05000410/LER-2023-001, Supplement to LER 2023-001-00, Automatic Reactor Scram on Low Level Due to Partial Loss of Feedwater
| ML24060A053 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 01/30/2024 |
| From: | Crawford C Constellation Energy Generation |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NMP2L2867 LER 2023-001-01 | |
| Download: ML24060A053 (1) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 4102023001R00 - NRC Website | |
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~1\\Y.t:~1r.~ -'~~if~~~t?7-] _ ~-:. , *'-~1---{i, ,,,19 ~-,)_*, <:,,;",",* -~~,~~~~,-~__:_:_ _ 10 CFR 50.73 NMP2L2867 January 30, 2024 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
Subject:
Nine Mile Point Nuclear Station, Unit 2 Renewed Facility Operating License No. NPF-69 Docket No. 50-41 O Supplement to NMP2 Licensee Event Report 2023-001, Automatic Reactor Scram on Low Level Due to Partial Loss of Feedwater In accordance with the reporting requirements contained in 10 CFR 50.73(a){2)(iv)(A), please find enclosed Supplement to NMP2 Licensee Event Report (LER) 2023-001, Automatic Reactor Scram on Low Level Due to Partial Loss of Feedwater. There are no regulatory commitments contained in this letter. Should you have any questions regarding the information in this submittal, please contact Brandon Shultz, Site Regulatory Assurance Manager, at (315) 349-7012. Respectfully, Carl Crawford Plant Manager, Nine Mile Point Nuclear Station CC/MLR
Enclosure:
Supplement to NMP2 Licensee Event Report 2023-001, Automatic Reactor Scram on Low Level Due to Partial Loss of Feedwater cc: NRC Regional Administrator, Region I NRC Resident Inspector NRC Project Manager
Enclosure Supplement to NMP2 Licensee Event Report 2023-001, Automatic Reactor Scram on Low Level Due to Partial Loss of Feedwater Nine Mile Point Nuclear Station, Unit 2 Renewed Facility Operating License No. NPF-69
APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 03/31/2024 "!I" FORM 366 il.s. NUCLEAR REGULA TORr.cJMMISSION ltup1-2O23) Esbmaled
+""""""---~~~""!"""---""l""'!!~------1 1_ Facility Name rgj 050
- 2. Dockit Number
- 3. Page Nine Mile Point Unit 2 I
052 410 1 OF 4. ~T~e
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supplement to LER 2023-001-00, Automati9/ ~eactor Scram on Low Level Due to Partral Loss of Feedwater . I..
- 5. Event Date
- 6. LER Number/
- 7. Report Date
- 8. Other Facilities Involved Month Day Year 09 02 2023
. Year Sequential / Number, I 2023 * - 001 t 1
- Revision No.
01 Month 01 Day 30 Year Facility Name 2024 Facility Name
- 9. _Operatinq Mode
- Mode 1 Power Operation 1
- 10. Power Level 100%
Abstract
On:912/2023 at 0632 EDT, with Nine Mile Point Nuclear Station operating at 100% power, a Feedwater transient occurred resulting in a Reactor Protection Syster:n (RPS) :Automatic Reactor Scram on Low Level (Level 3, 159.3"). Following the scram,* reactor water _level dropped below Level 2 (108.8") resulting in a Group 2 Recirc Sample System Isolation, Group 3 Traversing lncore Probe (TIP) Isolation Valve lsqlation, Group 6 and 7 Reactor Water Cleanup Isolation, and Group 9 Containment Purge Isolations. All control rods inserted as expected. High Pressure Core Spray and Reactor Core Isolation Cooling (RCIC) initiated and injected as expected. Emergency Core Cooling Systems (ECCS) and RClC were secured and normal reactor pressure and level control was established for hot shutdown. This event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A).
- 2. DOCKET NUMBER
- 3. LER NUMBER YEAR SEQUENTIAL REV NUM!:IE~
- NO.
410 2023 - 001
- 01 THE ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) COMPONENT FUNCTION IDENTIFIED AND.SYSTEM NAME OF EACH COMPONENT OR SYSTEM REFERRED TO IN THIS LER ARE ENCLOSED:WITHIN [BRACKETS]
I.
DESCRIPTION OF EVENT
A. PRE.:Ev'ENT PLANT CONDITIONS: Prior to the event, NMP2 was in Mode 1 (Power Operation)* at_ 100% power.
- 8. EVENT:*
On ~/2/2023 at 0632 EDT, with Nine _Mile. Point Nuclear Station operating at 100% po_wer, a Feedwater [SJ] transient
- occurred resulting in an RPS AutomatiG ReaGtor Scram on. Low L~yel (Level 3, 159.3"). :Following the Scram,: reactor water level qrcipped below t,.ow Low Level (Level 2, _ 108.8") resulting in a Group 2 Recirc Sample System Isolation [AD];
Group 3 T:IP* Isolation Valve ls9Ja~ion, Group 6 and 7 Reactor Water Clearit,11,i Isolation [CE], aod Group 9 Containmen_t Purge Isolations [BB]. * * * *
- All contr9( rqds' inserted as expeGte'd. High Pressure Core Spray [BG] and Reactor Core l~olatipn Cooling [BN] JnitiaJed and injected:as expected. ECG~ _$ystems and RCIC wer.e secured and n.ormal reactor pressure and level control was.*
established for hpt sh_u_tdpwn. This event is reportable in accorda1:1ce with 10 CFR 50.7;3(a)(2)(iv)(A). . t_ INOPERABLE.SYSTEMS, STRUCTURES, OR COMPONENTS THAT CONTRIBUTED TO THE EVENT: - None. , D. :DATES AND APPROXIMATE TIMES:QF MAJOR occ~~ENCES AND OPERATOR ACTION~: The dc1tes, tirnes, major occurrences, and oj:>e~ator c1ctions for this ev~nt are as follows.:
- September 2, 2023: :
0632-Asudden reduction in* Feedwater flow causes level lowering to Low ~evel (Level 3, 159.), resultfng in an RPS Automatic Reactox Sq:~m. 0632-Shortly following the scram, Low* Low Level (Level 2, 108.8") is reached, resulting in Group 2, *3, 6, 7; 8, and 9 isolations. Recirculating Pumps trip, and High Pressure Core Spray (HPCS)- and Reactor Core lsolation*cooling (RCIC) automatically initiate....
- E. METHOD OF DISCOVERY:
- 2. DOCKET NUMBER
- 3. LER NUMBER YEAR
- SEQUENTIAL
'REV
- - NUMBER:
NO.* '4:10 2023 001 01 This event was self-rev.ealed when RPS Automatic Scram signal was received on Low Level (Level 3, 159"):: F: :SAFETY SYSTEM RESPONSES: All safety syste!'flsresponded per design.. II. CAUS:E OF THE EVENT . : The direct cause of the Feedwater transient was due to stem-plug separation of the 2FWS*L V1 bB, Level Control Valve* Feedwater Flow H, which resulted loss of flow from. :the.'B' Feedwater line. The B Feedwater . Level Control Valve.was dis.assembled for insp*ection to.determine if stem-to-,plug separation was the caus*e of. . the instantaneous loss of Feedwaterflow. It was observed-~h~t the plug wa~ ncflonger connected to the stem,*
- with the-last three threads on the stem having been visibly damaged. Discovery also noted the tab.lock washer was n*ot bent: up on the stem collar and the anti-rotation pin was absent from the tab washer. The investigatJ6i,
- confirmed the tab lock, washer bend was not performed *at the vendor facility prior to *installation. PowerLabs
- failure analysis concluded the anti.:.rotation pin was ccirrecHy *installed on *the tab lock washer. The bulk of the stem threags we.re* intact, su~jgesting the plug hac;j unthreaded frol'n the stem. Remriarits of the Teflon balance seal assembly we,re located on the top su_rface. of the plug:
Ill.
ANALYSIS OF THE EVENT
The scram did not have any impact to the health or safety of:the public. All safety systems responded per design. Failure analysis*of the sterri and plug did not identify any adverse co*nditi.ons from manufacturing that would
- h1;3ve caused this i~sue. The obser.vati,or:t of the locking t_ab:not being bent up t:<;> the flat of the. s.~em.is not causal because the tack welds between the stem and lock washer were intact. The failure analysis confirmed
.. the anti-rotation pin was originally in~talled. This is based on analysis of the pin weld and corresponding we'i1r marks on the inner di~rr,eter of the pll,1g where the pin is contained. In this case, it appears fatigue propagation
- was associated with low cycle and high *relative loading which is caused from hydraulic instabilities within the
- valve due tq balance seal failure: Severe degradation of the Teflon balance seal allowec:f high-pressure fluid to travel past the* seal and across the stem-to-plug cohnectioo interface, resulting in the pin *shearing from the tab lock washer. The anti-rotation pin is nof designed to protrude up through the lock washer, therefor~ there is risk of iriadequatl? engagement betvve,eri the pin and \\al;:> lock washer that.could weaken the structural inte,grity of the design. This would provide minimal engagement between the pir-, and tab lock washer.increasing
!?Usceptibility to_ mechanicai failure: In addition,.the.fackweld associated.with the tab lock washer pin hole was
- further investigated* under the scanning electron micro$cope. Examination of the fracture surface revealed a dimple structure indicative of an overload condition. In addition, beach marks were id_entified which are.
- indicative of a progr~ssive failure mechanism, like.fatigue. Tt,e root cause of ~FWS-LVt0B stE;im to disk
. separation event is attributed to marg_inal design of the stem and plug connection provided by the vendor. IV.
- CORRECTIVE ACTION,S A'* ACTION TAKEN.TO RETURN AFFECTED SYSTEMs:to PRE-EVENT NORMAL STATUS:.
The station completed repairs:to the 2FWS*LV108, Level Control Valve Feedwater Flow 8. Additionally, a Lost Parts Evaluation was.completed for the identified foreign material in the FW system. * *
- 8.* ACTION TAKEN'dR.PLANNED TO PREVENT OCClJR'F~ENCE:-
- An engineering charige-(ECP-23-000326) was developed. in the*,interim to add welds between the tab lock
- washer <ind plug to provide. ac;Jde!d*protection for a future* stem/plug separation.:The station plans:to in.st.all a..
revised (integral/single piece) _valve anti-rotation de~ign which eliminates the potential for stem-t.o-plug.... separation. *
- V.
ADDITIONAL INFORMATION
2FWS*LV108; Level Control Valve Feedwater Flow 8. s.* PREVIOUS LERs on SIMILAR EVENTS: .. Npne* * ..... Page_4_ of ___ 4_ }}