Text

Insp Rept 50-483/87-23 on 870908-11.Violations Noted.Major Areas Inspected:Licensee Action Following 870709 Surveillance Test Which Indicated That Both Independent Control Room Emergency Ventilation Sys Inoperable
	ML20235J798
Person / Time
Site:	Callaway
Issue date:	09/25/1987
From:	Gill C, Greger L; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	;
Shared Package
ML20235J789	List: ML20235J788; ML20235J798;
References
	50-483-87-23, NUDOCS 8710020100
	Download: ML20235J798 (13)
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U.S. NUCLEAR REGULATORY COPMISSION

REGION III

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Report No. 50-483/87023(DRSS)

Docket No. 50-483 License No. NPF-30

Licensee: Union Electric Company

Post Office Box 149

St. Louis, MD 63166

Facility Name: Callaway County Nuclear Station

Inspection At: Callaway Site, Callaway County, Missouri

Inspection Conducted: September 8-11, 1987

Inspector: C. F. Gill b.Y h AT h' _

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Approved By: L. R. Greger, Chief ,5

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Facilitios Radiation Dat6

Protection Section

Inspection Summary

Inspection on September 8-11, 1987 (Report No. 50-483/87023(DRSS))

Areas Inspected: Special, announced inspection of licensee action following a

July 9,1987 surveillance test which indicated that both independent Control

Room Emergency Ventilation Systems were inoperable. l

Results: The licensee's failure to have both independent Control Room Emergency

Ventilation Systems operable violated regulatory requirements (Section 4). 1

The appropriate enforcement action for this failure will be determined and l

communicated to the licensee by separate correspondence,

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DETAILS

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1. Persons Contacted

R. Affolter, Superintendent, Systems Engineering

J. Blosser, Plant Manager

H. Bono, QA Engineer

R. Cage, Outage Supervisor, Modifications

W. Campbell, Manager, Nuclear Engineering

G. Czeschin, Superintendent, Planning

D. Hollabaugh, Supervising Engineer, Systems Engineering

L. Kanuckel, QA Supervising Engineer

J. Laux, Superintendent, QA Technical Support

J. Little, QA Assistant Engineer

C. Naslund, Manager, Operations Support

K. Schweiss, System Engineer

T. Sharkey, Supervisor, Compliance

W. $heppard, Superintendent, Outages

B. Stanfield, QA Engineer

D. Wingbermueble, Supervising Engineer, Modifications

R. Wink, Systems Engineer

B. Little, NRC Senior Resident Inspector

The inspector also contacted other licensee and contractor employees.

Denotes those present at the exit meeting.

2. General

This inspection which began at 3:00 p.m. on September 8,1987, was

conducted to review the circumstances surrounding a July 9, 1987

surveillance test which indicated that both independent Control Room

Emergency Ventilation Systems were inoperable while the plant was in

operational Mode 1 (power operation).

3. Licensee Event Reports (LER) Followup

Through direct observations, discussions with licensee personnel, and

review of records, the folinwing event report was reviewed to determine

that deportability requirements were fulfilled, immediate corrective

action was accomplished, and corrective action to prevent recurrence had

been accomplished in accordance with Technical Specifications. The LER

listed below is considered closed: ,

Description

LER NO.

87013 Inoperable Control Room Emergency Ventilation

System and T/S 3.0.3 Entry When Personnel Error

Caused Pressure Boundary Breach

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The LER was reviewed as part of the inspection into the apparent inability

of both independent Control Room Emergency Ventilation Systems to meet

their design requirements; this matter is discussed in Section 4.

4. Inability of both Independent Control Room Emeroency Ventilation Systems

to Perform their Design Requirements

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a. Event Summary ,

Contract maintenance personnel breached' electrical penetration seals

located along the control room (CR) pressure boundary.to support a

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design change to area radiation monitoring system annunciators on

about June 8 June 11, and July 6, 1987. On July 9,1987, the '

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!- licensee performed a scheduled 18-month technical ~ specification

surveillance test to verify that Control Room Emergency Ventilation

System (CREVS) Train A maintained the CR at a positive pressure.cf !

greater than or equal to .25 inch water gauge (wg) relative to the j

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outside atmosphere as required by Technical Specification (T/S) i

Surveillance Requirement 4.7.6.e.3. The surveillance test measured

positive pressure was approximately .01'" wg. The plant was in

operational Mode 1 (power operation) from June 8 through July 9,1987. j

After four breached electrical penetration seals were found and

plugged with a temporary seeling cumpound, the CR pressure increased,

but only to approximately .15" wg; at 2000 CDT on July 9,1987,

approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the surveillance test began, CREVS

Train A was declared inoperable. A fifth breached electric

penetration seal was discovered at 2350 CDT; the licensee concluded

at this time that breached electrical penetration seals were a common

mode failure to both CREVS trains. Since the inoperability of both

CREVS trains is prohibited by T/S 3/4.7.6, entry into T/S 3.0.3

was declared by the licensee. At 2400 CDT, the fifth penetration

was plugged with a temporary sealing compound, CR pressure increased '

to .26" wg, CREVS Train A was declared operable, and T/S 3.0.3 was

subsequently exited. Or July 10, 1987, between 0200 and 0235 CDT,

CREVS Train B was tested; the CR pressurized to .62" wg. The. licensee .

then made the assumption that Train B had sufficient capacity to.

accommodate the previously discovered breached electrical penetration

seals, and the previous entry into T/S 3.0.3 was declared to have been

in error. (This was an erroneous conclusion as noted below.) 1

On July 13, 1987, an engineering evaluation was initiated by the f

licensee to confirm that CREVS Train 8 had been operable on July 9, j

1987, during the period when the CR pressure boundary electrical

penetration seals were breached. On Aug'ust 6,1987, the engineering.

evaluation determined that the breached electrical penetration seals,

identified on July 9, 1987, rendered both trains of the CREVS

inoperable per T/S 3/4.7.6. A four-hour ENS phone call was made to

the NRC at 1135 CDT pursuant to 10 CFR 50.72(b)(2)(iii).

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On a licensee followup inspection, one CREVS' Train A isolation damper

(GKD0283) was.found to be leaking but still operational; the licensee

believes this leakage did not significantly contribute to the failure

on July 9, 1987. After damper repair on August 13, 1987, the

surveillance test on CREVS Train A measured a control room positive

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pressure ofi.28".wg.

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b. Event Causation

The following occurrences contributed to the failure of the Control

' Room Emergency Ventilation Systems to meet their design requirements:

(1) The nuclear safety evaluation (10 CFR 50.59 review) of Callaway

Modification Package (CMP No. 85-0178), which was prepared in

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response to Work Order No. 7451' to add reflash capability to

area monitor system annunciators, failed to consider the effect

which CR pressure boundary breached electrical penetration seals *

would have on the ability of the CREVS to maintain the CR at the

required positive pressure of .25 inch wg relative to the.outside

atmosphere. The breaching of the CR pressure boundary electrical

pen (tv-ation seals during this event, and earlier reportedly less

significant events, was only considered from a fire protection

viewpoint.

(2) The woik requests associated with CMP No. 85-0178 failed

to indicate that this modification could piece the plant in

violation of a technical specification LCO (T/S 3.7.6).

Licensee and contractor personnel . apparently failed to

recognize that the electrical penetration seals were located

at the CR pressure boundary and that breaching them would

therefore potentially effect. compliance with T/S 3/4.7.6. ;

(3) As confirmed by the licensee on August 6,1987, the evaluation J

made on July 9, 1987, that CREVS Train 8 was operable between

June 8 and July 9,1987, was faulty. The licensee should have f

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entered T/S 3.0.3 as soon as the common mode CR breached

electrical penetration seals were discovered on July 9, 1987.

c. Corrective Actions

(1) As noted in Section 4.a above, all five breached electrical

penetration seals were plugged with a temporary sealing

compound. Trains A and B of CREVS were tested and found to

maintain adequate control room positive pressure (0.26" wg and

0.62" wg, respectively). The licensee plans to replace the

temporary electric penetration seals with permanent seals during

the September / October 1987 refueling outage while the plant is

in operational Mode 5 (cold shutdown) or Mode 6 (refueling).

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(2) On July 24, 1987, an Operations Night Order was issued by the

licensee to prohibit CR boundary breachs until furthet-

investigation was completed.

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(3) .The licensee stated that this event was discussed with the

personnel involved and that personnt) involved in planning

design changes and corrective maintenance also reviewed this

event.

(4) - On September 4,1987, portions of the Callaway Civil Structural

Manual (CSM) were revised to incorporate the lessons learned

from this event into the post pour installation ins.tructions.

Specific revisions to the CSM include the post pour installation

form (PPIF)'to include system engineering review and sign-off

for penetrations through pressure boundaries, additional

instructions for completing the PPIF, and the addition of the

control room, fuel building, and the auxiliary building to the :

engineering criteria and guidelines for post-pour . installations.

(5) By mid-October 1987, the licensee expects to complete administrative

procedural revisions to accommodate fire barriers in a manner similar

to that discussed above for post pour installations.

d. Control Room Personnel Dose

In 'an attempt to estimate the theoretical radiological consequences

to control room personnel under the CREVS inoperability condition

found during the T/S surveillance test on July 9, 1967, the licensee

requested a consultant (Bechtel) to calculate Callaway control room

personnel doses during a LOCA without control room pressurization.

The Bechtel calculation was completed on September 11, 1987. Assuming

a containment leak rate of .051% per day (Callaway containment

integrated leakage rate test, January 1984). for the first' day and

one-half that rate thereafter,- Bechtel calculated a total LOCA

control room personnel thyroid dose of 20.62 rem (17.62 rem due to

containment leakage and 3.0 rem due to ECCS leakage).

Assuming the FSAR specified containment leak rate value of 0.2%/ day

.(FSAR Table 15A-1 and T/S 3.6.1.2) a thyroid dose of approximately

72 rem instead is predicted based on the Bechtel calculation. FSAR

Table 15.6-8 lists the control room personnel thyroid LOCA dose as

18.7 rem, also assuming a containment leak rate value of 0.2%/ day;

thus, the consequences of the analyzed accident to control room

personnel increase significantly if the CR is not maintained at an

adequate positive pressure.

In addition to significantly exceeding the FSAR specified control

room personnel thyroid LOCA dose (18.'7 rem), the 72-rem thyroid dose

exceeds the design basis for the CREVS. ,The Basis for T/S 3/4.7.6

states that the operability of the CREVS, in conjunction with

control room design provisions, is based on limiting the radiation

exposure to personnel occupying the control room to 5 rems or less

whole body, or its equivalent, and that thit, limitation is consistent

with the requirements of General Design Criterion 19 of Appendix A,

10 CFR Part 50. Standard Review Plan (NUREG-0800) Section 6.4,

" Control Room Habitability Review," states that a thyroid dose of

30 rem is compatible with the GDC-19 dose guideline.

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I. e. Safety Review Requirement

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l The Callaway Final Safety Analysis Report (FSAR), Section 6.4.5

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states that the emergency mode of the control room HVAC system will

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maintain a .25 inch wg positive pressure in the emergency zone.

Table 6.4-2 of the FSAR states, in response to Regulatory Guide 1.95

Position a.3, that the.1ow-leakage construction details for maintaining ,

the control room pressure at .25 inch wg are shown in FSAR Figure 6.4-1. ]

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FSAR Figure 6.4-1, Sheet 3 shows the typical detail sealing of cable

L tray penetrations which include specifications to fill the areas

between cables and cable tray and between the cable tray and the metal '

penetration sleeve with silicone foam.

As noted in Section 4.b above, when contract maintenance personnel

breached penetration seals in the control room pressure boundary on

about June 8. June 11, and July 6,1987, changes were made in the

facility as described in the safety analysis report without prior

Commission approval and the changes involved an unreviewed safety

question. With the penetration seals breached, the Control Room

Emergency Ventilation System trains were unable to achieve a control

room positive pressure of greater than or equal to .25 inth wg

relative to the outside atmosphere during system operation. The

reduced pressure in the control room would allow higher unfiltered

inleakage than that evaluated in the FSAR accident analysis, thus

increasing the consequence of the accident with regard to control

room personnel doses.

This is a violation of 10 CFR 50.59(a)(1) which prohibits a licensee

from making changes in the facility as described in the safety

analysis report, without prior Commission approval, if the proposed

change involves an unreviewed safety question. 10 CFR 50.59(a)(2)

states, in part, that a proposed change shall be deemed to involve

an unreviewed safety question if the consequences of an accident

previously evaluated in the safety analysis report may be increased.

(483/87023-01)

One violation was identified.

f. CREVS Operability Requirement

Technical Specification 3.7.6 Limiting Condition for Operation

states in part, that two~ independent Control Room Emergency

Ventilation Systems shall be operable during Modes 1, 2, 3, and 4.

Technical Specification Surveillance Requirement 4.7.6.e.3 defines

operability, in part, as the ability of,the systems to maintain the

control room at a positive pressure of greater than or equal to

.25 inch wg relative to the outside atmosphere during system

operation. The Basis for. Technical Specification 4.7.6 states, in

part, that the operability of the Control Room Emergency Ventilation

System ensures that the control room will remain habitable for

operations personnel during and following all credible accident

conditions.

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As noted in Section 4.a above, between approximately June 8,1987 and

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July 9,1987, the unit remained in Mode 1 although both independent

Cuntrol Room Emergency Ventilation Systems were inoperable during this

period in that neither system could maintain the control room at a

positive pressure of greater than or equal to .25 inch wg relative

to the outside atmosphere during system operation.

Inasmuch as the action statement for Technical Specification 3.7.6

does not address continued operation in modes 1, 2, 3, or 4 with both

CREVS' inoperable, Technical Specification 3.0.3 applied and required

that the unit be placed in mode 5 or 6 within a specified time frame. ,

This action was not taken by the licensee until July 9, 1987, beyond l

the specified time frame of Technical Specification 3.0.3. This is a

violation of Technical Specifications 3.7.6 and 3.0.3. (483/87023-02)

One violation was identified.

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5. Additional Regulatory Concerns

Although, this inspection concentrated on a review of the licensee's

failure to have both independent Control Room Emergency Ventilation

Systems operable when required by technical specifications, other

regulatory concerns were noted by the inspector and discussed with the

licensee. These additional regulatory concerns have not been reviewed f

in detail, thus they will remain unresolved items pending further review

during a future inspection.

a. Evidence was found which indicates that a CREVS Train 8 balancing

damper (GKD0324) was adjusted, without a work request, sometime

between the CREVS Train B pre-operational test and the first 18-month

T/S surveillance test. Damper GKD0324 is shown in SNUPPS design

drawing M-2H1521Q as required to be adjusted to produce a flow

of 350 cfm from the control room A/C duct to the control room A/C

and filtration room. The following information infers that this ;

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safety-related damper was improperly reset in a position contrary to

design specifications without formal authorization and QA review.

The pre-operation test report for CREVS Train B on May 25, 1984

showed a CR positive pressure of .26" wg. The first 18-month

technical surveillance test report for CREVS Train B on

March 4, 1986 showed a CR positive pressure of .62" wg. Damper

GKD0324 is located in the CREVS Train B, such that if the

damper reduces the flow rate into the CR A/C and filtration

room, the CR pressure will increase; the change in control room

pressure implies a change in the sgtting of damper GK00324.

At the inspector's request, the licensee completed a preliminary

search for a work request, or any other authorization, to change

the design setting of damper GK00324 for the time period between

the pre-operational and first T/S surveillance CREVS Train B

CR pressure tests; no documentation was found.

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On August 12, 1987, on a licensee CREVS system followup

inspection after LER 86-013-00 was issued, the licensee found

that CREVS Train 8 damper GKD0324 was virtually closed which

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produced very little flow from the CR A/C duct into the CR A/C

and filtration room.

Pending further review, this matter is considered an Unresolved Item.

This matter was discussed at the exit meeting and will be reviewed

further during a future inspection. (483/87023-03)

b. When CREVS Train B damper GKD0324 was found to be improperly closed

on August 12, 1987, an attempt was made to adjust the damper to

produce the design requirement flow of 350 cfm. It was found that

the required flow rate could not be achieved without reducing the CR

positive pressure below the T/S required .25" wg. The damper was

left in a position which produced a flow rate of 160 cfm into the CR

A/C and filtration room and a CR pressure of approximately .4" wg.

The licensee appears to have left an ESF component in a configuration

contrery to design specifications. In response to the incident, the

licensee issued Incident Report No.87-173, issued Request for

Resolution (RFR) No. 4210, and put Work Request No. W104769 on hold

pending RFR No. 4210 resolution. Pending further review, this

matter is considered an Unresolved Item. This matter was discussed

at the exit meeting and will be reviewed further during a future

inspection. (483/87023-04)

c. Because this inspection concentrated on the CREVS operability aspects

regarding the CR pressure boundary breached electrical penetration

seals associated with CMP No. 85-0178, which was prepared in response

to Work Order No. 7451 to add reflash capability to area monitor

system annunciators, the fire protection acceptability of the

modification was not reviewed. Pending further review, this matter

is considered an Unresolved Item. This matter was discussed with the

Plant Manager, and others, by telephone on September 18 and 21, 1987,

and will be reviewed further during a future inspection.

(483/87023-05)

Three unresolved items were identified. i

6. Exit Meeting

The inspector met with licensee representatives (denoted in Section 1) at

the conclusion of the inspection on September 11,1c87 and with the Plant

Manager, and others , by telephone on September 18 and 21,1987. The

inspector summarized the scope and findings pf the inspection, including

the unresolved items and the apparent violations. The inspector also

discussed the likely informational content of the inspection report with

regard to documents or processes reviewed by the inspector during the

inspection. The licepsee did not identify any such documents / processes

as proprietary.

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RP 0516B

. 5 ALP FUNCTIONAL AREA ASSES $ MENT AND PRELIMINARY INSPECTOR EVALUATION FORM

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Inspection Report No.: fd-(f3/89023

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l MAINTENANCE l l l l l l l l l l Pl

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l 50RVE]LLANCE l l l l l l l j, l l Rl

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l FIRE PROTECTION l l l l l l l l l -l Tl

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l EMERGENCY PREPAREONE55 l l ! l l l l l l l Of

l l l 1 1 I I I I I l Nl

l SECURITY l l l l l l l l l l 5l

l 1 l. I I I I I l l l l

l GUTAGES l l l l l l l l l l l

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l QUALITY PROGRAMS & ADMIN. l l l l l l l l l l

l CONTROLS AFFECTING QUALITY I l l l l l l l l l

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l LICENSING ACTIVITIES l l l l l l l l ,l l

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l TRAINING & QUALIFICATION l l l l l l l l l l

l EFFECTIVENESS l l l l 1 l l l l l

l 50lLS AND FOUNDATIONS l l l l l l l l l l l

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l CONTAINMENT, SAFETY-RELATED l l l l l l l l l l 0l

l STRUCTURES & MAJOR STEEL SUPPORT 51 l l I l l l l l l Nl

l PIPING SYSTEMS & SUPPORTS l l l l l l l l l l $l

l l l I I I I I I l l Tl

l SAFETY-RELATED COMPONENTS l l l l l l l l l l RI

l MECHANICAL l l I I l l l l l l Ul

l AUXILIARY SYSTEMS l l l l l l l l l l Cl

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l ELECTRICAL EQUIPMENT AND CABLES l l l l l l l l l l Il

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l INSTRUMENTATION l l l l l l l l l l Nl

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Afunctional areas for Construction and Operations

CRITERIA FOR DETERMINING CATEGORY RATING

1. Management Involvement in Assuring Quality.

2. Approach to Resolution of Technical Issues from a Safety Standpoint.

3. Responsiveness to NRC Initiatives.

Enforcement History.

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4. -

5. Operational and Construction Events. MW

6. Staffing (including management).

RATING KEY: (For Categories 2 - Declining and 3, provide narrative basis for

conclusion)

Category 1 Category.2 - Declining

Category 2 Category 3

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( spector(s) concerns adequately addressed o_r

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l l Inspection Evaluation Form being processed. g

Lead Inspector CA41.7, h/ 9/ 7, Section Chief I M E /k g

- (Signature) - (Ofte) [ Signature) (DateykJ5/h '

March 3, 1986

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l OTHER SIGNIFICANT ITEMS

1. Syatens and Components

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2. Facility Items

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3. Managerial Items

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U. S. NUCLEAR REGULATORY COMMIS$10N

REGION III

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Report No. 50-483/87028(DRP) j

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Docket No. 50-483 License No. NPF-30

Licensee: Union Electric Company

Post Office Box 149 - Mail Code 400

St. Louis, MO 63166

Facility Name: Callaway Plant, Unit 1

Inspection At: Callaway Site, Steedman, M0

Inspection Conducted: August 24 through September 11,1987

Inspector: B. H. Little

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Approved By: J. . Hinds, hief o9.zs.B7

Reactor Projects Section 1A Date

Inspection Sumary

Inspection on August 24 through September 11, 1987 (Report No.

50-483/87028(DRP))

Area inspected: A special unannounced safety inspection by the Senior

hesident Inspector regarding the essential service water system operability.

Results: One apparent violation was identified (failure to promptly identify

ard correct an incorrect essential service water valve position - paragraph 2).

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DETAILS

1. : Persons Contacted

'J. D. Blosser, Manager, Callaway Plant

J. R. Peevy, Assistant Manager, Technical Services

M. E. Taylor, Superintendent, Operations

J. C. Gerhart, Superintendent, Quality Assurance Operations Support

T. P. Sharkey, Supervisor. Compliance-

T. H. McFarland, Superintendent, Design Control

C. D. Naslund, Manager, Operations Support

J. L. Cunningham, Shift Supervisor

We F. Stubblefield, Superintendent, Administrative Services

Denotes those present at one or more exit interviews.

In addition, other members of the operations and engineering staffs were

contacted,

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2. Inspection of the Essential Service Water System Operability

a. Background

On August 15, 1987, the licensee's operating crew determined that

valve EF-V-0117 (Essential Service Water [ESW) train "B" to the

UltimateHeatSink[ UHS)isolationvalve)waspartiallyshut,

resulting in a "B" train flow rate of 11,000 gpm. The design flow

specified in the Final Safety Analysis Report (FSAR) is 13,594 gpm.

The valve actuator position indicator showed that vtive EF-V-0117

was in the locked open position; however, the valve stem indicator

indicated that the valve was partly closed. This valve position

mismatch was previously identified on Work Request (WR) No. 024695

dated May 14, 1984. It appears that knowledge of the valve's l

position deficiency was lost through subsequent WR reissue, review ]

oversights, and WR voiding.

On August 20, 1987, an engineering evaluation concluded that,

although it was functional, the ESW "B" train had been inoperable

since 1984 on the basis that the FSAR design flow rate had not

been met, and that the plant had unknowingly entered Technical

Specification (T/S) 3.0.3 during occasional periods when the ESW

"A" train had been inoperable. Upon making the inoperability

determination, the licensee notified the NRC of the event using the

Emergency Notification System (ENS).

A special safety inspection by the senior resident inspector was

performed to assess the licensee's activities associated with the

incorrect position of the ESW isolation valve.

The inspection included the following:

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Review of associated documentation: plant records, work

requests, a temporary niodification, surveillance procedures,

and an incident report.

Interviews with licensee personnel from the following

departments: Compliance, Engineering, Planning, Scheduling,

and Operations.

b. Inspection Findings

(1) Event Chronology

April 1984 The ESW system preoperational test was

completed.

May 11, 1984 The manual actuator on valve EF-V-0117

(WR No. 8517) was replaced. The WR

specified that the valve be stroked as

a " retest."

May 14, 1984 WR No. 024695 was issued to adjust the

actuator stops of valve EF-V-0117. The

WR identified that the " valve would not

shutoff" ar.d that the " position indicators

on gland (stem) and on manual operator

contradict each other." Condition Tag

No. 1556 was hung on EF-V-0117.

May 25, 1984 The retest (stroking) of valve EF-V-0117

(WR No. 8517) was deleted. Followup action

was specified as WR ho. 024695. 1

June 30, 1984 The ESW B" pump baseline test (ASME

Section XI) was performed.

August 10, 1984 The plant first entered Mode 4 (ESW required

operable per T/S).

May 16, 1986 WR No. 58783 was issued to repair the broken

output shaft on the manual operator for

valve EF-V-108 (E5W "A train isolation

valve). The WR was revised to install

Temporary Modification (TM) No. 86M-041 in

lieu of repair of the valve actuator.

May 17, 1986 TM No. 86M-041 was installed on EF-V-108.

June 10, 1986 WR No. 59421 was written to remove the

manual operator from valve EF-V-0117 for

inspection to assist the engineering

evaluation of the broken operator on

EF-V-0108. The initiating document was

Nonconforming Material Report (NMR) No.

86-I-00328.

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September 9, 1986 WR No. 024695 was voided; the reason

provided was " work not required per system

engineer." This voiding may have been done

because the later issued WR No. 59421 would

negate the adjustmer.ts. The deficiency

relating to the indicator mismatch was not

documented on the later WR and was apparently

overlooked. Voiding the WR resulted in

removal of the Condition Tag No. 1556 and

precluded the licensee's ability to track

and resolve the valve EF-V-0117 position

deficiency.

September 10, 1986 WR No. 59421 was voided; the reason stated

as "not required per new disposition of NMR

86-1-00328." The initial disposition was

to repair the valve operator (EF-V-108)

onsite; the disposition subsequently changed

to send the operator to the manufacturer for

failure analysis and replacement or repair.

October 24, 1986 Work on WR No. 58408 was completed. The

work performed included the removal of TM

86M-041 from and the replacement of the

actuator on EF-V-10E.

August 15, 1987 EF-V-0117 was found partially closed by

equipment operators during verification of

the ESW valve lineup. The valve lineup was

in response to observed low pressure values

during the containment cooler flow verifica-

tion test.

Valve EF-V-0117 was opened, and the correct

valve position was verified by stem indica-

tion and system flow. The event was

documented in Incident Report (IR) No.87-148.

August 20, 1987 An engineering evaluation of ESW "B" train

operability concluded that the ESW "B"

train had been functional but inoperable

l since 1984, and that the plant had

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unknowingly entered T/S 3.0.3 on those

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occasions when the ESW "A" train was

I inoperable for routine surveillance

testing and maintenance.

The licensee notified the NRC of the event

using the ENS.

(2) Document Review

The SNUPPS FSAR, Chapter 9 provides the description of the ESW

l system and the ESW system flow requirements. During normal

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plant operations, the ESW receives water from the Service

Water System (SWS) and supplies water to the safety-related

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components and air compressors. After cooling the equipment,

the heated water is returned to the SWS. Following a design

basis accident or a loss of offsite power, the safety-related

signals will isolate the ESWS from the SWS by closing the j

associated motor-operated isolation valves. Also, the ESW {

pumps will automatically start receiving power from the j

preferred power supply or from the emergency diesel generators i

and supply water from the ultimate heat sink to the safety- )

related components and air compressors. The minimum flow rates j

required to remove heat from the containment and necessary 1

safety-related components from a postulated loss of coolant

accident (LOCA) or main steam line break and dissipate it to

the ultimate heat sink are listed in Table 9.2-3 of the FSAR.

The minimum flow requirement specified for ESW train "B" is l

13,594 gpm.

Technical Specification (T/S) Limiting Condition For Operation

(LCO) 3.7.4 requires that "at least two independent essential I

service water (ESW) loops shall be operable" while in Modes 1,

2, 3, and 4. T/S action requires that "with only one ESW loop

OPERABLE, restore at least two ESW loops to OPERABLE status

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ."

The Callaway Plant first entered Mode 4 on August 10, 1984. {

T/S surveillance requirements do not require ESW train or system !

flow verification. The specified surveillance associated !

with the ESW system have been satisfactorily performed. These J

include monitoring of ESW pump performance (obtained by closing ]

down on the pump discharge valve), verification of containment )

cooler flow rates, and routine verification of valve position

(excluding locked valves). Operation Department Procedure

ODP-4 (Locked Component Control) specifies that for locked open ]

valves the handwheel or operator be moved in the close direction j

enough to verify valve movement. Then the valve is returned to l

its original position. This procedure has routinely been l

performed each plant outage. The operator indicator, the most l

visible and convenient indication of valve position, is

utilized during thir activity. The valve position mismatch

would not likely be identified during performance of this

procedure.

On August 15, 1987, the ESW train "B" degraded flow condition

was identified during an ESW syrtem valve lineup review. The

lineup was performed in response to low pressure readings

observed during performance of the containment cooling water

flow surveillance (T/S 4.6.2.3.a.2). Operating crew members

found valve EF-V-0117 locked open as indicated by the valve

actuator position indicator, but actually partially shut (valve

stem indication). The as-fougd ESW train "B" flow rate (control ,

room flow meter) was 5.5 x 10 Lbm/hr (approximately 11,000 )

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gpm). The valve was opened and verified open by correct flow

indication. Incident Report (IR) No.87-148 was issued to

document the event. f

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A preliminary engineering evaluation of the as-found flow

rate was performed to assess ESW train "B" operability. Based

on the evaluation, the licensee determined.that the degraded

flow condition resulted in train "B" being inoperable but

functional since 1984. Based on the preliminary engineering

assessment and infomal discussions with' Bechtel, the licensee

believes that the as-found flow rate was adequate to perform

the specified safety function, although less than the FSAR

design value. The licensee has authorized Bechtel Corporation

to perform a detailed safety analysis.

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Through'the review of plant records and interviews with

licensee personnel, the inspector was not able to detemine

the actual date valve EF-V-0117 was incorrectly positioned.

However, a satisfactory pre-operational test of the ESW system

was performed in April 1984. A subsequent ESW "B" pump

baseline test, perfomed in June 1984, provides pump data

which indicates that the valve was partially shut. There was

no requirement to compare "A" train and "B" train data. Such a

comparison would have detected the "B" train flow deficiency.

Multiple work requests were issued to document deficiencies on

valves EF-V-0117 and EF-V-0108 (see Event Chronology). It

appears that some WRs were voided to eliminate duplicate WRs,

and in the process all deficiencies were not retained on

replacement WRs. WR No. 024695, issued May 14, 1984, documented

the deficient valve position indication; however, the deficiency

was not corrected, and on September 9, 1986, the WR was voided.

The incorrect actuator position indicator caused the valve to

be placed in, and locked in, a partially closed position and

resulted in a degraded flow condition of ESW train "8". The

degraded flow condition existed from approximately May 1984 to

August 15, 1987, and resulted in an unanticipated reduction in

the safety margin. The licensee's failure to promptly correct

the valve position deficiency is a violation of 10 CFR 50 l

Appendix B, Criterion XVI, " Correct Action."

3. Licensee's Corrective Action

Completed Action

Tbe valve actuator has been caution tagged to alert operators to

the problem and to ensure that the local indicator on the valve

stem is used for position indication until the actuator is repaired

during Refuel II.

Engineering personnel i~volved

n in voiding the WRs have been

counseled. It was emphasized that thorough research be done prior

to authorizing the voiding of any WR. Thorough research prior to

voiding a WR is presently a plant-wide accepted policy. This would

include retests. In addition, it is now required by procedure that

the reason for voiding a WR be given along with the name of the

person voiding the WR. This will also provide traceability between

WRs.

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Planned Action l

This event will be added to the required readine list for all

system engineers. Engineers involved with ASME Section XI u

evaluations were reminded to recognize the effect on the total i

system when re-establishing pump baselines. This event will be

reviewed by the planners.

The failure of personnel to properly prioritize the work on the l

valve indicators initially is considered an isolated case. To

provide additional assurance that this is an isolated case, a review '

of voided and open WRs will be performed. This review will ensure

that operability concerns have been properly identified and prior-

itized.

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A review of the WR control program indicated weakness in the WR

voiding process. The current' procedure requires additional l

information on the reason for voiding work documents. In addition, '

the procedure will be revised to require that appropriate

information is transferred from a voided WR to the current WR which j

implements previously uncompleted work. '

An ESW System Total Flow Verification test will be performed during

Refuel II to verify total ESW system flows for trains "A" and "B" ,

in the event of a LOCA. -4

4. Safety Significance

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The ESW system consists of two independent 100-percent capacity trains. I

The ESW "A" train was fully operable while the ESW "B" train flow was j

restricted, except for durations of approximately one day per month {

to support maintenance and surveillance testing. A preliminary assess-

ment by the licensee indicates a high probability that the "B" train

flow rate was adequate to perform the specified safety function.

The event posed no significant threat to the public or plant safety,

based on the independent capacity of the ESW trains and the low

probability of an event involving the ESW system occurring during the

periods that only the "B" train was available. However, the event is

considered significant based on the licensee's failure to control and

correct a known deficiency in a timely manner, the unanticipated

reduction in margins of safety, and the entry into a condition prohibited

by Callaway Technical Specifications.

5. Exit Interview

The inspector met with licensee representatives (denoted under Persons

Contacted) at intervals during the inspection period. The inspector

summarized the scope and findings of the inspection. The licensee

representatives acknowledged the findings as reported herein. The

inspector also discussed the likely informational content of the

inspection report with regard to documents or processes reviewed by

the inspector during the inspection. The licensee did not identify

any such documents / processes as proprietary.

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