ML20043G354

From kanterella
Jump to navigation Jump to search
LER 90-008-00:on 900513,automatic Scram Resulting from Load Rejection Occurred While at Full Power,Resulting in Trip of Generator Field Breakers.Caused by Fault on Offsite 345 Kv Transmission sys.Loss-of-field Relay replaced.W/900612 Ltr
ML20043G354
Person / Time
Site: Pilgrim
Issue date: 06/12/1990
From: Bird R, Ellis D
BOSTON EDISON CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
90-074, LER-90-008, NUDOCS 9006200183
Download: ML20043G354 (9)
v • d • e


Text

.

L ;;

r 3 4 t.,:  ;

10 CFR 50.73 c

g, m k Pilgrim Nuclear Power station

' Rocky Hill Road '

Plymouth, Massachusetts 02360 June 12, 1990 bl hP G. Bird BECo Ltr.90-074 Senior Vice President - Nuclear U.S. Nuclear Regulatory Commission ,

Attn: Document Cont ol Desk '

Hashington,'D.C. 20555 a

Docket No. 50-293 i License No. DPR-35 l

Dear Sir:

i The enclosed-Licensee Event Report (LER) 90-006-00, " Automatic Scram Resulting {

From Load Rejection at Full Power", is submitted in accordance with 10 CFR a

Part 50.73.  !

Please 'do' not hesitate to contact me if there are any questions regarding this report. j) l Ya h 1

DHE/bal l

Enclosure:

LER 90-008 q cc:' Mr. Thomas T. Martin

% Regional' Administrator, Region I-U.S. Nuclear Regulatory Commission 4 475 Allendale Rd. h King of-Prussia,'PA4 19406  :

1 Sr. NRC Resident-Inspector --Filgrim Station i Standard BECo LER ' Distribution p l u <

g y 090 p4 A 9006200163 900612 6' PDR S

ADOCK 05000293 PDC'- li l l

r.q->

  • 1 3

senC poem'ans u s. NUCLE .n c.atutATORY ComeusssiON

# ' ' , APf ROvsp Oms t;O steocios . )

LICENSEE EVENT REPORT (LER) "aES "

  • l I

P1CILITY NAM nl DOCILEY NUMOtR til PAGE (3i Pilgrim Nuclear Power Station 016101010121 911 1 loFl 018 ft1LE4' Automatic Scram Resulting From Load Rejection At Full Power EVENT DAf f (El LtR NuMetR tt) REPORT DAf t (7) OTHf R F ACILITits INv0Lvt0 let '

MONTH DAY YiAR YtAR ** ['18,$ MONT H OAV YiAR 'aciLITv hauts DOcktT NUMetHis) u *d L"'

N/A 01510l010; ; l i

~ -

0l5 1l3 9 0 9 l0 0l d 8 0l0 ol 6 1l2 9l0 015101010IIl

,,,,y,,, tHis REPORY is susMartto PuRouANT TO THE REOuiReMENT: Os to Cp R 6. (ca c* ea, e, me,e or ra, toueweer ut:

"oot m N :o . cal.: ro easi.i oo.rsi.n:H=i 73.viini X

yno, to aosaimm so. maim so. anan.> tuu.i il 001 a 4asi.imm so. maim so T =n:Hn

_ gHgsgg,e, gs y _

20 a'wH*i susanam _

sosanan.mnas mi i  ?' '

to doelaH1Hi.i 60.731 112 H60 so.T3taH2Hviinnel

2. do6 ten 1Hul So,736en2Hild to.73laH2Hal LICENSEE CONT ACT POh THl3 LIR (12)

NLMt TELEPHONE NUMBER ARE A CODE Douglas W. Ellis - Senior Compliance Engineer siin 1 R 7 IA I 71 - l 81 11610 COMPLETE ONE LINE FOR ( ACh COMPONENT F AlLumt DitCRISED IN THIS REPORT H3)

REPO TA La F C-CAust $YSTEM COMPONENT "(hAC. mg mm mm M REP 5T a 3 X QL l l912 Will 210 Y l i I I I I I I I I I I I I s 1 I I I I I I

$UPPLEMtNT AL REPORT EXPECitD n41 MONTH DAY YEAR 4t$ lifVes. teelpme LKPICTIO StignH3stON CA tti NO l l l

..mACT -,se. .,a. ,.. -,,s - - n.,

t On May 13,1990 at 1603 hours0.0186 days <br />0.445 hours <br />0.00265 weeks <br />6.099415e-4 months <br />, an automatic scram resulting from a load rejection occurred while at 100 percent reactor power. The load rejection included a trip of the Generator Field Breaker, actuation of the Turbine mechanical hydraulic control speed governor, closure of the four Turbine Control Valves and opening of the three Bypass Valves, and the brief actuation of the Main Steam / Target Rock two-stage .

relief valves at approximately 1100 psig (low end of the 1115 psig setpoint range including tolerance).

The load rejection was caused by a momentary fault on the offsite 345 KV transmission system. The Generator's loss-of-field relay (240) detected the fault' and immediately tripped the Generator without an expected (inherent) 15 cycle time delay because one of it's components, the telephone relay ('X') coil, was defective. The relay (240) was last calibrated and functionally tested on October 26, 1989. At that time, the operation of the ('X') coil was tested in accordance with the vendor manual. The relay's- time delay was built-in and not adjustable, and was not required to be timed. The. relay was installed during plant construction (c. 1972). The cause for the open coil is being investigated but is believed to be random or age related failure. The relay is the only one of its type (Westinghouse type KLF-1) installed at Pilgrim Station and was replaced with another KLF-1 relay having an adjustable time delay. The (240) relay's calibration sheet was revised to. include a calibration of the adjustable time delay.

This event occurred with the reactor mode selector switch in the RUN position. The Reactor Vessel (RV) pressure was initially at 1035 psig with RV water temperature at 548 degrees Fahrenheit. This report is submitted in accordance with 10 CFR 50 73(a)(2)(iv) and this event posed no threat to the public health and safety.

jag, ~= =.

t sene se,W sooA . ' u s NucttAa KE2vtaiony comutssiom

. LICENSEE EVENT REPORT (LER) TEXT CONTINUATION m aoveoovs=o mo-om ,

, EXPIRES: 9?) BOB >

f ActLITY hAast He DOCKE T NVuSE R (21 ggn gvuggR 461 PAGE(3) v5^a " OtW. M Pilgrim Nucicar Power Station YEXT (# mene spece as meuwed ese 9Mesew/ NAC Feen JetWes 4th 0 l5 l0 j o lo 12 l 9l3 91 0

'0l0l 8 -

010 012 oF 0l8 EVENT DESCRIPTION i

On May 13, 1990 at 1603 hours0.0186 days <br />0.445 hours <br />0.00265 weeks <br />6.099415e-4 months <br />, an unplanned automatic Reactor Protection System (RPS) scram signal and reactor scram occurred while at 100 percent reactor power.

The scram signal occurred as a result of a load rejection that included a trip of the Turbine-Generator. 4

'The t' rip of the Generator lockout' relay (286-1) resulted in the following designed' responses:

  • Automatic opening of the Generator Field Breaker (41M). I l
  • Automatic opening of the 345 KV switchyard air circuit breakers ACB-104 (352-4) and ACB-105 (352-5).
  • Automatic transfer-of the source of 4160 VAC power for the Auxiliary Power Distribution System (APDS) from the Unit Auxiliary Transformer (UAT) to the l Startup Transformer (SUT).  ;
  • Automatic trip of the Turbine master trip solenoid (MTS-1) that resulted in the closing of the Turbine Stop Valves and Combined Intermediate Valves, and the-trip of'the Turbine lockout relay (286-2).

Concurrently, the Generator trip resulted in a momentary increase in Turbine speed

.that was caused by the sudden mismatch in Generator load (zero percent) and Turbine '

power (100 percent). The rapid acceleration actuated the mechanical speed governor

'in the Mechanical Hydraulic Control (MHC) portion of the Turbine Control System, and an adjustment in the mechanical linkage connected to the acceleration relay. 1 The actuation of the acceleration relay resulted in the following designed responses:

1

  • ' Fast closure of-the 4 (four) Turbine Control _ Valves and the subsequent opening of the 3 (three) Turbine Bypass Valves.

,-

L .

l . The Main Steam /RV pressure increased as a result of the fast closure of the Turbine  !

Control Valves. The opening of the Bypass Valves, controlled by the pressure Jregulator, mitigates the pressure transient; however, the pressure increased due to .

the 25 percent total bypass capacity of the Bypass Valves. The pressure, initially ~

.at 1035.psig, increased to approximately 1100 psig and resulted in the automatic actuation of the Main' Steam /Terget Rock two-stage relief valves RV-203-3A (s/n

,, 1040), RV-203-3B (s/n 104P.), and RV-203-3C (s/n 1046). Relief valve RV-203-3D (s/n <

L 1025)'did not actuate. l l

i l

NMC PORM 366A -

i@k ' , _,

. .. - -- - --- - -- - - . - - - _ - . + - . -- - . u unc poesn ageA '

u 3 NUCLEAR E.tivtATORY COMusseeced LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Anaovfo ove =o mo-oio.

+

s EXPinf 8: l'31/W F AC4LITY 884488 tu DOCKli NUllSER 83) Llh NUMDIR tel PA06 t3l Pilgrim Nuclear Power Station "'" '

"M' -

Ow*aU TEXT M mese ausse 4 #seased, see anWomist NNC 7em Wel117) o 15 l o l o l o b 1913 91 0 ol d s -

nin ! 9 3 or 01 8 '

As expected, the RV water level decreased in response to the scram because of ',

-shrink-(i.e., decrease in the void fraction in the RV"wr.ter). The RV water level eventually decreased to approximately -10 inches (narrow range level).. The decrease in RV water level, to less than the low RV water level setpoint '

(calibrated att approximately +12 inches) resulted in automatic actuations of the Primary Containment Isolation Control System (PCIf,) and Reactor Building Isolation Control System (RBIS).

The PCIS actuation resulted in the following lesigned responses:

(PCS)/ Reactor Hater Sample isolation valves A0-220-44'and -45.

  • Automatic closing.of'the inboard and outboard PCS' Group 2 (two) isolation valves that were open. _g
  • The PCS Group 3 (three)/ Residual Heat Removal System . isolation valves, iin' the , '

closed position,: remained closed. ' 1 e- Automatic. closing of the inboard and outboard PCS Group 6 (six)/ Reactor Hater 4

- Cleanup (RHCU) System isolation valves and a temporary interruption in RHCU System operation, i m The RBIS actuation resulted in the automatic closing of the Reactor i Building / Secondary Containment System (SCS) supply and-exhaust ventilation dampers.

~(Trains. 'A' and ~ 'B'), and the automatic start of Trains 'A' and 'B' of the

.SCS/ Standby Gas-Treatment System (SGTS).

? Initial Control Room operator response was orderly and included the.following. The-reactor mode selector switch'was-moved from the RUN position to the REFUEL position

and the ' reactor. feedpumps were' tripped in accordance with procedure 2.1.6, " Reactor Scram"'.-Emergency Operating Procedure (E0P)-01, "RPV Control", was.. initiated when the:RV water: level decreased to less than +9 inches (narrow' range) and was terminated when the RV water level increased to greater'than +9 inches. Meanwhile, .

the High Pressure Coolant Injection System (HPCIS) was manually started in the full-flow test mode as.a. precautionary pressure control measure'and in accordance with n the guidance provided in E0P-01. The Residual Heat Removal System (RHRS) loop 'A' l< (pump 'A'),was placed in the Suppression Pool Cooling (SPC) mode in accordance with.

procedure 2.2.86, " Residual Heat Removal", because of the heat addition from the steam discharged into the Suppression Pool via the discharge piping from the-relief valvesiand HPCIS turbine. The Suppression Pool temperature was logged in '

+

accordance with procedure 2.1.19, " Suppression Chamber Temperatures". Procedures 2.1.5. Attachment 1, " Shutdown /Cooldown Checklist", and 2.1.7 Attachment 1, "RPV Temperature'and Pressure Checklist", were initiated. The PCIS circuitry was reset L1 and the RHCU System was returned to service. The RBIS circuitry was reset, the

'SGTS was returned to normal standby service and the Reactor Building ventilation system was returned to normal service.

p c T

Q.

'* senc F.rh setA U.S. NUCLEAR KEQULATORY COMMfSSION  !

"" ~

UCENSEE EVENT REPORT (LER) TEXT CONTINUATION - - maovio ove no mo-m4 ii g *

(XPlRES: $/3USil

(.,, 9.C4LITY NAME (1) DOCKET fdUMBER (U LIR NUMetR (6) PA08 (3) ~

Pilgrim Nuclear P e r Station " ' "

" E U "'

~ O l ,

0 [5 l0 l0 l0 l2 l 9l3 gj o --

nlglg -

gl 0 ' Ol 4 oF Ol8 7 Iw a ., . o .~ - m w., use,wn L/ The NRC Operations-Center was notified of the event in accordance with-10 CFR 50.72 b on May 13,.1990 at.1734 hours0.0201 days <br />0.482 hours <br />0.00287 weeks <br />6.59787e-4 months <br />. Failure and Malfunction Report 90-159 was written to document the event. A post trip review of the eventwas performed in accordance with procedure 1.3.37, " Post Trip Reviews".

BACKGROUND i

Prior to'the event, steady state operating conditions existed and included.the '

F following. The RV water level was +29 inches (narrow range) and the Feedwater E System was being controlled in the automatic three element control mode. The RV i -pressure .was 1035 psig and was being controlled via the Electric Pressure <

-Regulator. The Turbine speed was approximately 1800 rpm and Turbine first stage ,

pressure was approximately 730 psig. The Recirculation System pumps were being controlled in the local-manual control mode. The Condensate System and'Feedwater System pumps were-all in service. Except for Bus A6, the APDS was energized by-the

.UAT. The preferred source of offsite power, 345 KV transmission lines 342 and 355, were in service. The 345 KV switchyard air circuit breakers ACB-102 (352-2),

JACB-103'(352-3), AC8-104 (352-4), and ACB-105 (352-5) were in service. T,te 23 KV ,

backup ~ source of offsite power was in service. Thunderstorm activity wa*, reported i f in the' region, j

Just prior to the event on May 13 1990 at 1603 hcurs,~ numerous Main Control Room

. alarms ' occurred in a short interval of time and included Panel C-3R, " Unit #1 Generator: Low Excitation"..and " Unit #1 Generatot nelay Trip". After the event,.

g' #

the regional power agency (REMVEC) reported thir the Pilgrim Station; trip occurred-at the same time that'a 345 KV transmission _ syn .m line (322) fault occurred.

Transmission line 322 is related to the nearby. canal Station and, via transmission, line' 342, to Pilgrim ~ Station. A static wire (lightning' protection) for line 322 3 fell onto its-conductors and caused a fault. The fault.resulted in,an electrical

' disturbance that was detected and isolated by protective devices on line 322. The

%.' momentary disturbance was sensed by the transmissi.on system prior to the isolation L

offline 322. .The disturbance did not result in a trip of the Canal Station i

. generators but did result in the trip of Pilgrim Station, t g The cause for the Pilgrim Station trip was the momentary fault on line 322 (transmitted via line 342)'and the absence of the telephone relay time delay ('X'

, coll) function' of-the loss-of-field relay (240) at(Pilgrim Station. The

disturbance actuated the impedance (Z) and directional (D) and voltage (V) units of

, *. .the relay.(240) that is part of the Generator's protective circuitry. -.The .'X' coi?

is' normally energized and is designed to change state (drop.out) 15 cycles (i.e.,

0.25 second) after it de-energizes due to a disturbance (Z and D and V). The time delay is provided to prevent a-trip of the Generator due to a momentary fault on ,

the 345'KV transmission system. The failure of the 'X' coil resulted in.no time delay for relay 240 and consequently, the Pilgrim Station Generator Lockout Relay  ;

(286-1) actuated at the time of the momentary fault. The loss-of-field relay (240) was manufactured by Hestinghouse Electric Corporation, type KLF-1, style 2928333A10, 125 VDC, 5 Amps, 69 volts, 60 Hertz.

i NRC FORM 34ea .

' M] ~ -

. ~ - . . -- .- . - - - - . . . -

1 EftC form 3044 V 5 NUCLEAR RE2VL&f 0AY COMul4Ss08s j LICENSEE EVENT REPORT (LER) TEXT CONTINUATION A m ovieousNo mo-om txmes ows:

9dCILITY fsAf tt (H DOC 8LET NUMBER Gl LE R NUMSER (61 PA04 (31 Pilgrim Nuclear Power Station " ' "

" W a" '

WaN

_ 0 15l0101012 1913 91 0 ol d s -

olo ol 5 0F 0 l8 The non safety-related loss-of-field relay (240) is functionally tested in accordance with procedure 3.M.3-39 Attachment 1 "Tufbine/ Generator Lockout Test and Associated Annunciator Verification" and calibrated in accordance with j Attachment 2 " Calibration of Turbine / Generator Relays". The functional test includes a check (operation) of the relay's alarm functions and trip functions, i The relay was most recently functionally tested and calibrated with satisfactory '

.results while shutdown on October 26, 1989. The operation of the telephone relay-was tested in accordance with the vendor manual (V-0250) that does not include or specify a calibration (timing) of the relay's inherent time delay. The time delay  ;

of the relay was built-in and not adjustable, and therefore was not checked as part  !

of the calibration. The relay (240) was installed during original plant I construction (c. 1972). A search of the Nuclear Plant Reliability Data System (NPRDS) revealed no other failures of a KLF-1 relay. Therefore, the cause for the open ('X') coil of the telephone relay is believed to be random or. age related failure.. The cause for the open coil is being investigated'and an update to this t report will be submitted if the investigation reveals significant newlinformation.  !

l CORRECTIVE ACTION The unit returned to commercial service on May 15, 1990 at 1030 hours0.0119 days <br />0.286 hours <br />0.0017 weeks <br />3.91915e-4 months <br />.

The loss-of-field relay (240) was replaced on May 16, 1990 with another relay (KLF-1) having an adjustable time delay. Interim measures taken until the relay-was replaced consisted of modifying the relay via a Temporary Modification (TM 90-11) that was supported by a Safety Evaluation (SE 2470). Essentially,'the relay's-(240) trip function was disabled without affecting the relay's alarm functions.

The relay's.(240) calibration sheet has been revised to include a calibration of the new' relay's adjustable time delay. The loss-of-field relay (240) is the only.

.one of'its- type (KLF-1) installed at Pilgrim Station.

An Engineering Service Request (ESR 90-327) has been written to determine if any other relay (s), included in procedure 3.M.3-39 or 3.M.3-40 (Relay House Testing),

is equipped with a built-in ' time delay that may need additional testing.

SAFETY CONSEOUENCES This event. posed no threat to the public health and safety.

The Technical Specification 2.2.B limiting safety system setting for the Main Steam System / Pressure Relief System (PRS) relief valves is 1095 to 1115 psig with a tolerance of +/- 11 psi. The setpoint of the relief valves is 1115 psig.

Therefore, the setpoint range of the relief valves including tolerance is 1104 psig to 1126 psig. During the event, the highest RV/ Main Steam System pressure that occurred was approximately 1100 psig. The two-stage Target Rock relief valves (RV-203-3A/3B/3C/30) are installed in Main Steam pipelines 'A' (RV-203-3A and -3D) and 'O' (RV-203-3B and -3C).

3...

gC ,o.

, t penC form 304A , U.S NUCLEAR KE0VLATO2Y COMMIS8 toss

" - . LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Ae*Rovio oMe No mo-o*

,. EXP1RES: S/3USB DOCKET NUMBER 121 eA04 (3)

  1. pCILify NAME tu 48R NUMSER 46)

./

naa "up  :

5?.?:

i Pilgrim Nuclear Power Station 0 l5 l0 l 0 l 0 l9 lql 3 gj o -

n ln l R -

ol 0 n l6 oF glp j Review of RV pressure and. lain Steam pressure data indicates that the pressure '

increased rapidly (i.e., 38 psig in 0.5 second) until RV-203-3A (s/n 1040).and RV-203-3B (s/n 1048) actuated. After RV-203-3A and -3B actuated, the pressure continued to increase, but in a,less rapid manner, for another 0.5 second until RV-203-3C (s/n 1046) actuated. -After RV-203-3C actuated, and with RV-203-3A and

-3B still actuated, pressure remained relatively constant for another 1.2 seconds and RV-203-3C then closed. After RV-203-3C closed, and with RV-203-3A and -3B-still actuated, pressure decreased (i.e., 30 psig in 2 secondt) and RV-203-3A and n-3B then closed. The duration of the pressure transient, fron initial to maximum to initial (all relief valves closed), was approximately 5 (five) seconds. Relief valve RV-203-30 (s/n 1025) did not actuate for the pressure relief function. The t relief valve (s/n-1025) was.most recently tested at an offsite t u t facility on August 2,1988 with certified as-left popping pressures of 1114 ps!g +/- 1 (one) psig. The load rejection with. bypass experienced during this event is bounded by the transient analysis described in the Updated Final Safety Analysis Report section 14.4.3, " Generator Load Rejection Hithout Bypass". The actuation (opening) of some or all of the _4 (four) two-stage relief valves (RV-203-3A/3B/3C/30) is an expected response to a load rejection with bypass at 100 percent power.

The. Technical Specification 2.2.C limiting safety system setting for the Main Steam / PRS safety valves (RV-203-4A and -48) is 1240 +/- 13 psi. . During the event, the' highest RV' pressure that occurred (1100 psig) was approximately 140 psig less than the. safety valves' setpoint (1240 psig).

The HPCIS was . manually started in the full flow test code in accordance with procedure 2.2.21 [HPCIS] section 7.4.2'and in accordance with the guidance of "

  1. E0P-01. In the full flow test mode, steam from Main Steam pipeline 'D' is supplied to the HPCIS turbine and is exhausted to the Suppression Pool, and water from the ,

Condensate Storage Tanks (CST) is supplied to the HPCIS pump and is returned to the CST via'.the. test return line. The injection function of the HPCIS was not used.

The scram signal was the designed response to the Turbine Control Valves Fast Closure (i.e., opening of PS-37/38/39/40). The closure was the expected designed
response to a load -rejection with the Turbine first stage pressure at approximately 730 psig, i.e. greater'than the scram bypass setpoint (calibrated at 108 psig +/- 3 psig)Lcorresponding to 25 percent of the normal first' stage pressure).

The decrease in the RV water level was the expected response to the scram and accompanying shrink in the RV water. The PCIS and RBIS actuations were the expected-designed responses to a low RV water level condition, i.e. +12 inches (narrow range).

The Technical Specification 2.1.I limiting safety system setting for actuation of the Core Standby Cooling Systems (CSCS) is -49 inches. During the event, the lowest RV water level that occurred (-10 inches) was approximately 36 inches above the CSCS setpoint (calibrated at approximately -46 inches). In addition, the level

(-10 inches) was approximately 117.5 inches above the level that corresponds to the top of the active fuel zone (-127.5 inches).

.'i$ 0""*"^

T' v

'. =ac e.!= ansa u s. avetsan s:sovutorr couuission

$m . LICENSEE EVENT REPORT (LER) TEXT CONTINUATION uemoveo rws no mo-em emmis: sewse p kt #Acmtr samt m DOCKET NUMBER W LER NUMetR W PA06 m '

'k' vsan se p,4 L a,eyy,a .l h Pilgrim-Nuclear Power Station 0l5l0l0l0l,,lgl3 glg -

0l0 l8 -

Olt Ol'7 0F 0 l 8. -

.f . isneu m au e-o.i . m unc m .nm \

[& 1 This report'is submitted in_accordance with 10 CFR 50.73 (a)(2)(iv) because the kPS l d<i 3 .y <

was actuated.

H k SIMILARITY TO PREVIOUS EVENTS 1

'}. .

'{ - A review was conducted of Pilgrim Station Licensee Event Reports. (LERs) submitted 1 since January 1984. The review focused on LERs submitted in accordance with l' 10 CFR 50.73(a)(2)(iv) that involved the loss-of-field relay, a load rejection or i similar scram. The review identified similar events reported in LERs- l I 50-293/85-025-00 and 89-026-00.

For LER 85-025-00, an automatic scram occurred on September 1,1985 at 0521 hours0.00603 days <br />0.145 hours <br />8.614418e-4 weeks <br />1.982405e-4 months <br />

+ Lwhile at 32 percent reactor power. At the time of the event, the Main. Condenser twas being backwashed and a live washdown of the 345 KV switchyard insulators was being performed to reduce arcir;g due to salt from a heavy ocean storm. A 345 KV phase.'B' insulator, located on the Main Transformer side of the switchyard ACB-104, disintegrated:and resulted in a load rejection. The cause for the scram was high RV pressure that resulted from the load rejection.- The cause for the

~

event was due to thesforces of nature (i.e. high winds'and salt air). Please note 1 that the event occurred while at 32' percent reactor power. At-that power level, Lthe~ Turbine first stage pressure was approximately-200 psig. An RPS scram signal due to a Turbine Control-Valves Fast Closure (i.e., PS-37/38/39/40) or Turbine Stop l Valves. closure (i.e., less than 90 percent open) would have occurred if the Turbine. ,'

first-stage pressure was greater than approximately 280 psig (i.e., scram bypass setpoint for 45. percent of ~the normal first stage pressure). The scram bypass setpoint was changed from 280 psig to 108 psig (+/- 3 psig) via a modification

. (PDC 87-48) during Refueling Outage number 7.

For LER 89-026-00,- an automatic .RPS scram signal and . scram occui red. on -

August-30,_1989_at 1917-hours.while at 65 percent reactor. power. _The cause for_the ,

scram signal was'high RV pressure (ultimately 1069 psig) that occurred as a result of an automatic Turbine: runback. The runback included the automatic adjustment of the Turbine Control Valves and sequential opening of'the Turbine Bypass Valves.

The runback occurred as a result of the failure of the primary winding of the Main

+

Generator'24 KV phase "A' potential transformer and a Generator-Voltage Balance-Relay (260) wiring' error that:affected the transfer function'of the Generator's Voltage.Regelator. The wiring error was due to a drawing error. 'The error was not prev nusly detected because the surveillance test procedure (3.M.3-39) used to ifunctionaily. test the relay (260), although demonstrating the voltage balance relay

" function and alarm functions, did not include a step (s) to identify the auxiliary T  ; relay U.60X1 or 260X2) that actuates the same alarm (Panel C-3R, " Generator

. Potential Fuse Blown") during the test (s).

NRC FORM 34GA -

r NO - .,

s,. , c. .

-( .

M

sene e4. sesA - u s Nuctst.n s.ssutAToRv e=== 3 - ,

., .OCENSEE EVEN'T REPORT (LER) TEXT CONTINUATION maovio ove no aisoaion

{ . g; ,

EXPlRi$ $/31/98 f,. 5ACILITY NAME 111 DOCKET NUMetR (27 LtR NUMetR (6) PA06 (3)

TI'h v mIm NuwaYa j LP11 grim Nuclear Power Station -

o ls'j o 1010 62l913 01 d 8 Ol 8 OF 0 91 0 Ol 0 TEXT M mes spese a neweef, was .eestemaer WC form JeEA W lth ;

ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) CODES:

The EIIS codes for this report are as follows:- ,

COMPONENTS CODES Relay, Voltage and Power Directional (240)- 92  ;

Valve, Relief' RV y ,

SYSTEMS Containment Isolation Control System (PCISsRBIS) JH

= Engineered Safety Features-Actuation System JE (PCIS, RBIS, RPS)~

Hain Generator Output System EL Main' Steam System SB Hain Turbine System TA

  • Plant Protection; System (PRS) JC  !

1: _ Reactor Hater Cleanup (RWCU) System CE-  !

Standby Gas-Treatment System (SGTS) 'BH-Switchyard System (345 KV)- 'FK-

[

4

,k

'. NRC FORM 366A

. JCI . _ __ _ . . _ - - - _ . _ - --

Retrieved from "https://kanterella.com/w/index.php?title=ML20043G354&oldid=2562334"