ML20141F703

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Insp Repts 50-424/97-05 & 50-425/97-05 on 970427-0531. Violations Noted.Major Areas Inspected:Operations,Maint, Engineering & Plant Support Re Scenario for off-hour Emergency Response Drill
ML20141F703
Person / Time
Site: Vogtle  Southern Nuclear icon.png
Issue date: 06/24/1997
From:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20141F696 List:
References
50-424-97-05, 50-424-97-5, 50-425-97-05, 50-425-97-5, NUDOCS 9707030144
Download: ML20141F703 (33)


See also: IR 05000424/1997005

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U. S. NUCLEAR REGULATORY COMMISSION (NRC)

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REGION ll

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Docket Nos. 50-424 and 50-425

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License Nos. NPF-68 and NPF-81

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Report No:

50-424/97-05, 50-425/97-05

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Licensee:

Southern Nuclear Operating Company, Inc.

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Facinty:

Vogtle Electric Generating Plant (VEGP) Units 1 and 2

Location:

7821 River Road

Waynesboro, GA 30830

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Dates:

April 27 through May 31,1997 -

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Inspectors:

C. Ogle, Senior Resident inspector

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M. Widmann, Resident inspector

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K. O'Donohue, Resident inspector (in training)

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W. Kleinsorge, Reactor inspector, Region ll (Sections M1.3, M1.4, M1.5

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and M1.6)

P. Kellogg, Reactor inspector, Division of Reactor Safety (Section E1.1)

N. Merriweather, Reactor Inspector, Region ll (Sections E1.2, E2.2, E6.1

and E7.1)

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J. Kreh, Radiation Specialist, Region ll (Sections P4.1, P4.2, and P4.3)

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Approved by:

P Skinner, Chief

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Reactor Projects Branch 2

Division of Reactor Projects

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Enclosure 2

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9707030144 970624

PDR

ADOCK 05000424

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PDR

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EXECUTIVE SUMMARY

~ Vogtie Electric Generating Plant Units 1 and 2

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NRC Inspection Report 50-424/97-05,50-425/97-05

This integrated inspection included aspects of licensee operations, engineering, maintenance,

and plant support. The report covers a five-week period of resident inspection, it also includes

the results of announced inspections by regional inspectors in the areas of engineering and

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maintenance.

Ooerations

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in general, the conduct of operations was professional and safety-conscious (Section

01.1).

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The Unit 1 startup on April 30 through May 1,1997, was performed well. The pre-

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briefing was thorough, reactivity adjustments were well controlled, control room

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distractions were actively reduced, and annunciator response procedures were used

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(Section 01.2).

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The 1E 480 Volt alternating current electrical distribution system was properly aligned.

However, two errors in the operation's procedures used to align the system were

identified. The procedures were not updated following implementation of a design

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change (Section O2.1),

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A violation was identified for failure to detect anomalous behavior of a Unit 2

containment sump level transmitter during a containment pressure release. This

anomalous behavior was not identified despite previous occurrences of this behavior

and a routine channel check of the instrument (Section O2.3).

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A violation was identified for failure to document entry into an applicable Technical

Specification Limiting Condition for Operation following detection of anomalous behavior

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of a containment sump level transmitter (Section O2.3).

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A non-cited violation was identified for failure to follow the procedure for placing a

demineralizer cation bed in service. This action resulted in an unplanned negative

reactivity addition (Section 08.2).

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Maintenance

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Maintenance and surveillance activities were generally comp leted thoroughly and

professionally (Sections M1.1 and M1.2).

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Reinstallation of the main generator bearing and bearing housing was unnecessarily

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delayed by a personnel error that resulted in the loss of foreign material exclusion

(Section M1.3),

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Enclosure 2

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A weakness was identified related to the control of lever hoist misuse and the

associated latent damage (Section M1.3).

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A pwr work practice was noted relating to informal, undocumented work authorization

for wnrk on important reactor components (Section M1.4).

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The observed / reviewed welding activities were conducted in accordance with

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procedures, licensee commitments, and regulatory requirements (Section M1.5).

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-A poor work practice was noted relating the issuance of multiple grades of welding filler

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materials to a single welder (Section M1.5).

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Material condition deficiencies were noted associated with the Unit 2 pre-purge supply

system (Section M1.6).

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A poor work practice was identified concerning maintenance work conducted on the Unit

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1 control room annunciator power supply inverters. This contributed to the loss of all

Unit 1 control room annunciators for approximately three minutes (Section M1.7).

Enaineerina

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- As a result of system walkdowns the inspectors identified an Unresolved item (URI)

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regarding the potential lack of missile protection on the exhaust piping for the Turbine

Driven Auxiliary Feedwater Pumps (Section E1.1),

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One example of a lack of attention to detail was identified in which, for almost two years,

operators never questioned that the numbers listed in the surveillance procedures for

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four chart recorders did not match the plant instrument labels (Section E1.2).

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The 50.59 safety evaluation for Design Change Package (DCP) 94-V1N0013-0-1 was

insufficiently supported and weak in concluding that a submittal to NPC was not

necessary for substitution of recorders with digital treno;ng, multi-point indicators, or

- dual meters in that, it was partially based on the assumption that the recorder

references would be deleted from the Technical Specifications (TSs) by MERITS,

without putting in place appropriate controls to ensure that the actions were completed

(Section E1.2).

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Operations failed to adequately support engineering in the closeout of the Chart

Recorder Upgrade Modifications on Units 1 and 2 in that, Operations Procedures

14000-1,14000-2, and 14490-1 were not revised to incorporate the plant changes

(Section E1.2).

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Enclosure 2

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Despite errors in the description of the facility, the safety evaluation appropriately

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supported continued operation with a loose part in the Unit 2 reactor coolant system

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(Section E2.1).

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Engineering support to maintenance and operations on the steam generator blowdown

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control system replacement was gooo (Section E2.2).

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The licensee's fuel oil storage tank calculations were adequate and supported level

instrumentation in the control room (Section E3.1).

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The licensee has been effective in redL cing and managing the engineering backlog, and

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the performance of the engineering si.pport staff normally met licensee established

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goals for timeliness (Section E6.1).

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Safety Audit and Engineering Review (SAER) audits of engineering were effective in

identifying deficiencies in engineering performance (Section E7,1).

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A non-cited violation was identified for an inadequate safety evaluation performed on a

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procedure used to transfer the contents of the spray additive tank. The safety

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evaluation failed to adequately consider all the hazards associated with blocking open

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plant doors to perform the procedure (Section E8.1).

Plant Sunoort

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The scenario developed for the off-hour emergency response drill on May 27,1997

effectively tested a combination of some of the principal functional areas of the onsite

emergency response capability (Section P4.1).

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During the conduct of the May 27 drill, the performance of the emergency response

organization was generally satisfactory, and emergency facilities and equipment were

obse.Ned to be adequate. However, several deficiencies were identified (Section P4.2).

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The licensee's critique of the May 27 drill was probing and thorough (Section P4.3).

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A violation was identified for failure to perform a fire hose station and fire extinguisher

surveillance within the required 31-day interval (Section F2.1).

Enclosure 2

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Reoort Details

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Summarv of Plant Status

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Unit 1

At the beginning of the inspection period, the unit was in Mode 3, with repairs to the turbine

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generator in progress. Following completion of this maintenance on April 30, unit startup

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commenced. Mode 2 was achieved on April 30, with entry into Mode 1 on Mav 1. Nominal

power was attained on May 4,1997. The unit operated at throughout the remainder of the

inspection period.

Unit 2

The unit operated at f ail power until May 30, when power was reduced to approximately 63% to

support repairs to the isophase bus duct cooler number 2 fan. At the completion of repairs,

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power ascension activities were initiated. Nominal full power was achieved on May 31,1997.

l. Operations

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Conduct of Operations

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01.1 General Comments (71707)

The inspectors conducted frequent reviews of ongoing plant operations. In general, the

reviews indicated that the conduct of operations was professional and safety-conscious.

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01.2 . Unit 1 Startm2

a.

Insoection Scoce (71707)

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The inspectors observed portions of the Unit 1 reactor startup conducted April 30

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through May 1,1997. These observations included the transition from Mode 3,

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approach to criticality, and entry into the power range. The inspectors reviewed startup

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Procedures 12003-C, " Reactor Startup (Mode 3 To Mode 2)," Revision (Rev.) 29, and

12004-C, " Power Operation (Mode 1)," Rev. 43.

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b.

Qharvations and Findinas

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The pre-briefing was thorough, reactivity adjustments were well coordinated, and the

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licensee actively reduced control room distractions. Appropriate reference to

annunciator response procedures was made. Control room communications were clear.

c.

Conclusions

The startup was performed well.

Enclosure 2

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Operational Status of Facilities and Equipment

O2.1

Safetv-Related Walkdowns

a.'

Insoection Scoce (71707)

As part of the routine inspection effort to verify availability and overall condition of the

safety-related systems, the inspectors walked down the 480 Volt 1E Alternating Current

(AC) Electrical Distribution System (Units 1 and 2).

b.

Observations and Findinas

The inspectors verified proper system configurations through walkdowns of all 1E 480

Volt AC safety related switchgears and motor control centers. The inspectors also

observed the overall material condition of the system components. During the

walkdown, the inspectors identified errors in Procedures 11429-1, "480V AC 1E

Electrical Distribution System Alignment," Rev.11, and 11429-2, "480V AC 1E Electrical

Distribution System Alignment," Rev. 7. Specifically, the inspectors noted that these

procedures listed breakers 1 ABA-06 and 2ABA-06 as power supplies to Control Building

- Control Room Return Air Fans 1-1531-B7-005 and 2-1531-B7-005, respectively. The

breakers were labelled as spares. These loads were made spares by Design Change

Package 89-V1N0297, Control Room and Technical Support Center Chlorine Detection

System. However, the procedures were not update;d. This was identified to the

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licensee for resolution. Given that the breakers we.re open, this error was insignificant.

Other minor issues identified were forwarded to the licensee for resolution.

c.

Conclusions

The inspectors concluded that the systems reviewed were available to perform their

intended function and were properly aligned.

O2.2 Containment Penetrations Walkdown

a.

Insoection Scooe (71707)

The inspectors walked down accessible portions of the following containment

penetrations to verify proper valve lineups:

PenetratioD

. Unit

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1,2

Nuclear service cooling water (NSCW)

supply

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to reactor cavity coolers

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NSCW return from reactor cavity coolers

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Instrument air

103,104

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Auxiliary feedwater

Enclosure 2

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b.

Observations and Findings

Proper valve lineups were observed for all penetrations. Minor administrative errors in

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Updated Final Safety Analysis Report (UFSAR) Table 6.2.4-1, Containment

Penetration / Isolation Valve Information, were identified and provided to the licensee for

resolution. A minor labelling error and a procedure error identified by the inspectors

were also provided to the licensee.

c.

Conclusion

The detected errors were minor and did not impact the proper lineup of the penetrations.

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O2.3 Unit 2 Containment Sumo Level Transmitter Performance

a.

Insoection Scoce (71707)

The inspectors reviewed the licensee's discovery of changes in indicated Unit 2

containment sump levels during a containment pressure release. This included reviews

of Technical Specification (TS) Limiting Condition for Operation (LCO) status sheets;

control room operator log entries; and corr puter trend charts and strip charts of

containment sump level transmitter behavior. The inspectors also reviewed surveillance

log entries made in accordance with Procedure 14000-2, " Operations Shift and

Surveillance Logs," Rev. 41. In addition, cognizant management was intentiewed.

b.

Observations and Findings

At approximately 9:50 p.m. on April 28,1997, the licensee entered TS LCO 3.4.15,

Reactor Coolant System (RCS) Leakage Detection Instrumentation. This entry was

made in response to anomalous behavior of 2L-7777, Unit 2 Containment South Sump

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Level and 2L-7778, Unit 2 Containment Reactor Cavity Sump Level. Both instruments

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displayed upward shifts in indicated sump level during a containment pressure release

that occurred between 12:59 p.m. and 1:47 p.m. earlier that day. Transmitter 2L-7777

exhibited about a 7-inch increase in level, while 2L-7778 exhibited a 1-inch increase in

level. Given the nominal 48-inch span of these instruments, this represented a 15%

upward shift in 2L-7777 and a 2% upward shift for 2L-7778. The required action for this

TS LCO required daily performance of a RCS water inventory balance surveillance and

restoration of at least two of the three containment sump monitors to an operable status

within 30 days.

On the following day, engineering evaluated the behavior of the transmitters during the

containment pressure release. Based on the stability of the level shifts over the last six

months, the change in level being limited to 2%, and indicated level eventually trending

to normal over several hours, the conclusion was made that 2L-7778 did not require

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replacement. However, the same analysis concluded that 2L-7777 required

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Enclosure 2

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replacement. Based on this,2L-7778 was determined to be operable and 2L-7777 was

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replaced. This LCO was exited et 2:03 p.m. on May 1,1997.

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The inspectors noted from their review of the LCO tracking sheet and the Unit 2 Shift

Supervisor's Log, that the Unit 2 Shift Supervisor failed to document entry into another

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applicable TS LCO. An inoperable 2L-7777 also required entry into TS LCO 3.3.3, Post

Accident Monitoring (PAM) Instrumentation. Entry.into TS 3.3.3 was not documented on

- a LCO tracking sheet nor logged in the control room logs. The safety consequences of

this failure were minimal. Transmitter 2L-7777 was replaced well before the expiration

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of the action statement of TS 3.3.3.

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The inspectors also reviewed approximately a month's worth of strip chart printouts for

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the Unit 2 containment sump level transmitters. During this review, the inspectors

identified that during a pressure release on April 6,1997,2L-7777 exhibited about a 5-

inch change (10%) in indicated level. Similar, but smaller changes were also detected in

this instrument and 2L-7778 during other containment pressure releases in the interim.

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Following inspector questions on the implications of the April 6,1997, shift on the

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operability of the transmitter, a deficiency card (DC) was generated. The DC evaluation

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concluded that the 2L-7777 transmitter should have been replaced after this 5-inch shift '

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in the level on April 6,1997. Smaller shifts as far back as January 16,1997, were

documented by the licensee.

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The inspectors reviewed the work order history for the containment and reactor cavity

sump level transmitters for both units. The inspectors noted that seven previous work

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orders were written documenting similar occurrences of changes in indicated sump level

with variations in containment pressure. A root cause and corrective action report in

response to a July 1996 occurrence attributed this behavior to gas accumulation in one

or both of the liquid-filled capillary _ lines between the differential pressure transmitter

bellows and the sensing bellows. In addition to transmitter replacement, the licensee

included as recommended corrective actions in the root cause evaluation, reviews to

evaluate a replacement differential pressure transmitter and a corrective design change

to the system if appropriate. Licensee personnel stated that they previously had

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planned to modify the transmitter design to incorporate a containment sump level

transmitter which is not susceptible to this phenomenon.

The inspectors were also aware that an informal transmitter performance review

process had been established following the July 1996 occurrence. In this process,

operations personnel provided engineering personnel data on containment level

transmitter performance during containment pressure releases. However, this process

was not proceduralized and was not being used at the time of the most recent

occurrence.

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Also, the April 6,1997, shift in 2L-7777 was not discovered during the channel check of

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containment normal and reactor cavity sump level monitors, performed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

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in accordance with TS surveillance requirement 3.4.15.1. From the strip chart and log

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entries, the inspectors noted that a day shift channel check was performed prior to the

pressure release and hence prior to the shift in 2L-7777. However, the evening shift

channel check was completed after the pressure release. The anomalous behavior of

2L-7777 was visible on the stripchert recording but not detected by the operator

performing the evening channel check on April 6,1997.

c.

Conclusion

The failure to detect the anomalous behavior of 2L-7777 on April 6,1997, was contrary

to the requirements of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions. Two

factors contributed to the licensee's failure to promptly ident:1y and correct a condition

adverse to quality. The first was the lack of a process to evaluate containment sump

level transmitter parformance in the face of previous transmitter failures. The second

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was a channel check which failed to detect unexpected transmitter performance. This is

identified as Violation (VIO) 50-425/97-05-01, Anomalous Containment Sump Level

Transmitter Behavior Not identified by Licensee.

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Additionally, the failure to document entry into TS 3.3.3, PAM Instrumentation on April

28,1997, was contrary to the requirements of Procedure 10008-C, " Recording Limiting

Conditions For Operation," Rev.19. This is identified as VIO 50-425/97-05-02, TS LCO

Entry Not Documented Properly.

O3

Operations Procedures and Documentation

O3.1 Walkdown of Clearances (71707)

During the inspection period, the inspectors walked down the following clearances:

19700313

Centrifugal Charging Pump (CCP) B preventive maintenance outage

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19700358

Fuel handling ,nsole power 480 Volts Alternating Current (VAC)

19700427

NSCW Makeup Well Water Level Control Valve

29700074

NSCW tower fans 1 and 3 outage

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29700213

2HV-1210 and 2HV-1210A Turbine Driven Auxiliary Feedwater (TDAFW)

outside air damper

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b.

Observations and Findings

The inspectors identified one minor error in clearance 19700427 for licensee resolution.

No other problems were identified.

c.

Conclusion

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Control and implementation of clearances were generally well controlled.

Enclosure 2

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08

Miscellaneous Operations issues (71707)

08.1 (Closed) Licensee Event Reoort (LER) 50-424/96-06 Rev.1: Reactor Trip Due to Blown

Fuse in Main Feedwater Isolation Valve

This revision to the LER provided additional information resulting from the licensee's

review of this event. The inspectors reviewed the licensee's event review team report

for this issue. No additional inspection is warranted. This item is closed.

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08.2 (Closed) Unresolved item (URI) 50-424/97-04-02: Unit 1 Unplanned Negative Reactivity

Addition

a.

Insoection Scooe (71707)

This URI documented an unplanned negative reactivity addition to Unit 1 on April 10,

1997, after placing a demineralizer cation bed in service.

The inspectors reviewed the licensee's corrective actions for this event. In addition, the

inspectors reviewed measures implems ited by licensee management to improve overall

reactivity management.

b.

Observations and Findinas

Following the April 10 negative reactivity addition event, the licensee developed a new

procedure for the control of demineralizers, Procedure 11912-C, "Demineralizer Control

Log", Rev. O. Based on their review, the inspectors concluded that this procedure

should preclude repetition of a similar event.

As part of a comprehensive review of the event, the licensee personnel stated their

intention to evaluate their methods and procedures used to control reactivity. As a

result, the licensee has revised several additional procedures dealing with reactivity

control, including Procedure 10000-C, " Conduct of Operations," Rev. 37; and

Procedure 13009-1/2, " Chemical and Volume Control System (CVCS) Reactor Makeup

Control System," Rev.17, and Rev.12, respectively.

c.

Conclusions

The inspectors concluded that the licensee's enhancement of reactivity control was a

positive step. The inspectors also concluded that the lack of documentation for the

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removal of the Unit 1 CVCS cation bed on March 31,1997, was contrary to the

requirements of Procedure 13006-1, " Chemical Volume and Control System," Rev. 37,

Steps 4.4.2.7 and 4.4.2.9 of that procedure required that each time the cation

demineralizer beds were placed in or out of service, the date, time, and boron

concentration be documented. The status of the demineralizer bed on April 10, was not

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Enclosure 2

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documented in accordance with this requirement. However, consistent with Section Vil

of the Nuclear Regulatory Commission (NRC) Enforcement Policy, this is identified as

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Non-Cited Violation (NCV) 50-424/97-05-03, Failure to Follow Procedure for Control cf

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CVCS Demineralizers. Based on this review, URI 50-424/97-04-02 is closed.

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08.3 (Closed) VIO 50-424/96-10-02: Unit 1 Post Accident Sampling System (PASS) Valve

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Mispositioned

a.

IDARRdirg Scooe (71707)

This violation documented the licensee's Sertember 19,1993, identification that 1HV-

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8220, RCS hot leg PASS sample isolation valve, was mispasitioned.

The inspectors reviewed the licensee corrective actions resulting from this issue.

b.

Observations and Findings

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As a result of several mispositioned valves, the licensee performed partial system

walkdowns of various safety-related systems to verify proper lineups. Licensee

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management has also revised Procedure 10000-C, " Conduct of Operations," Rev. 37, to

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clarify expectations for periodic observation of control board system alignments. In

addition, a modification was performed on the Unit i valve to alter the seal-in circuitry to

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reduce the potential for valve 1HV-8220 to inadvertently open.

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c.

Conclusions

The licensee's corrective actions should preclude repetition. This item is closed.

ll. Maintenance

M1

Conduct of Maintenance

M1.1 Maintenance Work Order Observations

a.

Insoection Scota (62707)

The inspectors observed portions of maintenance activities involving the following work

orders:

19700729

Diesel generator train B fuel oil level transmitter IL-9024/lL-9025

calibration

19701205

Nuclear service cooling water (NSCW) containment coolers 1 and 2

supply valve 1HV-1806 preventive maintenance

19701270

Replace inverlor end static switch on Unit 1 annunciator system with new

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Exettech invertor assembly

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Enclosure 2

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19701523

Steam generator blowdown fuse holder replacement

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19701664' . NSCW train A tower makeup will not come full open

'29700151

28 DG power light replacement on air dryer 22403G4001K01

29702992

2BAO3-15 NSCW pump 6 breaker not charging

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Observations and Findinos

The observed maintenance activities were performed satisfactorily.

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. M 1.2 Surveillance Observation

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Insoection Scoos (61726)

The inspectors observed the performance or reviewed the following surveillances and

plant procedures:

14000-1

Operations shift and daily surveillance logs

14030-2

Power range calorimetric channel calibration

14495-2

Auxiliary feedwater (AFW) system flow path verification

14545-2

Motor driven auxiliary feedwater (MDAPN) pump train A monthly

operability test

14634-2

Solid state protection system (SSPS) slave relay K630 train A

containment isolation

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14803-1

Component cooling water (CCW) train A pumps (1, 3, and 5) and check

valve inservice test (IST)

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Observations and Findinas

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The observed surveillance activities were performed satisfactorily.

M1.3 Reoair to Main Generator

a.

Insoection Scone (62700)

The inspectors observed portions of maintenance activities involving Maintenance Work

Order (MWO) 19701399 associated with the reinstallation of the main generator bearing

and bearing housing. Observations were compared with applicable procedures and the

Updated Final Safety Analysis Report (UFSAR).

b.

Observations and Findinas

The licensee experienced very high levels of frictional resistance in their initial attempt to

reinstall the main generator bearing and bearing housing. The licensee attempted this

by lubricating the bearing housing cavity and rotating the bearing and bearing housing

into place. A 3000-pound force lever-operated hoist (chain fall or come-a-long) was

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Enclosure 2

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installed such that it would apply a tangential load to the bearing housing, thus

encouraging the housing to rotate into the desired position. The inspectors observed a

mechanic pull repeatedly on the lever of the hoist until he was unable to cause the lever

to move. This was followed by a mechanic repeatedly striking the radial end of the

housing with a " dead" mallet. Some perceptible movement was noted. This was

i

followed by repeated cycles of applying force to the lever hoist until the lever would not

move, and then repeated blows with the " dead" mallet. Circumferential movement was

reported as about one inch per hour. When no further rotational movement of the

bearing housing could be achieved, the licensee removed the housing and inspected the

housing and cavity. The licensee identified a piece of floor matting, approximately 4"x

6"x %"in size, wedged between the housing and cavity. Sections of this matting had

,

,

been used to protect finished machined surfaces on the generator. The disassembly

l

and reassembly was being accomplished under HOUSEKEEPING ZONE IV

-

requirements defined in Procedure 00254-C, " Foreign Material Exclusion and Plant

Housekeeping Programs," Revision (Rev.) 17. Procedure 00254-C, paragraphs 4.4.2.2

and 4.4.2.3, required that inspections be made before, during, and after work to identify

and either remove or protect against potential debris. The licensee indicated that the

foreign material exclusion efforts did not meet expectations. The preceding was an

example of personnel error resulting in failure to properly implement site foreign material

exclusion measures.

Review of the "Tugit@ Operations, Service and Parts Manual" revealed the following:

the average handle force (in pounds) required to raise the rated load of the model 245

(3000-pound) hoist, is 34 pounds (Ibs.), loads to the handle by any means in excess of

34 lbs. will result in an overload condition of the hoist and will result in a dangerous

condition for the user and nearby persons and property, it was clear that a large

,

mechanic pulling repeatedly on the lever of the hoist until he was unable to cause the

lever to move, would apply significantly more than 34 lbs. to the lever. Procedure

i

20426-C," Control of Lifting and Rigging Equipment," Rev. 5, required an annualload

test for lever hoists, with a two-month grace period. Under the licensee's program, this

I

hoist may not be tested again for up to nine months. As the magnitude of the load

applied to the bearing housing sufficient to cause the housing to move was

indeterminant, the operator of the hoist was unable to determine whether he had applied

i

a load in excess of the hoist's rated load of 3000 lbs. Thus, this non-cognitive misuse of

the hoist and the associated latent damage could be a precursor to a significant event.

The inspectors consider the aforementioned a weakness in the licensee's program for

the control of lifting and rigging. Licensee personnel indicated that they would look into

this matter further.

Procedure 25040-C,"VisualInspection of Lifting and Rigging Equipment," Rev. 6,

Paragraph 4.3.1.3, stated, " Repair of shackles by means of welding is not

recommended." This statement does not assure that appropriate controls will be in

place if the option of welding is taken. Licensee personnelindicated that they would

look into this matter further.

Enclosure 2

. _ _

_

.. ..

_ _ _ _ . _ _ . _ . . - - _ _ _ __ . . .

__. __._ __

,

.

!

10

c.

Conclusion

!

'

The effort to reinstall the main generator bearing and bearing housing was

unnecessarily delayed by a personnel error that resulted in the loss of foreign material

exclusion. A weakness was identified related to the control of lever hoist misuse and

the associated latent damage.

M1.4 Calibration of Tachometer Relavs

a.

Insoection Scoos (62700)

l

The inspectors examined maintenance activities involving MWO 19701284 and

Deficiency Card (DC) 2-97-097, associated with the verification of calibration of points 7

and 8 on diesel generator tachometer relays. The calibration of these points was

inadvertently dropped from the Preventive Maintenance (PM) program as a result of a

procedure change, even though the points were used for surveillances. This verification

l

was dor,e to validate data taken at those points after the passing of the calibration recall

period. Observations were compared with applicable procedures and the UFSAR.

b.

Observations and Findinas

The work was done on all but one of the tachometer relays under MWO 19701284. The

remaining tachometer relay (1 A) had been removed from the plant and operations

control, and subsequently placed in the plant's rebuild program, as the result of a

damaged terminal strip. The Root Cause and Corrective Action Report attached to DC

2-97-097 recommended, in part, that a MWO be issued to collect as-found data for the

removed tachometer relay and the relay be sent to the rebuild program. The licensee

determined that formal written work authorization was not required for work on

components in the rebuild program, and issued a verbal request to collect the as-found

data. The inspectors consider informal, undocumented work authorization of work on

important reactor plant components to be a poor work practice. Licensee personnel

indicated that they would look into this matter further.

Procedure 22583-C, "AIRFAX 300 Series Control Tachometer Calibration," Revision 5,

contained three drawings of the tachometer relay test set-up for: meter calibration;

output calibration; and relay calibration. These drawings depicted the electrical

connections to the two terminal strips. All three drawings showed T20 Volts Alternating

Current (VAC) or 240 VAC connected across terminals 1 and 2. The eight tachometer

relays in the plant had nothing connected to terminals 1 and 2. Therefore, the drawings

did not reflect the actual plant configuration. Licensee personnelindicated that they

would make appropriate changes to the procedure.

l

Enclosure 2

l

r.

-.

-

_ _ _

_ _ _ _

_ - . _ _ . _ _ _ _ _ _ _ _ - - _

. _ - .

_ _ . . _ _

'

.

11

i

c.

Conclusion

Work was satisfactorily done. However, informal undocumented work authorization for

work on important reactor plant components is considered to be a poor work practice.

!

M1.5 Weldina

i

a.

Insoection Scooe (62700)

To evaluate the licensee's welding program and its implementation, the inspectors

reviewed procedures, observed work in progress, and reviewed selected records.

Observations were compared with applicable procedures, the UFSAR, and American

Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code

Section IX, latest at the time of qualification.

The specific area examined was NSCW pump 2-1202-P4-004 support authorized by .

MWO 29601076.

The inspectors reviewed records for welders and materials utilized in the MWOs. These

records included: Welding Procedure Specifications (WPSs) and their supporting

Procedure Qualification Records (PQRs); Welder Performance Qualification (WPQ)

i

records; records attesting to the maintenance of welder qualification; and inspection

reports and Certified Material Test Reports (CMTRs) for welding filler materials.

b.

Observations and Findinas

The inspectors noted that welder AAAA had been issued two cmdes of welding filler

material: E-308L-16 (for welding stainless steel to stainless steel) and E-309L-16 (for

- i

welding carbon steel to stainless steel). Welder AAAA was assigned to make both

stainless steel to stainless steel welds and carbon steel to stainless steel welds. The

inspectors noted that the welding filler material was segregated by size (3/32 inch and

.l

1/8 inch) and grade (E-3C8L-16 and E-309-16) in separated pouches, but the pouches

'

were all in the same bucket. Procedure GEN-25, "VEGP Welding Manual Control of

Welding Consumeble - Section 6," Revision 5, was silent regarding the issuance of

multiple grades of welding filler materials to a single welder. This is considered a poor

i

practice due to the opportunity for material mixup and its potential use in an

inappropriate application.

Examined welding activities were conducted by quahfied and certified welders using

j

correct and certified welding filler materials in accordance with qualified WPSs. PORs

,

l

were reviewed and determined to be adequate.

i

,

,

Enclosure 2

1

.-

,

.

.

-

_ _ . _ _ .___ _ .. _ _ -._ _

.. _ _ . _ . _ . _ _ . - . _ _ . _ _ . _ _ . . . _ _ _ _ _ . . .

,

.

.

l

l

12

L

l

'

I

(-

c.

Conclusion

The observed / reviewed welding activities were conducted in accordance with

procedures, licensee commitments and regulatory requirements. A poor work practice

i

was noted relating the issuance of multiple grades of welding filler materials to a single

i

welder.

l

M1.6 Plant Material Condition

i

a.

insoection Scooe (62700)

The inspectors conducted walkdown inspections of selected portions of the plant to

evaluate plant material condition. Observations were compared with applicable

procedures and the UFSAR.

b.

Observations and Findinas

in Unit 2, room 113, the inspectors noted in excess of a dozen examples of missing,

loose, or not engaged closure devices and fasteners on ducting and electrical panels

';

associated with the Unit 2 pre-purge supply system. Improperly sealed weather tight

electrical panels are of concern because the environmental qualification of the panel

!

may have been compromised. The improperly secured openings into the containment

pre-purge ducting are of concem because the efficiency of the pre-purge supply system

has been compromised. The licensee took immediate corrective action.

'

c.

Conclusion

,

Material condition deficiencies were noted associated with the Unit 2 pre-purge supply

system.

M1.7 Loss of Unit 1 Control Room Annunciators

. a.

Insoection Scooe (62707) (71707)

,

The inspectors reviewed the loss of the Unit 1 control room annunciators which occurred

during maintenance activities on May 22,1997. The inspectors reviewed MWO

19701270, for replacement of the power supply inverters and electrical draw:ngs of the

s. .unciator power supply system. The inspectors attended portions of the licensee's

meetings to review this event. In addition, the inspectors interviewed involved

maintenance and operations personnel.

Enclosure 2

_,

. - ._ _ _ . _ _ _ _ _ _ _ .___ _ . . . _

_ . _ - . . _ _ . . . _ _ _ _ . _

_ .

.

-

>

,

l-

13

1

b.

Observations and Findinas

' On May 22, in support of maintenance activities involving replacement of the number 2

inverter in the Unit 1 control rocm annunciator system, the operations shift crew

inadvertently de-energized both the 115 VAC and 125 Volts Direct Current (VDC) power

j

i

supplies to the control room annunciators. Operations personnel took immediate

[

corrective actions to restore the Unit 1 control room annunciator functions. A loss of

!

control room annunciation for a period greater than 15 minutes would have constituted a

Notice of Unusual Event. However, based on the available information provided by the

,

licensee, this event was estimated to have lasted approximately three minutes.

Prior to commencement of the maintenance activities, the Unit 1 Shift Supervisor (USS)

'

was briefed by an instrumentation and Controls (l&C) foreman. A review of the MWO

indicated that the sequence for manipulating the power supply breakers and lifting the

necessary leads during the maintenance was not documented in the work package.

Instead, the planned sequence was handwritten by the I&C foreman on one of the

electrical drawings and verbally communicated to the USS. Based on the inspectors'

discussion with the USS, the philosophy to de-energize one source at a time was not

fully understood by the USS. Hence, this was not communicated to the Balance of Plant

(BOP) operator or Plant Equipment Operator (PEO). During review of this issue by the

licensee, it was noted that a breakdown in the communication between maintenance

and operations personnel had occurred. In addition, the operations portion of the

inverter replacement, de-energizing the inverter, was performed out of sequence as

desired by the l&C work group,

c.

Conclusions

The inspectors concluded that attempting an evolution of this complexity without

adequate procedural guidance is a poor work practice. However, since the annunciators

alarm system is a non-safety related system, and not covered by 10 CFR 50 (Code of

Federal Regulations) Appendix B requirements, this does not constitute a violation of

regulatory requirements.

Lil. Engineering

E1

Conduct of Engineering

E1.1

Missile Protection of Turbine Driven Auxilis-v Feedwater (TDAFW) Pumo Exhaust

Pioina

a.

Insoection Scooe (37550)

The inspectors observed missile protection of various systems during walkdowns of

external portions of buildings and structures.

Enclosure 2

. .

- .. -

_-

.

- _ _ - - - - . - - - - . .

- . - . - . - . .

-

!

'

.

14

b.

Observations and Findinas

On May 8,1997, the inspectors observed the exhaust pipe for the TDAFW pump exiting

the side of the Auxiliary Feedwater (AFW) building and extending upward for about four

feet.' The pipe had no missile protection once it exited the side of the building. The

inspectors held discussions with the system engineer. The system engineer indicated

.

that at one time the exhaust from the TDAFW pump turbine had exhausted through two

j'

i

pipes and that one of them had been deleted during startup testing. The drawing for the

l

modification to remove the one pipe indicated that missile protection for the remaining

pipe was not required in that if a missile hit the exhaust pipe and a single failure was

taken on one motor driven AFW pump, then the one remaining motor driven pump

would be sufficient.

The inspectors reviewad the Updated Final Safety Analysis Report (UFSAR). The

UFSAR, section 10.4.9, described the system, in part. as being "a safety grade system,

Seismic Category 1, redundant system with class 1E electrical components." Section

10.4.9.1 stated, in part, that, " Protection of the AFW system from wind and tornado is

discussed in section 3.4"..." missile protection is discussed in section 3.5." Section

10.4.9.1.1, Safety Design Bases indicated, in part, that, "The AFW is protected from the

i

effects of natural phenomena. This uptem is designed to remain functional after a safe

shutdown earthquake or fonowing a postulated hazard such as fire, internal missile, or

high-energy line break." No specific reference to an external missile being considered

was made in this section. Section 10.4.9.2.2.2 indicated that, "In the unlikely event that

neither offsite nor onsite ac power is available, the turbine-driven pump can function

normally for up to four hours, at which time the batteries can sustain Direct Current (DC)

power during a Station Blackout (SBO)." Section 3.5 indicated, "In accordance with the

requirements of 10 CFR 50, Appendix A, General Design Criteria (GDC) 2 and GDC 4,

adequate missile protection is provided to ensure that those portions of the safety-

related structures, systems, or components whose failure would... reduce to an

unacceptable level the functioning of any plant feature required for safe shutdown,...are

designed and constructed so as not to fail or cause such a failure in the event of a

postulated credible missile impact." Section 8.4.1.1.2, Station Blackout Coping Analysis

Assumptions, indicated that, "No design basis events or additional single failures are

assumed prior to or during the station blackout event, other than the loss of one

emergency diesel on the non-blacked out unit."

c.

ConclusioD

Upon further discussions with the licesce staff, the inspectors determined that the

licensee's position to be that the exhaust pipe did not have to be missile protected

. because each utor driven pump was a full capacity pump. The inspectors indicated

that this item would be identified as Unresolved item (URI) 50-424,425/97-05-04,

Missile Protection for the Turbine Driven Auxiliary Feedwater Pump Exhaust Line,

perding a review of the design basis with the Office of Nuclear Reactor Regulation.

l

Enclosure 2

-

_

__

__

._.

_-_ _ _ __ _._ _ ____ _ _ _

_ _ _ _._____._ ___ _

'

.

15

E1.2 Review of Comoleted Desian Chanaes and Plant Modification Packaaes

a.

Insoection Scone (3755Ql

L

The inspectors evaluated the adequacy of Design Change Packages (DCPs) to ensure

compliance to 10 CFR 50.59 (Code of Federal Regulations),10 CFR 50.71,10 CFR 50

Appendix B, Criterion til and V, licensee commitments, Technical Specifications (TS),

and applicable licensee procedures. The inspection focused on the design controls for

l

incorporating design basis information into plant procedures. The programmatic

controls were described in design control procedures 00400-C, " Plant Design Control,"

'

Rev. 23, Procedure 50006-C, " Design Change Initiation, Cancellation and Revision,"

Rev.12, Procedure 50007-C, " Engineering Review of Design Change Packages, Rev.

-

8, and Procedure 50008-C, "DCP Implementation and Closure," Revision 8.

b.

Observations and Findinas

The DCPs reviewed and a description of the modifications are listed below:

e

92-VAN 0205

Modified the Hydrogen Recombiner and Monitoring

System

e

94-VAN 0029

Replaced Post Accident Sampling System (PASS)

Containment isolation Valve,1 HV-8220

e

94-V1N0059

Added Solid State Protection System (SSPS) Containment

Ventilation Isolation Block / Reset Indication

e

93-V1N0071

Modified the Piping Penetration Fan Start Control Logic

e

94-V1N0013

Chart Recorder Upgrade

The inspectors noted that the above DCPs had been reviewed and approved for

implementation in accordance with the design control procedures. The 10 CFR 50.59 '

safety evaluations for the changes and the technical bases that an unreviewed safety

question did not exist were reviewed and found to be adequate, with the exception of

the safety evaluation for DCP 94-V1N0013. - It concluded that, without appropriate

controls being put in place, that the reference to certain recorders would be deleted by

the submittal for MERITS, making it unnecessary for any submittal to the Nuclear

Regulatory Commission (NP,C) for substitution of the recorders with indicators. The

submittal for MERITS, dated May 1,1995, did not reflect the design changes as a result

of DCP 94-V1N0013. It simply relocated the recorder references, as identified in the old

technical specifications, to the proposed improved TS (ITS) Bases. The ITS were

approved by NRC on Septembe. 25,1996, with the old recorders still being referenced.

DCP 94-V1N0013-0-1 provided the design to remove, replace or upgrade various chart

recorders located in the main control room. Approximately 58 recorders were deleted,

15 recorders were replaced with dual meters or Multi-point indicators, and an additional

6 recorders were replaced with new digital recorders. Four of the recorders that were

replaced with dual panel meters or multi-point indicators were identified as 1FR-1818

l

Enclosure 2

.

_

_

_

, _ - . _

_

-_

. .

-._ _ _ _ .

. _ _ . _ _ _

_ . _ . _ . _ _ _ _ _ _ _

_ _ _ _ _ _ . _

- _ _ .

d

'

.

!

>

2

16

1

and 1FR-1819, Containment Fan Coolers Cooling Water Flow, and 1TJR-1690 and

'

1TJR-1691, Nuclear Service Cooling Water (NSCW) Tower Basin temperature. In the

assessment for 50.59 applicability, it was noted that Section 3/4.7.5 of the TS refers to

'

recorders TJR-1690/1&2 and TJR-1691/1&2, and Section 4.6.2.3 refehs to recorders

FR-1818/1&2 and FR-1819/1&2. The 10 CFR 50.59 applicability evaluation concluded

!

that any submittal to NRC for substitution of the subject recorders with digital trending,

<

'

. multi-point indicators, or dual meters was not necessary because the reierence to these

'

instruments was being deleted by MERITS and the fact that the instruments were

referenced in parentheses in the TS and the footnote to TS 3/4.0 clarified this to mean

that the references were for "information only."

,

i

The licensee's MERITS submittal was dated May 1,1995. Contrary to the above, the

MERITS submittal did not delete the references to the subject recorders. Instead, the

references were moved to the Bases of TSs, and the footnote to Section 3/4.0 was

deleted. Subsequent to the licensee's submittal for MERITS, on July 13,1935,

4

recorders 1TJR-1690 and 1TJR-1691 were replaced with multi-point indicaters and on

August 1,1995, recorders 1FR-1818 and 1FR-1819, were replaced with new dual panel

meters. The submittal on MERITS was not revised to add the new indicators and

meters and delete the references to the recorders. The inspectors found that TS Bases

5

surveillance requirement (SR) 3.6.6.3 refers to recorders FR-1818/1&2 and FR-

1819/1&2 and SR 3.7.9.2 refers to recorders TJR-1690/1 and TJR-1691/1, which are no

longer installed in the plant. The new indicators or meters have almost identical plant

identification numbers as the old recorders, with one exception, the "R" was reolaced by

"l" so that for example, deleted recorder 1TJR-1690 is now indicator 1TJI-1690. Later

during this inspection, the licensee determined that Procedures 14000-1,14000-2 (Unit

1

2 procedure), and 14490-1 had not been revised to delete references to the old.

i

recorders and add the new indicators. The licensee documented this problem oa

Deficiency Card (DC) 1-97-242 to investigate the root cause for the procedures and its

bases not being revised, and to identify corrective actions. The inspectors noted that

the affected procedures were required to be performed every shift change and that for

almost a two-year period, operations had not questioned the fact that the instrument

numbers listed in the procedures did not match the labeling in the plant. The inspectors

- considered this to be an example of a lack of attention to detail by the operators. The

examples of TS bases and procedures not being revised to delete the recorder

. references demonstrate a weakness in the design controls for ensuring that design

basis information is incorporated into plant procedures to reflect plant modifications. As

part of the Return to Service evaluation for the new indicators, operations was to ensure

that procedures had been correctly revised to incorporate the changes in the plant. The

closure package for the above modification indicated that operations had been informed

of the status of the modification and that operations informed engineering that the

appropriate procedures had been revised.

!

,

1

Enclosure 2

-

..--

__

,

.-

. - . -

. . _ . .

O

-

.-

~.

--.

. - . -

-

- - - .

.-. . _ - .

-

- .

l

'

.

l

!

17

l.

c.

Conclusions

l

One example of a lack of attention to detail was identified in which for almost two years

operators never questioned that the numbers listed in the surveillance procedures for

four chart recorders did not match the plant instrument labels.

The 50.59 safety evaluation for DCP 94-V1N0013-0-1 was insufficiently supported and

weak in concluding that a submittal to NRC was not necessary for substitution of

recorders with digital trending, multi-point indicators, or dual meters in that it was

partially based on the assumption that the recorder references would be deleted from

the TSs by MERITS without putting in place appropriate controls to ensure that the

actions were completed.

Operations failed to adequately support engineering in the closeout of the Chart

Recorder Upgrade Modifications on Units 1 and 2 in that Operations Procedures 14000-

1,14000-2, and 14490-1 were not revised to incorporate the plant changes.

E2

Engineering Support of Facilities and Equipment

'

E2.1

Loose Part in Unit 2 Reactor Coolant System (RCS)

a.

Insoection Scooe (37551)

The inspectors reviewed the licensee's actions taken in response to a potential loose

part detected in the Unit 2 RCS.

b.

Observations and Findinas

On May 16,1997, a vendor analysis of tape recordings made from the Unit 2 digital

metal impact monitoring system (DMIMS) identified a probable loose part in the Unit 2

RCS. The object was identified by the vendor as a small metal object, less than a

j

quarter pound in weight. The object was detected on DMIMS channel 753, indicating

that the object was probably located in the lower plenum of the reactor pressure vessel.

The presence of this probable loose part was reported to the NRC Operations Center in

!

a one-hour non-emergency report later that day. Additionally, the licensee submitted a

written followup report to the NRC on May 29,1997.

'

The inspectors reviewed a written safety evaluation performed by the vendor which

concluded that continued operation with the loose part was acceptable. The safety

evaluation was also reviewed by NRC personnel. The inspectors also attended a Plant

Review Board (PRB) meeting on May 22,1997, during which this safety evaluation was

reviewed. The inspectors found that the safety evaluation, while very methodical in its

l

approach, contained numerous errors regarding the integrated arrangement of Plant

Vogtle. This was the subject of considerable discussion at the PRB meeting. Plant

l

Enclosure 2

. . - _ .

_ -

_ . - . _ - . _ - - ~ _

- . _ - - . - . - . - . . . . . . -

- - ~ _ . ,

'

.

l

18

!

l

management informed the inspectors that they intended to discuss this issue with the

vendor.

j

!

l

c.

Conclusion

!

!

The inspectors concluded that, despite errors, the safety evaluation appropriately

i

supported continued operation with a loose part in the Unit 2 RCS.

E2.2 Modification imolementation

a.

Insoection Scooe (3755D}'

The inspectors reviewed and observed modification implementation activities associated

with the Steam Generator Blowdown Control System Replacement per Minor Design

Change (MDC) 97-V1M019.

b.

Observations and Findinos

After an operating event that resulted in the Steam Generator Blowdowr. canel being

flooded with water, which caused several control cards in the racks to hase electrical

shorts, engineering recommended that a modification be implemented to replace the old

analog controllers in the panels with new digital controllers. MDC 97-V1M0 ?9 was

developed by Engineering Support to replace the analog control card racks and

controllers and to install digital electronic single loop controllers. The inspectors

reviewed the MDC package. It contained the required assessment by 10 C.rR 50.59,

unreviewed safety question criteria, required functional testing, and job ta<a activities.

The inspectors found it to be adequate. The as-built modifications were examined and

determined to be consistent with the modification scope. The electrical wiring and '

terminations inside the panel were examined and found to be satisfactory,

c.

.Qonclusions

The inspectors concluded that engineering support to maintenance and operations on

the steam generator blowdown control system replacement was good.

E3~

Engineering Procedures and Documentation

1

E3.1

Diesel Generator (DG) Fuel Oil Storace Tank Calculations

a.

insoection Scone (37551)

As a result of a DG fuel oil storage tank low level annunciator in the Unit 1 control room,

the inspectors reviewed the fuel oil storage tank level instrument calibrations

procedures, engineering tank calculations, system drawings, vendor instrumentation

specifications, and fuel oil chemistry control procedures. This effort also included

!

!

Enclosure 2

1

.

r-

.

!

19

discussions with Instrument and Controls (l&C) technicians, instrumentation engineers,

and l&C and engineering supervision as to the determination of the TS minimum fuel oil

storage tank level indicated in the control room.

b.

Observations and Findinas

The inspectors reviewed the DG fuel oil storage tank capacity calculation provided by

the licensee. The inspectors determined that in their original calculation (number

X4C2403V01,1982) the licensee appropriately considered only the useable portion of

the fuel oil storage tank to be available. To ensure that enough useable fuel was

available to meet the seven-day inventory requirement, the licensee determined that the

!

minimum level in the tank needed to be maintained above 76% indicated level (68,000

l

gallons). The TS minimum was also 76% of span indicated. The level indicated in the

control room included an unusable portion of fuel (the bottom 15 inches of the tank,

3058 gallons). The inspectors questioned the validity of indicating av:!!able fuel when,

in fact a portion of that fuel was unusable.

Based an a revised calculation (number X4C2403V08, Standby Diesel Generator Fuel

l

Oil Cont umption and Storage Tank Capacity,1992), performed to support an

'

Engineering Distribution System Functional inspection (EDSFI), the licensee determined

that more fuel was necessary to support operation of one diesel for seven days than

was contained in one tank (one storage tank holds approximately 80,000 gallons; the

l

new minimum volume required approximately 85,350 gallons). This new fuel level

l

required taking credit for fuel contained in the opposite train tank. Therefore, although

the licensee's instrumentation in the control room indicated an unusable portion of fuel,

the accuracy of the level indicated for one tank becomes a less significant issue. Due to

the revised requirement, the licensee was only required to ensure that 85,350 gallons

were available. Total available fuel oil onsite with two tanks each at 100% capacity

equaled approximately 129,884 gallons per unit.

c.

Conclusiont

The inspectors concluded that the licensee's fuel oil storage tank calculations were

adequate and supported level instrumentation in the control room. Onsite fuel

inventories met TS minimum requirements.

E6

Engineering Organization and Administration

E6.1

Review of Enaineering Backloa

a.

insoection Scooe (37550)

_

The inspectors reviewed the backlog of open items in the Engineering Support and Plant

l

Modifications and Maintenance Support departments. The combined open items

backlog for the above Engineering departments included Design Change Requests

!

Enclosure 2

l

s

_ .. .

4

l.

'

.

p.

!

20

l(DCRs), MDCs, Temporary Modifications, Request for Engineering Review (RER)

+

Reports, Engineering Maintenance Work Orders (MWOs), Vendor Document Reviews,

.

DCs; open items, and open commitment items.

i

b.

Observations and Findings

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The licensee provided the inspectors historical trend reports showing the numbers of

l

l

open items during the periods from' January 1993 through March 1997, and December

j

1996 through March 1997. By reviewing these reports, the inspectors found that the

backlog of open DCRs and MDCs for the period of December 1996 through March 1997

remained relatively stable. However, the number of open Temporary Modifications

t

{.

Increased slightly, but none were older than one fuel cycle, which met department team

,

goals of less than one fuel cycle. The average number of days required for engineering -

)

-

to disposition open items (i.e., DCs, MWOs, and Vendor documents) usually met team

{.

goals. The only exception was the time required to disposition RERs, which on average

took 30 days, and the team goal was less than or equal to 24 days. The long term

,

performance over the period from January 1993 to March 1997, showed an overall

a

!

downward trend in all categories of open items. The inspectors noted that open items

and open commitment items were not included in the licensee's historical reports on

j

open items. These items were assigned due dates and were tracked by the licensee's

Commitment Tracking Program. The status of these items was reported to

i

management twice weekly The inspectors reviewed the status report on open items

dated May 12,1997, and determined that no engineering open items were past due.

.

.

c.

Conclusions

1

The inspectors concluded that, the licensee has been effective in reducing and

managing the backlog. The performance of the engineering support staff normally met

j

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team goals for timeliness.

j

E7

Quality Assurance (QA)in Engineering Activities

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E7.1

Review of QA Audit Reoorts

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F

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a.

Insoection Scooe (37550)

]

The inspectors reviewed Safety Audit and Engineering Review (SAER) Audits of onsite

engineering and technical support to verify that audits were being conoucted in

'

accordance with the requirements of 10 CFR 50 Appendix B, Criterion XVill.

1

b.

Observations and Findings

The inspectors reviewed recent audits performed in the areas of Environmental

Qualification (EQ), inservice Inspection Program, Programmatic Sampling for Licensing

Basis Conformance, and Materials Control. All of the audits concluded that the Quality

Enclosure 2

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. . - - . _ . - . - . . - . . - . .

.

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21

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Assurance Program was being adequately implemented. However, there were some

audit findings, comn:ents, and recommendations identified in the areas examined. For

example, in the EQ area, three Audit Finding Report items were identified involving

Environmental Summary Sheets, and System Component Evaluation Work Sheets that

were not revised as required, EQ Post-Accident Monitoring System components that

l

were not listed in the EQ preventive maintenance program, and EQ Checklists were not

l

being completed in accordance with procedural requirements. The Audit of

Programmatic Sampling for Licensing Basis Conformance identified an inadequate

50,59 safety evaluation that had been performed to support a change to the Updated-

,

Final Safety Analysis Report. The inspectors noted that the audit findings had been

reported to the appropriate management.

c.

Conclusions

SAER Audits of engineering were effective in identifying deficiencies in engineering

performance.-

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E8

Miscellaneous Engineering lasues

!

E8.1

(Closed) URI 50-424/97-04-06: Adequacy of Licensee's Safety Evaluation For Sodium

Hydroxide Transfer

.

a.

Insoection Scooe (37551)

This URI documented the inspectors' concerns with a safety evaluation performed for

Temporary Procedure T-CHEM-97-01, " Transfer, Neutralization, and Disposal of Spray

Additive Tank Contents," Revision (Rev.) 0.

The inspectors reviewed Procedure 00056-C, " Safety and Environmental Evaluations,"

Rev.15; Procedure 00432-C, " Penetration Seal Control," Rev. 7, and a hazards analysis

performed after questions were raised by the licensee and the inspectors about the

procedure. Additionally, the inspectors interviewed cognizant personnel as to the

licensee's review of this issue.

bc

Observations and Findinas

The inspectors noted from their review that the safety evaluation performed for

Procedure T-CHEM-97-01 contained detailed information, indicating that the review

performed by the licensee was not superficial. However, the safety evaluation failed to

appropriately consider the impact of doors blocked open for the transfer on the

operation of the piping penetration area filtration and exhaust system and equipment

flood protection. This was contrary to the requirements of Procedure 00056-C.

As corrective actions, licensee personnel informed the inspectors that they have or will

conduct hazards awareness training for key engineers, fire protection technicians, and

i

Enclosure 2

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22

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licensed operators; inform plant personnel through notices and electronic mail of the

need to be sensitive to altering door positions without adequate evaluation; and

developing a database and associated procedures to enhance door control.

The failure to adequately evaluate the hazards associated with the performance of

Procedure T-CHEM-97-01 was contrary to the requirements of Procedure 00056-C.

However, consistent with Section Vil of the NRC Enforcement Policy this was identified

.

"

as Non-Cited Violation (NCV) 50-424/97-05-05, Safety Evaluation Does Not Adequately

Consider Hazards Associated With Open Plant Doors.

1

C.

Conclusions

The inspectors identified a non-cited violation for an inadequate safety evaluation

i

performed on a procedure to transfer the contents of the spray additive tank.

'

IV. Plant Suonort

)

P2

Status of EP Facilities, Equipment, and Resources

P2.1

Eacility insoection (71750)

The inspectors conducted tours of the Emergency Operations Facility (EOF), Technical

Support Center (TSC), and the Operations Support Center (OSC). The conditions found

l

were acceptable. The inspectors noted that the housekeeping in the mechanical

equipment room for the EOF was markedly improved over that noted during their last

tour.

P4

Staff Knowledge and Performance in EP

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P4.1

Drill Scenario

a.

Insoection Scooe (82302)

In preparation for evaluation of the emergency response drill on May 27,1997, the

inspectors reviewed the licensee's scenario to determine whether provisions had been

made to test a combination of at least some of the principal functional areas of the

onsite emergency response capability, as required by Section IV.F.2.b of Appendix E to

10 CFR Part 50.

b.

Observations and Findinas

The drill scenario involved an off-hour mobilization of the emergency response

organization (ERO) based on the simulated declaration of an Alert classification at

approximately 8:00 p.m., with no further escalation of the emergency classification

Enclosure 2

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anticipated. The scenario included an aircraft crash with damage to s nuclear service

cooling water tower, a fire, and an offsite release of radioactivity (all casualty conditions

l

were simulated). The scenario was designed to effect activation of the TSC and OSC,

l

and a standby status (staffed but not fully operational) for the EOF.

c.

Conclusions

l

The inspectors concluded that the scenario developed for this drill effectively tested a

combination of some of the principal functional areas of the onsite emergency response

!

capability.

P4.2

Emeroency Resoonse Drill

a.

Insoection Scooe (71750. 82301)

The inspectors witnessed licensee performance during an after-hours recall drill on May

27,1997. Inspectors observed portions of licensee activities in the control room, TSC,

OSC, and EOF to determine whether adequate facilities and equipment were available

and maintained to support an emergency response, and to determine whether the ERO

was adequately staffed and capable of responding to an emergency situation at the

Vogtle facility. The inspectors also observed a portion of the activities of one field

monitoring team.

b.

Observations and Findinas

The drill started at approximately 7:30 p.m. and las'ed for approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

Specific inspector observations included:

The first three emergency notifications had lim 4ed descriptions of the

emergency. More detailed information was available and should have been

provided.

The licensee's initial categorization of the event as an Aler' eas appropriate.

The event progressed to a Site Area Emergency despite the fact that an Alert

was the highest emergency classification anticipated by the scenario (a

development attributable to faults with controller data inputs rather than player

performance).

e

Notifications of state, local, and NRC authorities met regulatory requirements.

The staffing of the TSC and OSC was accomplished in accordance with

e

,

emergency plan requirements.

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Enclosure 2

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_ - _ . _ . _ . _ __- _ _ __ _ _._ _ _ _ ___ _ ._ _..._ _ __ ._

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EP facilities and equipment were adequate to support the response activities of

the ERO.

The inspectors determined that at least six individuals recalled to the site, who

l

responded to the TSC and OSC, failed to sign in on the appropriate emergency

.

l

response facility roster data sheet. This sheet includes a question as to whether the .

l

recalled individual has consumed alcohol in the previous five hours. Pending additional

i

inspector review of this issue, this is identified as Unresolved item (URI) 50-424,425/97-

05-06, Fitness For Duty Statements Potentially Missing for Recalled Individuals.

.

c.

Conclusion

The inspectors concluded that the EP facilities and equipment were adequate, and ERO

performance was satisfactory. However, several deficiencies were identified.

i

P4.3 Drill Critiaue

j

a.

Insoection Scoos (71750.621011

The inspectors observed the player critiques immediately after the drill and the

controller / evaluator critique the following day to determine whether weaknesses and

deficiencies in the drill were identified and formally presented to licensee management.

b.

Observations and Findinas

The licensee's critique process successfully identified the significant areas that created

response problems during the drill. The major findings were in the areas of dose

assessment, field monitoring, content of offsite notification messages, and the

assembly / accountability process. The critique results were presented to plant

management on May 29,1997. The licensee's Emergency Plan specified the

development of corrective actions for identified deficiencies and the tracking of same to

ensure resolution.

c.

Conclusions

l

The inspectors concluded that the critique process was probing and thorough.

S3

Security and Safeguards Procedures and Documentation

S3.1

(Closed) Violation (VIO) 50-424. 425/96-10-05: Designated Vehicle Left Unattended

inside the Protected Area

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Enclosure 2

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a.

Insoection Scoce (71750)

This violation involved the licensee's identification of a designated vehicle not being

properly secured inside the protected area on August 26,1996. In response to this

'

issue, the inspectors reviewed the licensee's corrective actions.

b.

Observations and Findinas

The licensee has implemented a key control program. Coiled lanyards have been

issued that attach to the driver and to the ignition keys of designated vehicles. This

should reduce the potential for leaving the key in a designated vehicle. In addition, the

licensee has revised the Physical Security and Contingency Plan to reduce the number

of designated vehicles approved for entry into the protected area.

)

c.

Conclusions

l

i

The inspectors concluded that the licensee's corrective actions were adequate. Based

j

on this review, VIO 50-424,425/96-10-05 is closed.

F2

Status of Fire Protection Facilities and Equipment

F2.1

Fire Protection Surveillanes Not Performed

a.

Insoection Scoce (71750)

During routine tours of the plant on May 12 and 13,1997, the inspectors identified

approximately six fire hose stations and five fire extinguishers with out-of-date

inspection stickers. The inspectors reviewed the surveillance requirements contained in

Procedure 29100-C, " Portable Fire Extinguishers and Fire Hose Stations Visual

j

Inspections," Revision (Rev.) 8; and Procedure 29134-C, " Portable Fire Extinguishers

Annual Surveillance," Rev. 4; associated surveillance task sheets; and Updated Final

Safety Analysis Report (UFSAR), Section 9.5.1," Fire Protection Program;" and Table

'

9.5.1-10c, " Fire Hose Stations." The inspectors also interviewed fire protection

personnel regarding the fire protection surveillance program.

,

1

b.

Observations and Findinos

During inspection activities in the Unit 2 Nuclear Service Cooling Water (NSCW) towers

and the fire pump house, the inspectors questioned out-of-date inspection stickers on 11

fire extinguishers. Specifically, six fire hose stations in Unit 2 NSCW tower tunnels and

five fire extinguishers located at the fire pump house were found with expired stickers.

,

The requirement to inspect fire hose stations was contained in UFSAR Table 9.5.1-10,

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i

Section 5.4.1, which stated that each of the fire hose stations given in Table 9.5.1-10c

shall be demonstrated operable at least once per 31 days by a visualinspection of the

Enclosure 2

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fire hose stations accessible during plant operation. The requirement to inspect fire

l

extinguishers was contained in NFPA 10 (National Fire Protection Association).

Procedure 29100-C, Checklist 1, was used by the licensee to accomplish these visual

!

inspection requirements. This procedure checklist was applicable to the fire

'

l

extinguishers identified by the inspectors.

,

L

Once questioned, the licensee reviewed the surveillance status of all fire hose stations

i

and fire extinguishers contained in Procedure 29100-C, Checklist 1. A total of 20 fire

.

!

hose stations located in the both Unit 1 and Unit 2 NSCW tunnels failed to have the

' l

required inspection performed in accordance with the inteival specified in UFSAR Table

9.5.1-10 (31-day interval). In addition, all 157 fire extinguishers listed in Procedure

l

29100-C were also found to not have been inspected within the required 31-day time

.

.

frame. As immediate corrective action, the licensee inspected the lapsed fire hose

stations and fire extinguishers within the fire protection program's action statement

,

!

requirements of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with satisfactory results.

t

Licensee personnel later determined that due to incorrect assumptions and use of an

annual inspection procedure for the previous month's inspection, they failed to

accomplish the visualinspection during the month of April. The monthly surveillance

was last performed on April 3, and not completed again until May 13,1997. A total of 40

,

days elapsed between successful performances of the fire hose station and fire

extinguisher inspections,

c.

Conclusions

The inspectors concluded that the licensee failed to perform the fire extinguisher and fire

hose stition surveillance within the required 31-day period. This was contrary to

nrocedure requirements and identified as VIO 50-424,425/97-05-07, Fire Extinguisher

and Hose Station Missed Surveillance.

F2.2

Disabled Fire Protection Comouter Trouble Alarm

a.

Insoection Scooe (71750)

On May 9,1997, the inspectors were informed by licensee management that the fire

computer trouble alarm had been temporarily disabled by the on-shift Unit 1 Shift

Supervisor (USS) on May 4,1997. As part of a limited review, the inspectors reviewed

Procedure 17103A-C, " Annunciator Response Procedures for Fire Alarm Computer,"

Rev.1. In addition, the inspectors met with the involved USS and had limited

discussions with licensee management as to the review of this event.

b.

. Observations and Findinos

l

l

On May 4,1997, the Unit 1 USS disabled the fire computer alarm function for

i

approximately 10 minutes by placing a piece of paper in the " silence button" to maintain

i

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27

it in the depressed (silent) position. According to statements made to the inspectors, the

USS disabled the fire computer silence button in an effort to maintain the focus of the

Reactor Operator (RO) on plant reactivity. The USS stated that the fire computer

trouble alarm had been received four to five times within a few minutes. The alarm, in

.

his opinion, became a nuisance and was potentially distracting the RO at the controls.

t

The Unit 1 RO, at the time of the fire computer trouble alarm, was the sole licensed

reactor operator in the control room. The Balance Of Plant (BOP) operator was

temporarily out of the control roorn.

]

The applicable guidance provided for control room personnel was contained in

Procedure 17103A-C. Section 3.0, Response To a Computer Trouble Alarm, step 3.1,

which stated to " DEPRESS ALARM SILENCE button and red ACK [ Acknowledge) key

until alarm silences." The USS did not follow the procedure guidance for response to a

computer trouble alarm.

Licensee management is currently reviewing the circumstances surrounding this issue.

V. Manaaement Meetinas and Other Areas

X

Review of Updated Final Safety Analysis Report (UFSAR)

A recent discovery of a licensee operating its facility in a manner contrary to the UFSAR

description highlighted the need for a special focused review that compares plant

practices, procedures and/or parameters to the UFSAR descriptions. While performing

the inspections discussed in this report, the inspectors reviewed the applicable portions

of the UFSAR that related to the areas inspected. Except as noted above the inspectors

verified that the UFSAR wording was consistent with the observed plant practices,

procedures and/or parameters.

X1

Exit Meeting Summary

The inspectors presented the inspection results to members of licensee management at

the conclusion of the inspection on June 4,1997. The licensee acknowledged the

findings presented.

The inspectors asked the licensee whether any materials examined during the

inspection should be considered proprietary. No proprietary information was identified.

l

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Enclosure 2

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PARTIAL LIST OF PERSONS CONTACTED

]

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Licensee

!

J. Beasley, Nuclear Plant Geiiers.! Manager

B. Brown, Manager Emerge".cy Preparedness and Training

W. Burmeister, Manger Engineering Support

S. Chesnut, Manager Operations

J. Gasser, Plant Operations Assistant General Manager

)

K. Holmes, Manager Maintenance

P. Rushton, Plant Support Assistant General Manager

'

M. Sheibani, Nuclear Safety and Compliance Supervisor

C. Stinespring, Manager Plant Administration

'

C. Tippins, Jr., Nuclear Specialist l

l

INSPECTION PROCEDURES USED

lP 37550:

Engineering

IP 37551:

Onsite Engineering

'

IP 61726:

Surveillance Obsentations

IP 62700:

Maintenance implementation

IP 62707:

Maintenance Observations

IP 71707:

Plant Operations

IP 71750:

Plant Support Activities

IP 82301:

Evaluation of Exercises for Power Reactors

IP 82302:

Review of Exercise Objectives and Scenarios for Power Reactors

ITEMS OPENED AND CLOSED

Ooened

50-425/97-05-01

VIO

Anomalous Containment Sump Level Transmitter Behavior

Not Identified by Licensee (Section O2.3)

50-425/97-05-02

VIO

TS LCO Entry Not Documented Property (Section O2.3)

50-424/97-05-03

NCV Failure To Follow Procedure for Control of CVCS

Demineralizers (Section 08.2)

50-424,425/97-05-04

URI

Missile Protection for the Turbine Driven Auxiliary

Feedwater Pump Exhaust Line (Section E1.1)

,

i

50-424/97-05-05

NCV Safety Evaluation Does Not Adequately Consider Hazards

Associated With Open Plant Doors (Section E8.1)

l.

Enclosure 2

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50-424,425/97-05-06

URI

Fitness for Duty Statements Potentially Missing For

Recalled Individuals (Section P4.2)

50-424,425/97-05-07

VIO

Fire Extinguisher and Hose Station Missed Surveillance

(Section F2.1)

l

Closed

50-424/96-06, Rev.1 LER

Reactor Trip Due to Blown Fuse in Main Feedwater Isolation

Valve (Section 08.1)

50-424/97-04-02

URI

Unit 1 Unplanned Negative Reactivity Addition (Section

O8.2)

50-424/97-05-03

NCV Failure to Follow Procedure for Control of CVCS

Demineralizers (Section O8.2)

50-424/96-10-02

VIO

Unit 1 Post Accident Sampling System (PASS) Valve

Mispositioned (Section 08.3)

50-424/97-04-06

URI

Adequacy of Licensee's Safety Evaluation For Sodium

Hydroxide Transfer (Section E8.1)

50-424/97-05-05

NCV Safety Evaluation Does Not Adequately Cons. der Hazards

Associated with Open Plant Doors (Section E8.1)

50-424,425/96-10-05

VIO

Designated Vehicle Left Unattended inside the Protected

Area (Section S3.1)

Enclosure 2

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