ML17347A621

LER 87-030-00:on 871027,determined That Existing Min Recirculation Design Configuration for RHR Sys Potentially Inadequate.Recirculation Flow Path for RHR Pumps Original Design of Plant.Mods developed.W/871123 Ltr

ML17347A621
Person / Time
Site:	Turkey Point
Issue date:	11/23/1987
From:	Hart R, Woody C FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
L-87-483, LER-87-030, LER-87-30, NUDOCS 8711300254
Download: ML17347A621 (6)

Text

REGULA. Y INFORMATION DISTRIBUTI YSTEM (RIDS)

ACCESSION ABR: 8711300254 DOC. DATE: 87/11/23 NOTARIZED: NO DOCKET 0 FACIL; 50-250 Turkey Point Planti Unit 3i Florida Power and Light C 05000250 AUTH. NAME AUTHOR AFFILIATION HART'. D. Florida Power h Light Co.

WOODY, C. Q. Florida Power 5 Light Co.

REC IP. NAME RECIPIENT AFF ILI AT I ON

SUBJECT:

LER 87-030-00: on 871027'esign basis reconstitution discovers RHR recirculation line not designed to assure adequate flow for each pump. Caused bg plant or iginal design.

Plant mods will be completed. W/871123 ltr.

DISTRIBUTION CODE: IE22D COPIES RECEIVED: LTR j ENCL ) SIZE:

TITLE: 50. 73 Licensee Event Repor t (LER) i Incident Rpti etc.

NOTES:

REC IP IENT CQP IES REC IP IENT COPIES ID CODE/NAME LTTR ENCL ID CODE/MANE LTTR ENCL PD2-2 LA 1 1 PD2-2 PD 1 1 Mc DONALD' 1 INTERNAL: ACRS MICHELSON 1 1 ACRS MOELLER 2 2 AEOD/DOA 1 1 AEQD/DSP/NAS 1 1 AEOD/DSP/ROAB 2 2 AEOD/DSP/TPAB 1 1 ARM/DCTS/DAB 1 1 DEDRO 1 1 NRR/DEST/ADS 0 NRR/DEST/CEB 1 1 NRR/DEST/ELB 1 1 NRR/DEST/ ICSB 1 1 NRR/DEST/MEB 1 1 NRR/DEST/MTB 1 1 NRR/DEST/PSB 1 1 NRR/DEST/RSB 1 1 NRR/DEST/SGB 1 NRR/DLPG/HFB 1 1 NRR/DLPG/GAB 1 NRR/DOE*/EAB 1 1 NRR/DREP/R*B 1 1 NRR/DREP/RPB 2 2

/SIB 1 1 NRR/PMAS/ ILRB 1 1 REG FIL 02 1 1 RES DEPY GI 1 1 RES TELFORDi J 1 RES/DE/EIB 1 RGN2 FILE 01 1 EXTERNAL: EGS(G GROHI M 5 5 H ST LOBBY WARD 1 1 LPDR 1 1 NRC PDR 1, 1 NSIC HARRIS' 1 NSlC MAYS> G 1 TOTAL NUMBER OF COPIES REQUIRED: LTTR 45 ENCL 44

NRC Form 355 US. NUCLEAR REOULATORY COMMISSION (943) APPROVEO OMB NO. 31500101 EXPIRES: 5/$ 1/SS LICENSEE EVENT REPORT (LER)

DOCKET NUMBER (2) PA E FACILITY NAME (I) 0 5 0 0 0 2 5 7 of P 4 Design Basis Reconstitution Discovers Residual Heat Removal Recirculation ed to Assure Ade uate Flow for Each Pum EVENT DATE(51 LER NUMBER IS) REPORT DATE (7I OTHER FACILITIES INVOLVED (5)

SEQUENT!AL IIACILITYNAMES DOCKET NUMBER(S)

MONTH DAY YEAR YEAR Pg NUMSEII .r~oP~ NUMBER MONTH OAY YEAR 0 5 0 0 0 2 5 ]

00 11 23 87 0 5 0 0 0 OPERATINO THIS REPORT IS SUBMITTED PURSUANT TO THE REOUIREMENTS OF 10 CFR (Ir IChecl one or more of the follovp'nPI III)

MODE ISI 20.402 (5 I 20A05(cl 50 73(s)(2) liv) 73.7101) l(v)ill POWER 20.405(s) III II) 50.35(c) (1) 50,73(s) (2)(vl 7$ .71(c)

LEVEL p p 20A05(s) Ill(vl (1)(S I 50.35(cl(2) 50.73(sl)2)(vill OTHER ISpeclfy In Asrtrect (re/ow end In Test. NRC Form 20,e05 ( ~ Ill ) (IIII 50.73(s) (2) III 50.73(s I (2 (Al 50.73(sl(21(vill)(SI JEEAI 20A05(el()I (iv) 50.73(s)(21(N) 20AOS(s) 50.7 3 (s) (2)(l 5 I 50.73( ~ ) (2)(el LICENSEE CONTACT FOR THIS LER (12)

NAME TFLEPHONE NUMBER AREA CODE D Hart Licensin En ineer 30 24 - 6559 COMPLETE ONE LINE FOR FACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

MANUFAC. EPORTABLE COMPONENT MANUFAC EPORTABL CAUSE SYSTEM COMPONENT TO NPRDS SYSTEM TURER TURER TO NPRDS s s'-'AUSE ys ego.

!,r)~C.

SUPPLEMENTAL REPORT EXPECTED (Iel MONTH, DAY YEAR EXPECTED SUBMISSION DATE HEI YES IIIyer, complete EXPECTED SUBMISSION DATEI NO ABSTRACT (Limit to Ie00 tpscsL I.e., sppmslmerely fifteen tl tpeepece typervrftmrr Ence) 115)

On October 27, 1987, while Unit 3 and Unit 4 were in mode 5 (cold shutdown), it was determined that a design discrepancy existed in the residual heat removal (RHR) system. During the design basis reconstitution of the RHR system, it was discovered that the existing minimum recircula-tion design configuration was potentially inadequate . The present RHR system design has two (2) RHR pumps discharging flow through a shared mini flow recirculation line . If the performance of one RHR pump is slightly better than the other, then it discharge pressure to deadhead the RHR pump with the lower discharge is possible for the RHR pump with the higher pressure if the reactor coolant'system (RCS) pressure is above RHR pump shutoff head. The existing plant emergency operating procedures require the operator to terminate RHR if the RCS pressure is above RHR pump shutoff head . However, it cannot be assured based on existing plant procedures that this step will be accomplished prior to the potential damage of a RHR pump. This potential failure (by design) coupled with a single failure of the other operating RHR train could result in complete loss of RHR pump capability. The recirculation flow path for the RHR pumps was an original design of the plant. Plant modifications will be completed on both units to correct this discrepancy. Appropriate procedure changes and training has been completed.

87ff93 87 g ggo0254 ADQCK 5000250 pDR PDR .

NRC Forn.

(903)

NRC Form 3ddA U.S NUCLEAR REOULATORY COMMISSION tt'OCKET IBWI LICENSEE EVENT REPORT ILERI TEXT CONTINUATION APPROVED OMB NO. 3ISO~ICd EXPIRES: B/31/BS

~ FACILITY NAME NUMBER ISI I.ER NUMBER IS) PACE I3I YEAR go)'EOVENTIAL :P/R'EVISION NVMSER  :<PN NVM Ell Turkey Point Uni t 3 o o o o o 2 5087 030 0 02 00 4 TEXT fit moto t/Metis todo/tod, oto oddi5ono/ H/IC %%dtm 3SBA't/ I I 3 I EVENT:

On October 27, 1987, while Unit 3 and Unit 4 were in mode 5 (cold shutdown), it was determined that a design discrepancy existed in the residual heat removal (RHR) system. During the design basis reconstitution of the RHR system, it was discovered that the existing minimum recircula-tion design configuration was potentially inadequate . The present RHR system design has two (2) RHR pumps discharging flow through a shared mini flow recirculation line . If the performance of one RHR pump is slightly better than the other, then it is possible for the RHR pump with a higher discharge pressure to deadhead the RHR pump with lower discharge pressure the reactor coolant system (RCS) pressure is above RHR pump shutoff head.

if Events where the RHR system may be required to operate on recirculation include spurious safety injection (SI), small break LOCA, or any postulated scenario where the SI signal was actuated and the RCS pressure remains above RHR pump shutoff pressure. The existing plant emergency operating procedures require the operator to terminate RHR if the RCS pressure is above RHR pump shutoff head, however, it cannot be assured based on existing plant procedures that this step will be accomplished prior to potentially damaging a RHR pump.

Our NSSS vendor evaluated this concern and determined based on present conditions that a RHR pump could operate deadheaded for 10.4 minutes without pump degradation or damage. This potential failure (by design) coupled with a single failure of the other operating RHR train could result in complete loss of RHR pump capability.

CAUSE OF EVENT:

The recirculation flow path for the RHR pumps was an original design of the plant.

ANALYSIS OF EVENT:

This condition was discovered while both Units were in cold shutdown for a maintenance outage . At this time the concern identified for the RHR pumps is not applicable . Permanent changes to install individual minimum recirculation flow paths for each of the RHR pumps will be completed on each unit before the unit enters mode 4. The modified recirculation system will allow operation of both pumps for at least 30 minutes while operating in a closed loop without cooling the recirculating flow (i.e., with the pumps operating only on miniflow), thus meeting the requirements of the current emergency operating procedures. In addition, the minimum flow recirculation lines will be designed such that the in)ection flow assumed in the current Turkey Point Units 3 and 4 LOCA analysis will not be reduced.

For the new design of the recirculation line, maximum RHR pump flows for one and two pump cases during both in)ection and recirculation operation were calculated . These flows were used to generate RHR pump KW to be used in the emergency diesel generator (EDG) loading evaluation (LER 250-85-40).

The increased flows resulted in increased RHR pump KW values . However, when these values were used in the EDG loading evaluation, the results indicated NRC FORM SddA I9 B3I

NRC Forrrr 3ddA VS. NUCLEAR REOVLATORY COMMISSION IS'83)

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OM8 NO, 3150MI Oi EXPIRES: 8/31/88

~ FACILITY HAMf Il) OOCKET NVMEER 13) LER NVMBER Id) PACE 13)

YEAR SEOVENTIAL IId V Id IO N NVMddR ~<we NVMddR p p p5P 87 p 3 0F p 4 that the worst case remains bounded by the previous EDG loading evaluation.

Therefore, the EDGs can still be safely operated during design basis events as described in the final safety analysis report (FSAR) and the EDG loading evaluation.

The as discovered design condition required one RHR pump to be stopped to prevent potential pump failure due to deadhead operation and associated overheating. Existing plant emergency procedures required operators to secure RHR pumps to prevent overheating of a pump. The condition that was determined to be not acceptable was that it could not be assured the operator would be at the appropriate procedural step in less than the 10.4 minutes. However, even though this condition was found to exist, the health and safety of the public was not affected and it is judged that this condition was not a significant safety hazard based on the following:

1) The condition as discovered was not a problem in the design basis or large break loss-of-coolant accident (LOCA) because both RHR pumps, would be delivering flow to the reactor in less than 10.4 minutes.

if running,

2) The condition as discovered is not a problem if any plant single failure would have resulted in one RHR pump not operating on demand .

3) Plant historical data indicates that for actual inadvertent SI signal actuation, the operator typically had reached the .procedura3."step to secure RHR in less than 10.4 minutes. If the operator had not and one RHR pump had been damaged due to overheating, sufficient time would be available to restore one RHR pump prior to proceeding to a cold shutdown condition.

4) For a small break LOCA or other SI system actuation for which the RCS pressure remains above the RHR pump shutoff head pressure, the RHR pumps are not required on the short term.

CORRECTIVE ACTIONS:

Plant change/modifications (PC/Ms) have been developed for each unit to install independent recirculation line for each RHR pump. The lines will be designed to allow operation of the RHR pumps for at least 30 minutes without affecting pump operability or current FSAR accident analysis. These PC/Ms will be installed prior to each unit entering mode 4.

2) The following plant procedures have been revised to reflect implementation of the PC/Ms:

a) 3(4)-EOP-E-0 Reactor Trip or Safety Injection b) 3(4)-EOP-ES-1.4 Transfer to Hot Leg Recirculation c) 3(4)-GOP-503 Cold Shutdown to Hot Standby d) 3(4)-OSP-050.2 Residual Heat Removal Pump Inservice Test e) 3(4)-OP-050 Residual Heat Removal System f) 0-ADM-205 Administrative Control of Valves, Locks, and Switches NRC FORM dddA

$ 4)3)

r NRC Form 355A U.S. NUCLEAR REOVLATORY COMMISSION (I'OCKET (94'J)

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROV EO OMS NO. 3(50M(04 EXPIRES: 8/31/88 FACILITY NAME NUMBER (2) LER NUMBER (8) PACE (3)

YEAR Pg: 55QUCNTIAL gW REVr5ION NUM 5R Orerr NUM 5R Turke Point Unit 3 o s o o o 2 5 8 7 0 3 0 4 DF o4 TEXT /// more 4/reoe /4 rtqw'nkvd, ore eddrr/orro///RC Form 3584'4/ (17)

3) Training Brief number 213 was issued on November 12, 1987, describing the condition, the proposed modifications, and the 'procedures affected by the modifications.

4) As a part of the confirmatory'order associated with EA 86-20 issued August 12, 1986, Turkey Point is currently performing a Selected Safety System review to assure that the Turkey Point Plant as built condition is consistent with the current licensing basis and has the capability within the systems to mitigate any of the design basis accidents and/I'or shutdown the plant.

ADDITIONAL DETAILS:

The RHR pumps are single stage centrifugal pumps manufactured by Ingersoll-Rand.

Similar Occurrences: None NIIC CORM 355A (94)3)

P. Ox 14000, JUNO BEACH, FL 33408.0420 rIPyEMBER 2 o 1987

. L-87-483 IO CFR 50 73 U. S.'Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gent I emen:

Re: Turkey Point Unit 3 Docket No. 50-250 Reportable Event: 87-30 Date of Event: October 27, I 987 Design Basis Reconstitution Discovers Residual Heat Removal Recirculation Line Not Desi ned to Assure Ade uate Flow for Each Pum The attached Licensee Event Report is being submitted pursuant to the requirements of IO CFR 50.73 to provide notification of the subject event.

Very truly yours, Executive Vice President COW/SDF/gp Attachment cc: Dr. J. Nelson Grace, Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, Turkey Point Plant SDF I /063/ I an FPL Group company