05000354/LER-2015-005, Regarding Reactor Scram Due to Invalid RRCS Actuation
| ML15329A093 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 11/24/2015 |
| From: | Carr E Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N15-0240 LER 15-005-00 | |
| Download: ML15329A093 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| 3542015005R00 - NRC Website | |
text
LR-N15-0240 NOV 2 4 2015 PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Hope Creek Generating Station Unit 1 10CFR50.73 Renewed Facility Operating License No. NPF-57 Docket No. 50-354
Subject:
Licensee Event Report 2015-005-00 In accordance with the requirements of 10 CFR 50. 73(a)(2)(iv)(A),
PSEG Nuclear LLC is submitting the enclosed Licensee Event Report (LER) Number 2015-005-00, "Reactor Scram Due to Invalid RRCS Actuation."
If you have any questions or require additional information, please contact Mr. Thomas MacEwen at (856) 339-1097.
There are no regulatory commitments contained in this letter.
Sincerely, Eric S. Carr Plant Manager Hope Creek Generating Station ttm Attachment: Licensee Event Report 2015-005-00
LR-N 15-0240 Page 2 of 2 Document Control Desk cc:
Mr. Daniel Dorman, Regional Administrator-Region I, NRC Mr. Tom Wengert, Project Manager - US NRC Justin Hawkins, NRC Senior Resident Inspector - Hope Creek (X24)
Mr. Patrick Mulligan, Manager IV Bureau of Nuclear Engineering New Jersey Department of Environmental Protection PO Box 420 Trenton, NJ 08625 10CFR50.73 Mr. Thomas MacEwen, Hope Creek Commitment Tracking Coordinator (H02)
Mr. Lee Marabella - Corporate Commitment Tracking Coordinator (N21)
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0113112017 (02-2014)
- ti:M"<'f\\
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 6. LER NUMBER
- 3. PAGE YEAR I SEQUENTIAL I REVISION NUMBER NUMBER 3 OF 3 2015
- - 005
- - 000 A review of this condition determined that a Safety System Functional Failure (SSFF) as defined in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Performance Indicator Guideline," did not occur.
PREVIOUS EVENTS The cause evaluation will review similarity to previous events. The result of that review will be included in the supplement.
CORRECTIVE ACTIONS
The technician involved in the event was disqualified from performing any surveillance testing or other plant maintenance duties.
Other corrective actions will be determined by the cause evaluation.
COMMITMENTS
This LER contains no regulatory commitments.
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor (BWR/4)
Reactor Protection System - EllS Identifier {JC}*
Redundant Reactivity Control System - EllS Identifier {JC}*
Reactor Recirculation System - EllS Identifier {AD}*
APPROVED BY OMB: NO. 3150*0104 EXPIRES: 01/31/2017
, !he NRC may not conduct or sponsor, and a person Is not required to respond to, the information collection.
YEAR 2015
- 6. LER NUMBER SEQUENTIAL REVISION NUMBER NUMBER
- - 005
- 000
- 3. PAGE 2 OF 3
- Energy Industry Identification System {EllS} codes and component function identifier codes appear as {SS/CCC}
IDENTIFICATION OF OCCURRENCE Event Date: 09/28/15 Discovery Date: 09/28/1 5 CONDITIONS PRIOR TO OCCURRENCE Hope Creek was in Operational Condition 1 at 100 percent rated thermal power (RTP). Redundant Reactivity Control System (RRCS) {JC}, Division 1, surveillance testing was in progress.
DESCRIPTION OF OCCURRRENCE On 9/28/2015 at 20:46, a Hope Creek Instrument and Controls technician was performing a surveillance test of RRCS division 1, channel B, to simulate a high reactor pressure condition. The RRCS system is designed to detect and respond to an Anticipated Transient Without Scram (ATWS) condition. One indication of this condition is high reactor pressure, at or above 1071 psig. Under these conditions, the RRCS is designed to trip both Reactor Recirculation Pumps (RRPs) {AD} and initiate Alternate Rod Insertion (ARI). The RRPs are tripped to reduce core flow and increase the formation of core voids, thus reducing power. ARI provides an alternate path for control rod insertion by depressurizing the scram air header through valves separate from the RPS {JC} scram valves.
During the test, a keypad on the local RRCS panel is used to enter the test parameter, the test signal value and the channel being tested. The technician was expected to enter a test pressure signal of 1400 psig into the B channel of division 1. Plant data indicate the test pressure signal was also entered in channel A of division 1. With the 1400 psig test signal in both the A and B channels of logic, division 1 of the RRCS system actuated, causing RRPs to trip and ARI to begin control rod insertion by depressurizing the scram air header.
The change in reactor power caused a reactor water level transient which reached the RPS trip set-point of +12.5 inches. Although the control rods were already moving inward due to ARI actuation, the RPS functioned as designed to ensure reactor shutdown was completed via a scram signal. After the initial transient, plant operators stabilized reactor pressure and water level using turbine bypass valves and the feed water system, respectively.
CAUSE OF EVENT
A cause evaluation is being conducted to determine the causes associated with the event. A supplement to this LER will be submitted to report the results of the cause evaluation.
SAFETY CONSEQUENCES AND IMPLICATIONS
There were no consequences to nuclear safety as a result of this event. The RRCS and RPS system operated as designed to shut down the reactor. All necessary support systems functioned as needed to support plant stabilization and recovery post transient.
SAFETY SYSTEM FUNCTIONAL FAILURE APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 6. LER NUMBER
- 3. PAGE YEAR I SEQUENTIAL I REVISION NUMBER NUMBER 3 OF 3 2015
- - 005
- - 000 A review of this condition determined that a Safety System Functional Failure (SSFF) as defined in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Performance Indicator Guideline," did not occur.
PREVIOUS EVENTS The cause evaluation will review similarity to previous events. The result of that review will be included in the supplement.
CORRECTIVE ACTIONS
The technician involved in the event was disqualified from performing any surveillance testing or other plant maintenance duties.
Other corrective actions will be determined by the cause evaluation.
COMMITMENTS
This LER contains no regulatory commitments.