ML18038A390

LER 88-014-00:on 880313,reactor Scram & ESF Actuations Occurred.Caused by Equipment Failure Due to Design Deficiency.Transmitter Replaced W/Upgraded Model & Temporary Mod Performed to Bypass Logic for valves.W/880412 Ltr

ML18038A390
Person / Time
Site:	Nine Mile Point
Issue date:	04/12/1988
From:	Ronaldo Jenkins, Perkins T NIAGARA MOHAWK POWER CORP.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
LER-88-014, LER-88-14, NMP3218, NUDOCS 8804190038
Download: ML18038A390 (18)

Text

NRC Form 366 UJ. NUCLEAR REGULATORY COMMISSION (94)3)

APPROVED OMB NO. 3150410l LICENSEE EVENT REPORT (LER) EXPIRES: SI31/65 FACILITY NAME (1) DOCKET NUMBER (2l PA 6 3I o s o o o 410 1 OF 08 Reactor Scram and Emergency Core Cooling System Actuation due to a

,EVENT DATE (5) r Flow Caused b a Desi n Deficienc LER NUMBER (6) REPORT DATE IT) OTHER FACILI1'IES INVOLVED ISI

. SEOVENTIAL MONTH DAY YEAR YEAR +i?2 NIIMSEII T6gp IIIIMSER MONTH DAY YEAR FACILITYNAMES DOCKET NVMBERIS)

N/A 0 5 0 0 0 00 0 12 88 N/A 0 5 0 0 0 THIS REPORT IS SUBMITTED PURSUANT T0 THE REGUIREMENTs oF 10 cF R 5: Icnech one or more ot the Ioiiowrnpi (11)

OPERATING MODE (6) 20A02(bl 20A05(c) 60.73(e l(2)osl 73.71III)

POWER 20A06(el)1)D) 50.36(cl(1) 50.73(e I l2 l(v) 73.71(cl LEVEL (10) 20A05 (e I (I) I i) I 60.36(cl(2) 50.73(e I (2((vii) OTHER ISpecify In Abstract below end In Test. HIIC Form 20A06 ( ~ I (1)(E ii) 50.73(el(2) li) 60.73(e l(2)(viEI)(Al 366AI 20A06 ( ~ ) ( I ) (iv) 50,73( ~ l(2)(iil 60.73( ~ I (2)(v i)i I (6 I 20A06(e)(1)(vl 50.73 (e I (2 I (iiil 50.73( ~ l(2) lel I.ICENSEE CONTACT FOR THIS LER (12I NAME TELEPHONE NUMBER AREA CODE Robert E. Jenkins, Assistant Supervisor Technical Support 315 349-.4220

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COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBEO IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANVFAC EPORTABLE MANUFAC. PER TAEEE TVRER TO NPRDS CAUSE SYSTEM COMPONENT E TVRER AD PT R369 Y

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SUPPLEMENTAL REPORT EXPECTED (Iel MONTH DAY YEAR EXPECTED SUBMISSION DATE (16)

YES (if yes, compsere EXPECTED SVBMISSIDH DATEI NO ABSTRACT ILImlt to Ie00 speces, I e., epprossmerery Hrreen sinpieepece rypewritren Iiness (16)

On March 13, 1988 at 17:39 with the reactor mode switch in Run (Operational Condition 1) and at a power level of approximately 43K (see note) rated thermal capacity, Nine Mile Point Unit 2 experienced an automatic reactor scram, an automatic initiation of the Division 3 Emergency Core Cooling System (ECCS) with a subsequent coolant injection, and the automatic actuation of several Engineered Safety Features. These events were the result of low water levels in the reactor vessel caused by a total loss of feedwater flow. An Unusual Event declared at 17:45 was terminated by 18:00 that day. The ECCS injection was manually terminated and a normal reactor shutdown was commenced by the NMP2 operators. (NOTE: Reactor power was at 9% two minutes prior to this event.

However, an instrument failure caused the reactor recirculation pumps to downshift to low speed operation, decreasing reactor power to 43K.)

The ir(mediate cause for this event is an equipment failure. However, the root cause for this event is a design deficiency.

The corrective actions for this event are; (1) a temporary modification has been performed to bypass the "seal inH logic for the low pressure heater string outlet isolation valves, (2) a permanent modification will be implemented to reroute the piping for the Second Point Feedwater Heaters level switches, and (3) the failed pressure transmitter, which was replaced with an upgraded model, will be sent to the vendor for a failure mode analysis..

88041c)003S 880012 po~"(~ go 89. IE NRC Form 366 (96>>

NRC form 388A U.S. NUCLEAR REOUI.ATORY COMMISSION (983)

LICENSE VENT REPORT (LER) TEXT CONTINUATION APPROVE 0 OMS NO 3150&(04 EXPIRES: 8/31/88 FACILITYNAME I'l OOCKET NUMBER (2)

LER NUMBER (8) PACE (3)

YEAR SEOVENZ/AL CPu REY SION NVM ER NVMSER Nine Mile Point Unit 2 o 5 o o o 410 88 014 00 02 oF 08 TIKT/8/AINe RMOP N nyule4 Cer aRRFabne/NRC form 3(I/)AS / OT)

I. DESCRIPTION OF EVENT On March 13, 1988 at 17:39 with the reactor mode switch in Run (Operational Condition 1) and at a power level of approximately 434 rated thermal capacity, Nine Mile Point Unit 2 (NMP2) experienced an automatic reactor scram, an automatic initiation of an Emergency Core Cooling System (ECCS), and the automatic actuation of several Engineered Safety Features (ESF). These events were the result of low water levels in the reactor vessel which were caused by a total loss of feedwater flow.

The sequence of events for this incident is as follows:

At 17:37:03, pressure transmitter 2ISC*PT122 failed, causing an erroneous low differential temperature signal for the steam dome/recirculation pump suction interlock. As a result, the Reactor Recirculation Pumps (RCP) automatically downshifted from high speed to low speed operation. Reactor power which was approximately 98K prior to this event decreased to approximately 43K within 15 seconds after the RCP downshift.

Between 17:37:38 and 17:39:00, the sharp reduction in reactor power caused a low pressure condition in the Extraction Steam System (ESS). Reduction in the extraction steam pressure caused steam flashing in the Second Point Feedwater Heaters (SPFH) affecting the level instrumentation for that equipment. This resulted in intermittent false high level signals for several of the low pressure feedwater heaters, which initiated an isolation for two of the low pressure feedwater heater strings. [These spurious signals were of an extremely short duration, typically alarming and cle'aring within one s'econd. However, the outlet isolation valves (2CNM-MOV32A(B,C)) for the low pressure feedwater heater strings have a "seal in" feature requiring only a single instantaneous high level signal to close these valves.]

At 17:39:00, the "A" and "C" low pressure feedwater heater strings were completely i sol ated.

At 17:39:33, the Reactor Feed Pumps (RFP) tripped on a low suction pressure condition. This condition resulted from a reduction of flow'o the RFP's after the "A" and "C" low pressure feedwater heater strings isolated. Tripping of the RFP's resulted in a loss of feedwater flow to the reactor.

At 17:39:38, the reactor water level began to decrease as a result of the loss of feedwater flow. At this time the reactor low water level'Level 4) alarm was annunciated in the NMP2 control room.

NRC FORM 385A

+U.S.OPO:19854.624 538/155 (945)

NRC Form 38SA (083) U.S. NUCLEAR REQULATORY COMMISSION LICENS VENT REPORT (LER) TEXT CONTIN TION APPROVED OMS NO. 3150MI OC EXPIRES: 8/31/88 FACILITY NAME (ll OOCKET NUMBER (2)

LER NUMBER (6) PACE (3)

YEAR 5EQV ENTIAL REVISION NUM S R 'I'N'i NVMSER Nine Mile Point Unit 2 p 6 p p p 410 88 014 00 03 08 TEXT EN'awe <<mco Ir /PSMIoc( oco AEI/ooo/NRC For/I/ 3(/543/ (ITl At 17:39:50, the reactor scramed on a reactor low water level (Level 3) trip.

At 17:39:57, the NMP2 licensed operators placed the reactor mode switch to s hut down.

Due to the continued loss of reactor water inventory (due to boiling by decay heat), reactor water level reached the low-low level trip setpoi nt at 17:40:02.

A reactor low-low water level (Level 2) trip signal was generated (as expected) which initiated the automatic actuation of the following systems:

1. The High Pressure Core Spray (HPCS) system (Note: HPCS is a Division 3 ECCS system.)

2. The Division 3 Emergency Diesel Generator (EDG2)

3. The Reactor Core Isolation Cooling (RCIC) system

4. The Division 1 and Division 2 Standby Gas Treatment Systems (GTS)

5. The Division l,and Division 2 Reactor Building Ventilation (HVR) Unit Coolers

6. Recirculation pump trip

7. Alternate Rod Insertion

8. The Division 1 Control Building Special Filter Ventilation (HVC) system

9. Isolation of the Normal Reactor Building Ventilation system

10. The Division 2 HVR Emergency Recirculation Unit Cooler (2HVR*UC413B) ll. Isolation of the primary containment except for the Main Steam Isolation Valves (MSIV's)

(Note: Primary Containment Isolation Valve Groups 4 and 5 isolate on a Level 3 signal.)

At 17:40:04, the HPCS system started to inject water from Condensate Storage Tank (CST) "B" to the reactor vessel. As a result, reactor water level began to increase.

At 17:40:05, EDG2 attained its normal operating parameters. Additionally, the ventilation system for EDG2 automatically started up. (The diesel generator ventilation systems are considered ESF systems.)

At 17:40:15, the RCIC system injected to the reactor from CST "A". As a result of the RCIC injection, a simultaneous Main Turbine-Generator trip was initiated.

NRC PORLI SSSA IS 83) oU.S.OPO.I SSSO 62C 538/455

NRC Form 388A (94)3) US. NUCLEAR REOULATORY COMMISSION LICENSE VENT REPORT (LER) TEXT CONTINUATION APPROVED OMB NO. 3150-0104 EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER 12)

LER NUMBER IB) PACE (3)

YEAR C~g BEDVBNTIAL REVISION whV,: NVMBBII NVMBBR Nine Mile Point Unit 2 p p p p p 410 88 014 00 04 oF 08 Tm(T O'AKBB <<>>CB EI /BBVBBB( VBB BIIt/BOn4////IC%%dnn 3()84'4/ (17)

At 17:42:27, the NMP2 operators restored partial feedwater flow to the reactor.

At 17:42:49, the reactor water level was restored to its normal level by the feedwater system and the HPCS and RCIC injections. All low water level annunciation had cleared by this time.

At 17:42:57, the HPCS injection was manually terminated by the NMP2 operators.

At 17:45, in accordance with Eme'rgency Action Procedure EAP-2, an Unusual Event was declared for NMP2. Additionally, the NMP2 operators placed the Scram Discharge Volume (SDV) bypass switches to bypass. (This is done to prevent another scram signal on high SDV water level.)

At 17:48:03, the alternate rod insertion initiation was reset by the NMP2 operators.

At 17:51, the reactor scram was reset by NMP2 operators in accordance with Operating Procedure N2-0P-101C.

At 17:55, NMP2 operators secured the HPCS system.

At 18:00, the Unusual Event was terminated.

At 18:12, the primary containment valve group 6 and 7 isolations were bypassed by Operations in order to place the Reactor Water Cleanup (RWCU) system back into service.

At 18:42, the RCIC injection was secured by the NMP2 Operations Department.

Between 18:43 to 18:50 the Division 1 and Division 2 GTS systems were secured and normal HVR was restored by the NMP2 Operators.

At 19:56, the Main Turbine-Generator trip was reset by the NMP2 Operations Department.

At 23:35, the NMP2 operators secured the Control Building Special Filter Ventilation System.

The duration for this event, from the initial event transient (failure of pressure transmitter 2ISC*PT122) to the termination of the Unusual Event was approximately 23 minutes; It is estimated that approximately 11,500 gallons of water were injected into the reactor vessel by the HPCS system from CST "B".

Add'itionally, it is estimated that the RCIC system supplied approximately 38,000 gallons of water from CST "A" to the reactor vessel. A small increase in reactor water conductivity (well within Technical Specification limits) was noted after this event.

The individual systems and components functioned as designed. There were no other inoperable systems which contributed to this event. No other plant system or component failure resulted from this event.

NRC FORM 344A *U.S.OPO:1985&824 538/455 (94)3)

NRC form SSSA U.S. NUCLEAR RECULATORY COMMISSION 1043)

LICENSE cVENT REPORT (LER) TEXT CONTINUATION APPROVED OMB NO. 3150&104 EXPIRES: B/31/BB PACILITY NAME 11) OOCKET NUMBER )2)

LER NUMBER )S) PACE 13)

YEAR V$) SEOVENTIAL NVM EII

>Mr REVISION I:<5 NOMBEII Nine Mile Point Unit 2 0 5 0 0 0 410 88 014 00 05 OF 08 TEXT %ewe <<>>oo /r /SENSe4 INo EEESOo>>/NRC fonrr SSSES/ )Ill To satisfy the reportability requirement of TS Section 3.5.1(f) the following information is being provided for the ECCS high pressure coolant injection event:

Total accumulated initiation cycles for the HPCS system (from receipt of the NMP2 operating license up to and including the March 13, 1988 event) = 3 The usage factor value for the HPCS injection nozzle (as of March 13, 1988) remains significantly below 0.70.

II. CAUSE OF EVENT The immediate cause for this event was the failure of pressure transmitter 2ISC*PT122. This instrument failure (which i s considered to have been a random incident) generated a spurious signal which caused the RCP's to switch to low speed operation; this in turn caused the sharp reduction in reactor power.

These initial events subsequently led to the reactor scram and the Division 3 ECCS actuation. However, this instrument failure is not considered to be the root cause for this event since the loss of feedwater flow (as it occurred in this event) is not an anticipated result for a transient involving a decrease in the reactor coolant system flow rate. (See the NMP2 Final Safety Analysis .

Report (FSAR) Section 15.3.)

Therefore, the most probable root cause for this event (and for the loss of feedwater flow) is a design deficiency. Spurious high water level signals, generated by the level switches for the SPFH's, initiated the "A" and "C" feedwater heater string isolations. These signals were the result of perturbations (caused by 'steam flashing) in the level switch instrument lines.

LIt is thought that the steam flashing (caused by. the reduction in the ESS system pressure) may have forced water slugs through the sensing lines for the SPFH's level instrumentation.]

Niagara Mohawk Engineering has determined that the physical installation of the sensing lines for the SPFH level instrumentation made these instruments particularly susceptible to the transients caused by steam flashing.

Engineering has concluded that a different installation configuration would make these level switches less susceptible to similar disturbances.

III. ANALYSIS OF EVENT ICOSI This event is considered reportable via 10CFR50.73(a)(2)(iv) because the reactor scram and th'e various safety system actuations (such as the automatic startup of the Division 3 ECCS and of the Division 1 and 2 GTS systems, the automatic alignment of HVR and HVC to their emergency modes, and the primary containment isolation (except for the MSIV's)) were automatic ESF actuations.

NRC FORM OSSA o U.S.CPO:10554-52E 535/455

Perm 388A (883) US. NUCLEAR REOULATORY COMMISSION LICENSEE ENT REPORT (LER) TEXT CONTINUATION APPROVED OMS NO. 3(50-0(04 EXPIRES: 8/31/88 I'ACILITYNAME (1) OOCKET NUMSER 121 LER NUMBER (8) PACE (31 YEAR SEOUENTIA(, >XP REVISION NUMB E R i??3 NUMSER Nine Mile Point Unit 2 p 6 p p p 410 88 014 00 06 OF 08 TE)(T 8'mw Nmoc 4 /CSUN< Imc Redone/NRC Fc/IR 388A'c/ (IT)

A reactor scram occurred on a Level 3 reactor low water level trip as a direct result of the loss of feedwater flow. A reactor scram is a conservative plant response which does not pose any safety consequences. The spectrum of events (which include the Division 3 ECCS and all the ESF actuations discussed above) that occurred as a result of the loss of feedwater flow are bounded within the analysis of the "Loss of Feedwater Flow" event discussed in the FSAR Section 15.2. 7.

Reactor water level decreased slightly below the Level 2 trip setpoint.

(Actually, the lowest water level attained was 100 inches above instrument zero.) HPCS and RCIC initiated as designed at the Level 2 setpoint and injected coolant from the CST's "B" and "A" respectively. Reactor water level was restored to its normal level approximately 3 minutes after the injection comenced. After the reactor water level was restored to normal, operators manually secured the HPCS and RCIC injections.

The temperature difference between the coolant injected i nto the reactor vessel and the reactor coolant was approximately 460 degrees. However, Niagara Mohawk Engineering has determined that the injection did not cause a damaging transient to the reactor components.

The automatic actuation of the HPCS system with a subsequent coolant injection was a conservative plant response with minimal plant impact and no resultant impact on public safety. The HPCS actuation is considered conservative because the ECCS systems are designed to provide timely protection against onset and consequences of conditions that threaten the integrity of the fuel barrier and the Reactor Coolant Pressure Boundary.

The other ESF actuations (i.e., Division 1 and 2 GTS startup, lineup of the HVR and the HVC systems to their emergency modes, and the primary containment isolation) were also conservative plant responses, with minimal plant impact and

. no resultant impact on public safety.

The primary containment valve isolation is considered conservative since the primary objective of the isolation function is to provide protection to the plant'and public by preventing releases of radioactive materials to the envi ronment.

The GTS system is designed; (1) to limit the release of radioactive gases from the RB to the environment within the guidelines of 10CFR100 in the event of a loss of coolant accident and, (2) to maintain a negative pressure in the RB under accident conditions. (The emergency recirculation mode of HVR helps achieve these objectives.) Therefore, an automatic initiation of GTS and the emergency recirculation mode of HVR are considered conservative since their proper function serves to limit and contain radioactive releases from the primary and secondary containments.

NRC CORM SSSA (883) *U.S.OPO:1988&82l 538/455

NRC POIIII 388A (083) U.S. NUCLEAR REOULATORY COMMISSION LICENSEE VENT REPORT (LER) TEXT CONTINUATION APPROVED OMS NO. 3(50W104 EXPIRES: 8/31/88 fACILITYNAME (1) DOCKET NUMBER (2)

LER NUMBER (8) PAGE (3) yEAR gM SEOUENTIAL HUMEER

~riN II1 V IS IO II IIUM E R Nine Mile Point Unit 2 o s o o o 41o 88 014 00 07 OF 08 TSXT (8'aWe Red 8 IFSU)FIS car aESS(Fne/A//IC %%dnII 3///AS / (17)

Finally, the automatic alignment of the HVC system to the special filter, ventilation mode provides monitored and filtered air to the control room under accident conditions. This mode of HVC minimizes the influx of radioactive contaminants into the control room. Therefore, an automatic alignment of HVC to thi s mode of operati on i s a conservati've response.

The elapsed time for the event, from the initial event transient to the termination of the Unusual Event was approximately 23 minutes.

I V. CORRECTIVE ACTIONS (1) To minimize future losses of feedwater flow due to spurious feedwater heater string isolations, a temporary modification (¹88-104) has been implemented to bypass the "seal in" feature for the low pressure feedwater heater string outlet i sol ati on val ves (2CNM-MOV32A(B,C)) . (The decision to remove or to permanently implement this modification will be based upon the operating experience of NMP2.)

(2) As an effort to desensitize the SPFH level transmitters to short term perturbations such as those caused by steam flashing, Modification (PN2Y88MX055) wi 11 be performed rerouting the piping for these instruments. It is anticipated that this modification will be installed no later than the mid-cycle outage (scheduled for September, 1988).

'(3) The failed pressure transmitter (2ISC*PT122), a Rosemount Model ¹1152GP, was replaced with a Rosemount Model ¹1153GB via Work Request (WR

¹138201). This defective transmitter will be sent to the manufacturer for an analysis of its failure mode.

V. ADDITIONAL INFORMATION LER's 88-01 and 88-12 also discuss events where the reactor scrammed on a low water level (Level 3) trip and the Division 3 ECCS system actuated. However, the causes for those events are not similar to the event discussed in this report. Therefore, there are no previous events similar to that discussed in this LER.

Failed Component Identification: . Pressure Transmitter Model ¹1152GP8 Manufacturer - Rosemount Vendor - General Electric (GE)

GE Part Number - 169C8393P882203 NRC FORM 305A (883) e U,S.GPO:10850.824 538/455

NFC Form 388A 083 l U.S. NUCLEAR REOULATORY COMMISSION LICENSE ENT REPORT (LER) TEXT CONTINU ION APPROVEO OM8 NO. 3150-0164 EXPIRES: IU31/88 FACILITYNAME 111 OOCKET NUMSER 12l LER NUMSER (81 PACE 13) yEAR @'C' 8 CUE NTIAL NUM ER ?j9? REVOIQN NUMSER Nine Mile Point Unit 2 o s o 4lO 88 014 00 08 08 o o oF TEXT EEr more <<coo k roEIem5 Ieo aAt5onel NRC Form%54'Fl lITI V. ADDITIONAL INFORMATION (Cont')

Identification of Components Referred to in this LER IEEE 803 IEEE 805 Component EIIS Funct System ID Pressure Transmitter PT AD Level Transmitter LT SJ Reactor Recirculation Pump P AD Feedwater Pump P SJ Low Pressure Feedwater Heater (SPFH) HX SJ .

Piping (Tubing) TBG SJ Diesel Generator DG EK Scram Bypass Switches HS JC Scram Discharge Volume COL AA Isolation Valves (Primary Containment 5 Feedwater Heaters) ISV JM,S J Main Steam Isolation Valves (MSIV's) ISV JM HVR Unit Coolers Cl R VA Condensate Storage Tank TK KA Reactor Mode Switch HS JC Extraction Steam System N/A SE Feedwater System N/A SJ High Pressure Core Spray System N/A BG Reactor Core Isolation Cooling System N/A BN Standby Gas Treatment System N/A BH Reactor Building Ventilation System N/A VA Reactor Water Cleanup System N/A CE Emergency Diesel Generator Ventilation System N/A VJ Control Building Ventil ation System N/A VI Primary Containment N/A NH Reactor Recirculation System N/A, AD NRC FORM 306A *U.S,OPO'.1986&62A 538/455 1843 I

7 NlAGARA NMP 32818 U MOHAWK NIAGARAMOHAWKPOWER CORPORATION/301 PLAINFIELDROAD, SYRACUSE, N.Y. 13212/TELEPHONE (315) 474-1511 April 12, 1988 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Re: Docket No. 50-410 LER 88-14 Gentlemen:

In accordance with 10 CFR 50.73, we hereby submit the following Licensee Event Report:

LER 88-14 Is being submitted in accordance with 10 CFR 50.73

('a)(2)(iv), "Any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF), including the Reactor Protection System (RPS)."

A 10CFR50.72 (b)(2)(ii) report was made at 1810 hours0.0209 days <br />0.503 hours <br />0.00299 weeks <br />6.88705e-4 months <br /> on March 13, 1988.

This report was completed in the format designated in NUREG-1022, Supplement 2, dated September 1985.

Very truly yours,

, 6n4 'i Thomas J. P / kins Vice President - Nuclear TJP/POB/meed Attachments cc: Regional Administrator, Region I Sr. Resident Inspector, W. A. Cook

IS