Text

Special Rept:On 900320,unexpected Pressurization Transient of Reactor Coolant Sys Occurred.Caused by Mgt Deficiency Due to Incomplete Review of Equipment Status Indicators Prior to Condition Change.Walkdown of Affected Piping Performed
	ML20034C057
Person / Time
Site:	Catawba
Issue date:	04/26/1990
From:	Owen T; DUKE POWER CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	NUDOCS 9005020021
	Download: ML20034C057 (23)
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Duke her Corropary gg y yyny Catuwba Nudeur Station 3

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Clowr. S C 2Y'10 DUKEPOWER April 26, 1990 Document Control Desk U. S. Nuclear Regulatory Commission Washington, D. C.

20555

Subject:

Catawba Nuclear Station Special Report - Unit 3 Technical Specification 3.4.9.3.c.

Gentlement pursuant to Technical Specification 3.4.9.3 Action Statement c.,

a special Report is to te submitted within 30 days of an event in which the "PORVs or the Reactor Coolant System Vents are used to mitigate a Reactor Coolant System pressure transient". An event occurred on Unit 1 on March 20, 1990 in which the PORVs controls were unable to respond but the Residual Heat Removal System's pump suction relief valves mitigated the transient.

As described in the Sthtion FSAR and the NRC's Safety Evaluation Report, the auction relief valves are a part of the LTOP protection when that system is aligned to the primary coolant system as it was that day.

Therefore, we are submitting this Special Report to describe the circumstances of the event, the effect of the relief valves on the transient and corrective actiona taken to prevent recurrence. The contents of this report were described in an Enforcement Conference held in Atlanta on April 25, 1990.

Very truly yours, 1

b Tony B. Owen Station Manager keb\\ REPORT.SP xc:

Mr. S. D. Ebneter American Nuclear Insurers Regional Administrator, Region II c/o Dottie Sherman, ANI Library U. S. Nuclear Regulator Commission The Exchange, Suite 245 101 Marietta Street, NW, Suite 2900 270 Farmington Avenue Atlanta, GA 30323 Farmington, CT 06032 l

M & H Nuclear Consultants Mr. K. Jabbour 3221 Avenues of the Americas U. S. Nuclear Regulatory Commission l

New York, NY 10020 Office of Nuclear Reactor Regulation s

Washington, D. C.

20555 INPO Records Center Suite 1500 Mr. W. T. Orders 1.100 Circle 75 Parkway NRC Resident Inspector Atlanta, GA 30339 Catawba Nuclear Station 9005020021 900426-PDR ADOCK 05000413, if S

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DUKE POWER COMPANY CATAWBA NUCLEAR STATION PROBLEM INVESTIGATION REPORT NO. 1-C90-0094 REACTOR COOLANT AND RESIDUAL HEAT REMOVAL SYSTEMS UNEXPECTED PRESSURIZATION TRANSIENT DURING FILL AND VENT ABSTRACT on March 20, 1990, with Unit 1 in Mode 5, Cold Shutdown, while in the process of filling and venting the Reactor Coolant (NC) System, an unexpected pressurization transient of the NC System occurred at approximately 0940. hours.

Control Room Operators (CROs) had closed the Pressurizer power Operated Relief-Valves (PORVs):and were increasing NC System pressure to 100 psig by adjusting the Chemical and Volume control'(NV) System charging and Residual Heat Removal (ND). letdown., CRos were closely monitoring the NC System pressure indications in the Control Room when they noticed that the~ Pressurizer Relief Tank (PRT) level was increasing and that the ND pump.1A discharge pressure was' abnormally high. CRos then realized that the NC System was pressurized above 100 psig; but NC System pressure' indications all read ~zero.

Investigation. confirmed that;the ND pump 1B suction relief valve lifted, NC System was pressurized to greater ti.tn 100 psig, and root valves were isolated to all the Control Room NC pressure instrumentation. This incident is attributed to a Management Deficiency due to an incomplete review of equipment status indicators prior to a condition change.

A walkdown of affected piping on the ND System, testing of the ND suction relief valve, an engineering safety evaluation, and an operability. detenmination of ND and NC were performed before the NC System fill and vent activities were rer med. This report is submitted as a Special Report.

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Page'2 BACKGROUND The primary purpose ofthe Residual Heat Removal [EIIS:BP) (ND) System is to remove heat from the Reactor core and Reactor Coolant [EIIS:AB),(NC) System during pla.nt cooldown and refueling operations. 1In addition, the ND. System I

secondary functions include transfer of refueling water between the. Refueling Water Storage Tank and the Refueling.. Cavity for refueling operations, providing.

-overpressure protection.to the NC System, and providing NC letdown flow for; pressure control and purification during plant shutdown and refueling.

The ND System is used as.a part of the Emergency Core Cooling System (ECCS) and consists of two residual heat removal' heat exchangers [EIIS:HX),: two residual heat renoval pumps-[EIISIP), and the associated piping [EIISIPSP], valves

[EIIS:V), and instrumentation necessary for operational control. The inlet.-

lines to the ND. System are connected to' the> hot legs of two Reactor coolant loops, while the return lines are connected to the cold legs of each of the Reactor coolant loops.

These return lines are also the Emergency Core Cooling System (ECCS) low head injection lines.

In its capacity as the low head portion of the ECCS, the ND System provides long term recirculation capability for core cooling following the injection phase of the loss of coolant accident (LOCA).

The purpose of the NC System is to transport heat from the Reactor core to.the Steam Generators (S/Gs), where heat is transferred to the Main Feedwater

-[EIIS:SJ) (CF) and Main Steam [EIIStSB] (SM) Systems of the secondary side.

The NC System concists of four identical heat transfer loopa connected in parallel to the Reactor vessel [EIIS:VSL).

Each loop contains an NC pump and a S/G. In addition, the system includes a pressurizer, a pressurizer relief tank, interconnecting piping, valves, and instrumentation necessary for operational control.

NC System pressure is controlled by the use of the pressurizer where water and steam are maintained in equilibrium by electrical heaters [EIIS:EHTR] and water sprays. Steam can be formed (by the heaters) or' condensed.(by'the pressurizer Y

spray) to minimize pressure variations due to contraction and expansion of the

.l Reactor coolant spring-loaded safety valves and Power Operated. Relief Valves

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(PORVs) from the pressurizer provide for steam discharge to the pressurizer relief tank, where the steam is condensed.and cooled by mixing with water.

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Overpressure protection for the NC System'is afforded by a coinbination of-procedural controls as well as ' design features of several components. The Low Temperature Overpressure Protection (LTOP) System is placed into operation when NC temperature is reduced below a setpoint (285 degrees F) value. This provides

. rotection against nonductile fracture (10CFR50, App. G) by reducing the p

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pIRs1-C90-0094/ Speci:1,R: port 1

-Page 3 setpoihts of PORVs NC32B.and NC34A, per Technical Specification requirements.

In addition, the HD System is unisolated from the NC System at pressures less than approximately'400 psig until heat removal.via the ND System is no. longer required._ In this unisolated state, the'ND System is a natural extension of the-basic function to control-NC System temperature and pressure._-In this state; ND' suction,line relief valves ND3 and ND38 provide _NC overpressure protection.

Overpressure protection for the ND System is assureo by_a combination of relief valves and auto-isolating suction valves. The inlet. isolation-valves (ND1B, ND2A, ND36B, and ND37A) have interlocks to prevent opening until NC pressure is.'

below a setpoint-(approximately,400 psig) and to auto-isolate.at a; higher

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setpoint (approximately 600 psig). This higher setpoint serves to protect against the potentialLfor an intersystem LOCA.; With these valves open, the-above mentioned relief valves, ND3 and ND38, provide not only NC but'ND-overpressure protection as well. These are 4 inches-by 6 inches Dresser type 1910 safety relief valves. These~ valves have a nominal set relief pressure of 450 psig at operating temperatures and a corresponding cold set pressure of 463 poig. At temperatures below 250 degrees F, these valves.would : ideally relieve at the cold set. pressure. _ They also have an accumulation pre sure of 45, psig.

which means that the valves would not achieve full lift until a pressure of 45 psig greater than the relief pressure is-reached.- At full ~ lift,;aach relief.

valve has the capacity to relieve at a flowrate of 1040 gpm..These valves would ideally rescat at approximately 22.5 psig below the set pressure; This flow is in excess of the combined flow of the charging pumps at that pressure. Only one j

charging pump is allowed operable in Mode 5, Cold shutdown, per Technical Specifications. Also, relief valves-(ND31, ND35, and ND64)'in the ND discharge d

lines provide overpressure protection against back leakage from the NC System 1

through the discharge flowpath check valves.-

j Technical Specification (T/S) 3.4.9.3 requires at'least one of the.following overpressure protection systems shall be OPERABLE:

Two PORVs-with a lift setting of less than or equalfto 450 pcig, or a.

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The NC System depressurized with an HC vent of greater ~than or equal

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to 4.5 square inches.

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T/S 3.4.9.3 is applicable in Mode 4, Hot Shutdown, when the temperature-of!any NC System cold leg is less than or equal to 285 degrees F, Mode 5, Cold Shutdown, and Mode 6, Refueling, with the Reactor vessel. head on.

When these conditions cannot be met, the following actions are required.

a.

With one PORV inoperable, restore-the inoperable PORV to OPERABLE status within 7 days or depressurize and vent the-NC System through=at i

least a 4.5 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

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With both PORVs inoperable, depressurize and vent the NC System-through at least a 4.5 square inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c.

In the event either the PORVs or the NC System vent (s) are used to; mitigate an NC System pressure transient, a Special Report shall be-prepared end submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The_ report'shall describe the circumstances initiating the transient, the effect of the PORVs or-NC System vent (s) on the transient, and any corrective action necessary to prevent-recurrence.

The Integrated Scheduling (I/S) group is responsible for. scheduling and tracking; all outage related work and activities. To perform thisffunction, I/S receives blocks of-work from each station group and' integrates:the work into~'a schedule to be used'by station' groups involved in an' outage.; Project'2,:the: outage scheduling computer program, is the mechanism used to develop-the schedule. I/S.

depends on the Operations _ (OPS) group to identify scheduling logic based on

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operability and Technical' Specification requirements for specific plant condition and mode changes.

Operations is responsible for performing reviews which ensure that all required work and activities are completed, all systems and components are operable, and Technical specification requirements have been met prior to allowing a plant-condition or mode change to occur. To perform this-review, OPS uses the following tools: Nuclear Maintenance. Data Base (NMDB), Project 2, Removal and Rostoration (R&Rs, i.e. tagouts), Periodic Tests (pts), and System Alignment Procedures. OPS Support personnel are expected to_ utilize all the above tools necessary.to assure appropriate plant configuration' control.

EVENT DESCRIPTION on March 20, 1990, with Unit 1 in Mode 5, Operations (OPS) was conducting the NC System Fill and Vent per procedure OP/1/A/6150/01, Filling and Venting the Reactor Coolant System. A Chemical and Volume Control [EIIS:CB] (NV). System centrifugal charging pump was operating to pressurize the NC System to approximately 100 psig while letdown was being accomplished to the NV' System by i

means of the Residual Heat Removal (ND) System. The ND System was aligned to the NC System with Train A operating to remove core residual heat. ' Refer to r.for incident sequence of events and Attachment 2 for fill and vent 4

sequence.

The initial fill of the NC System was complete and preparation for increasing NC pressure was being performed per the fill and vent procedure. With the PORVs closed, Control Rcom Operators (CRos) noticed the Pressurizer Relief Tank (PRT) level was still increasing.

Per their training and procedural guidance, they recognized the PRT increasing level as abnormal (a noticeable level increase in the PRT is an indication that the Pressurizer (PRZ) and NC System piping are l

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water solid).. Because the Operators believed that p0RV leakage was.the cause for the increasing level in the-PRT, they reduced charging flow and isolated the PORVs one at.a time. However, PRT level continued to increase..

At-approximately 0957 hours0.0111 days 0.266 hours 0.00158 weeks 3.641385e-4 months , CROs discovered that ND' pump 1A discharge pressure was indicating approximately 375 psig. Again, they recognized that this was also abnormal since normal ND pump. discharge pressure should have been approximately 200 psig when NC pressure is zero psig. After subsequent discussion with their

~ supervision, it was concluded,that the NC Syutem was pressurized.to approximately 175 poig, although NC System pressure indications:in the Control-Room read zero and~the ND suction relief; valves may have lifted.

A' decision was made to dispatch a Non-licensed Operator (NLO) in Containment to visually-inspect the ND suction relief valves while depressurizing to approximately 200 pulg ND pump discharge pressure.

While in Containment, at approximately 1008 hours0.0117 days 0.28 hours 0.00167 weeks 3.83544e-4 months , the NLO found the Train B ND

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suction relief value passing flow. The ND suction relief valves are routed to the PRT and provide overpressurization protection.- At 1030 hours0.0119 days 0.286 hours 0.0017 weeks 3.91915e-4 months , CRos then isolated the Train B ND suction relief valve from the NC~ System by closing the ND loop suction isolation valves to rescat the ND. suction relief valve..At approximately 1030 hours0.0119 days 0.286 hours 0.0017 weeks 3.91915e-4 months , the PZR PORVs were declared' inoperable per T/S..

3.4.9.3, placing Unit 1 in an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months action statement..(the T/S inoperability time was moved back to 0708 hours0.00819 days 0.197 hours 0.00117 weeks 2.69394e-4 months to reflect when the~PORVs. wore originally closed).

The ND suction relief was reseated, and~ Train B'of ND realigned to the loops at 1205 hours0.0139 days 0.335 hours 0.00199 weeks 4.585025e-4 months .

Operations Engineer A notified Instrument and Electrical (IAE) Supervisor A and

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requested an investigation of the NC -System pressure indications problem (all:

t indication reading zero). After subsequent investigation, IAE Supervisor A confirmed that the root valves to NC System pressure instruments were valved out l

for previously scheduled maintenance compression fittings outage. work (wide j

range pressure loops INC5120, 5121, and INC 5140, 5141 per Work Requests 5491 i

IAE-2 and 1493 MES-1, respectively).

I At approximately 1345 hours0.0156 days 0.374 hours 0.00222 weeks 5.117725e-4 months , NC pressure had;been reduced'to atmospheric'and the-NC PORVs reopened to establish the 4.5 square inch vent space required by Technical Specifications in the absence of operable PORVs. At approximately 1420 hours0.0164 days 0.394 hours 0.00235 weeks 5.4031e-4 months , IAE completed-unisolating and restoring the Control' Room NC pressure l

instrumentation. 7 Operations then realigned Train B ND to the NC System and

-l declared the PORVs operable, exiting T/S 3.4.9.3 and the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months action statement.

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t CONCLUSION This incident is attributed to a Management Deficiency'duef o an inadequate-t review of_ equipment status indicators by Operations to assure that all required' work was completed prior to a plant condition change during the outage. During.

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this outage and in the past,~ Operations had reviewed the following documents to-cnsure equipment operability' prior to changing modes or sonditions within a-mode r P

1)

Nuclear Maintenance Data Base'(NMDB) f 2) project 2 3)

Removal and Restoration (R&Rs, i.e. tagouts) j 4) periodic Tests 1

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procedure System Alignments' 6)

Detailed knowledge of work status by Unit Manager's Group Engineers?

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Technical Specifications Action Item Log.

ll Operations had been successful in enter.ing conditions such as loops. drained, initiation of refueling, and mid ' loop operations through :these specific reviews.

L In-this event, the Operations Unit 1 Manager's Group did-not review the Nuclear o

Maintenance Data Base'.for setting the head and PORV related items. Had this database been reviewed, the incomplete status of the original work request for welding the fittings on these instrument loops _and of the supplemental 1would have been noted. The relationship of the NC pressure instruments t'o pORV operability was understood by Operations.

The review of;NMDB for outstanding i

work requests prior to condition change was not a formalized process, i,

q The program that'had been used in the past was not a formal' written' program requiring review of.the above noted components. As a corrective action', this program that Operations follows to assure equipment operability during mode and 3

condition changes will be formalized and established within' the Operations Management procedures. The program will be written and a review session cor. ducted prior to the Unit 2 EOC3 ' outage.

i A contributing cause is assigned to the scheduling mechanism that permitted, setting the Reactor Vessel Head with the pressure instruments isolated. The-work requests to replace compression tube fitting with socket welds on NC-System Wide Range pressure instrumentation were scheduled to be worked:and completad prior to setting the Reactor Vessel Head.

Integrated Scheduling scheduled ~these l

(primary) work requests as an activity in the outage schedule, which:was required to be complete prior to fill and vent.

These primaryJwork requests had, as a prerequisite for closeout, a functional verification involving a visual inspection while the NC System was pressurized.

The supplemental work requests to isolate and restore the NC System pressure instrument loops were identified only as notes associated with the primary work request and were_not identified as stand alone items within the outage schedule.

per the supplemental work request, the restoration of the instrumentation was~to be

-performed and then documented on a Standing Work Request (SWR), 6114 SWR, which'

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PIRx1-C90-0094/..Specicl Report

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s is used.to assure that the NC pressure instrumentation is properly aligned and in-service' prior _to Mode.4.

When the welding work identified by,the. primary-work request was= reported as completed, the primary.and' supplemental work-requests were dropped from the Project 2 (outage scheduling computer, program)

U schedule.- These supplemental _ work requests' wore obtained by an IAE Planner who-recognized that their remaining work scope (restoration of instrumentation):

paralleled-that_of the SWRs used to assure NC pressure instrumentation _ alignment prior to Mode 4.: Lacking an indication on-the supplemental work requests that restoration was required prior to setting the Reactor: Vessel Head, thouIAE Planner filed the SWR with the supplemental' work requests so,that-they could be-scheduled and performed at the same time.

This work would have been scheduled 1

to be performed prior t'o Mode'4 (expected for March 24).-

1 6114 SWR is used to perform a valve checklist assuring certain NC System l

instrumentation, required in Mode 4 and for plant operation, is properly ' aligned-prior to Mode 4~.

Immediate corrective actions included stopping.all activities associated with a

plant heat-up isolation.of the PORVs to determine leakage into the PRT, visual inspection of ND suction relief valves in Containment,_-and, isolation of ths ND 4

suction relief valve from the NC System to rescat the relief valve. -In addition, IAE unisolated and restored NC System wide' range pressure instruments in order to restore PORV operability.

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Additional, subsequent corrective actions included a walkdown of the'ND System,.

review of NC and ND System instrumentation, relief valve pressure testing, and initiation of a 10CFR50.59 Evaluation, an Operability Evaluation, and PIR-1-C90-0094.

1 As corrective actions to.the. scheduling and planning of outage work the following actions are being taken:

4 1)

The outage schedule will be coded such that conditon changes will.be -

handled similar to mode changes.

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Operations will, develop the listLof_ conditions and. equipment, required 1

operable for each conditon, 3)

The IAE group will. revise their administrative controls for root valve f

status <in condition changes to be similar to their present approach to mode changes.

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IAE and OPS will develop a_ program to clearly-identify instruments out of service or calibration that will require a sticker to be placed on Control Room instruments.

5)

The Planning group will consider establishing a program.on supplemental work requests that will tie them to the original as -1 or i

-2 rather than as separate group work requests.

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Pag] 8' CORRECTIVE ACTION SUBSEQUENT

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1)

Control Room Operators isolated the PORVs one at a time to dotermine -

the leak path into'the PRT.

PORVa.were not leaking by<to:the PRT.

2)

NIO in Containment ~ inspected'the.ND Pumps auction relief valves to determine if valves were passing flow. The NLO.found ND Pump 1B i

suction' relief valve passing flow to.the~PRT.9

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CRos depressurized the NC System and. isolated ND Pump 1B suction relief valve from the NC System in. order to.roseat tho' valve.

4)

IAE determined that NC System pressure instrumentation root valves were isolated. Using Work Requests 5491 IAE-1, 1491-MES-l',,1492 MES-1, and 1493 MES-1, IAE opened the root valves and returned the.NC Syatem pressure instrumentation to service.

5)

IAE performed a calibration on INDPT5090 Loop.(ND Pump 1A discharge pressure) and discovered:that the instrument was out of calibration

.(65 psig high) thus indicating.an erroneous ND and:NC: pressure.

6)

CRos realigned ND Pump 1B suction relief'to the NC System and verified that the relief valve had. reseated.

7)

Plant Startup was held at this point until Management. understood the-event and.necessary follow-up actions.

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Operations performed a walkdown of both trains of the ND System to

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inspect for leaks.

9)'

Operations reviewed the fill and vent procedure and' developed a list of instruments required to fulfill the procedure. Operations identified instruments which were indicating correctly, needed maintenance, and needed functional verification.

10) IAE reviewed and determined the <tatus of ecch-instrument identified by' operations as being essential to the, fill and vent procedure.

11) Operations reviewed work requests to ensure equipment operability.

N before proceeding with fill and vent.

Surveillances'were reviewed to:

j ensure all necessary equipment tests were up-to-date.

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12) MES replaced the ND Pump 1B suction relief valve with a valve previously pressure tested to 625 psig.

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13) CNS initiated a 10CFR50.59 Safety Evaluation to ensure that an unreviewed safety question-did not exist.

14) PIR 1-C90-0094 was initiated to request from Design Engineering an operability evaluation for the NC and ND System.

PLANNED 1).

Operations will formalize the program to assure equipment operabilityL during mode and-condition changes; this program will be established within.the Operations Management Procedure.

'2)

Operations will make changes to the controlling procedure for Unit shutdown that prior to installing the Reactor Vessel Head, IAE signoffs=must be obtained.

3)

Operations will assume the lead in identifying Unit conditionLchanges within Modes 1 through 6.

A final list that includes equipment requirements for each condition will be provided to the responsible station group. The station sections will use this list-to determine when signoffs for condition changes need to be provided in the Operations procedure. The Operations procedures group'will incorporate these signoffs in their procedures.

4)

Integrated Scheduling will provide logic in the outage schedule for' the scheduling of Technical Specification plant conditions as.

addrosaed in item #2 above.

Codes will be provided for these plant conditions similar to the Mode codes presently in use.: This will allow lists of identified work, required to be completed prior to a specific plant condition change, to be generated for review by the outage management team.

5)

Instrument and Electrical will revise the existing adminsitrative controls for instrument root valve position verification prior to a Mode change. The revision will address those instrument root valves required to bo1 verified prior to a Technical Specification plant condition change.

6)

IAE and OPS will develop a program that will clearly identify instruments within the Control Room that are either out of service (i.e, root valves isolated) or known to be out of calibration.

7)

Maintenance Engineering Services will change Standing Work Requests-for instrument realignment and ensure they match Unit condition and mode change requirements.

8)

Planning will consider a program to ensure that a work request writteii

.q as a supplemental or in support of another work request will bela

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-2, etc. rather than one of an initiating group, j

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9) planning will review the work request pre-review process.and procedure to ensure that they comply with the outage scheduling mechanism-j changes.

10)

To. correct the Human performance deficiencias identified'in this-and f

other recent' events, additional corrective actions are to be taken to I

-limit the opportunities for error.-

1 SAFETY ANALYSIS:

l While charging to pressurize the NC System, the CROs had been monitoring PRT level.and.NC System wide-range pressure instrumentation per the NC Fill and Vent procedure. The NC System wide range pressure instrumentation had-been isolated-for previously scheduled outage work and never. restored.. Therefore, Operators were unaware that NC pressure eventually reached approximately 460 psig.

The NC System wide range pressure instrumentation also'provides input to pZR,pORV logic. Because this instrumentation was isolated, the ND loop suction valves-would not have been capable of auto isolation and Low Temperature overpressure-Protection (LTOP) was not functional; leaving the ND suction relief valves; as the only pressere relief path.

(Refer to Attachment 3, Operability Determination.)

i On March 20, Catawba performed a 10CFR50.59 Safety. Evaluation of~this incident-to ensure that an unreviewed safety question did not exist. -The results of the evaluation showed that the peak pressure experienced by NC and.ND Systems-did not exceed the pipe ratings, Technical Specifications, or past' system hydrotesting.

In addition, neither the probability nor the consequences of an accident or malfunction of equipment important to safety evaluated or not.in the l_

Final Safety Analysis Report (FSAR), was increased by this incident. 'Therefore, no unreviewed safety question exists. It is concluded that the' margin of safety as defined by Technical Specifications has not been reduced by this incident.

On March 21, Design Engineering responded to PIR 1-C90-0094,:with an operability-7 evaluation for_the NC and ND Systems. It was determined that the ND safety relief valves responded properly to the pressurization transient event caused by the NV pumps. After a review of ND piping, valves, instrumentation, and' components, there.were no adverse impact on the.ND System resulting.from the 7

peak pressure created during the incident. Design Engineering concluded that the peak pressure reached was within.the normal response range for a relief valve, with a maximum design setting of 600 psig plus 10%-accumulation.

In addition, the NC System Nil Ductility Transition Temperature (NDTT) limit'for Reactor startup was reviewed. At no time did the NC System exceed the pressure allowed for its corresponding temperature. Therefore, it was concluded'that the NC and ND Systems were operable following this incident.

The health and safety of the public were not affected by this incident.

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ATTACHMENT 1 E

SEQUENCE'OF EVENTS i

DATE.

TIME EVENT 1 3//20/90'0700

. Fill of-the Reactor. Coolant System.(NC) in progress.

0700 pressurizer is full as identicated by Pressurizer Relief' Tank (PRT) level increase. Closed Pressurizer PORVs.:

Charging flow was adjusted to approximately 100 gpm'and letdown flow to approximately 30 gpm,to begin pressurizing

'i the NC System to'100.psig.'. (This evolution,normally takes'4 4

.to.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .) Operators are closely observing the 3 NC System pressure indications?in the Control. Room.

y 0945 Control Room Operator notices PRT level increasing.

Recognized this as abnormal,and' reduced the charging rate :

4 towards the letdown rate.

Supervision ~ notified.

0950 Operators felt pressurizer PORVs may be leaking into the PRT.

Began isolating them one at a-time.

PRT level still increasing.

'~0957 Operators, noticed RHR Pusp-(ND)(dischargeLpressure Indicating 375 psig '(normal pressure with zero NC. pressure 3 would be approximately 200 psig). Theyfthenirealized NC-System was pressurized'to approximately 175-paig at this--

time, but NC System pressure 11ndicationscalleread zero.

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Notified Operator in Contal' ment'to'lookfat!ND' suction n

reliefs and continued depressurizing to 200 psig ND pump discharge pressure.

1008 Operator in Containment reported B ND suction relief passing flow.

L 1030 With ND Pump Discharge reduced toward normal, Operators isolated B ND suction from.theNC System to reseat'the relief valve. Also entered action statement for-inoperable pressurizer PORVs.(8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months action. statement).

Clock was started at 0708 hours0.00819 days 0.197 hours 0.00117 weeks 2.69394e-4 months . Notified'IAE to invest'igate NC.

System pressure indication problems, i

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DUKE: POWER COMPANY / CATAWBA NUCLEAR STATION '

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sa TPIR 1-C90-0094/ Speci21 R:portt i

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-DATE TIME' EVENT' 3/20/90 ~1100 ~

Obtained printout from AllePointsiData Base,(APD)l that:

showed;that'NC pressure had reached approximately 520: !

l psig.- Notified Management and. Design Note:

t-Later on i

s i that eveningj IAE' performed.a calibration on1the-

~

p

- INDPT5090_ loop. i' The "As = Found calibration, of this

' instrument;.(Analog! Operator Aid Computer Channel) was 65i psig. highl: : The - actual. NC pressure,,therefore. was' 455'psig.>

_~#

ii 4

1112 Operators attempt tio restore "B" Train ND suctions

'i relief but valve would not openiduesto interlock withL

.1FW-55B.: Dispatched operator per; procedure to' manually open.IND-36B.

b ?!

4 1205-

. Realigned B ND suction 1to'theLNCl System.and verified-

'I relief ~had reseated.:

d 1345 PORVs reopened and establishod; Technical Specification.

required 4.5 squareLinch~ vent' apace.

>i 3

'1420:

NC Pressure! instruments.were_unisolat'ed-and' pressurizer.

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PORVa>were' declared operable.

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1PIR'l-C90-0094/, Specisl~ Report?

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v A7T/4 IMENT-2' cThe following is a' sequence.of events which must be: accomplished to properly-Q

' fill and vent the NC System.'

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NC' System Fill (OP/1/A/6150/01, EnclosuroL4.1)'

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1) - Review limits:and precautions ~.

~ 2)-. Establish' initial. conditions.,

f 3)

. Verify' breaker and valve positions perl Enclosure 4.1.

4). ' Vent PRT to Containraent -

jj dj 5)- - Pzr spray' control" manual'and openi jj

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6)

Open: RV: head and Par vents to PRT -

7)'

Establish fill source i

-g) 8)

Fill NC System i,'

Ch A

9)

Isolate individual components < vents'as a solid stream of water;

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10) Purge N': pumps -se'al bypass lines-A

11) Fill NC cold leg injection lines' 1

S 7

12) At ~22%: Par level isolate other'NCleveliinstrumentation-i (tygon tubing) 13). Reduce flow when PRT level increases noticeably {(Par 1 full).

'14). Close Pzr PORVs and. place:in " low press" position.

3

15) Maintain 50 psig NC system press A

1

'16) Isolate fill water' source-.

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17).-.Establishnormal:makeupalignment:(ifQpplicable);

-18) Close RV head vents to PRT

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NC System Venting.(Enclosure 4'2);

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Establish 325 psig (minimum)-NC System press.

2)

Open NC pump-seal returns,,,...~

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Run.one NC. pump 20-30 seconda~,lthen=stop. pump;and wditL10 minutes

't 4)

Vent RV. head to PRT.-

s 5)

Start one pump 20-30 seconds, then stop pumpc.y 6)

Reduce NC press to 100'psig;and wait'15 minutes 7)

' Vent RV: heat to PRT 8)

. Vent; Par to PRT; 9): ' Repeat steps 1);through.8)<for each NC pump-10). Vent system instrumentation

~

i

11) Run each NC pump'one minute, reduce NV press to 100 psig'and' wait 15' minutes

12) Vent RV head-to,PRT.

13) Vent Pzr to PRT i

14) 'NC system press to 325 paig

15) Run-all'four:NC pumps for 5-10 minutes 16)..Stop NCPs

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-PIRL1-C90-0094/ Specicl: Report' cD p;g5 15' 4

ATTACHMENT 3 OPERABILITY EVALUATION Residual Heat Removal (ND) and Reactor Coolant (NC) Systems Revision 1 Statement of Problem During NC fill and vent operations, the ND pump 1A discharge pressure reached an apparent pressure of 690 psig. The HD loop suction relief valves appeared to.

1 11ft and relieve. The wide range NC pressure loops.5120, 5121, 5140, and 5141' were inolated for maintenance work, thus preventing any pressure indication-or LTop protection with 1NC32B or INC34A. This also would have' prevented'ND. loop suction valves automatic closure.

System Descr{gtjon.

)

apOverpressure protection for the NC (Reactor Coolant) SysWem is afforded by a' combination of procedural controls as well as design features of'several

~

components. The icw Temperature Overpressure protection (LTOP)LSystem is-placed into operation when NC temperature is-reduced /below~a_setpointivalue.yThis-provides1 protection-against nonductile fracture.(10CFR50I Appi G) by reducing the setpoints of two pressurizer power oper1ted relief valves (PORVs NC32B;and l

NC34A) from'2335 psig.to 400 psig LIn addition', the ND (ResidualLHeat-Removal)

System, which borders the-NC System, is unisolatedCfrom'the'NC' System at-pressures less than approximately 400 psig until heatLremoval-via'the'ND, System' is no longer. required. In this unisolated state,athe ND System is a natural cxtension of the basic function to control NC' System temperature and pressure.

i In=this state, relief valves ND3 and UD38 on Train A and B'respectively. provide

~

d~

NC overpressure protection.

Overpressure protection for the HD System is assured by a combination'of relief valves and auto-isolating suction valves-actuated by an Auto-Closure. Interlock (ACI). The' inlet isolation valves (ND1B, ND2A, ND36B,' and-ND37A).have interlocks to prevent opening until'NC pressure is below a setpoint (approximately.400 '

q poig) and to auto-isolate at a higher setpoint (approximately 600 psig). This.

t higher setpoint serves to protect against the potential for'an intersystem LOCA.

j With'these valves open,.the above mentioned relief valves, ND3 and ND38, provide j

not only NC but ND overpressure protection as well. These are 4" x 6" Dresser d

type 1910. safety relief valves. These valves ~have a nominal set relief pressure-t of 450 psig and e accumulation pressure of-45'psig which means that the valves-1 would not achieve. ?ntl lift until e pressure 45 psig greater than the relief, 1

pressure is reacaed. At full lift, ecchirelief valve has the capacity to relieve' i

1 1

4

'L

DUKE POWER COMPANY / CATAWBA _NUCLERR STATION >

pIR 1-C90-0094/ Special Report

' Page_16.

at a flowrate of at least 900 gpm. These valves would ideally reseat at i

approximately 22.5 psig.below the set pressure (blowdown.of 22.5 psig). This flow is in excess of the combined flow of the charging: pumps at that pressure,

_I not_all of which should be operable during HD operation. Also, relief valves (ND31, ND35, and ND64) in the ND discharge linea provide _ overpressure protection H

=

fagainst backloakage from the NC System through the' discharge flowpath check I

valves.

j operability Evaluation i

In preparation for the NRC telephone conference on=3-21-90, the ND pump 1A discharge pressure instrument previously reading 690 psig was found to be in j

error by 65 pulg. As a result, the HD pump discharge pressure never exceeded 625 psig. This is still a conservative estimate-due to the extrapolation of values between 5 minute readings. Because of the conservative uncertainty associated -

1 with this extrapolation, it is concluded that'the HD relief valves _ performed as expected and that ND pump suction pressure was approximately_460'psig, not 520

_k as originally interpreted. Suction safety relief valve IND3 apparently opened.

l and had reseated by the time operators were able to investigate. Additionally, ND pump discharge pressure of 625 psig was not in excess of' normal expected system response pressure. This is because even when ND pump discharge relief I

valve IND31 relieves at 600 psig, full lift and hence full flow is not achieved until 10% accumulation pressure (660.psig) iul reached. This response is acceptable under ASME Section III.

i

-l Although HD Train B was out of service, it was determined that ND. suction relief I

valve on Train B'(ND38) was passing flow,.thus. suction to'ND Train B was isolated in order to' reseat the valve. The flow passing Train B; relief valve was calculated to be approximately 155 gpm which is well below the rated capacity of.

1 the valve (at least 900 gpm). Full capacity of the relief valve was never challenged so the corresponding system suction pressure never exceeded the set pressure plus 10% accumulation (450 psig + 45 psig = 495 psig). Since this is the normally expected system response, and the design pressure of 540 psig was never exceeded, no question of operability exists for.either Train A or B suction side of ND System.

Because ND pump 1B was not operating, no pressure increase existed across the pump; therefore, its discharge piping and components were not challenged.

NC System Pressure Evaluation l

When the Unit 1 NC System was being filled and vented, the NC pressure sensing.

lines serving both the LT0p and ACI functions were blocked, forcing reliance on-

=

the ND suction line safety relief valves 1ND3-and 1ND38. These valves responded as designed when incoming ND pressure exceeded their set pressure; however, with the ND pump in operation, a maximum pressure of 625 psig was reached in the ND E

-y

DUT& POWER COMPANY / CATAWBA NUCLEAR STAT 10Ni

'N PIR 1-C90-0094/ Spechl Riport ito Pagy 17.:

9

s pump discharge piping on. Train A.

Using a; discharge pressure of 625 psig, and subtracting 165 psig for ND pump head'at 3100 gpm (Ref CNM-1201.05-318), the:

maximum'HC System pressure was approximately,460 psig. This is'a conservative-estimate since vessel pressure would be approximately:10 psig lower due to its higher sLatic-elevation. Therefore, the actual'NC-System conditions of;460 psig' at=114'F do not exceed the "Ileatup Curve" from Tech Spec 3.4.9.1, Figure 3.4-21 and Tech Spec limits were not exceeded. This provides.approximately a 70 psi-margini(530-460) at-114'F.

1 ND System Pressure Review-Pipe,: fittings, equipment, and instrumentation located within the area -

pressurized have been evaluated for the acceptability of being pressurized to" 625-psig..(It should be noted that the discharge: portion of,the ND System was-hydrotested to'900 psig except for~a portion'hydrotested to 803 psig.since~the ND heat exchanger was a limiting component)'. A detailed review was conducted for the following items:

j 1)

Piping materials from ND pump'1A discharge through valves 1ND26,-IND27, IND28A, INV135 and all normally closed vent, drain, and isolation valves.

were considered. This pressure event did 'not affect piping. stress. analysis' or hanger designs, j

2)

Valves which are located within.the piping boundaries mentioned above 3)

Instrumentation located within the piping boundaries mentioned'above 4)

Mechanical Equipment Results of those reviews, as well as the methodology followed,' are' described' below.

1) Piping Materials Piping materials (pipe, fittings, and flanges) subjected.to this pressure event i

are qualified for at least a design pressure of 792 psig.at 120'F which is j

greater than the actual pressure of'625 psig-(114'F) during the pressure event.

Pipe spec. PS-601'2, which applies to all of this' piping, permits a maximum pressure of 792 psig at 120'F.

Code calculations per Section III paragraph NC3640 are performed for the design conditions listed by the pipe spec subtable. Pipe spec. subtables are not system-specific and a single subtable is used for many applications in the plant.

Maximum design conditions listed by the pipe' spec. subtable-are always greater-than or equal to the design conditions listed by the flow diagram. Also design g

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DUKE POWER COMPANY / CATAWBA NUCLEAR STATION _

. W pIR l-C90-0094/7 Specici R: port C

-p:ga 18 ec ze conditions for the pipe spec. subtables are based onlthe liditing pipe size and i

-schedule-for-all sizes covered by the subtable. Smaller pipe sizes such-as the-sizes affected by-this pressure event have a higher code allowable pressure then ll listed by the subtable. design conditions. This conservatism qualifies these piping materials for higher design conditions than-are listed by the flow-diagram and even the pipe spec. subtable. Stress analysis and hanger design are-not affected by this pressure event.

2) Valves The valves in'the 1A ND pump discharge line have'been reviewed in light-of the actual;(extrapolated) maximum pressure of 625.psig reached.during the event; 1

This pressure was not-in excess of normal expecced' system response pressure'of l

600 psig set pressure plus 10% accumulation (660 psig) for this discharte line's.

relief valve IND31. Such a pressure response ja acceptable per.ASME Coda Section

]

III, and is not considered an overpressure event. Therefore, no further-analysis.

j is required. No adverse impact on valves occurred as a result of this event.

1 1

However all valves affected have been reviewed to determine the maximum conditions that each valve is designed for. These conditions are.ar, follows:

TAG NUMBER VALVE DESIGN CONDITION l

PSIG

. F, l

~

1ND008 2485 650 l

IND010 680 200 IND012' 2485 650 1ND014 2485 650 IND019 680 200 IND020 2485 650 i

IND021 2485 650 IND024A 2485 650 1

IND025A 2485 650 IND026 710 150' 1ND027 710 150 l

1ND028A 1420 150-

-l IND029 2485 650 l

1ND058B 2485 650 1ND067 680 200 j

1ND069 2485 650 1ND070 2485-650 l

JND071 2485 650 1ND073 2485 650

- l

DUKE POWER COMPANY / CATAWBA' NUCLEAR STATION PIR:1-C90-0094/ Special R: port

=.

0; (Pag) 19 j

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l IND074 2485 650 IND077 2485 650:

.1ND090 2485 650:

IND100 2485 650 1ND101.

2485-650 IND102 2485

.650 i

1NM039 2485 680 INS 043A 31365 200 INV135 945 150

3) Instruments The ECPE and ECLD groups have reviewed the instruments subjected to the recent.

inadvertent ND System overpressurization.:Below is a list of the. instruments that were reviewed.

Review of Instruments Requiring Replacement The' instruments reviewed would'notfhave been damaged. The one exception to_the.

above' statement is'1NDPG5201; however, this, instrument's root. valve was closed.

This instrument is used.for testing onl'yLand was isolat'ediatlthe' time of.tho event. None of these instruments require replacement.

~

Review of Instruments Requiring Calibration Check' It is recommended that a two or.three point check to verify.the calibration of-instruments 1NDPG5040 and 1NDPG5041. 'The purpose ofLthis check-is to determine if a zero or span shif t has occurred. The results of this check will determine if a full calibration is required.'These are.the only~ instruments-that'actually were pressurized above normal conditions.

, 5 INSTRUMENTS REVIEWED 1NDTW5000 1NDFE5190 INDTW5020 1NDFT5190 1NDFE5040.INDFT5191' 1NDPG5040 1NDPT5090

.1NDPG5041 1NDPG5200 1NDTW5060 1NDPG5201-INDPS5200

- c s

DUKE' POWER COMPANY / CATAWBA NUCLEAR STATIONL p

.p PIR_.1-C90-CD94/ SpeciCl R: port e '

-Pigy20 s-e.

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i.

L4) Mechanical' Equipment

'The vendor manuals provide.both design conditions and hydrotest pressures to

which this equipment was tested. This information is presented'below

Ref Design (psig)

Hydrotest-(psig)_

l~

ND Pump' CNM-1201.05-318

'DCD 11-9-88 Casing only 600 936 Mechanical Seal 1

unit on1;r -

600 1200 l

ND Heat CNM-1201.06-83 Exchanger DCD'2-18-88 Shell Side (KC System) 150 225 Tube Side (ND System) 600-803 By inspection, the peak pressure of 625 psig reached during this pressure event ~

is enveloped by. allowable hydrotest pressures. Therefore, there was no adverse impact on mechanical components.

Impact on Technical Specifications:

Technical Specifications 3.4.1.4.1=and 3.4.1.4.2,,which specify the i

conditions required for the Residual: Heat Removal. (ND) System to perform a core cooling function, was not' adversely impacted.

Technical Specification 3.4.9.1, which specifies Reactor Coolant (NC) t System (except the pressurizer) temperature-and pressure limitations during heatup, cooldown, criticality, and inservice-leak and-i hydrostatic testing, was not violated since actual conditions reached L

(460 psig at 114~F) were within the limits-established by. Tech. Spec.

Although Technical Specifications _3.4.4-which specifies the conditions L

required for operability of the Power Operated Relief Valves (PORVs)

L and 3.4.9.3 wh4ch specifies the conditions" required for operability of-l' the Overpressure Protection System were not= observed leading-to the l

incident (described in PIR 1-C90-0094), the future ability of the L

PORVs to relieve in other modes or to provide Low Temperature Overpressure Protection (LTOP) was not adversely affected by this l

event; 1

l

LDUKE POWER COMPANY / CATAWBA _ NUCLEAR STATION

?-4 l I PIR-l-C90-0094/' Specicl R: port

, tai

~.Pagp 211

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s

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Technical Specifications 3.5.2 and'3.5.3_which;specify-the: limiting

?

conditions for operationa+ the Emergency Core CoolingL ystem (ECCS)

S Subsystems were not violatd since the Centrifugal Charging-(NV) Pump <

was unaffected,-and the ability of ND to. perform a core. cooling' '

p function was'not adversely impacted.

j t

]

conclusion Based on Design Engineering's-review of'PIR 1-C90-0094,- it wasidetermined that:

9 ND safety. relief valves responded properly _to a potential; overpressure-event:

caused by Centrifugal Charging (NV) pumps. Resulting pressures; created during.

f this eventfin the ND pump suction and discharge lines-have been evaluated. This,

review of ND piping,-valves, instrumentation, and equipment has identified,that:

there was=no adverse impact on ND System. In fact, the conservately extrapolated:

peak pressure of 625 psig reached during the event was.within' normal response-range for a relief' valve-set ~at 600 paig plus~10% accumulation."This response is.

l

-acceptable under:ASME-Section III.-

In addition, the Reactor Coolant (NC), System NDT limit for startup was~ reviewed.

i

-At no time did the NC-System exceed the pressure allowed:forJits' corresponding-temperature.

In conclusion, NC and ND Systems =are considered. operable.following=this event-.

References 1.

Flow Diagrams "r

i CN-1561-1.0 Rev 6 i:

CN-1561-1.1 Rev 3 CN-1554-1.7 Rev 5 9

2.

Telephone' conversations R.~C. Bucy with' p. W. Barrett,-

H Dennis Robinson,' Bill Hallman, Ron JonesLon-3-20-90.-,3-21-90 3.

- Duke Power Company conference call with NRC Site: Inspectors,. Region, and i

NRR 4._

Vendor Drawings n

. CNM-1201.05-318 DCD 11-9-88 CNM-1201.06-83 DCD 2-18-88 5.

Calc. by W. B. Hallman, Determination'of ND Suction Pressure-(Peak) sand Max Flow-through'1ND3 and 1ND38 During Potential Overpressure: Event of 3-20-90 i

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PIR 1-C90-0094/:SpeciCL Reporty d

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Lett.er.from'R._S'.Bondurant'to:R.-C. Bucy:concerning ND, System' Loverpreasurization,13-21-90 8.:

Memoifrom B.;J. Barbee to R.-C.'BucyI' add S. S. Lefl'er-concerning-PIR--l

'1-C90-0094', 3-22-90 4

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Telephone conference,. 3-21-90, 8 M S.'i Lefleriand R.: C.1Bucyf to :

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M;-B'. Purserfand D.;L.;Caldwell'regarding pipe? stress and~ hangers?

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