ML20134A345

Insp Rept 50-395/85-37 on 850907-30.Violation Noted: Feedwater Isolation Valve Rendered Inoperable for Approx 10 Days,Causing Isolation of Air Supply to Valve

ML20134A345
Person / Time
Site:	Summer
Issue date:	10/29/1985
From:	Cantrell F, Hehl C, Hopkins P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20134A334	List: ML20134A331 ML20134A335 ML20134A345
References
50-395-85-37, NUDOCS 8511040362
Download: ML20134A345 (12)

See also: IR 05000395/1985037

Text

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UNITED STATES

/ d uroq'o NUCLEAR REGULATORY COMMISSION

[ p REGloN li

g j 101 MARIETTA STREET,N.W.

• ATLANTA, GEORGI A 30323

%...../ r

Report No.: 50-395/85-37

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-Licensee: South Carolina Electric and Gas Company

Columbia, SC 29218

Docket No.: 50-395 License No.: NPF-12

Faci 1ity Name: V. C.--Summer

Inspection Conducted: September 7 - 30, 1985

. Inspector: b h b ,I 6 '

/0/2 7/gg

!

C.W.Hehl() Q Date Signed

0&&

Perry C. H6pkins u

&\) l0/29) $5

Da'te S'igned

Approved by: [

F. S. Cantrell, Sectioli(Cfb pf'

/ 9/b

Date Signed

Division of Reactor Proj(cts

SUMMARY

Scope: This routine, unannounced inspection entailed 385 inspector-hours onsite

in -the areas of plant tours; operatio_nal safety verifications; monthly surveil-

lance observations; monthly maintenance observations; a review of the licensee

program' for _. maintenance modifications and design changes; and a review of

operating events.

Results: -One violation was identified - with the unit at power, a feedwater

isolation valve was rendered inoperable for a period of approximately ten days.

851104036285Oh95

ADOCK

PDR PDR

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REPORT DETAILS

1. Persons Contacted'

Licensee Employees

• 0 Bradham, Director, Nuclear Plant Operations

• K. Woodward, Manager, Operations

B. Williams, Supervisor of Operations

• M. Quinton, Manager, Maintenance

• M. Browne, Manager, Technical Support .

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• B. Croley, Group Manager, Technical and Support Services

• A. Paglia, Manager, Nuclear Licensing

• H. Sefick,' Associate Manager, Station Security

• P'.LaCoe, Nuclear Licensing

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• G~ Putt, Manager, Scheduling and Maintenance

.

• R. Campbell, Engineer, ISEG

• H. :Donnelly, Senior Licensing Engineer, Licensing Engineering

'*A. Koon, Associate Manager, Regulatory Compliance

• M. Fowlkes, Engineer, Regulatory Compliance

• S. Hunt, Quality Assurance

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Other licensee employees contacted included engineers, technicians,

operators, mechanics, security force members, and office personnel.

• Attended exit interview

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2. Exit Interview

The inspection scope and findings was summarized on September 27, 1985, with

those persons indicated in paragraph I above. The inspector described the

areas. insp'ected and discussed the inspection findings. One violation was -

identified:

Violation. 50-395/85-37-01: With -the unit at power, a feedwater

isolation valve was rendered inoperable by isolation of its instrument

air supply for a period of approximately ten days.

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The licensee acknowledged the inspection findings and did not identify as

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

during the inspection.

During the exit interview, the licensee did state that they were concerned

-with a Trip Report discussing the results and findings from a site visit on

July 22 and 23, 1985, by two NRR persons and one Region II inspector

conducting a review of limited aspects of the licensee's training program.

The licensee expressed a concern that this report had been placed in the

Public Document Room without the licensee being forwarded a copy. The

inspectors acknowledged the licensee's comment.

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3. Operational Safety Verification (71707, 71710)

The inspector observed control room operations, reviewed applicable logs and

conducted discussions with control room operators during the report period.

The inspector verified the operability of selected emergency sy stems ,

reviewed removal and restoration logs, and tagout records, and verified

proper return to service of affected components. Tours of the control,

auxiliary, intermediate, diesel generation, service water and turbine

buildings were -conducted to observe plant equipment conditions including

potential fire hazards, fluid leaks. and excessive vibrations, and to verify

. that maintenance requests had been initiated for equipment in need of

' maintenance. The inspector, by observation and direct interview, verified

that the physical security plan was being implemented in accordance with the

Station Security plan.

'Within the areas inspected, no violations or deviations were identified.

4. Surveillance Observation (61726)

During the inspection period, the inspector verified by observation / review

that selected surveillances of safety-related systems or components. was

conducted in accordance with adequate procedures, test instrumentation was

calibrated, limiting conditions for operation were met, removal and

restoration of the affected components were accomplished, test results met

requirements and were reviewed by personnel other than the individual

directing' the test, and that any test deficiencies identified during the

testing were properly reviewed and resolved by appropriate management

personnel.

Within the areas inspected, no violations or deviations were identified.

5. Maintenance Observation (62703)

Station maintenance activities of selected safety-related systems and

components were observed / reviewed to ascertain that they were conducted in

accordance with regulato ry requirements. The following items were

considered in this review: the limiting conditions for operations were met;

activities were accomplished using approved procedures; functional testing

and/or calibrations were performed prior to returning components or systems

to service; . quality control record were maintained; activities were

accomplished by qualified personnel; parts and materials used were properly

certified;- and radiological controls were implemented as required.

Maintenance Work Requests were reviewed to determine status of outstanding

jobs to assure that priority was assigned to safety-related equipment which

might affect system performance.

Within the areas inspected, no violations or deviations were identified.

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6. Maintenance / Modifications and Design Changes (37700)

The inspectors reviewed four design changes and modifications, not

previously approved by NRR, (along with four that had been approved by NRR)

to ascertain that the changes had been appropriately reviewed and approved

by the licensee in accordance with section 6 of the Technical Specifications

(TS), and 10CFR50.59.

Item Area Modification No. Subject

1 Reactor Control MRF 20208 Reactor Trip Switch-

gear

~2 Reactor Coolant MRF 20249 Overpressure Protection

System Modification

3 Plant Electrical MRF 20448 Diesel Generator (DG)

Cor. trol and Relay Panel

Relocation

4 Radwaste MRF 20172 Mixer Added to Waste

Monitor Tank

5 Nuclear Blowdown MRF 20571 Nuclear Blowdown Holdup

System Tank Stabilizer

Supports

6 Protection System MRF 20354-A Design Review, Westing-

(startup) house CAB Drawing

Review

7 Reactor Heat MRF 20435 Reactor Heat Removal

Removal Calibrations Set Point

8 Electrical MRF 20560 Electrical Fire Carrier

Certification

During the review, the inspectors verified (1) that the design changes were

reviewed and approved in accordance with TS and established QA/QC controls,

(2) that design changes were controlled by established procedures, (3) that

the licensee conducted a review and evaluation of test results, that these

test results were within previously established acceptance criteria, that

any test deviations were resolved and necessary retesting was accomplished

as appropriate, (4) that operating procedure modifications were made and

approved in accordance with Technical Specifications (TS), and (5) that

as-built drawings were changed to reflect the modifications. For

modifications, the inspectors observed (1) that change activities were

conducted in accordance with the appropriate specifications, drawings, and

other requirements, (2) that acceptance and startup testing of modifications

were conducted in accordance with technically adequate and approved

procedures, and (3) the implementation of appropriate controls (e.g.,

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firewatch, portable fire fighting equipment, welding and cutting permit,

etc.). Additionally, the inspectors reviewed the outstanding facility

change requests (FCR) and determined that an excessive backlog was not

developing.

The following Technical Services Procedures and Station Administration

Procedures were reviewed:

Station Administration Procedure SAP 133, Design Control / Implementation,

Revision 3, April.25,' 1985

Technical Services TS 128, Initiation, Evaluation, and Approval of Design /

Modification Requests, March 1, 1984

Technical Services TS 129, Design Development / Design Package, March 1, 1984

Technical Services TS 130, Design Analyses and Calculations, March 1, 1984

.

Technical Services TS 131, Design Verification, March 1, 1984

Technical Services TS 132, Technical Services Disposition of Modification

Change Notices, March 1, 1984

Technical Services TS 133, Preparation of Interim Drawings / Sketches, March

1, 1984

Technical Services TS 134, Temporary Bypass, Jumper, and Lifted Lead Review,

October 12, 1984

Technical Services TS 135, Station . Design Change Interface and Implemen-

tation, March 1, 1984

A review was conducted of the Engineering Project Printouts to include

technical _ services tracking system.

It appears that the licensee has established and scheduled training and

familiarization with current and past changes / modifications etc., affecting

plant conditions 'and training for appropriate licensee personnel prior to

these personnel assuming responsibilities in the plant.

Based upon interviews with licensee personnel in engineering, licensing,

regulatory compliance, maintenance . and operations, interfaces appear to be

functioning adequately. Generally, engineering had a good handle on the

total. design / modification process.

Further, a review / sampling of records reflect adequate control at all

levels.

Within the areas inspected no-violations or deviations were identified.

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7. On-site Followup of Operating Events (93702)

During . this inspection period, the inspectors reviewed operating events

which occurred at the facility. The focus of this review was to ascertain

the safety significance of the event; evaluate performance of safety systems

and actions taken by the licensee; confirm that proper notifications of the

event, if required, had been made in accordance with 10CFR50.72; and

evaluate the need for further or continued NRC response to the event.

a. Loss of Instrument Air Event

On September 8, 1985, the licensee reported to the NRC that, at

4:40 p.m. on September 8,1985, the Turbine Driven Emergency Feedwater

Pump, (TDEFW) was determined to be rolling at approximately 200 - 250

revolutions per minute (RPM) with no TDEFW start signal present and the

steam inlet valve (XVG-2030) indicating closed. The main steam header

supply valves to TDEFW pumps A and B, from B and C steam generators

respectively, were closed and the TDEFW speed was reduced to zero. The

determination that the TOEFW pump was rolling was made by the control

operators following problems witl. the instrument air (IA) compressors

which resulted in a short term reduction in IA header pressure.

At 4:40 with the unit at 100 percent power "B" instrument air

compressor tripped on an indicated high vibration, high temperature and

high lube oil temperature. Upon indication that the "B" IA compressor

trips, the operator started "A" IA compressor. "A" IA compressor

started, but apparently failed to supply air as the IA header pressure

, continued to decrease. The operator then attempted to place a third,

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supplemental air compressor, in service. The supplemental air

compressor started, but tripped almost immediately on high discharge

pressure. At 4:47 "B" air compressor auto started on low header

pressure, control was transferred by the operators local manual

control, and IA header pressure was restored to a near normal valve of

95 PSIG. Normal IA header pressure is approximately 110 PSIG.

Subsequent to the restoration of "B" IA compressor, a mobile diesel-

driven air compressor was hooked up to the IA header, but would not

start due to a dead battery. Following stabilization of IA header

pressure, the TDEFW pump was declared inoperable at 5:30 p.m. on

September 8,1985 based on apparent excessive seat leakage through the

steam inlet valve PVG-2030 allowing the turbine to roll at low RPM.

The low RPM makes the turbine unreliable due to concerns with poor

bearing lubrication and pressurizing the Woodward governor, with the

potential for an overspeed trip following receipt of an automatic start

signal. During subsequent trouble shooting by operations to determine

the scope of the TDEFW pump problems, the pump was observed to roll

even with the-trip throttle valve closed.

Subsequent evaluation and trouble shooting determined that steam inlet

valve PVG-2030 had experienced steam cutting of the valve seat allowing

excessive seat leakage. The valve seat was subsequently repaired and

the valve returned to service. The observed phenomenon of the TDEFW

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pump rolling with the trip throttle valve closed was determined to have

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- resulted from the " normal" valve alignment of maintaining the XVT-2803

A and B, TDEFW pump turbine throttle valves drain valves open two turns

, to allow draining of moisture collecting in the above and below seat

areas of this valve. The maintaining of these drain valves two turns

open effectively short circuited the steam around the trip throttle

valve. The licensee's valve alignment procedures have been changed to

require the XVT-2803 B below seat drain to be maintained fully closed,

with any moisture collecting in this area now being drained through the

turbine casing drains. The TDEFW pump was successfully tested and

returned to service at 10:10 a.m. on September 11, 1985.

Although the above problems with the TDEFW pump were identified

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- following the September 8,1985 partial loss of instrument air, the

licensee has not been able to definitively determine whether the TDEFW

pump was rolling at low RPM prior to the event. The steam inlet valve,

PVG-2030, fails open on loss of IA, but the air system for this valve

has an accumulator to allow limited operation of the valve following a

- loss of IA. Since the as found testing of the leak tightness of this

' accumulator and~ associated piping was not performed prior to its being

disturbed during repair of valve PVG-2030, a determination of the

effect of the partial loss of IA on September 8,1985 cannot be made.

Post maintenance testing of the PVG-2030 valve included testing of the

valve's ability to function upon isolation of the IA supply; excessive

leakage in the'IA system was not identified during tnese tests. Review

of the performance of Surveillance Test Procedure (STP) 120.002, TDEFW

Pump Test, a monthly STP requiring the running of the TDEFW pump and

includes operator verification that the TOEFW turbine rolls to a stop

following closure of valve PVG-2030, indicates that on August 26, 1985 *

the turbine rolled to a stop following closure of PVG-2030. Aside from

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the performance of STP 120.002, routine periodic determination that the

TDEFW pump is not rolling, is not performed.

With' regard to the TDEFW pump reliability concerns associated with the

turbine rolling at-low RPM, prior to declaring the TDEFW operable on

September 11,1985, -the licensee's engineering organization evaluated

the bearing' lubrication concern and determined that no adverse affect

on system reliability was expected. The licensee engineering organi-

zation was pursuing but had ; not completed their evaluation of the

potential for a turbine trip on overspeed occurring as a result of the

TDEFW pump rolling at low RPM supplying oil to the governor speed

setting cylinder. The potential reliability concern associated the

governor's ability to control turbine speed during a start when

previous shaft motion was present (as was the case on September 8,

1985) is. potentially generic concern identified following a December

1984 event at Crystal River Unit 3 and discussed in IE Report

50-302/85-04.

As identified above, 'following the initial trip of the running IA

compressor "B" on September 8, 1985, apparent multiple sequential

failures prevented the installed standby and the supplemental air

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. compressor from performing their intended function. The plant

1 instrument air systems, although not a safety related system, are of

concern in their potential 1 affect on the safety related components

, served. By design, the plant instrument ai.r systems are not required

. to . function to mitigate the consequences of an accident in that all

. valves are either designed to fail to the required accident mitigation

position or are provided with air accumulators. A loss of instrument

air should not affect the ability of safety systems ' to perform their

safety functions. Unfortunately, . instrument air systems that are not

properly maintained have been known to precipitate corrosion and

' fouling related failures in safety related systems. Thus maintenance

and reliability related problems in IA systems must be examined' closely

to ascertain if they are indicative of a lack of adequate preventative

maintenance on the IA system which could precipitate safety system

failures. A review of the V.C. Summer program to assure quality of IA

determined that the IA systems were periodically monitored only by the

F Health Physics organization to assure the air met breathing air

l' requirements. The amount of moisture (dewpoint) in the IA system is

not periodically monitored to assure proper air dryer ope' ration. A

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review of the maintenance performed on the IA compressors following the

. September 8, 1985 event did not identify any failure mechanisms

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directly attributed to IA quality. Despite the absence of a program

for monitoring the quality of instrument _ air, no apparent degradation

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of safety related components has resulted.

b. September 20, 1985 Reactor Trip

On September 20, 1985, the licensee notified the NRC that at 8:58'p.m.

a reactor trip from 100 percent power had occurred in response to a

turbine trip initiated by a trip of all main feedwater (MFW) pumps.

The MR/ pump trip initiated during the conducted of a special test to

determine condensate pump net positive suction head (NPSH) in

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preparation for a planned modification to convert these pumps from

variable speed to constant speed units. Following the reactor trip,

the plant responded as expected with the exception that the MFW

isolation valve XVG-16118 failed to close in response to the expected

automatic feedwater isolation on low reactor coolant average

temperature coincident with P-4 (reactor trip breakers open), and

intermediate range nuclear instrumentation channel (NI) 36 failed

downscale. The feedwater isolation valve and NI-36 were repaired and

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tested and the unit was returned to power on September 22, 1985.

The condensate pump test in progress which resulted in the reactor trip

was conducted to determine the NPSH of the condensate pumps at design

pump speed and maximum pump speed by taking manual speed control of

each pump and increasing pump speed while monitoring pump parameters.

To avert a trip of the condensate pumps on high discharge pressure

during the testing, the pressure transmitter in the common condensate

pump discharge header was isolated. The isolation valve for this

transmitter apparently allowed sufficient leakage to allow- the

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transmitter to sense the elevated header pressure expected during the

testing-and-1nitiate a trip of the condensate pumps.

Subsequent . evaluation of the above identified failure of N1-36

downscale, following the reactor trip, was determined to have resulted

from excessive detector compensating voltage. The compensating voltage

was adjusted,'the channel tested and was returned to an operable status

-prior to plant restart.

During plant startup following this reactor trip, a misaligned

instrument block equalizing valve for flow transmitter IFT 3571

rendered the emergency feedwater (EFW) flow control valve IFV 3546

incapable of automatically isolating EFW flow from the TDEFW pump.

This finding and a discussion of the failure of the main feedwater

valve XVG-1611 B to close as identified above are discussed in detail

below,

c. Misaligned Equalizing Valve

On September 20, 1985, the licensee found emergency feedwater header

flow transmitter IFT 03571 inoperable because of an open equalizing

valve on the transmitter manifold. With this transmitter inoperable

-isolation of emergency feedwater to steam generator B on high. flow was

also inoperable.

The licensee identified the discrepancy and took immediate action to

rectify the situation. The design analysis descriptions mitigate the

conse'quences of such an event. That the licensee's Emergency Procedure

further ' mitigates the situation, and by review of related documents

indicate that this is a highly isolated case'with minimum consequences.

There appears to be no programmatic deficiencies, relative to this

occurrence, therefore, no notice of violation will be issued.

d. Failure of MFW Isolation Valve to Close

Following the reactor trip on September 20, 1985, MFW isolation valve

(FWIV) XVG-16118 failed to close in response to a feedwater isolation

signal on low reactor coolant average temperature coincident with

permissive reactor trip breakers open. Subsequent licensee evaluation

determined that the manually operated instrument air supply valve to

valve XVG-16118 had been isolated and the associated air receivers had

bled down to a pressure insufficient- to allow valve closure. The

instrument air supply valve was determined to have been closed on

September 10, 1985, in conjunction with the disposition of a previously

identified problem in a portion of the valve hydraulic circuit not

required for FWIV 1611B to close in performance of its accident

mitigation function.

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The FWIV at V.C. Summer use hydraulic pressure to open and close the

valve. Instrument air pressure is necessary to position a 4-way valve

that directs the hydraulic pressure to the under or above valve

actuating piston area. The 4-way valve does not move and the FWIV

fails as is on loss of instrument air. The valve operator has two

redundant air / hydraulic circuits labeled active and stand-by. The

active side circuit alone receives signals from the Engineering Safety

Features (ESF) system to provide FWIV functions required for safe

shutdown and accident mitigation. The inactive side receives signals

for waterhammer and steam generator baffle bolting protection as well

as a permissive for forward flushing (feedwaterline warming) opening.

~ Thus the accident mitigation functions are performed by the active side

alone. This particular design' the valve receiving only a train "A"

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ESF signal, is an accepted configuration in that the feedwater

regulating valves receive both train "A" and train "B" ESF signals and

the main feedwater pumps receive a train "B" ESF (safety injection)

trip signal. The use of non-safety related components (feedwater

regulating valves and the main feedwater pumps) for redundant accident

mitigation in a steam break senario has been accepted by the NRC as

discussed in NUREG 0138, page 1 - 11.

The previous problem with the FWIV noted above which resulted in the

isolation of IA to valve XVG-1611 B involved a defective 4-way valve in

the stand-by hydraulic circuit. Pending replacement of this defective

component at the next shutdown the licensee performed an engineering

evaluation in accordance with 10CFR50.59 and determined that since the

active side of the hydraulic circuit was unaffected by the defect, the

valve remained operable. The 10CFR50.59 evaluation, attached to

Non-conformance Notice (NCN) 2032, determined that the FWIV would meet

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its functional requirements if the active side (train "A") hydraulic

accumulator is pressurized to at least 4800 psig.

Although not addressed in the disposition of the inactive side problem,

the defective 4-way valve caused the inactive hydraulic circuit to be

at a reduced pressure resulting in the air driven hydraulic pump

running constantly in an attempt to change this side. To reduce

unnecessary wear on this pump, on September 10, 1985, a decision was

made to isolate IA air to the pump and establish a twice per shift

surveillance of the pressure in the hydraulic accumulator in the active

side to ensure pressure was maintained via the hydraulic accumulator

check valves. The IA supply isolation valve was subsequently closed.

Review of the IA isolation aspect of this event and the decision

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process was indeterminate in that documentation was not available to be

examined. The IA valve was not removed from service under the tag out

program. Its closure was not addressed on the disposition of NCN 2032.

Thus apparently no formal evaluation was performed or deemed necessary

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at the time. Concern 'over the bleed down of the air accumulators was

apparently not considered. It is noteworthy to mention that currently

there is no installed instrumentation to monitor air reservoir

pressure.

By design, a loss of IA (a non-safety system) should not preclude the

ability of this valve to perform its accident mitigation function. _ Air

reservoirs (accumulators) with associated check valves are provided to

maintain the required approximately 75 psig air pressure necessary to

position the 4-way valves to allow valve closure. With IA isolated,

either leakage through the IA reservoir check valves or other FWIV

components allowed the IA pressure to bleed down to a pressure that was

insufficient to close valve XVG-1611B. No data is available to.

determine the actual cause of this air bleed down as the licensee

performed no leakage testing on valve XVG-1611B's air circuits prior to

replacement of the defective inactive side 4-way vahe. Subsequent

" snooping" of the IA circuits on September 24, 1985, did not identify

any leakage.

Technical Specification' (TS) 3.7.1.6 requires that with the plant in

Modes 1, 2 or 3, each feedwater isolation valve be operable. For a

component to be considered operable, TS definition 1.18 specifies that

it must be capable of performing its specified functions. One of the

safety functions of the FWIV is to close and isolate feedwater in

response to an ESF signal calling for feedwater isolation.

As identified above, during the period from about September 10,

1985, until September 20, 1985, FWIV XVG-1611B was rendered inoperable

fo.r performance of its safety functions by isolation of its instrument

air supply. During this period, the unit was operated at 100's power

and when called upon to function to limit cooldown of the reactor

coolant system subsequent to the September 20, 1985 reactor trip, it

did not function. This is a violation (50-395/85-37-01).

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On April 30, 1985, the NRC issued Information Notice (IEN) 85-35,

Failure of Air Check Valves to Seat". IFN 85-35 discussed a

potentially significant problem concerning Parker-Hannifin check valves

supplied as part of the valve operators on Anchor / Darling Main Steam

Isolation Valves (MSIV) and FWIV. Tnese check valves, which are

installed on the IA supply 'line to the air reservoirs, have been

demonstrated to fail to seat during a slow depressurization of the IA

supply line. As a result the MSIV or FWIV air resersoir would bleed

down rendering the valve incapable of being closed. IEN 85-35,

Attachment 2, identified V.C. Summer Sta_ tion as a plant with FWIV

supplied by Anchor / Darling Company with these Parker-Hennifin air check

valves.

Inspection review of the licensee's evaluation and actions associated

with this IEN determined that the licensee's Nuclear Engineering Group

had determined that FWIV-1611A, B and C had been supplied with these

type check valve as identified in their August 21, 1985 letter to the

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plant and recommended that these check valves be replaced at the

- "ea rl iest opportunity". This initial evaluation of IEN 85-35 by

Nuclear Engineering did not address or put forward any justification

for continued operation of the facility pending replacement of these

check valves. In light of the above bleed down of the air reservoirs

for valve XVG-1611B, and the lack of data to exclude failure of the

~ check valve to seat as a cause, the licensee on September 25, 1985

completed an engineering evaluation to justify continued operation

until the valves are replaced. NRC review of this evaluation was still

in progress. Replacement of these valves is scheduled during the

October - November 1985 refueling :;utage. Completion of NRC review of

this justification and replacement of valves is an Inspector Followup

Item 50-395/85-37-02.

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