ML20151Z882

Insp Repts 50-369/88-20 & 50-370/88-20 on 880625-0722. Violations Noted.Major Areas Inspected:Operations Safety Verification,Surveillance Testing,Maint Activities & Followup on Previous Insp Findings

ML20151Z882
Person / Time
Site:	Mcguire, McGuire
Issue date:	08/19/1988
From:	Croteau R, David Nelson, William Orders, Peebles T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20151Z872	List: ML20151Z868 ML20151Z874 ML20151Z882
References
TASK-1.B.1.2, TASK-TM 50-369-88-20, 50-370-88-20, NUDOCS 8808300240
Download: ML20151Z882 (13)

See also: IR 05000369/1988020

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION

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, REGION il

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j 101 MARIETTA STREET. N.W.

's ATLANTA, GEORGIA 30323

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Report Nos. 50-369/88-20 and 50-370/88-20

Licensee: Duke Power Compary

422 South Church Street

Charlotte, NC 28242

Facility Name: McGuire Nuclear Station 1 and 2

Docket Nos.: 50-369 and 50-370

License Nos.: NPF-9 and NPF-17

Inspection Conducted: June 25,1988 - July 22,1988

i Inspectors: #1/f/ b , ~ [

DatV Signed

W.O'rdirs, Senior)residentInspector

, WJ W f

4atV Signed

jDT tM1's'oT1, ~ResidptInspector

,/ A1W/

T. Cr'~tiau, Ris dent Ins ector

f

Da~tV Signeid

Approved by: /A + MbAn

T, K. Pedbles,~ Section Chief'

, 8"//4 @

Date' Signed

Division of Reactor Projects

SUMMARf

Scope: This routine unannounced inspection involved the areas of operations

safety verification, surveillance testing, maintenance activities,

and follow-up on previous inspection findings.

Results: In the areas inspected, two violations were identified. Activities

observed indicate a weakness in determining system status during ,

outages (see paragraph 8). A strength was noted in licensee response

to inspector concerns generated from events occurring at other

facilities (see paragraph 10).

Within the areas inspected, the following violations were identified:

Inadequate procedure / failure to follow procedure with respect to

draining of steam generator s without blocking the automatic start of

an auxiliary feed pump, and to removing the 28 off site busline from

service without an adequate procedure and without properly aligning

switch gear assemblies.

Failure to meet the intent of Technical Specifications with respect

to the McGuire Safety Review Group surveillance of plant activities.

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g ADOCK 05000369

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REPORT DETAILS

1. Persons Contacted

Licensee Employees

• N. Atherton, Compliance

D. Baxter, Operations

L. Bost, Design Engineer

J. Boyle, Superintendent of Integrated Scheduling

S. Copp, Planning Engineer

• J. Day, Compliance

• J. Foster, Health Physicist

• B. Hamilton, Superintendent of Technical Services

• S. LeRoy, Licensing, General Office

T. McConnell, Plant Manager

W. Reeside, Operations Engineer

• M. Sample, Superintendent of Maintenance

R. Sharp, Compliance Engineer

• A. Sipe, Safety Review Group Chairman

• J. Snyder, Performance Engineer

• B. Travis, Superintendent of Operations

R. White IAE Engineer

Other licensee employees contacted included construction craftsmen,

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

personnel.

"Attended exit interview

2. Unresolved Items

An unresolved item (UNR) is a matter about which more information is

required to determine whether it is acceptable or may involve a violation

or deviation. There were no unresolved items identified in this report.

3. Plant Operations (71707, 71710)

The inspection staff reviewed plant operations during the report period to

verify conformance with applicable regulatory requirements. Control room

logs, shift supervisors' logs, shift 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.

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Plant tours taken during the reporting period included, but were not-

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

electrical equipment rooms, Units 1 and 2 cable spreading rooms, Unit 2

reactor building, .and the station yard zone inside the protected-area.

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The Unit 2 reactor builaing was walked down prior to entry into mode 4  ;

following the refueling outage. Several minor deficiencies were noted l

which the licensee took action to correct. Overall - cleanliness was

adequate.  ;

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Ouring the plant tours, ongoing activities, housekeeping, security,

equipment status _and radiation control practices were. observed.

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a. Unit 1 Operations

Unit 1 began the period at full power. On June 26 at 12:05 p.m., the

unit sustained a 50% load rejection due to the loss of one main feed

pump. A non-safety load center, SMXL, de-energized due to several

electrical grounds. This load center powers the main feed flow

elements which control the feed pumps' recircult +. ion valves. The

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de-energized flow elements indicated zero flow causing the recire ,

valves to fully open. This created an actual high flow condition

causing feed pump suction pressure to decrease. One feed pump

tripped on low suction pressure. Sufficient suction prt ssure

] remained to supply the remaining feed pump. Operaters Gnually '

star'.ed both motor driven auxiliary feed pumps to assist in

maintaining steam generator levels during the transient. The unit

returned to full power early the next day. Except for several brief '

periods of load following, the unit remained at full power for the

remainder of the period, l

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1 b. Unit 2 Operations

Unit 2 began the period in the end-of-cycle 4 refueling outage which

commenced on May 27. The outage is on schedule with a planned return

to service on July 27.

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On June 24, with the unit defueled, a loss of off site power

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occurred. In preparation for maintenance on the B off site power

j switch gear, plans were made to realign incoming power such that the

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A busline would be supplying all loads. Normally, each busline

supplies two 6.9 KV buses, but alternate alignments can be made to ,

allow a single busline to supply all four 6.9 KV bases. T o prepare  !

for the maintenance, the two 6.9 KV buses normally supplied by the B

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busline were to be aligned to their alternate source, the A busline.

With this intent, the control room operator mistakenly aligned the

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two 6.9 KV buses normally supplied by the A busline to their

alternate source, the B busline. Therefore, the B busline, not A,

was supplying all electrical power. The B busline Primary Circuit

Breakers (PCB) were then opened from the switchyard causing all off

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site power to be lost to Unit 2. Both emergency diesel generators

(DG) automatically started to supply power to the emergency _ buses, y

however, the A DG tripped on an indicated overspeed condition. (The-

licensee determined later that an actual overspeed did not occur, but a

was simulated by blockage _ in a pressure instrument sensing line.

This problem was corrected during on going outage maintenance.)' _No

other abnormalities occurred during the event. l

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The licensee considers the cause of~the loss of off site power to be-

attributed to the operator's error. .The operator was performing a

section of operating procedure OP/2/A/6350/05, AC Electrical

Operation Other Than Normal Lineup, when initially aligning the 6.9

r 'KV buses. This procedure provides generic instructions for switching

any of the four 6.9 KV buses to their alternate power supply. It i

refers to the 6.9 KV buses as the "applicable" or "respective" buses i

and to the byslines as the "normal" or "standby" power. supplies so

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that this single procedure can provide instructions for any of the

numerous lineups possible. In order to successfully use this

procedure, the operator must have additional- instructions, either j

written or verbal, to provide the desired final lineup. In this case t

the desired lineup was provided by the operator's supervision. The

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licensee considers that the instructions to the operator were  ;

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accurate. The operator understood the intended alignment, but

performed the opposite action than he had intended to perform. The

control board mimic bus for this system is accurate, removing any  !

cause for confusion as to which breakers should have been operated to ,

make the desired lineup. A second operator made a similar error in

that an entry into the reactor operators' log was made stating that  :

- the lineup was about to be performed. The lineup described was the l

same incorrect lineup that the first operator was making. *

, The general operating proceaure used by the operator does not provide  ;

)i instructions for operating the PCBs to de-energize the busline once  ;

the initial lineup is complete. Investigation by the inspector  :

i revealed that a second document, Removal and Restoration (R&R) tagout

28-616, was specifically prepared for this maintenance. This

document provided specific sequence and components for tagging in

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order to ensure that the B busline remained de-energized during the

,_ maintenance. R&R's are considered by the licensee to be procedures

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and are to be followed as such. As stated in Operations Management '

Procedure 2-17, Tagout/ Removal and Restoration (R&R) Procedure, one

, of the purposes of R&Rs is "to allow the removal and restoration of )

equipment to be accomplished in a specific manner by directing the

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sequence of the steps involved in repositioning the equipment and

indicating the desired removal and return position." In this case,

the R&R did not direct the sequence of steps involved to' achieve the

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desired alignment but only provided specific instructions for tagging

j once the busline was de-energized. It referred to the generic

! instructions contained in the operating protetture (presumably to

j align the 6.9 KV buses) but did not provide the desired alignment of

1 the buses nor did it provide for opening the PCBs.

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The NRC considers that the operator did not follow the operating

procedure _ in that the incorrect 6.9 KV buses were aligned to their '

alternate power supplies. Also both the operating procedure and the

R&R procedure were inadequate in that neither specified the desired

alignment of the 6.9 KV buses nor provided for opening the PCBs. A~

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more specific procedure would have aided the operator: in the-

repositioning of equipment. This constitutes two examples of a

violation of TS 6.8.1 for failure to follow procedures and for an

inadequate procedure. (Violation 370/88-20-01) 1

1 While draining the Unit 2 steam generators on July 18 to establish

proper chemistry for heatup, the auto start signal to the turbine

driven auxiliary feedwater (TDCA) pump was received opening the steam

admission valves SA-48 and SA-49 and isolating blowdown. The Unit

was'in mode 5 at the time with no steam pressure so the TDCA pump did ,

not inject water into the steam generators.

The steam generators were being drained in accordance with

OP/2/A/6250/MA, Steam Generator Cold Wet Layup Recirculation. Step  :

2.10 of the prot.edure directed IAE to defeat the feedwater isolation '

signals but no step was included to defeat the TDCA auto start signal

on low-low steam genvator level. The procedure was subsequently

changed to block the auta start signal and the systems were returned

to normal. ,

This event demonstrates that OP/2/A/6250/03A was inadequate in that .

performance of the OP resulted in an unplanned ESF actuation. This  !

constitutes a third example of a violation of TS 6.8.1 for an

inadequate procedure (Violation 370/88-20-01). l

4. Surveillance Testing (61726) l

4 Selected surveillance tests were analyzed and/or w;enessed by the l

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inspector to ascertain procedural and performance adequacy and conformance I

with applicable Technical Specifications.

Selected tests were witnessed to ascertain that current written approved

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procedures were available and in use, that test equip 6ent in use was

2 calibrated, that test prerequisites were met, that system restoration was ,

completed and test results were adequate. 1

Detailed below are selected tests which were either reviewed or witnessed: i

MP/0/A/7150/7 On Line Ice Basket Weight Determi.a. ation Process

PT/0/A/4200/18 Ice Bed Analysis

TT/2/A/9100/269 DG Starting Air / Instrument Air Blackout Header

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Test

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No violations or deviations were identified. I

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5. Maintenance Observations (62703)

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Routine maintenance activities were reviewed and/or witnessed by the s

resident inspection staff to ascertain procedural and performance adequacy

and conformance with applicable Technical Specifications.

The selected activities witnessed were examined to ascertain that, where-

applicable, current written approved procedures were available and in use,

that prerequisites were met, that equipment restoration was completed and

maintenance results were adequate.

No violations or deviations were identified.

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6. Licensee Event Report (LER) Followup (90712,92700)

The following LERs were reviewed to determine whether r'aporting require- ,

ments have been met, the cause appears accurate, the corrective actions

appear appropriate, generic applicability has been considered, and whether

the event is related to previous events. Selected LERs' were chosen for

! more detailed followup in verifying the nature, impact, and cause of the

event as well as corrective actions taken.

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(CLOSED) LER 370/86-02: Reactor Trip on Intermediate Range High Flux  !

Signal During Unit Shutdown. This item is associated with Unresolved Item

370/85-46-01 discussed in paragraph 7 of this report. This item is

closed.

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(CLOSED) LER 369/86-14: Train B Safety Injection In Mode 5. This event

, involved an inadvertent safety injection while testing reactor trip

breakers. The licensee has made appropriate procedure changes to prevent  !

recurrence. This item is closed.

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7. Follow-up on Previous Inspection Findings (92702)

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l The following previously identified items were revirwed to ascertain that

the licensee's responses, where applicable, and licensee actions were in

compliance with regulatory requirements and corrective actions have been

completed. Selective verification included record review, observations,

and discussions with licensee personnel.

(CLOSED) Unresolved Item 370/85-46-01: Intermediate Range Detector Not

Calibrated Following Replacement. This item involved a reactor trip

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caused by an intermediate range detector during a plant shutdown. The

detector at fault had been recently replaced and subsequent calibration

was deemed unnecessary by the licensee due to the increased sensitivity of

the new detector. As a result, the intermediate low power reactor trip

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i set points were not verified. The safety significance of this event was

small in that more conservative set points resulted. The Itcensee has

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changed the procedure for detector replacement to require setpoint

calibration in all cases. This item is closed,

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8. Maintenance of System Status During Outages

During this reporting period two problems developed in which inadequate

maintenance of system status contributed to the cause: ,.

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The first example involved an inadequate containment integrity verifi-

cation. Prior to commencing Unit 2 fuel load (entering Mode 6) a

containment integrity verification was performed. During this process

valve NV-245, charging line containment irolation valve, was determined

to be shut by the operator's observation cnat the Operator Aid Computer

r (OAC) indicated that it was shut. Neither the red or green control board

indicators were illuminated which prompted the operator to consult the j

OAC. The licensee considers this to be an acceptable practice in normal

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situations. The valve, however, was undergoing maintenance, was disassem-

bled and was unable to provide containment integrity. The OAC indicated

the valve to be shut because the valve actuator remained in the shut

position - following disassembly. Fortunately, a check valve in the same i

line provided a second containment integrity barrier. Upon discovery of

the problem, Operations took action to revise the containment integrity

procedure to require additional verification if the OAC is the only means

available for determining valve position. This corrective action is

appropriate for this specific occurrence, but does not address the

! underlying problem of entering an operational mode without meeting all

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the mode change requirements. In this case, Operations did not have means

for determining that all maintenance required to be complete was actually  ;

completed. '

The second example involved a test of the Unit 2 diesel generator starting

air system. During the Unit 2 outage, a test (TT/2/A/9100/269) was

performed to assess the diesel starting air (VG) system's ability to I

supply the blackout header of the instrument air (VI) system. The first J

phase of this test consisted of lining up VG to the VI blackout header in i

a static mode and monitoring VG and VI pressure to determine the VI demand I

on VG. The test was deemed successful when very little VI demand i.e. l

leakage occurred. The licensee had expected more significant leakage and '

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was surprised at the results. The VI blackout header supplies pressurized

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air to numerous air operated valves whose actuators are known to'have air

leaks. Regardless of the unexpected results, plans were undertaken to i

i assess system performance in a dynamic state. The inspector conducted a I

walkdown of the portions of the systems involved with the test and

discovered that most of the VI blackou't header was isolated for outage

maintenance. Since most of the valve actuators supplied by the blackout

header were isolated, the test did little more than demonstrate that the ,

VI piping did not leak. The licensee was unaware of the system status {

prior to the test. When it was reperformed later in the outage, I

significant',y dif ferent results were obtained.

In these two examples it is obvious that actual system status was not

i known. The NRC considers the licensee's inability to accurately track

maintenance during an outage to be a weakness. The licensee has

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recognized this weakness and stated that steps are being taken to improve,

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According to the licensee the "Projects 2" computer program, when fully

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implemented, will provide a means by which outage maintenance items can be

tracked to assess the status of the maintenance at any time. The NRC will ,

continue to monitor tha licensee's performance in this area.

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9. NAMCO Limit Switch Environmental Qualification Concerns

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A concern arose at the Catawba Nuclear Station regarding the environmental

l qualification (EQ) of NAMCO limit switches. Identical limit switches are

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installed on numerous safety and non-safety related vahes at McGuire.

The concern was that incorrectly installed gaskets on the switch covers

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would permit moisture to enter the switch internals thereby possibly

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shorting electrical contacts. A consequence could be that safety related

valves that shut on an Engineered Safety Features (ESF) actuation would

reopen when the safety signal is reset instead of remaining in the safety

position.

The licensee conducted an inspection of all installed NAMCO limit switches

on Unit 2. Four were found with incorrectly installed gaskets and

repaired. Design Er;ineering analyzed that these four would have had no

adverse impact dur i ng an ESF actuation.

Due to Unit 1 being in operation, not all switches were accessible for ,

inspection. Therefore, the licensee conducted a consequence analysis for .

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i all Unit 1 inaccessible limit switches. In one case compensatory measures

l were taken: Valve 1RV76A is one of several valves that provide Phase B f

inside containment isolation for the containment ventilation headers. The

seismically designed boundary for this system is limited to the piping l

between these valves and the outside containment isolation valves.

Assuming a seismic event and a single active failure of the outside

containment isolation valve (s), the potential exists for radioactive

release if these valves failed to remain shut upon resetting the Phase B '

containment isolation signal. The licensee made compensatory Emergency

Procedure changes to verify that 1RV76A remains shut after Phase B reset. '

Additionally, the corresponding outside containment isolation valve will i

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be assured shst prior to the phase B signal being reset. The licensee

l stated that 1RV76A will be inspected and/or repaired at the next Unit 1

trip or during the next outage. All remaining inaccessible limit switches '

were deemed to be of no risk to safety by the licensee's consequence

analysis. All inaccessible switches will be inspected either at the next

unit trip or during the next unit outage. Several accessible Unit i limit

switches were found with incorrectly installed gaskets and were repaired t

but in each case the licensee determined through consequence analysis that

there were no past safety concerns. The licensee documented all  :

conclusions in a Justification for Continued Operation (JCO).

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10. Emergency Core Cooling System (ECCS) Sump Line Verification

During the inspection period, the Resident Inspection Staff informed the

McGuire plant manager of an incident at another facility involving debris

in a containment sump line which caused a loss of pump suction during a 2

test. McGuire unit two was in the final stages of a refueling outage,

which is the only time an inspection of similar lines could be performed.

During a resultant review of the documentation associated with the

maintenance history of the lines, the licensee noted that maintenance on a

valve on one of these lines had resulted in the detection of a piece of

debris in the valve seat.

! The licensee performed an inspection of the applicable piping. No debris

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was found. A similar inspection is to be performed on Unit I during the

l upcoming 1988 refueling outage.

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The effort expended to resolve the status of the McGuire. sump suction

lines is indicative of a positive attitude toward safety and is considered

a strength. ,

11. McGuire Safety Review Group Operation l

Background

In report 50-369,370/88-14 it was reported that on November 30, 1987, a

l Diagnostic Evaluation Team (OET) began an initial two-week evaluation at

the station and corporate offices.

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The DET report was transmitted to the licensee on April 8,1988. One of I

the findings documented in that report concerned the McGuire Safety Review

Group (MSRG). Specifically, it was determined that (1) the MSRG had not

been performing all functions identified as part of the McGuire licensing

basis and resultantly did not appear to have been meeting the intent of

McGuire TS 6.2.3.3 and 6.2.3.4, and (2) the scope and focus of current

MSRG activities had evolved to the point that the majority of the group's

time was spent on investigation of plant events, with little or no time

spent on surveillance of plant operations and maintenance activities.

Resident Inspector Staff Review

The resident inspection staff review of the MSRG and requirements

pertaining thereto included a review of:

a. Station Directive 3.1.32

b. Supplement 4 to the McGuire Safety Evaluation Report (SER)

c. Charter of the Station Safety Review Group (SSRG)

d. Section 6.2.3 of the McGuire Station Technical Specifications

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e. The DET Team Report

f. The results of an Office of Nuclear Reactor Regulation (ONRR) review

of commitments relative to the MSRG

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Based on a review of the references listed above and discussions with ONRR

and NRC Region II staff, the resident staff review was refined to deal  ;

exclusively with MSRG compliance with applicable Technical Specifications.

Discussions with the MSRG chairman, review of the SRG Work Assignment Log i

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covering the period of 1986 through the present, and review of the Inplant

Reviews conducted by the MSRG during that time, led to the following

conclusions: ,

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a. The majority of the MSRG's time is spent generating incident

investigation reports (IIR's). The Work Assignment Log revealed that

during the period 1986 - June 1988, 203 IIR's were generated but only ,

15 Inplant Reviews were performed. (Inplant Reviews are independent

reviews of station activities, ind are not necessarily coupled with

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IIR's; these come closer to meeting what was intended by Three Mile  !

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Island (TMI) action item I.B.1.2). This reveals that 93% of the L

MSRG's efforts were devoted to the generation of IIR's.

! b. Other than those reviewed in the course of performing an incident  !

investigation, procedures are not reviewed programmatically to t

determine adequacy.

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i c. Other than those reviewed in the course of performing an incident  ;

investigation, design changes are not programmatically reviewed to

l insure all safety concerns are properly addressed,

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l Ultimately, Technical Specification 6.2.3 Station Safety Review Group, is

the culmination of the negotiations of commitments relative to TMI Action

i Item I.B.I.2. To reiterate, the intent of an ISEG (MSRG) is to examine

l plant operating characteristics, NRC issuances, Licensing Information  ;

I Service advisories, Licensee Event Reports, and other appropriate sources '

which may indicate areas for improving plant safety. It is expected that

this group develop detailed recommendations for revised procedures,

equipment modifications, or other means of achieving the goal of improved

i plant safety. A principal function of the independent safety engineering

l group is to maintain surveillance of. plant operations and maintenance

activities to provide independent verification that these activities are

performed correctly and that human errors are reduced as far as practical.

These findings were discussed with the Nuclear Safety Review Board (NSRB)

and MSRG Chairmen on June 13, 1988. The NSRB Chairman indicated-that the

current operation of the SRG's at the Duke facilities was patterned after

the description of an ISEG in NUREG 0800, Standard Review Plan, (SRP)

Section 13-4., and that the resident's findings were a redefinition of the

requirements.

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A subsequent re-review of that section of the SRP confirmed the following:

a. The ISEG is to perform independent reviews of plant operations in

accordance nith the guidelines of item I.B.1.2 of NUREG-0660 and

NUREG-0737.

b. The groups function is to examine plant operating characteristics,

NRC issuances, Licensing Infc mation Service advisories, and other

appropriate sources of plant design and operating experience

information for areas for improving plant safety; and to maintain

surveillance of plant operations and maintenance activities to

provide independent verification that these activities are performed

correctly and that human errors are reduced as far as practicable.

j c. The group .is _ to perform independent reviews and audits of plant

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activities including maintenance, modifications, operational

! problems, and operational analysis, and aid in the establishment of

pragrammatic requirements for plant activities.

Based on the resident inspection staff review, it was concluded that the

intent of technical specifications 6.2.3.3 and 6.2.3.4 were not being met

by the current operation of the MSRG.

This was identified as an unresolved item (50-369,370/88-14-03) pending

results of a meeting which was held on July 18, 1988. The maeting which

involved participants from ONRR, NRC Region II, and the licensee was held

in order to conclusively identify the intent of T.S. 6.2.3.3 and 6.2.3.4

l and to determine if the current operation of the MSRG meets those

requirements.

Conclusions

Based on the results of the July 18, 1988, meeting, and the input from an

ONRR staff representative present at the meeting and involved in the

generation of the requirements, the following conclusions were reached:

a. The current operation of the MSRG does not meet the Intent of T.S.

6.2.3.3 which requires that the MSRG maintain surveillance of plant

activities to provide independent verification that these activities

are performed correctly. This finding is predicated on the review of

the MSRG Work Assignment Log and discussions with the MSRG chairman

and NSRB Director. During the period spanning 1986 until June of

1988, the MSRG did not perform routine independent surveillance of

plant operations and maintenance activities to provide independent

verification that these activities were performed correctly. The NRC

does not consider the performance of Incident Investigations to

adequately satisfy this requirement.

This is a Violation of T.S. 6.2.3.3 (50-369, 370/88-20-02)

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The current operation of the MSRG does not appear to meet the intent l

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of T.S. 6.2.3.4 which requires that the MSRG make detailed recom-

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mendations for revised procedures and equipment modifications to the  !

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Director of the Nuclear Safety Review Board. The licensee indicated

that since T.S. 6.2.3.4 is titled "Authority" they are authorized to

perform the actions of T.S. 6.2.3.4 but not required to perform them. ,

This area will remain Unresolved pending the resolution of the .

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licensee's position relative to their interpretation of the require-

ment and discussion with ONRR staff. (50-369/88-14-03) ,

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12. Information Meetings with Local Officials (94600)

On July 19, 1988 a seminar was held with local public officials. The

meeting was held with the cooperation of the licensee in the McGuire

Nuclear Station Energy Explorium auditorium and was comprised of two

sessions.

The first session was a private meeting between NRC representatives and

the local officials. The objectives of the session were:  ;

To familiarize public officials with the mission of the NRC. '

To introduce key NRC personnel associated with the McGuire facility.

To discuss lines of communication between the public officials and

the NRC.

To discuss the status of the facility and related community concerns

with public officials. -

Representatives from four surrounding counties were in attendance as well

as the NRC Section Chief responsible for McGuire, the NRC Region II

Director of State and Government Affairs, the Resident Inspector from the

Catawba facility and the McGuire Resident Inspectors.

The second session was comprised of a private meeting / plant tour with the

McGuire Plant Manager.

I Following the tour, Duke sponsored a working dinner which was attended by

l local officials, Duke and NRC personnel.

l The meeting, which was the first of this format in Region II was I

l successful in conveying the necessary information and improving public  !

I relations. l

l

13. Containment Spray Heat Exchanger 2B  !

On July 15 a heat balance test was run on containment spray (NS) heat

exchanger 2B in accordance with PT/2/A/4208/048, Train 2B Containment l

3

Spray Heat Exchanger Performance Test. Test results indicated that heat

transfer capability had decreased to approximately one million BTV/hr-F

l (approximately 35 percent of design but 75 percent of what is actually

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necessary according to the FSAR). FSAR section 6.2.1.1.3.1 specifies an

NS heat exchanger capacity of 1.47 million BTV/hr-F. Test results from

May 25, 1988, indicated heat transfer. capability at approximately 45

percent of design, The licensee has written an operability determination

stating that the heat exchanger is c.onsidered operable based on the ice  :

inventory of the ice condenser and the current heat transfer capability of  ;

the NS heat exchangers. The licensee stated that calculations show peak

containment pressure is predicted to remain below 15 psig.

4

Following the July 15 heat balance test, the service water side of the

heat exchanger was cleaned but only minor amounts of material was found.

The heat balance was again run on July 19 and. the results were very

similar to the July 15 test.

The licensee intends to inspect the primary side divider plate gasket to

determine if bypass flow is hindering performance and a new gasket design

,

is being evaluated. Several long term actions are also being evaluated.

The inspectors will continue to monitor the licensee's actions in this

area.

14. Exit Interview (30703)

The irispection findings identified below were summarized on July 22, 1988,

a with those persons indicated in paragraph 1 above. The following items

were discussed in detail:

i (OPEN) Violation 370/88-20-01, Inadequate Procedure / Failure To Follow

1 Procedure with respect to Draining of Steam Generators without blocking

input to ESF actuation and with respect to Removing an Off Site Busline

From Service without an adequate procedure or without properly aligning i

switch gear assemblies (see paragraph 3).  ;

l (OPEN) Violation 369, 370/88-20-02, McGuire Safety Review Group Failing To i

l Meet Intent Of Technical Specification Requirements (see paragraph 11).

(OPEN) Unresolved Item 369, 370/88-14-03, McGuire Safety Review Group

i Failing To Meet Intent Of Technical Specification Requirements (see

paragraph 11). ,

Weakness with respect to determining the status of plant systems during

outages (see paragraph 8).

'

Strength with respect to responding to inspector concerns arising from  !

events occurring at other facilities (see paragraph 10). l

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The licensee representatives present offered no dissenting comments, nor

did they identify as proprit.tary any of the information reviewed by the

1 inspectors during the course of their inspection.

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