ML20045G728

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LER 93-002-00:on 930608,inadvertent Arw Actuation Sys & RPS Actuations Experienced During Performance of Awf Sys Large Flow Surveillance Testing.Caused by Failure to Note Differential Pressure Condition.Valve opened.W/930708 Ltr
ML20045G728
Person / Time
Site: Calvert Cliffs Constellation icon.png
Issue date: 07/08/1993
From: Cruse C, Gradle R
BALTIMORE GAS & ELECTRIC CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-93-002-01, LER-93-2-1, NUDOCS 9307150106
Download: ML20045G728 (8)
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B A LTIMORE G AS AND

, ELECTRIC -

CALVERT CLIFFS NUCLEAR POWER PLANT 1650 CALVERT CLIFFS PARKWAY

  • LUS8Y, MARYLAND 20657-4702 CHARLES H. CRUSE DLANT GENE AAL MANAGER cAtvent ct;rrs July B,1993 U.S. Nuclear Regulatory Commission

'Ja shing ton , D.C. 20555 ATTENTION: Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit No. 2; Docket No. 50-318; License No. DPR 69 Licensee Event Report 93-002 Inadvertent ESF Actuations 'Jhile Performing Surveillance -

Testing N

The attached report is being sent to you as required under 10 CFR 50.73 guidelines. Should you have any questions regarding this report, we vill be pleased to discuss them with you.

Very truly yours, s

f}r" y'j '., f l / l /.

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_ ~$lif 7.h l-:V CHC/RCG/bj d Attachment cc: D. A. Brune, Esquire J. E. Silberg, Esquire R. A. Capra, NRC D. C. Mcdonald, Jr., NRC T. T. Martin, NRC P. R. 'Jilson, NRC R. I. McLean, DNR J . H . 'Ja l t e . PSC Director, Office of Management Information and Program Control 9307250106 930708

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NRC FORM 366 U. S. NUCLEAR REGULATORY COMMISSION RE l N ESTIMATED BURDEN FER RESPotCE TO COMPT,YWIT)4 THtS INFORMATION i Cou.ECTON EoVEST: 50.0 HRS. F0FMIARD CC%tMENTS REGAlote0 i MW EGNATE TO NE INFORMATON AND RECcm3 MMEMENT LICENSEE EVENT REPORT (LER) OR4CH (MNB8 7714). uS6 NUCLEAR REGULATOfN COMMISScit J

- WASHINGTON, DC affbao1 AND TO mE PAPERWORK REDUCTON PRQ)ECT 01500104). OFFICE OF kW4GEMENT AND DVDGET,WASHINGTOtt DC 3rlu fSee reverse for reaosred number of dia'ts/charsc'ers for each block)

FActuTY NAME (1) DOCKET NUMBER (2) PAGE (3)

C alvert Clif f s, U nit 2 05000 318 1 OF 07 l TITLE (4)

Inadvertent ESF Actuations While Performing Surveillance T.esting EVENT DATE (5) LER NUMBER 16) REPORT DATE (7) { oTHER FACluTIES INVOLVED i8)

SEQUENTIAL REVISION MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NUMBER 06 08 93 93 - 002 -

00 07 08 93 05000 OPERATING THIS AEPO AT l$ SUOMITTED DURSUANT TO THE AEoVIREMENTS OF 'O CFR Ched ona or more) (11)

MODE (9) 20.402M 20.405(c) l )( l 5013(an2)(rv) 73 7,433 POWER 73 MCI LEVEL 0 20.4054an1160 l 50.36tc)(2) 150.73(a)(2)(vic gg l ~

IlO 20.405/a)(1)6i4 l 50 73(aH2}6) l 5033(aH2)(vni)(A) (Specty in Abstract below and in i 50 73(a)(2)(viiO(B) Toxt. NRC Form 366A) 20.40$ial(1)0v) l l 50.73(aH2)(ii) l 20 405(a)(1)M l l 5033(aH2)(ii0 l 50.73(aH2)<x)

UCENSEE CONTACT FCR THIS LER (12) l NAME TELEPHONE NUMBER (include Area Code) l R. C. Gradle 410 260-3738 COMPLETE ONE UNE FOR EACH COMPONENT FAILURE DESCRIBE 0 IN THIS REPORT (13) i MAN RENRTARE TO '

CAUSE COMPONENT NPROS CAUSE SYSTEM COMPONENT l SYSTEM TURER TURER NPROS l l SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH I DAY I YEAR i 3 SUBMISSION i YES X NO 4f ves comolete EYPECTED SUOMIS$10N DATE) DATE (15)

ABST RACT (Larmt to 1400 spaces,i.e approximateey15 sing #e-space typewntten knes) (16)

On June 8, 1993, at 9:59 a.m., Calvert Cliffs Unit 2 experienced inadvertent Auxiliary Feedwater Actuation System (AFAS) and Reactor Protective System actuations during performance of Auxiliary Feedwater (ATJ) System large flow  !

surveillance testing when excessive differential pressure developed between  !

the steam generators. At the time of the event, Unit 2 was in MODE 3, HOT  ;

STANDBY, with the Reactor Coolant System at 2250 psia and 532 degrees Fahrenheit. The main steam isolation valves (MSIVs) were Shut. There were no safety consequences resulting from this event. The AFAS actuation signal immediately stopped ATJ flow to the affected (lower pressure) steam generator.

The cause of this event was that neither the test procedure nor the operators' j assessment of the expected plant response identified the potential for the {

development of this plant condition. Appropriate procedure changes will be  !

l made to warn operators of the potential for this condition to occur. Both  !

MSIVs will be required to be open during the test. Management expectations

that operators identify and evaluate potential problems prior to task I

{i performance, has been re-emphasited.

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_,. :r f NRC FCRM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 315o-0104 Sea . . EXPIRES 5/31/95 EStMATEo BURoEN PER RESPONSE To compt,Y wrTH THS fNFoRMATION Col.15C*loN REoVEST: 50 0 HRS. FoRWARo LICENSEE EVENT REPORT (LER) coMuESTs rec 4RoiNo eURoEN EsTruATE To THE NFeRurTioN ANo RECoRoS MANAGEMENT BRANCH (MNB0 77141. U.S. NUCLEAR TEXT CONTINUATION Recut.AroRy couu:sses wASHiNoTcN. oc rosss Coot. ANo to THE P&PERWoRK REoUCTON PROJECT (3150 0104). oFFCE OF MANAGEMENT ANo SuoGET.WASHINoToN. oC 20S03. ,

FActuTY NAME (1) DOCKET NUMBER (2) . L.ER NUMBER (3) P AGE (4) l I

.. Calvert Cliffs, Unit 2 05000 3 1.8 93 -

002 - 00 02 0F 07 i rtxt or == we. . . .omeoc o, NRC ro, asum I DESCRIPTION OF EVENT On June 8,1993, at 9:59 a.m. , inadvertent actuations of the Auxiliary Feedwater  ;

Actuation System (AFAS) and the Reactor Protective System (RPS) occurred at Calvert Cliffs Unit 2. The AFAS and the RPS are Engineered Safety Fcatures (ES F) . The actuations occurred when a high differential pressure developed between No. 21 and No. 22 steam generators during performance of Auxiliary Feedwater (AW) System large flow surveillance testing. The AFAS actuation initiated an "AFAS BLOCK" signal that immediately isolated AW flow to the affected (lower pressure) No. 21 steam generator. At the time of the event, Unit 2 was in MODE 3, HOT STANDBY, with the Reactor Coolant System (RCS) at 2250 psia and 524 degrees Fahrenheit in the No. 21 steam generator associated loop. All control element assemblies (CEAs) were fully inserted with the RPS trip circuit breakers closed to support previous CEA rod-drop testing that had been suspended. The RPS asymmetric steam generator transient (ASGT) actuation signal tripped opened the RPS trip circuit breakers.

3 The AW system is designed to provide feedwater to the steam generators to cool the RCS if the Main Feedwater System is inoperable. Two AW trains , consisting of one of two selected steam driven pumps, one motor driven pump, and associated flow paths, are capable of automatically initiating flow to either steam generator. In the event of a ruptured steam generator (i.e., main steam line break), blocking valves automatically shut and stop AW flow to the affected steam generator to pravent a continuation of steam generator blowdown to ,

containment. The AFW pump turbines (No. 21 and No. 22) are normally driven by '

j main steam (MS). The MS piping for the AFW pump turbines comes off each MS line upstream of the corresponding main steam isolation valve (MSIV).

An AFAS actuation automatically starts the AW pumps upon detection of Ic - level.

in either steam generator. An AFAS actuation caused by high steam generator differential pressure causer an "AFAS BLOCK"' signal to be generated shutting the -

blocking valves to the low pressure steam generator. This action mitigates the effects of a MS line rupture.

The RFS ASCT trip utilires steam generator pressure inputs to the RPS thermal  !

margin / low pressure (TM/LP) trip calculator. When the difference in pressure between the two steam generators exceeds the trip serpoint then the RPS ASGT trip causes a reactor trip.  ?

On June 3,1993, Unit 2 was in the final week of a scheduled refueling outage.

Plant heacup to MODE 3 was complete. Both MSIVs were shut with RCS temperature f

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NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 315o-0104  !

We21 EXPIRES 5/31/95

. ESSMATEo BURCEN PER RESPONSE To COMPLY WITH THis WFoRMATON COLLECT oN REoVEST: 50.0 HRS. FCPWARo LICENSEE EVENT REPORT (LER) cowuEuts RcaARoma suncEN EsnMATt To TwE mFoRuArcN ANo REcoRoS MANAGEMENT BMNCH (MNB8 7714L U.S. NUCLEAR TEXT CONTINUATION aEcutAtoav ceuwsscN. wAssmoroN. oc rosss.000t Ano To THE PAPEPWoRK REOUC*loN pro.iEc7 Qt50 0104). OFFICE CF WANAGEMENT AND BUDoET. WASMNGToN. oC 20$01 F Aclu TY NAME (1) DOCKET NUMBER p) LER NUMBER {3) PAGE (4)

Calvert Cliffs, Unit 2 05000 3 1 8 93 -

002 - 00 03 0F 07 TEXT m . c. o. .at,, e., co o, NRc F.,- seeA iiri being maintained at approximately 532 degrees Fahrenheit by passing steam ,

through the bypass valve around No. 21 MSIV to the main condenser. t At approximately 3:00 a.m., operations shift personnel conducted a pre-evolution briefing for the scheduled performance of Surveillance Test Procedure ,

(STP) 0-73H-2, " Auxiliary Feedvater Pumps No. 21/22/23 1.arge Flow Test." This test procedure is required to be performed every refueling outage. The test ,

coordinator, a licensed Control Room Operator, conducted a complete review of STP 0-73H-2. This review included a discussion of the test procedure's purpose, scope, prerequisites (initial required conditions), precautions and a walk-through of the steps in the procedure. The brief also included actions to be taken in the event of unexpected conditions.

The test coordinator also reviewed the current plant conditions and expected ,

plant responses for the performance of STF 0-73H-2 as required by the procedure.

The procedure required that a single steam driven A W pump be supplied from a single steam generator. The AW pump would feed approximately 300 gallons per minute of relatively cool condensate storage tank water to both steam generators. The procedure requires maintenance at this flowrate for a 5 minute stabilization period prior to recording test data (AFW pump suction flowrate, temperature, and pressure, pump discharge pressure, and pump bearing vibration .

measurement). In addition, test personnel were to record 10 A W pump and A turbine vibration data points for trend analysis. The procedure has a precaution which states that the steaming and feeding rates affect.RCS cooldown rate and references the plant's Technical Specifications for permissible RCS  ;

cooldown rates. Personnel involved in the test were sensicl:ed to the cooldown effect that the test would have on the RCS temperature and steam generator ,

pressure. The following precautions were reviewed: (1) RCS cooldown rate limits, (2) maintaining steam generator levels, and (3) close monitoring of RCS ,

temperatures, pressurizer level, and steam generator pressures during the steaming / feeding transient. Plant operators were aware that, following the 5 minute stabilization period, the recording of test data was to be conducted expeditiously to minimize RCS cooldown.

i At approximately 9: 40 a.m., the test was commenced. During the 5 minute stabilization period it was discovered that a vibration monitor test equipment connector was missing. The connector was obtained and installed, but this -

action delayed the recording of test data for an additional two or three minutes. Control Room personnel informed AW pump room personnel that RCS  ;

temperature was decreasing and to expedite the data collection. ,

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NRC FCRM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 315o-01o4 (Sea .

EXPIRES 5/31/95 ESTIMATE 3 BURoEN PEA RESPONSE To CouPLY WITH THis WFoRMATioN COLLECTION REcuE37 $0 0 MHS. FoRwAAo LICENSEE EVENT REPORT (LER) CouuCNTs nE24aoWo auncEN EsTuArE To TsE nFonuATioN ANo AECoRCS MANAGEMENT SAANCH (uNB9 7714). U S NUCLEAR TEXT CONTINUATION aEcutATosY CouusscN. WASMWoToN, oC 2055S4@1. ANo To THE PAPEAwoM AEouCTioN PROJECT (31504104). oFFCE of r l

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FACluTY NAME (1) DOCKET NUMBER (23 LIR NUMBER (3) PAGE g4)

Calvert Cliffs, Unit 2 05000 3 1 8 93 -

002 -

00 04 0F 07 TEXT [lf more spue is reowred wae adetsana, copies of NRC Form :i66A) (17)

At 9:59 a.m., as final test data were being obtained, the AFAS BLOCK to No. 21 steam generator and the RPS ASGT (TM/LP) trips actuated. Alarm annunciation was received and acknowledged by Control Room personnel. The test coordinator observed that the AFJ blocking valves to No. 21 steam generator went shut. He verified that a valid high steam generator differential pressure condition existed as he observed pressure indications for No. 21 and No. 22 steam generator to be approximately 810 and 920 psia, respectively. The licensed Control Room Supervisor (CRS) directed the test coordinator to open the bypass valve around No. 22 MSIV which lowered No. 22 steam generator pressure. The '

test coordinator secured No. 21 AFJ pump.

II. CAUSE OF EVENT  :

Neither the procedure nor the operators' assessment of the expected plant response identified the potential concern for the possible development of a high steam generator differential pressure condition. -

A cause of this event was the lack of procedural measures to prevent creation of an excessive differential pressure between steam generators. Surveillance Test

, Procedure 0-73H-2 did not warn operators of the potential for steam generator differential pressure to approach safety trip setpoints. The procedure did not specifically recognite that testing could occur with the MSIVs and the MSIV bypass valves shut. At the time of the event, with these valves closed and

We recognize that all plant conditions and configurations cannot be accounted for in a written procedure. We expect the pre-evolution brief to identify and evaluate potential problems. Despite the operators' review of the test procedure and expected plant response, they did not identify the potential for '

the subsequent excessive steam generator differential pressure condition.

An additional factor that contributed to the event was extension of the test duration on No. 21 AFJ pump due to a missing test equipment connector.

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III. A'IALYSIS OF EVENT
  • 4

here were no safety consequences resulting from this event. No plant systems ar component failures resulted from this event. 'The.RPS is designed to respond

) to transients while the reactor is operating at power. The differential pressure between steam generators is not a safety concern. The test procedure

MRC FCRM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO,315o-o104 4 02)

EXPIRES 5/31/95 ESTMATEo BURCEN PER RESPONSE To COMPLY V4TH THrs NFoRMATioN CoLLECTTON REoVEST, $0.O HRS. FoHWAAo '

LICENSEE EVENT REPORT (LER) CeuMEuTs neGAnoso aunoEN EsTuATE To THE mroAMATCN ANo AECORDS MANAGEMENT BRANCH (MNBS 77141, U.S NUCLEAR TEXT CONTINUATION nEcut4TCav CoMMrsscN. wAssmoToN. oc 2055$4001, ANo To

'HE PAPEHWoRK REOUCD PAoJECT (3150 01041. OFFICE OF MANAoEMENT ANo BUoGET,WASHINoToN. oC 20503.

  • l FACluTY NAME (1) DOCKET NUMBER (2) LER NUMBER (3) PAGE (4)

Calvert Cliffs, Unit 2 05000 3 1 8 93 -

002 - 00 05 0F 07 ,

TEXT no. .pc. . a. ean., com at NaC p mAnm requires the plant to be in MODE 3, therefore, this event could not have occurred during plant power operation. .

The Updated Final Safety Analysis Report (UFSAR) Chapter 14, " Safety Analysis" defines the asymmetric steam generator event as any initiator that affects only one of the two steam generators. The UFSAR describes the most limiting case as  !

a loss of load to one steam generator at hot full power. This event is initiated by the inadvertent closure of the MSIV for the affected steam generator. The resultant non-uniform core inlet ti aperature distribution produces local power peaking in the core. The analysis concludes that no safety  ;

limits are exceeded. At the time of this event, Unit 2 was in MCDE 3 with all CEAs fully inserted. The R?S ASGT trip actuation was not required to prevent t

exceeding any safety limits.

The UFSAR Chapter 14, " Safety Analysis" describes the most limiting steam line ,

break (SL3) event to be a guillotine-type non-isolable main steam line break  ;

between the steam generator and the MSlV at hot full power. The purpose of the  ;

AFJ BLOCK signal is to limit plant cooldown during the SLB event (limiting the '

potential return to power peak following reactor trip CEA insertion). The safety analysis takes credit for the AFJ isolation (AFAS BLOCK) signal stopping i AF'4 flow to the affected steam generator terminating the RCS cooldown when the affected steam generator subsequently blows dry. Since the procedure required I the plant to be in MODE 3 and all CEAs were fully inserted, then this event is 'l; bounded by our current safety analysis. Based on this. it is concluded that this event resulted in no real or potential significant safety consequences.

This event is reportable in accordance with 10 CFR 50.73(a)(2)(iv) in that.it resulted in automatic actuation of engineered safety features. These occurred as unplanned AFAS and RPS actuations. '

IV. CORRECTIVE ACTIONS 1 Immediate A. Control Room Operators confirmed that a valid high steam generator -

differential pressure condition existed, verified that AFJ-flow was ,

stopped to No. 21 steam 5enerator and that the RPS trip circuits 4 ,

breakers had opened, and shutdown No. 21 APJ pump. They opened the I bypass valve around No. 22 MSIV to reduce pressure in No. 22 steam j generator. This action alleviated the steam generator differential i pressure condition.

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FACluTY MAME (1) DOCKET NUMBER (2) LER NUMBER (3) PAGE (4)

Calvert Cliffs, Unit 2 05000 3 1 8 93 -

002 - 00 06 0F 07 s

' EXT W more space is Hwed, use scations copies at NRC Form 3tWi.4 p /)

B. Control Room Operators identified and discussed the need to closely monitor steam generator differential pressure during performance of STP 0-73H-2. Subsequent testing on No. 22 AFW pump was completed -'

without incident.

Actions to Prevent Recurrence: .

C. A caution statement will be-incorporated into STP 0-73H-1 (the similar procedure for Unit 1) and STP 0-73H-2 warning plant operators of the potential for steam generator differential pressure to approach safety setpoints.

D. A requirement will be added to STP 0-73H-1 and STP 0-73H-2 for both MSIVs to be open during this test.

The General Supervisor-Nuclear Plant Operations has re-emphasized to E.

operators management expectations that tasks be fully evaluated to ensure that potential problems that could exist or dev,elop during

, implementation are considered. ,

F. Management expectations that all appropriate test equipment be available prior to test initiation will be emphasized to appropriate ,

test personnel.

V. ADDITIONAL INFORMATION A. Identification of Components and Systems Referred to in this LER:

IEEE 803A/83 IEEE 805/84 Component or System Funct Ident System Code Reactor Protective System NA JC Engineered Safety Feature NA JE Auxiliary Feedwater System NA BA  !

Reactor Coolant System NA AB Steam Generator SG NA Main Steam Isolation Valve ISV NA AFW Pu=p P BA

1 NRC FORM 366A U.S. NUCLEAR REGULATCRY COMMISSION APPROVED BY OMB NO. 3150 0104  !

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EXPIRES 5/31/95 EST1 MATED BURCEN PER RESPONSE To CouPLY WITH 74:s WFoRMATCN COLLECTION REQUEST: 20 HRS. FORWARD UCENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN EST1 MATE TO THE WFORMATCN AND RECORCS MANAGEMENT BRANCH (MNBS 7714), U.S NUCLEAR TEXT CONTINUATION REGULATORY COMMISSON WASHWGTON, DC 20$$$@01. MD To THE PAPEPWORK REDUCTION PROJECT (31504104i. OFFICE oF MANAGEMENT AND BUDGET. WASHINGTON. DC 20503.

F ACIUTY NAME (1) DOCKET NUMBER (3 LER NUMBER {3) PAGE (4)

Calvert Cliffs, Unit 2 05000 3 1 8 93 -

002 - 00 07 0F 07 TEXT i n - .o.c. a .aomone. or NRC F sasAi o 7,

3. Previous Similar Events  :

There has been one previous reported ecent. involving an inadvertent.

ESF (RPS ASGT) trip actuation due to high steam generator differential pressure conditions, but the cause (MSIV hydraulic  :'

system malfunction) was not related to this event. LER 318/92-006 documented the event.

LER 317/85-05 reported an inadvertent ESF actuation due to not having the handswitch keys in the handswitches used for blocking an

ESF signal prior to reaching the setpoint during plant cooldown.

The lack of procedure guidance to prepare for the required ESF signal blocking during plant cooldown centributed to the event. The i appropriate procedures were revised to ensure adequate guidance is provided for the anticipated plant response. This event has not recurred.

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