05000364/LER-2019-002, Manual Reactor Trip Due to Misaligned Rod During Low Power Physics Testing
| ML19178A238 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/27/2019 |
| From: | Kharrl C Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-19-0734 LER 2019-002-00 | |
| Download: ML19178A238 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(e)(2)(iv)(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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(1) 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(iv)(B), System Actuation |
| 3642019002R00 - NRC Website | |
text
~ Southern Nuclear JUN 2 7 2019 Docket No.:
50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Charles Kharrl Vice President-Farley Joseph M. Farley Nuclear Plant-Unit 2 Licensee Event Report 2019-002-00 Joseph M Farlc) Nuck"ar Pl:~nl 7388 Nonh S1a1c H"Y 95 Columb1a, Alabama 363 I 9 33-1814-1'11 ld 33-1 s 1-1 -1575 r;,~,
clharrl a soulhcrnco com NL-19-0734 Manual Reactor Trip due to Misaligned Rod during Low Power Physics Testing Ladies and Gentlemen:
In accordance with the requirements of 10 CFR 50.73(a)(2)(iv)(A), Southern Nuclear Company is submitting the enclosed Licensee Event Report for Unit 2.
This letter contains no NRC commitments. If you have any questions regarding this submittal, please contact Gene Surber at (334) 814-5448.
Respectfully submitted, Charles Kharrl Vice President-Farley CK/rgs/scm Enclosure: Unit 2 Licensee Event Report 2019-002-00 Cc: Regional Administrator, Region II NRR Project Manager-Farley Nuclear Plant Senior Resident Inspector-Farley Nuclear Plant RTYPE: CFA04.054
Joseph M. Farley Nuclear Plant - Unit 2 Licensee Event Report 2019-002-00 Manual Reactor Trip due to Misaligned Rod during Low Power Physics Testing Enclosure Unit 2 Licensee Event Report 2019-002-00
NRCFORM366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0313112020 (04-2018)
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3.Page Joseph M. Farley Nuclear Plant, Unit 2 05000 364 1
OF 3
- 4. TIUe Manual Reactor Trip due to Misaligned Rod during Low Power Physics Testing
- 5. Event Date
- 6. LER Number
- 7. Report Date
- 8. Ottler Facilities lnvotved o.,
y-Yelt I s:::* I Rev MonOI 0.,
YMr FacllllyName OocMt Number lllonltl No.
05000 2019 002 00 OIJJ !J,'f ~t.caq F.cUIIy Maine DocketNu-.
05 01 2019.
05000
- 9. Operating Mode
- 11. Tills Report Is Submitted Pursuant to the Requirements of 10 CFR §: (Check all Urat apply) 0 20.2201(b)
D 20.2203(a)(3)(i)
D 5U3(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A) 2 0 20.2201(d)
D 20.2203(a)(3)(a) 0 50.73(a)(2)(ii)(B) 0 50.73(a)(2)(V1ii){B) 0 20.2203(8)(1) 0 20.2203(8)(4)
D 50.73(a)(2)(ill) 0 50.73(a)(2)(ix)(A)
O 20.2203(e)(2)(i)
O 50.38(c)(1)(1)(A)
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50.73(e)(2)(iv)(A) 0 50.73(a)(2)(x)
- 10. Power Laval O 20 2203(a)(2)(ii)
D 50 36{c)(1)(B)(A) 0 50.73(a)(2)(v)(A) 0 73 71<*><4>
D 20 2203(a)(2)(ili)
D 50.38(c)(2J 0 50.73(a)(2)(v)(B)
D 73.71(a)(5)
D 20.2203(a)(2)(iv)
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D 50.73(a)(2)(V11)
D 73 n(a)(2)(1i) 0 50 73(a)(2)(1)(C) 0 Other (Specify in Abstract below or in NRC Form 366A)
- 12. Ueensee Contact for this LER LJc1111aee Contact Telephone Number (Include Area Code)
Gene Surber, Licensing Manager 334-814-5448 CaUM Sptem I Comporett I Manuf*ctuiV Reporuble lo ICES I Cauae I Sptem I Co~-
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- 15. Expected Submission Date 0 Y* (If yes, complete 15. Expected Submission Date)
Abstract (Umn to 1400 spaces, I.e.. approximately 14 single-spaced typewritten lines)
At 16:43 COT on May 1, 2019, with Farley Nuclear Plant (FNP) Unit 2 in Mode 2 with power in the Intermediate Range the reactor was manually tripped. Operators manually tripped the reactor during Low Power Physics Testing due to a misaligned rod.
The trip was not complex with all systems responding normally. Operations stabilized the plant in Mode 3. Decay heat was removed by the Steam Generator Atmospheric Relief Valves due to Main Steam Isolation Valves being closed during Low Power Physics Testing.
This event is reportable under 10CFR50.73(a)(2)(iv)(A) due to manual actuation of systems listed in 10CFR50.73(a)(2)(iv)(B).
FNP Unit 1 was not affected during this event.
Corrective Actions included resistance readings in the power cabinet and inspecting local connectors on the reactor head.
Additionally, multiple control rod bank withdrawals and insertions were performed while collecting system operating data. No abnormal conditions were identified, and crud is the suspected reason for the misalignment.
NRC FORM 388 (04-2018)
NRC FORM 368A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03131/2020 (0..2018)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET (See NUREG-1022, R.3 for rnstruction and guidance for completing this form htto*//www.nrc.gov/readrng-nn/doc*collectronstnureqs!stafflsr1 022/r3J)
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Cotrmission. WashilgDl, DC 20555-0001, or by *mal 10 lnb:ollecls.Rescute@nrc.p, and ti the Desk O!lic:er. Olfice or lnbmabon and RegulaiDrf Allilirs, NEOB-10202, (315M104) Olfial o lola1agemant and IIOOgel. Washington, DC 20503.
If a means used to mpase IJl llllonnillon Clllleellon does not display a amntly Ylild OUB 001'11101 nllllber, the NRC may not oonduct or sponsor. and a person IS nol reqUrad to respond to, lha Information Cllllection
- 1. FACIUTY NAME
- 2. DOCKET NUMBER
.O.U:I<NUMI:It:l<
Joseph M. Farley Nuclear Plant, Unit 2 05000-1 364
~ NUMBER NO.
I YEAR SEQUENTIAL REV
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A. Event Description
On May 1, 2019, at 11:36 COT, Farley Nuclear Plant (FNP) Unit 2 entered Mode 2 with Reactor Coolant System (RCS)
[EIIS:AB] at 547 degrees Fahrenheit and RCS pressure at 2243 psig following a refueling outage. Upon entry into Mode 2 FNP began performing Low Power Physics Testing (LPPT). During this testing, Control Rods [EIIS:AA/Rod] are fully withdrawn and inserted into the core by the Operator at the Controls (OATC) to test their amount of negative reactivity. At the beginning of this event all Control Banks and Safety Banks were fully withdrawn per the LPPT procedure and Reactor Power was in the Intermediate Range. Per the LPPT procedure the OATC began testing the Bank B Control Rods (CBB).
During the insertion of CBS the OATC identified after approximately 9 seconds (16:32 COT) that Control Rod M6 (Part of CBB) had a greater than 12 step difference between the Digital Rod Position Indication (DRPI) and the group step counter. Upon recognition of the divergence in rod positions the OATC stopped all CBB rod movement. The OATC under direction of the Senior Reactor Operator (SRO) then inserted CBS rods at 16:36 COT to add additional negative reactivity to establish a slight negative startup rate to maintain power stable in the intermediate range. Following this CBS insertion, CBB was verified at 156 steps withdrawn (DRPI) while rod M6 was at 132 steps withdrawn (DRPI). After validating the DRPI indications, the Control Room crew entered the FNP abnormal operating procedure for the misaligned Control Rod at 16:40 COT. At 16:43 COT the crew procedurally tripped the reactor with Reactor Power low in the Intermediate Range.
This operator action manually initiated a Reactor Protection System actuation [EIIS:JC]. All Control Banks and Safety Banks fully inserted including Control Rod M6. The Main Feed Water system [EIIS:SJ] was not in service at the time and the Auxiliary Feed Water (AFW) System [EIIS:BA] continued to feed the steam generators. Decay heat was removed by the Steam Generator Atmospheric Relief Valves [EIIS:SB] due to the Main Steam Isolation valves being closed during Low Power Physics Testing.
B. Cause of Event
Control Rod M6 most likely became misaligned due to CRUD in the Control Rod Drive Mechanism (CRDM) based on site and industry operating experience. The cause of the crud being present in the CRDM is unknown.
C. Safety Assessment
When the Control Room Operator manually initiated the reactor trip, all systems responded as designed. The Reactor Protection System (RPS) actuated as designed, and all Control Banks and Safety Banks fully inserted into the core. A turbine trip was not necessary because the turbine was not online at the time. Because all systems responded as required and there were no adverse effects on the health and safety of the public, the safety significance is low and no loss of safety function occurred. Additionally, no dose limits were challenged.
D. Corrective Actions
Resistance readings were taken at the power cabinet for Control Rod M6 with no abnormal conditions identified. The connector on the reactor head for Control Rod M6 was also inspected. The resistance readings were checked before and after the connector was inspected. The resistance readings before and after were equivalent. During the inspection of the Control Rod M6 connector on the reactor head no issues were identified. Control Bank B was withdrawn to the full rods out position (229 steps) and inserted three times to ensure that any potential crud was removed. The other Unit 2 Control Banks and Shutdown Banks were also fully withdrawn and inserted three times to ensure no similar condition existed. All Unit 2 Control Rods and Safety Rods operated normally with no misalignment issues.
NRC FORM 3e6A (04-2018)
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E. Similar Events
There have been no recent events at Farley Nuclear Plant (FNP) that would demonstrate that there is a current concern in system operation. The following similar events were identified in FNP history which there was a similar cause:
07/28/2008 (CAR 168529)- During startup from forced outage Control Rod 806 became misaligned with the remaining group of rods by 12 steps. It was determined crud had spread into the CRDM housing causing the latching mechanisms not to latch up correctly. As a result, the 806 rod continuously slipped as it was being pulled out.
06/17/2005 (CR 2005105949 I OE20984) -While performing a Full Length Control Rod Operability Test, a control rod did not return to Digital Rod Position Indication (DRPI) of 228 as did the other rods. Several maneuvers of that particular control rod bank were conducted which ultimately resulted in two rods being misaligned with the remaining rods in the bank fully withdrawn. The investigation of this event along with the root cause analysis indicated that the most probable cause of the rod control anomaly resulted from service age crud transferred to the new CRDM via the rod drive shaft Page 3
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