i)(A)
J 5073(a)(2)(v)(C)
F OTHER 20.2203(a)(2)(vi) 60 5.73(a)(2)(i)(B)
I 50.73(a)(2)(v)(D)
Specify in Abstract below
____or In NRC Form 386A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER finclude Area Code) l Donald E. Jernigan, Site Vice President (757) 365-2001CAUSE SYSTEM COMPONENT FACTURER TO EPIX
CAUSE
SYSTEM COMPONENT MFACTU-R ElOTO BE X
AA ROD W120 l
Y+/-RTlEPI
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION 1 YES (If Ves, complete 15. EXPECTED SUBMISSION DATE)
E NO DATE ABSTRACT (ULmit to 1400 spaces, I.e., approximately 15 single-spaced Mypewitten lines)
On February 7, 2005 at 2024 hours0.0234 days <br />0.562 hours <br />0.00335 weeks <br />7.70132e-4 months <br />, with Unit 1 undergoing a reactor startup, a manual reactor trip was initiated due to a control rod misalignment on rod B1 0. At the time of the manual trip, reactor power was in the source range at approximately 210 counts per second. All required systems functioned as designed during the manual reactor trip and the operating staff acted promptly and appropriately to stabilize the unit at hot shutdown.
On February 8, 2005 at 0856 hours0.00991 days <br />0.238 hours <br />0.00142 weeks <br />3.25708e-4 months <br />, after ensuring no hardware concerns existed, control rod exercising began on the misaligned rod. At 1013 hours0.0117 days <br />0.281 hours <br />0.00167 weeks <br />3.854465e-4 months <br />, four tests were completed on the rod with no issues noted. After completing the test on rod B10, all control and shutdown rods were exercised, one bank at a time. The probable cause of this event is a build up of particulate (debris) in the control rod drive mechanism internals. Control rods will be exercised prior to any unit startup to mitigate the effects of particulate in CRDM internals. The unit was taken critical at 1859 hours0.0215 days <br />0.516 hours <br />0.00307 weeks <br />7.073495e-4 months <br /> on February 8, 2005 and placed on line on February 9, 2005 at 0420 hours0.00486 days <br />0.117 hours <br />6.944444e-4 weeks <br />1.5981e-4 months <br />. This report is being submitted pursuant to 10 CFR 50.73(a)(2)(iv)(A) for a manual actuation of the reactor protection system.
(if more space is required, use addRional copies of NRC Form 366A) 1.0 DESCRIPTION OF THE EVENT On February 7, 2005 at 2002 hours0.0232 days <br />0.556 hours <br />0.00331 weeks <br />7.61761e-4 months <br />, Unit 1 commenced a reactor startup following a planned feedwater heater [EIIS-SJ, HX] maintenance outage. On February 7, 2005 at 2017 hours0.0233 days <br />0.56 hours <br />0.00333 weeks <br />7.674685e-4 months <br />, while withdrawing Control Bank "A", control rod [EIIS-AA, ROD] B-10 indicated a rapid drop from approximately 42 steps to 17 steps on the Computer Enhanced Rod Position Indication (CERPI) panel [EIIS-AA]. Withdrawal of Control Bank "A" was stopped and the CERPI indication for control rod B-10 remained at 17 steps. The remaining CERPIs in Control Bank "A" were observed to be indicating 40 to 45 steps, which is within normal tolerance. The operating team initiated 0-AP-1.00, "Rod Control System Malfunction," and subsequently determined that the reactor should be tripped.
Therefore, at 2024 hours0.0234 days <br />0.562 hours <br />0.00335 weeks <br />7.70132e-4 months <br /> the reactor was manually tripped and procedure 1 -E-0 "Reactor Trip Response" was initiated.
The operating staff acted promptly and appropriately.
Proper response of the automatic protection systems following manual actuation of the reactor trip was verified. All required systems functioned as designed during the trip. Unit 1 was stabilized at hot shutdown with reactor coolant system temperature approximately 547 degrees Fahrenheit. Decay heat was removed via steam generator blowdown [EIIS-WI] and the Main Steam Dump Valves [EIIS-SB].
At 0014 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> on February 8, 2005 a non-emergency, eight-hour notification was made to the NRC pursuant to 10 CFR 50.72(b)(3)(iv)(A).
This report is being submitted pursuant to 10 CFR 50.73(a)(2)(iv)(A) for a manual actuation of the reactor protection system (RPS).
2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS
This event resulted in no significant safety consequences or implications. At the time of the manual trip, reactor startup was in progress with reactor power in the source range at approximately 210 counts per second.
All required systems functioned as designed during the manual reactor trip and there were no radiation releases due to this event. A risk impact concluded that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> integrated core damage probability increase was low at 1E-10 and the increase in potential large early release frequency was negligible.
This core damage estimate is slightly conservative because it is based upon a more severe transient risk from a trip at full power. Therefore, the health and safety of the public were not affected.
NRC Forn 366A 11 -20011U. S. NUCLEAR REGULATORY COMMISSION t-2001)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YE l I SEQUENTIAL REVISION Surry Power Station 05000- 280 NUMBER NUMBER 2
OF 4
2005 001 00
3.0 CAUSE
Inspection and testing conducted after the reactor trip found no hardware concerns that would have caused the B-10 control rod to drop. A Root Cause Evaluation (RCE) team reviewed internal and external operating experience which identified that unexplained control rod drops or partial drops are attributed to build up of particulate (debris) in the control rod drive mechanism (CRDM) internals.
Specifically, deposits of soluble and non-soluble particulate have previously been postulated by Westinghouse to inhibit the movement of CRDM magnetic armatures for the movable and stationary grippers (either by preventing full travel or slowing down full travel of the armature assembly). The RCE team concluded that the most probable cause for the B-10 control rod misalignment was a build up of particulate (debris) in the B-10 CRDM.
4.0 IMMEDIATE CORRECTIVE ACTION(S)
Upon recognition of a misaligned control rod, the operating team reviewed abnormal procedures, manually tripped the reactor and transitioned to emergency operating procedure 1-E-0, "Reactor Trip or Safety Injection." The team then verified that the reactor and turbine tripped, both AC emergency buses energized and all required systems functioned as designed.
All rods were verified on the bottom and the reactor was placed in stable condition.
Following the event, Engineering and Maintenance personnel conducted extensive troubleshooting of the rod control power and logic cabinets in an attempt to eliminate failure mechanisms due to hardware issues such as failed circuit boards, fuses, power supplies and connections. Maintenance electricians entered containment to conduct an inspection of the CRDM connector at the reactor head and the floor receptacle box. No issues were identified.
5.0 ADDITIONAL CORRECTIVE ACTIONS
Because it was concluded the drop of the control rod was caused by build up of particulate in the CRDM, an additional corrective action was to exercise the control rods.
Exercising the control rods has been an effective method of dislodging particulate from CRDMs.
On February 8, 2005 at 0856 hours0.00991 days <br />0.238 hours <br />0.00142 weeks <br />3.25708e-4 months <br />, after ensuring no hardware concerns existed, control rod exercising began on control rod B-10. At 1013 hours0.0117 days <br />0.281 hours <br />0.00167 weeks <br />3.854465e-4 months <br />, four tests were completed on B-10 with no issues noted.
After completing the test on rod B10, all control and shutdown rods were exercised, one bank at a time.
NRC For 366A (1.2001)
`NRC FORM 366A U. S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YE8 NUMBER NUMBER Surry Power Station 05000- 280 05 S
REVS I2005-001 00 O
6.0 ACTIONS TO PREVENT RECURRENCE Implement control rod exercising prior to any unit startup to mitigate the effects of particulate in CRDM internals.
7.0 SIMILAR EVENTS
Two similar events occurred in June 2003 during Unit 1 startup where the accumulation of particulate (debris) in the CRDMs was determined to be the source of misalignment on control rods G-13 and E-5. These events were thought to be isolated cases since they followed the replacement of the reactor vessel head, which required the removal of the CRDMs from the old reactor head and their installation on the new head. The control rods were exercised to correct both events.
8.0 MANUFACTURER/MODEL NUMBER The control rod B1 0 CRDM is a Westinghouse L-1 06A.
9.0 ADDITIONAL INFORMATION
Unit 2 was at 100% power and remained unaffected by the Unit 1 event.