ML20214V651

Insp Repts 50-369/86-21 & 50-370/86-21 on 860621-0820. Violations Noted:Reactor Coolant Cold Leg Injection from Charging Pumps Failed to Meet Design Criteria & Auxiliary Feedwater Pump Automatic Start Interlock Inoperable

ML20214V651
Person / Time
Site:	McGuire, Mcguire
Issue date:	09/18/1986
From:	William Orders, Peebles T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20214V591	List: ML20214V588 ML20214V604 ML20214V651
References
50-369-86-21, 50-370-86-21, IEB-85-003, IEB-85-3, IEB-85-321, NUDOCS 8610020357
Download: ML20214V651 (7)

See also: IR 05000369/1986021

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UNITED STATES

['grS Ff4q#o

. NUCLEAR REGULATORY COMMISSION

[ REGloN ll

j j 101 MARIETTA STREET.N.W.

• '- r ATLANTA, GEORGI A 30323

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Report Nos: 50-369/86-21 and 50-370/86-21

Licensee: Duke Power Company '

422 South Church Street

Charlotte, NC 28242

Facility Name: McGuire Nuclear Station

Docket Nos: 50-369 and 50-370 License Nos: NPF-9 and NPF-17

l Inspaction Conducted: June 21 through August 20,'1986.

Inspector: /M. _b f/ M

W. T. ~ 0'rders , Sen esident y p'ector 4) ate S4gned

Accompanying Personnel: S. F. Guenther

i- Approved by: ,, //d

T. A. Peebles, Section Chief

7-/ 8 W

.Date Signed

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} Division of Reactor Projects

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SUMMARY

Scope: This routine unannounced inspection was conducted on site in the areas of

operations, surveillance testing, maintenance activities, and event follow-up.

Results: Of the areas inspected, one violation was identified in each of three

areas: surveillance testing, maintenance activities, and equipment operability.

The two violations which are being considered for escalated enforcement are:

1. Violation (369,370/86-21-01) TS 3.5.2 requires two independent emergency

core cooling system (ECCS) subsystems, each with a coolant charging pump, to

provide high head safety injection protection. NI-9 and NI-10, reactor

coolant cold leg injection from the charging pumps, of both units failed to

meet design requirements.

TS 3.6.3 requires, in part, that NV-7, the outside containment letdown

isolation valve, on both units be operable with an isolation time of less

than 10 seconds, while in Modes 1, 2,~3 and 4 (paragraph 8).

2. Violation (369,370/86-21-02) TS 3.3.2, Table 3.3-3 (Engineered Safety

Features Actuation System Instrumentation) requires that the motor-driven

auxiliary feedwater pumps' automatic start interlock on a trip of all main

feedwater pumps be operable while in Modes 1 ard 2.

The 2B motor-driven auxiliary feedwater pump's " loss of all main feedwater

pump" automatic start interlock was inoperable (paragraph 9).

8610020357 860718

PDR ADOCK 05000369

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REPORT DETAILS

1. Persons Contacted

Licensee Employees

• T. McConnell, Plant Manager

• B. Travis, Superintendent of Operations

• D. Rains, Superintendent of Maintenance

• B. Hamilton, Superintendent of Technical Services

• L. Weaver, Superintendent of Administration

• M. Sample, Superintendent of Integrated Scheduling

• N. McCraw, Compliance Engineer

• N. Atherton, Licensing and Compliance

Other licensee employees coltacted included construction craftsmen,

technicians, operators, mechanics, security force members, and office

personnel.

• Attended exit interview.

2. Exit Interview

The inspection scope and findings were summarized on August 29 with those

persons indicated in paragraph 1 above. The licensee did not identify as

proprietary any of the materials provided to or reviewed by the inspectors

during this inspection.

3. Unresolved Items

No new unresolved items were identified during this inspection period.

4. Plant Operations

The inspector reviewed plant operations during the report period to verify

conformance with applicable regulatory requirements. Control room logs,

shift supervisors logs, shif t turnover records and equipment removal and

restoration records were routinely perused. Interviews were conducted with

plant operations, maintenance, chemistry, health physics, and performance

personnel.

Activities within the control room were monitored during shifts and at shift

changes. Actions and/or activities observed were conducted as prescribed in

applicable station administrative directives. The complement of licensed

personnel on each shift met or exceeded the minimum required by Technical

Specifications (TS).

Plant tours taken during the reporting period included but were not limited

to the turbine buildings, auxiliary building, Units 1 and 2 electrical

equipment rooms, Units 1 and 2 cable spreading rooms, and the station yard

zone inside the protected area.

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During the plant tours, ongoing activities, housekeeping, security, equip-

ment status and radiation control practices were observed.

Unit 1 Operations- ,

Unit I remained in a refueling outage throughout the inspection period.

Unit 2 Operations

Unit 2 began the report period in the final stages of a. refueling outage.

The unit was placed on line on June 28 and operated virtually unencumbered

until July 22, 1986 when it tripped on high pressurizer pressure. Details

of the reactor trip are delineated in Licensee Event Report (LER) 370-86-12,

and are currently under review. The unit returned to service the following

day and' completed the month at 100% power.

The unit operated until August 12, when it tripped subsequent to the inad-

vertent de-energization of a vital DC bus. Details of the trip are cur-

rently under review. The licensee will submit specifics associated with

the trip in an LER on or about September 12. The unit was restarted and

operated through the end of the report period with no major operational

difficulties.

5. Surveillance Te' sting

The surveillance tests categorized below were analyzed and/or witnessed by

the inspector to verify procedural and performance adequacy and conformance

with applicable T S. The selected tests witnessed were examined to ascer-

tain that current- written approved procedures were available and in use,

that test equipment in use was calibrated, that test prerequisites were met,

system restoration completed and test results were adequate.

Procedure Number Test

PT 2 A 4350 02 A 2A D/G OPERABILITY TEST

PT 2 A 2060 01 A 2A NI PUMP TEST

PT 2 A 4252 01 A 2A CA PUMP TEST

PT 2 A 4350 02 8 2B D/G OPERABILITY TEST-

PT 2 A 4204 11 28 ND PUMP AHU TEST ,

Apparent inadequacies associated with safety-related surveillance activities

were detected during this ~ report period. Details are delineated elsewhere

in this report.

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6. Maintenance Observations

Routine maintenance activities were reviewed or witnessed by the resident

inspector to ascertain procedural and performance adequacy and conformance

with applicable T S. The selected activities witnessed were examined to

ascertain that, where applicable, cu. rent written approved procedures were

i available and in use, that prerequisites were met, equipment restoration

completed and maintenance results were adequate.

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During a review of an event which transpired on' June 25, 1986, inadequacies

associated with maintenance on DC electrical circuitry were detected ~. These

apparent inadequacies are detailed elsewhere in this report.

7. Refueling Activities

Selected refueling activities on Unit I were reviewed and/or witnessed to

verify that:

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periodic testing and verification of the operability of refueling-

related equipment and systems were performed as required by TS and

licensee administrative procedures;

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fuel handling operations were performed in accordance with TS and

approved procedures;

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plant conditions were maintained as required by TS; and

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the licensee's staffing was in accordance with TS and approved pro-

cedures.

The inspector reviewed and/or witnessed selected portions of the following

activities to ascertain that they were being controlled and conducted as

required by TS and approved procedures:

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core monitoring during refueling operations;

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fuel accountability methods;

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vessel and spent fuel storage pool water levels;

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boron concentrations and appropriate dilution path valve check perfor-

mance;

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checks of containment penetrations;

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checks of decay heat removal system flowrate; and

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operability of the containment purge and exhaust isolation systems.

8. Inadequate Valve Operators

On May 7,1986, during a review conducted in accordance with NRC Bulletin

IE-85-03, it was determined by _ licensee personnel that the electric motor /

operators (EMO) for Unit 1 valve INI-10, and Unit 2 valves 2NI-9, and 2NI-10

. (Reactor Coolant Cold leg Injection from the Charging Pumps) were insuffi-

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ciently sized to guarantee . opening of the valves under worst. case design

conditions. Further, the torque switch setpoints for Units 1 and 2 valves

. NV-7 (Reactor Coolant Letdown Outside Containment Isolation) were found to

! be incorrectly specified. On May 8, 1986, during a visual inspection,it was

discovered that valve 1NI-9 also had an inadequate EMO installed. Unit 1

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was in Mode 1, Power Operation, at 80% power and Unit 2 was in Mode 6,

Refueling, at the time of the discovery. . Witi Unit 1 in power operation

status, a temporary modification was made to the motor operators of valves

INI-9 and INI-10 to ensure that sufficient operator torque was available to

open the valves as an interim solution to allow continued operation.

The Unit 2 motor operators for valves 2NI-9 and 2NI-10 were replaced and the

torque switch for valve 2NV-7 was set as required. Valve INV-7 was found to

have its torque switch setting actually set appropriately.

This issue was originally identified as an unresolved item (369/86-12-01),

awaiting the completion of the licensee's investigation, and the completion

of the inspection. With this report, the unresolved item is closed, having

been upgraded to a violation (369,370/86-21-01)

Tersely stated, the desi.gn specifications for the above valves are as

follows:

Valves NI-9 and NI-10, on both units, serve as redundant isolation

valves that admit.high pressure safety injection flow from the centri-

fugal charging pumps (CCPs) to the reactor coolant system. The normal

position of the valves is closed and their safety function is to open

within 10 seconds of actuation. The valves are designed to open

against a maximum differential pressure of 2735 psi. The motor

operators for the valves are designed to operate at 43 rpm and rated

to deliver'190 ft-lbs torque output.

Valve NV-7 (on Units 1 and 2) provides outside containment isolation of

the reactor. coolant system normal letdown flowpath. The valve is

required to close within 10 seconds of the actuation of a safety

injection signal . The valve is a three inch globe valve manufactured

by Walworth with an electric motor operator. The valve is designed to

close against a maximum differential-pressure of 600 psi. The motor

operators for the valves are designed to operate at' 43 rpm and rated

to deliver 100 ft-lbs of torque output.

Briefly restated, the problems identified with the above detailed valves

are as follows:

Valves INI-9, INI-10, 2NI-9 and 2NI-10 had . their operators replaced

with higher speed, lower torque units which rendered the valves

incapable of performing their intended function under the " worst case

design conditions." *

Valves INV-7 'and 2NV-7 had licensee specified torque switch settings

which were lower than required and specified by the valve manufacturer

to assure that the valves would perform their intended function.

Valve IN'V-7 was actually found with its torque switch set at the

manufacturer's specified .value, which was in disagreement with the

licensee's specification.

More complete details relevant to this issue are delineated in LER 369/86-09.

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The apparent problem areas associated with this issue are as follows:

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(a) Inadequate design control which allowed.the motor operators on valves

NI-9 and NI-10 to be replaced with deficient components.

(b) Inadequate design control which allowed the torque switch setpoints for

valve NV-7 on both units to be incorrectly specified.

(c) The actual setpoint of valve NV-7 on Unit 1 was not set in accordance

with the licensee's design document.

(d) Valves 1NI-9,1NI-10, 2NI-9 and 2NI-10 were not capable of operating

under. the worst case design conditions from 1980 (Unit 1) and 1983

(Unit 2) until May 1986.

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(e) Valve 2NV-7 was not capable of operating under the worst case design

conditions from 1983 until May 1986.

TS 3.5.2 requires two independent emergency core cooling system (ECCS)

subsystems, each with a coolant charging pump, to provide high head safety

injection protection. With NI-9 and NI-10 of both units unable to meet

design requirements, the ECCS systems have been unable to perform their

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designed function for specific types of projected accidents since 1980 on

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Unit 1 and 1983 on Unit 2.

9. Event Follow-up

a. On June 25,. 1986, Unit 2 was being started up with one condensate

booster pump (CBP 2A) and one main feedwater pump (CF pump 2A) running.

A trip of CBP 2A on a low condensate flow signal caused CF pump 2A to

trip which should, in turn, have caused both motor-driven auxiliary

feedwater pumps (CA pumps 2A and 28) to receive automatic start

signals. CA pump 2A automatically started but CA pump 29 did not

automatically start. Personnel in the control rocm responded

immediately tar manually starting CA pump 28.

Licensee personnel investigated the cause for CA pump 2B not automa-

tically starting and found normally closed DCB breaker 68, which

i supplies power to start the pump on the loss of both CF pumps, in the

open position; the breaker was promptly closed.

The licensee could not determine, with certainty, when DCB breaker 68

had been left in the open position. The breaker was known to have been

opened on June 11, 1986, when ground fault testing was performed on a

i number of breakers, including DCB breaker 6B. The work was performed

under Work Request No. 66150 and the procedure for Locating Grounds on

De-energized Electrical Circuits- (IP/0/A/3061/25). The procedure

required the technician's and the accompanying engineer's signatures to

independently verify that they were at the correct breaker and that the

, Senior Reactor Operator's permission had b~een obtained to de-energizing

l each individual load. The procedure also directed that Operations

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personnel restore power to the affected load using operating

procedures, however, no independent verifications were required to

ensure that the breakers were returned to their pretest positions.

The incident of June 25 disclosed that CA pump 2B was incapable of

! starting automatically upon a trip of all main feedwater pumps, as

required by TS 3.3.2. In accordance with TS 1.18 (Operability), CA

pump 28, one of the three independent steam generator auxiliary feed-

water pumps required to be operable in Modes 1, 2 and 3 per TS 3.7.1.2,

i must be considered to have been inoperable from June 11, the last date

on which DCB breaker 6B was known to have been' closed, until June 25,

the date on which DCB breaker 6B was reclosed after being found in the

open position. During the intervening 14-day period, none of the

associated TS Limiting Condition for Operation Action statements were

executed. This constitutes a violation of TSs 3.7.1.2 and 3.3.2

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(370/86-21-02).

b. On June 23, 1986, while Unit 2 was in Mode 3, Instrument and Electrical

(IAE) personnel were performing PT/0/A/4601/07A, the semiannual

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Response Time Testing of Reactor Trip Breakers procedure. This pro-

cedure has usually been performed while the unit was in Mode 1, but the

j procedure is written to allow it to be performed in -all Modes. While

j performing the test, IAE personnel opened reactor trip breaker RTA and

then opened bypass breaker BYA as directed by the procedure. The

procedure did not require control room personnel to block the feedwater

isolation signal. With breakers RTA and BYA open, the feedwater

isolation signal not blocked, and average' reactor coolant temperature

less than ~ 564 degrees Fahrenheit, the Solid State Protection System

(SSPS) initiated an Engineered Safeguard Features (ESF) train A feed-

t water isolation signal. Since Unit 2 was in Mode 3, with neither main

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feedwater pump operating, the only automatic action which occurred was

the closing of the feedwater containment isolation valves. Control

room personnel reopened the isolation valves and IAE personnel success-

fully completed the testing of the reactor trip breakers.

PT/0/A/4601/07A provided inadequate control of a Reactor Protection

System (RPS) surveillance test activity by allowing IAE technicians to

open both the RTA and BYA breakers without first having control room

personnel block the automatic feedwater isolation which would occur at

l the existing average reactor coolant temperature on June 23. This

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constitutes a violation of TS 6.8.1.a (370/86-21-03).

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