05000461/LER-1917-007, Re Manual Reactor Scram Due to Loss of Feedwater Heating

Re Manual Reactor Scram Due to Loss of Feedwater Heating

ML17223A187
Person / Time
Site:	Clinton
Issue date:	08/09/2017
From:	Stoner T Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SRRS 5A.108, U-604366 LER 17-007-00
Download: ML17223A187 (5)

LER-1917-007, Re Manual Reactor Scram Due to Loss of Feedwater Heating

Event date:
Report date:
Reporting criterion:	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)(ix)(A) 10 CFR 50.73(a)(2)(iii) 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(iv), System Actuation
4611917007R00 - NRC Website

text

Exelon Generation Clinton Power Station 8401 Power Road Clinton, IL 61727 U-604366 1 OCFR50.73 August 9, 2017 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001

Subject:

Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Licensee Event Report 2017-007-00 SRRS 5A.108 Enclosed is Licensee Event Report (LER) 2017-007-00: Manual Reactor SCRAM due to Loss of Feedwater Heating. This report is being submitted in accordance with the requirements of 1 O CFR 50.73.

There are no regulatory commitments contained in this report.

Should you have any questions concerning this report, please contact Mr. Dale Shelton, Regulatory Assurance Manager, at (217) 937-2800.

Respectfully, Theodore R. Stoner Site Vice President Clinton Power Station KP/cac Attachment: Licensee Event Report 2017-007-00 cc:

Regional Administrator-NRC Region Ill NRC Senior Resident Inspector - Clinton Power Station Office of Nuclear Facility Safety -

Illinois Emergency Management Agency

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2017)

, the htt(2://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3/)

NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. PAGE Clinton Power Station, Unit 1 05000461 1 OF 4

4. TITLE Manual Reactor SCRAM due to Loss of Feedwater Heating

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED I

SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.

MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 06 10 2017 2017 - 007

- 00 08 09 2017 05000

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)

D 20.2201 (bl D 20.22os(aJ(3J(iJ D 50.73(a)(2)(ii)(A)

D 50.73(a)(2)(viii)(A)

D 20.2201 (dJ D 20.22os(aJ(3J(iiJ D 50.73(a)(2)(ii)(B)

D 50.73(a)(2)(viii)(B) 1 D

D D

D 50.73(a)(2)(ix)(A) 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(iii)

D 20.2203(a)(2)(i)

D 50.36(c)(1 )(i)(A)

[gJ 50.73(a)(2)(iv)(A)

D 50.73(a)(2)(x)

10. POWER LEVEL D 20.2203(a)(2)(ii)

D 50.36(c)(1 )(ii)(A)

D 50.73(a)(2)(v)(A)

D 1s.11 (aJ(4l D 20.2203(a)(2)(iii)

D 5o.3s(cJ(2J D 50.73(a)(2)(v)(B)

D 73.11 (aJ(5J D 20.2203(a)(2)(iv)

D 5o.4s(aJ(3J(iiJ D 50.73(a)(2)(v)(C)

D 13.11(aJ(1).

093 D 20.2203(a)(2)(v)

D 50.73(a)(2)(i)(A)

D 50.73(a)(2)(v)(D)

D 1s.11(aJ(2)(iJ D 20.2203(a)(2)(vi)

D 50.73(a)(2)(i)(B)

D 50.73(a)(2)(vii)

D 1s.11(a)(2)(iiJ

~

D 50.73(a)(2)(i)(C)

D OTHER Specify in Abstract below or in SEQUENTIAL NUMBER 007 REV NO.

00 Troubleshooting using an ohmmeter found that high resistance in the circuit was eliminated after pulling Moore trip unit 1 L YHD103A, which is the trip unit that provides automatic actions on a Hi-Hi level for FW heater 4A. This indicated that the loss of power to rack CA-1 was caused by fuse FU-89 opening in response to a shorted condition on the Moore trip unit 1LYHD103A. When fuse FU-89 opened, power was also lost to the Moore trip units for low pressure FW heaters 2A, 3A, SA, and 6A resulting in a loss of extraction steam to the 'A' FW heater train.

C.

Cause of the Event

The root cause of this event is currently under investigation. The causal factors that created this condition and the associated corrective actions will be provided in a revision to this Licensee Event Report (LER).

D.

Safety Consequences

This event is reportable under the provisions of 1 O CFR 50.73(a)(2)(iv) due to the manual actuation of the reactor protection system.

An assessment of the safety consequences and implication of this event determined that the manual reactor scram ensured the plant remained in a safe and stable condition and no operating limits were exceeded.

The design basis loss of feedwater heating transient for CPS is based on a maximum temperature transient of 100°F. Should this event to occur at a lower reactor power level, the severity of the transient would be reduced commensurate with the reduction in feedwater heating.

E.

Corrective Actions

Prior to startup, CPS modified the circuit card locations in the panel that contains the Heater Drain system Moore trip units and thereby diversified the power supplied so that the trip units have less dependency on common fuses. In addition, the blown fuse FU-89 was replaced. Additional corrective actions included installation of temporary cooling and temperature loggers in the Panel 1 PA08J to monitor for elevated temperature condition.

Further corrective actions for this event will be provided in the follow-up LER.

F.

Previous Similar Occurrences LER 88-025 - Loss of Feedwater Heating System Transient Outside Design Basis Due to Inadequate Communication Between the Architect Engineer and the Nuclear Steam Supply System Supplier.

On July 28, 1988, CPS experienced a partial loss of FW heating. The FW temperature drop, excluding the change caused by a reduction in power, was greater than 102, but less than 112, degrees Fahrenheit (F). The design basis loss of FW heating transient for CPS is based on a maximum temperature transient of 100 degrees F. The cause of the loss of FW heating was the inappropriate setting of the FW heater level controllers. The cause of exceeding the design basis is attributed to the failure of the FW heating system design to meet design requirements. This was caused by a lack of adequate communication between the Nuclear Steam Supply System (NSSS) supplier and the architect engineer regarding the NSSS design requirements for the FW heating system. Feedwater heating system design changes, including changes to the level trip setpoint for closing the extraction steam valves and replacing power supply fuses, were made to ensure that the design basis is met.

G.

Component Failure Data

Failed card was determined to be a Moore Industries DCA alarm card.

Model Number: DCA/4-20ma/DH1 L2/45dC/-AD-100HB1 (PC)

Serial Number: 2412651 NR(; FORM 366A (04-2017)

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