05000410/LER-2010-001, Regarding Reactor Scram Due to Inadvertent Actuation of the Redundant Reactivity Control System During Maintenance
| ML100750244 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 03/08/2010 |
| From: | Lynch T Constellation Energy Nuclear Group, Nine Mile Point |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 10-001-00 | |
| Download: ML100750244 (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 10 CFR 50.73(a)(2)(iv)(B), System Actuation |
| 4102010001R00 - NRC Website | |
text
Thomas A. Lynch P.O. Box 63 Plant General Manager Lycoming, New York 13093 315.349.5205 315.349.1321 Fax CENG a~ joint ventture of 4O itenation o,0 -erF NINE MILE POINT NUCLEAR STATION March 8, 2010 U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION:
Document Control Desk
SUBJECT:
Nine Mile Point Nuclear Station Unit No. 2; Docket No. 50-4 10 Licensee Event Report 2010-001, Reactor Scram Due to Inadvertent Actuation of the Redundant Reactivity Control System During Maintenance In accordance with 10 CFR 50.73(a)(2)(iv)(A), please find attached Licensee Event Report 2010-001, Reactor Scram Due to Inadvertent Actuation of the Redundant Reactivity Control System During Maintenance.
The direct cause of this event was venting of Residual Heat Removal instrumentation during planned maintenance. A root cause analysis to identify programmatic/organizational weaknesses that contributed to this event is on going. A supplemental report will be submitted that identifies the root cause(s) for the event and corrective actions taken by May 7, 2010.
Should you have questions regarding the information in this submittal, please contact T. F. Syrell, Licensing Director, at (315) 349-5219.
Very truly yours, TAL/GNS
Attachment:
Licensee Event Report 2010-001, Reactor Scram Due to Inadvertent Actuation of the Redundant Reactivity Control System During Maintenance cc:
S. J. Collins, NRC R. V. Guzman, NRC Resident Inspector, NRC R. A. Hathaway, INPO if
- - 4 ATTACHMENT LICENSEE EVENT REPORT 2010-001 REACTOR SCRAM DUE TO INADVERTENT ACTUATION OF THE REDUNDANT REACTIVITY CONTROL SYSTEM DURING MAINTENANCE Nine Mile Point Nuclear Station, LLC March 8, 2010
ONRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08131/2010 (9-2007)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Nine Mile Point Unit 2 05000410 1 of 5
- 4. TITLE Reactor Scram Due to Inadvertent Actuation of the Redundant Reactivity Control System During Maintenance
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED FAcILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR None NA NUMBER NO.
MONTHDAYYEAR IFACILITY NAME DOCKET NUMBER 01 07 2010 2010 001 00 03 08 2010 j None NA
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
[1 20.2201(b)
El 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
[E 50.73(a)(2)(vii)
[ 1l 20.2201(d)
[I 20.2203(a)(3)(ii)
[] 50.73(a)(2)(ii)(A)
[I 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
El 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
[] 20.2203(a)(2)(i)
[I 50.36(c)(1)(i)(A)
Cl 50.73(a)(2)(iii)
[I 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A)
[3 50.73(a)(2)(x)
[I 20.2203(a)(2)(iii)
El 50.36(c)(2) 0l 50.73(a)(2)(v)(A)
El 73.71(a)(4)
[I 20.2203(a)(2)(iv)
[3 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
El 73.71(a)(5) 100 El 20.2203(a)(2)(v) 0l 50.73(a)(2)(i)(A)
[I 50.73(a)(2)(v)(C)
El OTHER [I 20.2203(a)(2)(vi)
[E 50.73(a)(2)(i)(B)
[E 50.73(a)(2)(v)(D)
Specify in Abstract below or in 01/04/10 - 01/07/10 01/07/10 @ 0100 Field walk downs of the planned work performed by maintenance technicians failed to detect the common reference leg shared by RHS differential pressure transmitter 2RHS*PDT24C and RPV level instrument 21SC*PT4B.
Workers opened and then reclosed the drain valve for RHS differential pressure transmitter 2RHS*PDT24C.
This action initiated a transient in RPV level instrumentation that caused an invalid Low-Low RPV water level signal (Level 2).
The Level 2 signal caused a Division II RRCS initiation signal that caused ARI scram initiation, trip of the Reactor Recirculation Pumps, and RCIC initiation.
E. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED
Reactor Recirculation Pumps could not initially be re-started following the plant trip. This condition was subsequently determined to be caused per design by the thermal shock prevention logic circuit.
Division II Automatic Depressurization System (ADS) Low RPV Level Confirmatory signal was not received following the scram. This condition was subsequently determined to be caused by the fill and vent activity on RHS differential pressure transmitter 2RHS*PDT24C that initiated this event.
F. METHOD OF DISCOVERY
The reactor scram was self-revealing via multiple control room indications.
After the plant scram, the technicians performed a hand over hand piping inspection and discovered the common reference leg of RHS differential pressure transmitter 2RHS*PDT24C with RPV level instrument 21SC*PT4B.
G. MAJOR OPERATOR ACTION:
Following receipt of the Division II RRCS initiation signal, the operating crew initiated a manual scram by placing the reactor mode switch in SHUTDOWN, verified all control rods fully inserted, and stabilized the plant in accordance with plant procedures.
H. SAFETY SYSTEM RESPONSES:
Division II Emergency Core Cooling System (ECCS) Residual Heat Removal (RHS) subsystems B and C were inoperable and unavailable due to planned maintenance.
Divisions I and III ECCS were operable and capable of performing their intended function during the event.
RCIC initiated due to the invalid RPV Level 2 signal. The RCIC system injected to the vessel and responded as designed.
1I.
CAUSE OF EVENT
The direct cause of this event was venting of RHS instrumentation during planned maintenance. A root cause analysis to identify programmatic/organizational weaknesses that contributed to this event is on-going.
Following completion of the root cause analysis, a supplemental report will be submitted that identifies the cause(s) for the event and corrective actions taken by May 7, 2010.
III. ANALYSIS OF THE EVENT
This event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph 10 CFR 50.73(a)(2)(iv)(B). The Reactor Protection System (RPS) and the RCIC system were actuated during this event. Both systems are listed in 10 CFR 50.73(a) (2) (iv)(B).
The actual consequences of this event were a reactor scram, initiation of RCIC, and exceeding the 100 Degrees F/hr. cooldown rate limit for the RPV. The maximum cooldown in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> was 102 degrees F. Cooldown rate was returned to less than 100 degrees/hr. within 30 minutes.
An engineering assessment of the cooldown concluded that RPV allowable stress loadings were not exceeded. Plant response to the initiation of Redundant Reactivity Control System was per design.
Based on the above, it is concluded that the safety significance of this event is low and the event did not pose a threat to the health and safety of the public or plant personnel.
This event affects the NRC Regulatory Oversight Process (ROP) Index for Unplanned Scrams. Due to this scram, the Unplanned Scram Index value will be 0.8 compared to a Green-to-White threshold value of greater than 3. This reduction in margin will not result in entry into the Increased Regulatory Response (White) Band.
IV. CORRECTIVE ACTIONS
A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:
The operating crew stabilized the plant in accordance with plant procedures.
B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:
Compensatory measures put in place as a result of this event include:
Requiring a separate plant impact determination and a second Senior Reactor Operator (SRO) review of multi-discipline work orders.
Requiring field walkdowns to be completed by each maintenance discipline for multi-discipline work orders.
Reinforcing the requirement that a work order be re-routed to Operations if text changes are made after the plant impact review is complete.
A description of the actions taken to prevent recurrence will be provided in a supplement to this Licensee Event Report after the root cause analysis for this event is completed.
V. ADDITIONAL INFORMATION
A. FAILED COMPONENTS:
None B. PREVIOUS LERs ON SIMILAR EVENTS:
Previous LERs will be reviewed for similar events after the root cause analysis for this event is completed.
The results of this review will be reported in a supplement to this LER.
C. THE ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) COMPONENT FUNCTION IDENTIFIER AND SYSTEM NAME OF EACH COMPONENT OR SYSTEM REFERRED TO IN THIS LER:
COMPONENT IEEE 803 FUNCTION IEEE 805 SYSTEM IDENTIFIER IDENTIFICATION Plant Protection System JC Engineered Safety Features Actuation System JE Reactor Core Isolation Cooling P
BN Residual Heat Removal PDIT BO Reactor Recirculation System P
AD D. SPECIAL COMMENTS: None