05000318/LER-2004-001, Regarding Reactor Trip Due to Low Steam Generator Water Level After Feed Pump Trip
| ML040860439 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  (DPR-069) |
| Issue date: | 03/23/2004 |
| From: | Nietmann K Constellation Energy Group, Constellation Generation Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 04-001-00 | |
| Download: ML040860439 (11) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3182004001R00 - NRC Website | |
text
Kevin J. Nietmann Plant General Manager Calvert Cliffs Nuclear Power Plant Constellation Generation Group, LLC 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 410 495-4101 410 495-4787 Fax Constellation Energy Group March 23, 2004 U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:
SUBJECT:
Document Control Desk Calvert Cliffs Nuclear Power Plant Unit No. 2; Docket No. 50-318; License No. DPR 69 Licensee Event Report 2004-001 Reactor Trip Due to Low Steam Generator Water Level After Feed Pump Trip The attached report is being sent to you as required under 10 CFR 50.73 guidelines. Should you have questions regarding this report, we will be pleased to discuss them with you.
Very truly yours, KJN/JKK/bjd
Attachment:
As stated cc:
J. Petro, Esquire J. E. Silberg, Esquire Director, Project Directorate I-1, NRC G. S. Vissing, NRC H. J. Miller, NRC Resident Inspector, NRC R. I. McLean, DNR
&r1--
Abstract
At 15:26 on January 23, 2004, Calvert Cliffs Unit 2 tripped from 100 percent power, initiated by the Reactor Protective System due to low steam generator water level caused by an erroneous over speed trip signal on the steam generator feed pump. The erroneous trip signal occurred because of a degraded digital speed monitor supply voltage caused by corrosion of an inline fuse and the fuse holder. The turbine bypass valves and atmospheric dump valves opened as designed, but the quick-open signal did not clear due to the failure of a relay in the reactor regulating circuit. The open valves resulted in over-cooling of the Reactor Coolant System, a Steam Generator Isolation Signal, and a Safety Injection Actuation Signal causing a loss of normal heat removal. During the recovery, a large insurge of subcooled water cooled the pressurizer, lowering reactor coolant pressure to produce a second Safety Injection Actuation Signal.
The corroded fuse, fuse holder, and failed relay were replaced. Operations staff was briefed on the effect of a large insurge of water on reactor coolant pressure, and Unit 2 was restarted and paralleled to the grid on January 25, 2004 at 21:53.
NRC FORM 36 (7.2W1)
(If more space Is required, use additional copies of (If more space is required, use additional copies of (If more space is required, use additional copies of (If more space Is required, use additional copies of (If more space is required, use additional copies of (If more space Is required, use additional copies of (If more space is required, use additional copies of (If more space Is required, use additional copies of NRC Forn 366A) pressurizer spray valve (2RC1 00F). Post-trip analysis demonstrated the disabled heaters and leaking spray valve did not have a significant impact on the recovery from the event.
The event resulted in automatic actuation of the RPS, safety injection, auxiliary feedwater, and DGs, and therefore is reported in accordance with 10 CFR 50.73(a)(2)(iv)(A). Immediate notification of the reactor trip and the SIAS (Event Number 40472) was made on January 23, 2004, in accordance with 10 CFR 50.72 (b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A), respectively.
On January 26, 2004, after post-trip review of data revealed the reactor trip was an automatic trip and not a manual trip, a follow-up notification was made.
This event is reported in accordance with all of the following criteria:
10 CFR 50.73(a)(2)(iv)(A); "Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) of this section,"
(a)(2)(iv)(B)The systems to which the requirements of paragraph (a)(2)(iv)(A) of this section apply are:
(1) Reactor Protective System (RPS) including reactor scram or reactor trip, (3) Emergency Core Cooling Systems (ECCS) for pressurized-water reactors (PWRs) including: high-head, intermediate-head, and low-head injection systems and the low-head injection systems and low pressure injection function of residual (decay) heat removal systems.
(6) PWR auxiliary or emergency feedwater systems.
(8) Emergency AC electrical power systems, including: emergency diesel generators (EDGs)
There was also a SGIS resulting in the MSIVs closing and the subsequent loss of normal heat removal capability.
No actual nuclear safety consequences were incurred from this event; however, plant equipment malfunctions that occurred following the trip were as follows:
The "Quick-open" signal to the ADVs and the TBVs did not clear due to the malfunction of the K7 relay. Because the ADVs remained open, excess energy was removed from the RCS, over-cooling the system. The Probabilistic Risk Analysis results in an increase in frequency of 9E-06 to the Core Damage Frequency (CDF) and a less than 9E-07 increase in the Large Early Release Frequency (LERF).
Control Room Operators could not reset SIAS channel "B" due to a loose connection in the "B" remote reset circuit. The SIAS channel "B" was reset from the ESFAS cabinet in the Cable Spreading Room. The inability to reset the SIAS "B" channel from the Control Room did not deleteriously affect event recovery.
(If more space Is required, use additional copies of NRC Form 366A)
During the recovery phase of the event, RCS pressure decreased to a point that a second SIAS actuation occurred. The equipment normally cycled by a SIAS was already in its demand state with the exception of safety-related pressurizer heaters which cutoff, and letdown isolation valves which closed. A Probabilistic Risk Analysis of the second SIAS initiation indicated no contribution to CDF or LERF.
Plant safety margins, design basis limits, and Technical Specification cooldown rates were maintained during the event. All appropriate compensatory and corrective actions were completed prior to the plant restart.
V.
ADDITIONAL INFORMATION
A.
Component Failures
IEEE 803 IEEE 805 Component EIIS Function System ID No. 22 SGFP Speed Control Power Supply Fuse FUB JK Reactor Regulating System K7 Relay RLY JD B.
Previous Occurrences
A review of Calvert Cliffs' events over the past several years was performed. There were similar events identified involving plant trip on low SG water level after SGFP trip, but the cause of the pump trip was tied to a feedwater control circuit that is no longer in use. In 1991, Licensee Event Report No. 317/91-003 "Reactor Protection System Actuation and Plant Trip Due to Low Steam Generator Water Levels Caused by Loose Electrical Fuse" documents a Unit 1 trip on October 1, 1991 caused by a loose electrical fuse in the power supply to the 12 Feedwater Regulating Valve and the 12 SGFP turbine speed controller. The root cause was an improperly installed fuse in the power supply circuitry. The affected circuitry is no longer in use. The causal analysis describes a plant trip due to human performance problems: workers did not exercise adequate caution when installing fuses to prevent damage to the fuse holder. The damaged fuse holder was replaced, other similar applications were inspected, and personnel who manipulate fuses were trained to properly install and remove fuses. There were no plant trips caused by electrical problems from failing to adequately assess environmental conditions in the current SG feed water control circuit.