05000410/LER-2003-002, Regarding Reactor Scram Due to Electric Grid Disturbance
| ML032970109 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 10/14/2003 |
| From: | Hopkins L Constellation Energy Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NMP2L 2103 LER 03-002-00 | |
| Download: ML032970109 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(x)(A) 10 CFR 50.73(a)(2)(iii) 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown |
| 4102003002R00 - NRC Website | |
text
P.O. Box 63 Lycoming, New York 13093 Constellation I
Energy Group Nine Mile Point Nuclear Station October 14, 2003 NMP2L 2103 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Subject:
Nine Mile Point Unit 2 Docket No. 50410 Facility Operating License No. NPF-69 Licensee Event Report 03-002, "Reactor Scram Due to Electric Grid Disturbance" Gentlemen:
In accordance with 10 CFR 50.73(a)(2)(iv)(A), we are submitting Licensee Event Report (LER)03-002, "Reactor Scram Due to Electric Grid Disturbance."
Very truly yours, Lawrence A. Hopkins Plant General Manager LAH/KLE/bjh Attachment cc:
Mr. H. J. Miller, NRC Regional Administrator, Region I Mr. G. K. Hunegs, NRC Senior Resident Inspector
<(?-1G1 1'
11-11"
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 7-31-2004 (1-2001)
, the NRC may not conduct or sponsor and a person Is not required to respond to, the Information collection.
FACILITY NAME (1)
DOCKET NUMBER (2)
PAGE (3)
Nine Mile Point, Unit 2 05000410 1 OF 3
TITLE (4)
Reactor Scram Due to Electric Grid Disturbance EVENT DATE (5)
LER NUMBER (6)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (8)
MO DAY YEAR YEAR SEQUENTIAL REV MO DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NO j05000 08 14 2003 2003 -
002 -
00 10 14 2003 FACILITY NAME DOCKET NUMBER MODE l
I I~~I j__05000 OPERATING THIS REPORT IS SUBMrrTED PURSUANTTO THE REQUIREMENTS OF 10 CFR §: (Checkall thatapply) (11)
MODE (9) 1 20.2201(b) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(B) 50.73(a)(2)(x)(A)
POWER LEVEL (10) 20.2201 (d) 20.2203(a)(4) 50.73(a)(2)(iii) 50.73(a)(2)(x) 100 20.2203(a)(1)
_ 50.36(c)(1)(i)(A)
X 50.73(a)(2)(iv)(A) 73.71 (a)(4) 20.2203(a)(2)(i) 50.36(c)(1)(ii)(A) 50.73(a)(2)(v)(A) 73.71 (a)(5)
_ 20.2203(a)(2)(ii)
_ 50.36(c)(2) 50.73(a)(2)(v)(B)
OTHER 20.2203(a)(2)(iii) 50.46(a)(3)(ii) 50.73(a)(2)(v)(C)
Specify In Abstract below or In
_ 20.2203(a)(2)(iv) 50.73(a)(2)(i)(A)
_50.73(a)(2)(v)(D)
(if more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A) (17)
Ill. Analysis of Event This event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A) because of the reactor scram while critical and because of the automatic start of the EDGs.
No Emergency Core Cooling Systems actuated or should have actuated.
All rods fully inserted on the scram.
All three EDGs started and ran as designed.
Based on the above, the event did not pose a threat to the health and safety of the public.
IV. Corrective Actions
None.
V. Additional Information
- 1.
Failed Components:
None
- 2.
Previous similar events
None
- 3.
Identification of components referred to in this Licensee Event Report:
Components IEEE 805 System ID IEEE 803A Function Feedwater System SJ N/A Main Steam System SB N/A Main Turbine System TA N/A Turbine Bypass System JI N/A Turbine Control System TG N/A Reactor Core Isolation Cooling BN N/A Emergency Buses EB N/A Non-Emergency Buses EA N/A Recirculation System AD N/A Reactor Core AC N/A Emergency Diesel Generator EK N/A 115 KV Off-site Power System FK N/A Control Rod Drive System AA N/A Turbine TA TRB Control rod AA ROD Bus EA, EB BU Pump AD, AA P
Valve SB, SJ, JI ISV, PCV Reactor Vessel AD RPV