ML20210N694

Insp Rept 50-482/97-11 on 970629-0809.Violations Noted.Major Areas Inspected:Operations,Maint,Engineering & Plant Support

ML20210N694
Person / Time
Site:	Wolf Creek
Issue date:	08/21/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20210N664	List: ML20210N660 ML20210N682 ML20210N694
References
50-482-97-11, NUDOCS 9708260090
Download: ML20210N694 (22)

See also: IR 05000482/1997011

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ENGLQELBEJ

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket No.: 50 482

License No.: NPF 42

Report No.: 50-482/97 11

Licensee: Wolf Creek Nuclear Operating Corporation

Facility: Wolf Creek Generating Station

Location: 1550 Oxen Lane, NE

Burlington, Kansas

Dates: June 29 through August 9,1997

Inspector: J. F. Ringwald, Senior Resident ;nspector

Approved By: W. D. Johnson, Chief, Reactor Projects Branch B

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I ATTACHMENT: Supplemental Information

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9708260090

PDR

970021

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ADOCK 05000482

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EKE _QUTIVE SUMMARY

Wolf Creek Generating Station

NRC Inspection Report 50 482/97 11

QDELS1LQDE

• The shif t supervisor effectively supervised the placement of a mobile drilling rig to

minimize the potential for impact on safety related equipment (Section 01.1).

• The inspector identified circumstances where the control room operators would not

be able to reliably contact the shift supervisor when the shif t supervisor was

outside the control room in high noise environments (Section 01.1).

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l * Operators displayed good control board awareness and responded appropriately to

the loss of an offsite power line (Section 04.1).

• Operators improved their communication during an annual requalification simulator

examination, and evaluators demonstrated even more challenging communication

standards by criticizing some examples of crew communications (Section 05.1).

• The conclusions contained in a self assessment report on emergency management

guideline setpoints were not consistent with the findings (Section 07.1).

Maintenance

• The licensee identified a procedure violation when an operator placed the wrong

pump in pull to-lock during the performance of a surveillance procedure

(Section M4.1).

• Instrumentation and control technicians unnecessarily extended the time operators

were in a Technical Specification Limiting Condition for Operation that required the

initiation of a plant shutdown within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, by initiating a surveillance test with

outstanding concerns about the proper interpretation of a procedure action verb

(Section M4.2).

Ennineerinn

• An inadequate flush following the stroke test of an essential service water system

supply valve to the turbine-driven auxiliary feedwater pump resulted in lake water

contamination of Steam Generators A and C during a pump test (Section E1.1).

• The licensee discovered that the control room ventilation process radiation monitors

had never met the time response test requirements described in the Updated Safety

Analysis Report and that a 10 CFR 50.59 evaluation of this condition had not been

completed (Section E8.2).

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• The licensee identified an inappropriate Technical Specification clarification that

resulted in Technical Specification violations as reported in Licensee Event

Report 9010 (Section E8.4).

Plant Sunnort

• The licensee identified two examples where radiation workers failed to comply with

their radiation work permit requirements (Section R4.1).

• Licensee personnel f ailed to properly evaluate the impact of an offsite power loss on

the function of the offsite ernergency plan sirens, resulting in a violation of reporting

requirements (Section P1.1).

• The licensee engaged in the good practice of screening smears prior to formal

counting to improvo radiation surveys (Section R3.1).

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Honort Details

.Summarv of Plant Status

The plant operated at essentially 100 percent power throughout the inspection period, with

the exception of two periods, on June 29 and August 5,1997, when the licensee reduced

power in response to events described in Sections 04.1 and 01.2.

LOperations

01 Conduct of Operations

01.1 Shift Sunervisor Contact Outside the Control Room

a. Insoection Scope f 71707)

The inspector reviewed the ability of control room personnel to contact the shift

supervisor whenever the shif t supervisor lef t the control room,

b. Observations and Findinns

On July 9,1997, the inspector observed the shif t supervisor direct the placement of

a mobils drilling rig during the performance of Work Package 120317, Task 1. The

shift supervisor provided very good direction to the maintenance personnel

responsible for the drilling rig. The final placement resulted in a situation wherein, if

the rig fell over, it would only impact one and not both of tne safety related

engineered safety features transformers supplying the 4.10 kVAC safety related

busses.

After the drilling began, the inspector noted that the noise of the crane interfered

with the ability of personnel in the vicinity of the crane to hear the Gal Tronics plant

announcement system. The inspector determined from discussions that the shift

supervisor could not understand announcements from the Gal Tronics system. The

inspector asked what method the contro! room operators would use to contact the

shift supervisor. The shift supervisor acknowledged that it would be difficult and

promptly added a note to the sh3t turnover form reminding all shift supervisors to

take a pager with them any time they left the control room to enable reactor

operators to contact them even when they were in high noise areas.

The operations manager acknowledged that contact with the shift supervisor could

be improved. He planned to add this tcpic to the agenda for an upcoming shift

supervisor and supervising operator meeting so they could identify an appropriate 1

means for control room operators to reliably contact the shift supervisor,

c. Conclusions

The shift supervisor effectively supervised the placement of a mobile drilling rig to

minimize the potential for impact on safety related equipment. The inspector

identified circumstances in which the control room operators would not be able to

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reliably contact the shif t supervisor when the shift supervisor was outside the

control room in high noise environments.

01.2 Control Room Ventilation Boundarv

a. insocction Scone (71707)

The inspector reviewed the licensee response to the discovery of damage to the

seal for the primary control room entry door,

b. Qhli. tty.gtlons and Findinas

On August 6,1997, the licensee discovered damage to the flexible seal for the

primary control room doorway, Since this door was part of the control room

ventilation envelope boundary, the shift supervisor declared the door inoperable,

entered Technical Specification Limiting Condition for Operation 3.0.3/ and initiated

a power reduction. Approximately 30 minutes after discovering the damage,

maintenance personnel replaced the damaged seal. Engineering personnel

completed the testing approximately 3__1/2 hours af ter the damage was discovered.

The shift supervisor exited Technical Specification Limiting Condition for

Operation 3.0.3 and initiated a power increase to return the plant to 100 percent

power.

Subsequent to this repair, engineering personnel performed additional testing to

- demonstrate that the miseite door would provide an adequate seal should the

primary control room access door seal fail again. In addition, engineering personnel

-coordinated with operations personnel to remove the seal that was damaged in

order to test whether the control room ventilation envelope was inadequate in that

condition. Engineering personnel determined that the control room ventilation

envelope remained sufficiently intact to meet the surveillance requirements even

with the affectad seal completely removed.

_ c.' Conclusions

Based on the information available to the shift supervisor when the damage was

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_ discovered, the shift supervisor responded in an appropriately conservative manner.

04- Operator Knowledge and Performance

04.1' Loss of an Offsite Power line-

a.- Insoection Scoce (71707)

The inspector obse_rved operator response.to the loss of an offsite power line

resulting from a storm.

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b. Qbservations and Findinas

On June 29,1997, during shift turnover, operators experienced a loss of the Benton

offsite power line. Operators noticed the open breakers immediately and responded

appropriately. Since the offgoing supervising operator and reactor operators had

not yet been relieved, they responded to the event. Operators commenced a power

reduction as required by off normal procedures. The crew coordinated their

activities with the power grid system operator and restored the site switchyard to a

normal lineup a short time later when power was restored to the Benton line.

Operators stabilized reactor power at 83.5 percent when power was restored at

7:53 p.m. and began a power ramp which returned the plant to 100 percent power

approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> later,

c. Cenclusions

Operators displayed good control board awareness in identifying the loss of the

offsite power line. Operator response to the_ power line loss was appropriate and

operators controlled the plant properly.

05 Operator Training and Qualification

( 05.1 Annual Ooorator Simulator ExaminatioD ,

a. lasoection Scono (71707)

The ins 1ector observed one of six crews respond to scenarios in the simulator as

pa.t ' their annual requalification examination,

b. Observ illnps and Findinas

On July ,0,1997, the inspector observed two simulator scenarios. While the

scenarios were adequately challenging to determine whether the crew met the

minimum stemdards, they were less challenging than evaluated training scenarios

observed previously. In particular, these scenarios did not contain the elements of

the complex scenario that training instituted approximately 15 months ago with

considerable success. The inspector asked the operations and training managers

why they did not use complex scenarios during annual examinations. The managers

responded by stating that they expected training mnarios to be more challenging

to operators than examination scenarios and that examination scenarios needed to

have the proper elements in order to adequately measure the adequacy of operator

performance.

The inspector noted that the crews generally responded appropriately to the

scenarios and successfully passed the examination. The evaluators observed every

operator performance deficiency identified by the inspector and dispositioned these

appropriately. The inspector noted considerably better crew communication during

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these scenarios than in previously observed scenarios, involving frequent usr. of

three way communication. Despite this improvement, the evaluators criticized the

crew's communication, demonstrating that operations management and training

personnel continued to raise the expected standard.

c. C.qnclusions

The observed simulator examination scenario accomplished the stated objectives

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and operators demonstrated adequate performance. Crew communication improved

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from that observed several months ago. Licensee evaluators criticized crew

l communication, demonstrating that communication expectations were still being

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07 Ouality Assurance in Operations

07.1 Emeroency Maneaement Guidelines Setoont Self Assessment

a. inspection Scone (71707) s

The inspector reviewed Self Assessment Report SEL 97 029, "EMG lEmergency

Management Guidelines) Sotpoint Verification."

b. Observations and Findinas -

On July 18,1997, the inspector reviewed the self assessment and noted that,

while the assessment found problems with 16 of the 37 setpoints reviewed, tho

report concluded that, "The accuracy of the EMG [ Emergency Management

Guidelines) setpoints in the database is considered good." While some of the

problems were very minor and only one problem was significant enough to warrant

a revision to the database, the fact that problems were found in nearly half of the

setpoints reviewed demonstrated the potential that more significant problems

existed.

After the inspector raised these q sestions, the operatiens manager initiated

Performance Improvement Request (PIR) 97 2212 to accress these concerns. The

initial corrective action will ba to review an additional 13 setpoints, then assess the

results of the reviews of 50 of the 269 setpoints. Further corrective actions will be

determined afMr these additional reviews are complete.

After the inspector discussed this issue with the operations manager, the

performance improvement and assessment manager raised similar questions with

the operations manager regarding this self assessment.

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c. Conclusions

The conclusions for Self Assessment Report SEL 97 029 were not consistent with

the findings. The questioning of this issue by the performance impiovemJnt and

assessment manager was appropriate.

-08 Miscellaneous Operations issues -

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08.1 (Closed) Violation 50 482/9624 01: Turbine driven auxiliary feedwater pump

operability. The inspector verified the corrective actions described in the licensee *a

L response letter, dated February 28,1997, to be reasonable and complete.- No

similar problems were identified.

08.2_ (Closed) Violation 50 482/9704-01: Operability of Valve EF HV0034. The I

inspector verified the corrective actions described in the licensee's response letters, )

dated April 23 and May 8,1997, to be reasonable and complete. No siruilar-

problems were identified.

08.3 (Closed) Unresolved item 50 482/9709-02: Containment Cleanliness. This item .

Involved the inspector's identification of fire hose covers on containment fire hose l

stations inside containment during power operation. Engineering personnel

addressed the open questions by repeating the experiment described in NRC

inspection Report 50 482/97 10, Section 08,4, with all hook and loop fasteners-

unfastened. The test results were similar. Engineere also described the calculated

maxirnum postaccident water levelin containment relative to the locations where

the covers were installed. The angineers demonstrated that, since none of the fire

hose stations in containment could be completely submerged,-they could not trap

air and float if they came off the hose stations, they would sink to the

containment floor. Engineers also demonstrated that calculated postaccident flow

-velocities would have insufficient force to transport the covers to the containment

sumps. The inspector concluded that these additional discussions answered all the

remaining open questions.

11. Maintenance

M1: Conduct of Maintenance

M1.1 General Comments on Maintenance Activities

a. Insoection Scone (62707)

The inspector observed all or portions of the following work activities.

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107034 Taskla NK system swing charger conduit installation

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117672 Task 3 BN HV8806B VOTES test

120317 Task 1 Cathodic protection modification installation

120406 Task 1 Appendix R emergency light quarterly

preventive maintenance

122466 Task 1 NK22 charger troubleshooting and repair

b. Observations and Findings

The inspector found no concerns with the maintenance observed.

c. Conclusions

The inspector concluded that the maintenance activities were being performed as

required.

M1.2 General Comments on Surveillance Activities

a. Inspection Scope (61726)

The inspector observed all or portions of the following surveillance activities.

STS IC 211 A, Revision 23 Actuation Log;c Test Train A solid state

protection system

STS NB-005, Revision 13 Breaker alignment verification

STS SE 001, Revision 21 Power range adjustment to calorimetric

b. Observations and Findinas

Except as noted in Sections M4.1 and M4.2, the inspector had no concerns with

the surveillances observed,

c. Conclusions

Except as noted in Sections M4.1 and M4.2, the inspector concluded that the

surveillence activities were being performed as required.

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M4 Maintenance Staff Knowledge and Performance

M4.1 Pun.o Handswitch Miscositioned Durino Surveillance Test

a. Inspection Scone (61726)

- The inspector reviewed the circumstances surrounding the licensee's identification

of a pump handswitch mispositioning during a surveillance tert,

b.- Observations and Findinos -

On July 31,1997, during the performance of Surveillance Procedure STS IC 643A,

the proceduro directed the reactor operator to place Containment Spray Pump A in

pull to lock. The operator inadvertently placed Residual Heat Removal Pump A in:

pull to lock. Approximately 1 minute later, the operator identified the error and

restoied Residual Heat Removal Pump A to the proper standby configuration. The

L shift supervisor noted the brief entry into Technical _ Specification Limiting Condition

for Operation 3.5.2 and the subsequent exit. The operator then placed Containment

._ Spray Pump A in pull to lock and completed the surveillance procedure without-

further problems.

The shift supervisor disciplined the operator involved and initiated PlR 97 2313 to--

s address this issue. -Prior to this occurrence, the licensee identified an' adverse trend

in mispositioning events on July 23,1997, and classified it at Severity Level 1. This-

prompted the_ formation of an incident investigation team to review these events

and formulate a roct cause with detailed corrective actions. The team conclusions

are due on August 22,1997. The team included PIR 97 2313 in the investigation

of these mispositioning events ( The failure of the operator to comply with the-

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surveillance _ procedure is a violation of Technical Specification 6.8.1. This

- nonrepetitive,; licensee identified and corrected violation is being treated as a

noncited violation, consistent with Section Vll B.1 of the NRC Enforcement Policy

-(50-482/9711 01).

c. - Conclusigna

The licensee identified a procedural violation when a reactor operator inadvertently

_.placed a residual beat removal pump in pull to lock _when the surveillance procedure

directed the containment spray pump be placed in pull to-lock.

M4.2 Work Delav Due to Action Verb Definition Contusion

a,- Insoectiori Scooe (61726)

The inspector observed technicians perform a surveillance in the control room.

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b. Qhservations and Findinas

On July 24,1997, technicians performed Surveillance Procedure STS IC 211 A,

" Actuation Logic Test Train A Solid State Protection System," Revision 23. This

test required the technicians to close a reactor trip bypass breaker. This placed the

plant in Technical Specification Lirniting Condition for Operation 4.3.1.1,

Table 3.31, item 19, Action 9, which required the initiation of a plant shutdown

within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if the reactor trip bypass breaker remained shut. Step 8.5.13 of i

Procedure STS IC 211 A read: " Verify that Blocking Function Test Switch is in R

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BLOCKS NOT INHIBITED Position." When the technicians performed Step 8.5.13,

! the blocking function test switch was in inhibit blocks position. Approximately.

l 3 weeks earlier, the technicians recalled that they had received training which

directed them te not reposition switches when the procedure step action verb was

" verify "~ Since the blocking function test switch was in the inhibit blocks position,

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and the technicians understood that they could not reposition the switch, the

technicians were unable to continue the test procedure.

The technicians contacted their supervisor anri asked for guidance. The supervisor

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initially directed the technicians to reposition the switch. After discussing the issue

with the inspector and the shift supervisor, the shift supervisor directed the

technicians to obtain an on the spot change to the rocedure to permit repositioning

the switch.

The technicians recognized the potential for this condition to occur prior to initiating

the test. As a result of inadequate communication between the technicians and the

supervisor, the situation did not get resolved prior to the start of the test. As a

result, the need for the on the spot change kept the plant in a condition where they -

would have to initiate a plant shutdown if the reactor trip bypass breaker remained

shut for more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. This delay lasted 35 minutes while the technicians

-halted the testing and waited for the prucedure change before continuing the test.

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When the technicians finished the testin(, and opened the reactor trip bypass

breaker, only 5 minutes remained before the operators would have had to initiate a

plant shutdown. While waiting for the procedure change, the shift supervisor

reviewed the plant procedures and could not find procedural definitions of action

verbs, The shift supervisor later found the action verb list on the site local area

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network.- This list defined the action verb " verify" to mean that the technicians

could reposition the switch if necessary. Since the technicians were not certain

that this list was controlled, they waited for the procedure change before continuing

the tost.

After the testing, the instrumentation and controls supervisor questioned the

training personnel, who acknowledged that they had discussed the action verb--

during a training laboratory class.- Their lesson plans did not specifically address

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this point and thu training personnel'never intended to communicate a change in the

interpretation of the action verbs. Concerns identified during this test resulted in

the licensee initiating PIRs 97 2290 2291, and 2293.

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c. Conclusiorsti

The inspector concluded that the test was performed correctly and that the

technicians should not have begun the surveillance test without resolving their

question about the action verb. In addition, the instrutnentation and control

supervisor should not have permitted the technicians to begin the test without

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ensurin[, that all questions regarding this action verb were resolved.

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M8 Miscellaneous Maintenance issues

MB.1 IClosed) Violation 50 482/9624-02t Missing main steam safety valve spindle nut

cotter pin. The inspector verifled the corrective actions described in the licensee'6

response letter, dated February 28,.1997, to be reasonable and complete. No

similar problems were identified.

M8.2 (Closedl Violation 50-482/9704-02: Turbine driven auxiliary feedwater pump

retest. The inspector verified the corrective actions described in the licensee's

response letters, dated April 23 and May 8,1997, to be reasonable and complete.

No similar problems were identified.

M8.3 (Closed) Violation 50 482/9704 Oh Turbine-driven auxiliary feedwater pump test. -

The inspector verified the corrective actions described in the licensee's response

-letters, dated April 23 and May 8,1997, to be reasonable and complete,- No similar

problems were ident!!.c<i,

M8.4 (Closed) Violation 50 482/9704-04: On the spot change for pump testing. The

inspector verified the corrective actions described in the licensee's response letters,

dated April 23 and May 8,1997, to be reasonable and complete. No similar

problems were identified.

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E1- Conduct of Engineering

E1.1 Turbine Driven Auxiliarv Feedwater Pumo issues

a. Insoection Scope (37551)

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The inspector reviewed the licensee's response to two issues during a recent

surveillance test.

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b.- Obsurvations and Findinng

On July 22,1997, operators and other licensee personnel performed Surveillance-

ProcMure STS AL 103, "TDAFW ITurbine driven auxiliary feedwaterl Pump

Inservice Pump Test," Revision 28. During the test, operators encountered two

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As soon as the pump started, operators noted that Steam Generators A and C and

! the condensate storage tank immediately Indicated increasing levels.of sodium. j

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After running the pump, operators asked the chemistry technician to sample the

water from Valves AL V138, and V139, tell tale drains for essential service water

supply to the turbine driven auxiliary _feedwater pump. Conductivity indicatlona for

both samples indica'ed the presence of lake water. The licensen conducted a feed

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and_ bleed cleanup ot the condensate storage tank and maximlad cleanup of Steam

Generators A and C. During the subsequent investigation, engineering personnel

determined that, during a recent stroke test of Valve AL HV0033, Train B essential  ;

service water supply to the turbine driven auxillary feedwater pump, leakage

through Valve AL V0014, manualisolation of Train B essential service water supply

, to the turbine-driven auxiliary feedwater pump, causbd lake water to flow through

I the pump, discharge piping,-and recirculation piping into the condensate storage

tank. This contaminated the condensate storage tank _and the licensee performed a

feed and bleed cleanup. Following the valve strokes, operators drained and flushed

the turbine driven auxiliary feedwater pump suction piping, but did not recognize the-

need to drain and flush the pump, discharge piping, and recirculation piping. The

licensee initiated PIR 97 2126 to address this concern. Corrective ' actions included

requiring routine sampling prior to starting the auxiliary feedwater pumps and

routine sampling and flushes following any stroke testing of the essential service

water supplies to the auxiliary feedwater pumps.

During this test, the proaedure required the pump to operate between 3850 and

3900 revolutions per minute. Using a hand held strobe tachometer, engineering

personnel observed a peak speed of 3780 revolutions per minute. Installed plant

equipment indicated that the pump operated at 3900120 revolutic.ns per minute.

During several meetings to evaluate the discrepancy, licensee personnel were not

. able to resolve the discrepancy. A check of the calibration of the installed plant

' equipment demonstrated satisfactory instiumentation performance. The licensee

1 subsequently repeated the test, declared the pump operable based on satisfactory

pump performance using the installed plant equipment, and documented a testing

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deficiency because the pump did not meet the requirements of Technical

Specification 4.0 5. The inspector will review the licensee's resolution of this issue

and will track this as an inspection followup item (50 482/971102).

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c. qqDGhiseta

Two equipment problems cornplieuted the completion of a surveillance test of the

1:arb*e driven auxiliary feedwWr pump. An inadequate flush following the stroke

test of hves Al HVOO32 and 33 resulted in lake water contamination of Steam

Generators A arW C. An inspectiN followup item will track the resolution of the

speed conuolist.w

E8 Miscellaneous Engineering inues

E8.1 (Closed) Violation SrhM2/9624-05: Review of Updated Safety Analysis Report

l commitments. The inspector verified the corrective actions described in the

licensee's response letter, dated February 28,1997, to be reasonable and complete.

No similar prob! ems were identified.

l EB.2 LClosed) Unresolved item 50 4Q2/9708 04: Response time discrepancy with

I control room ventilation radio tivity monitors. This issue involved the licensee's

discovery that they had never rnet the time response described in the Updated

Safety Analysis Report for Radiation Monitors GKRE04 and 05, the control room

ventilation process radioactivity monitors. These monitors isolate control room

ventilation from the outside environment in the event high levels of airborne

radioactivity are introduced into the control room heating, ventilation, and air

conditioning supply duct.

During startup testing, the liceneee tested the time response of these monitors and

found that they did not meet the 3 second specification. The only response time

test performed on the monitors was on February 8,1985, during preoperational

testing. The response time reported for Monitor GKRE04 was 5.167 seconds. The

response time reported for Monitor GKRE05 was 4.967 seconds. The licensee

-initiated action to resolve the discrepancy between the Updated Safety Analysis

Report and the preoperational test results.

The licensee could not identify a reason for the discrepancy and could not identify

any documentation that would indicate an effort to validate the 3 second response

time listed in the Updated Safety Analysis Report, Table 7.3-7.

Engineering personnel identified this issue during a review of the Updated Safety

Analysis Report described in the licensee's Letter ET 97-0010 of February 7,1997,

which committed them to review the Updated Safety Analysis Report in the letter,

the licensee stated that they will complete a formal review of the Updated Safety

Analysis Report by April 30,1997, to provide assurance that the Wolf Creek

Generating Station wou!J be operated according to the Updated Safety Analysis

Report.

The insp4ctor considered the acceptance of the as is condition of the response time

for Monitor <; GKRE04 and -05 to be a de f acto change to the f acility as described in

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the updated Safety Analysis Report. As such, the licensee was required by

10 CFR 50.59 to evaluate and dccument the acceptability of the change to ensure

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that an unreviewed safety question did not exist. The inspector considered the

acceptance of this de facto change to be a violation of 10 CFR 50.59. This

nonrepetitive, licensee identified and corrected violation is being treated as a

noncitcd violation in accordance with Section Vll.B.1 of the NRC Enforcement

Policy (50 482/971103).

The licenseo completed a preliminary investigation and did not identify any

operability issues. The licensee found that Updated Safety Analysis Report,

Section 15A.3, stated that only radiation dose to control room operators due to a

postulated loss of coolant accident was discussed in Updated Safety Analysis

Report, Chapter 15. This was because a study of the radiological consequences in

the control room due to various postulated accidents indicated that the

locs-of coolant accident was the limitirvj case. As such, the ventilation path

containing Radiation Monitors GKRE04 and -05 would isolate on a safety injection

I signal before Radiation Monitors GKRE04 and -05 would detect sufficient activity to

l initiate the control room ventilation isolation signal. The inspector asked whether

there were any design basis accidents that would rely on Radiation

Monitors GKRE04 and 05 to limit the exposure to control room operators. The s

inspector also questioned whether Updated Safety Analysis Report, Change

Request USARCR 97 0081, which proposed deleting the time response requirement

from Table 7.3 7 of the Updated Safety Analysis Report, would involve an

unresolved safety question. Pending answers to these questions and a review of

the licensee actions to resolve the discrepancy in Updated Safety Analysis Report

Table 7.3 7, this is considered an inspcction followup item (50 482/971104).

E8,3 (Closed) Licenseo Event Reoorts 50-482/96-11:0;JJ . <12 00/01. -13-00/01.

14 00/01, and -15 00/01 and Escalated Enforcement ltro,50-482/96-21-02:

Corrective action failures associated with Technical Specification clarifications.

These items involved examples wherein the licensee violated Technical Specification

requirements by following guidance they issued in thts foim of Technical

Specification clarifications which were not consistent with Technical Specification

requirements. The inspector verified the completion of the immediate corrective

actions described in the licensee event reports. The inspector verified the corrective

actions described in the licensee's response letter, dated April 30,1997, to be

reasonable and complete. No similar problems were identified.

E8.4 [ Closed) Licensee Event Reports 50-482/9610-09LQ1: Failure to comply with

Technical Specification Requirement 3.4.9.1. This item is similar to the issues

described in Section E8.3 of this report.- The corrective actions for the similar issue

apply to this issue as well. The report desenbes a violation of Technical

Spet'ication 3.4.9.1. This issue was not specifically addressed in Escalated

Enforcement item 50 482/9621 02. This nonrepetitive, licensee-identified and

corrected violation is being treated as a noncited violation, consistent with

Section Vll.B.1 of the NRC Enforcement Policy (50-482/9711-05).

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IV, Plant Suppm1

R3 Radiological Protection and Chemistry Procedures and Documentation

R3.1 Inanorontiate Survev instrument Use

a. Insection Scone (71750)

The inspector reviewed the circumstances surrounding a health physics technician's

use of a survey meter and probe without an operating procedure,

b. Observations and Findinas

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During the week of June 23,1997, a health physics technician used an ASP 1

l survey meter with an AC 3 detector to screen smears prior to formally counting

l them with a different instrument. Wh9e attempting to determine the efficiency of

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this instrument, the technician determined that, while licensee procedures contained

a calibration procedure for th:s instrument combination, there was no operating

procedure. The licensee initiated PIR 97 2003 to address this concern. Since the

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instrument was used for information or ly, and not for any formal surveys, the use

f of the instrument in this fashion was not governed by procedures or regulations.

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c. Conclusions ,

The inspector concluded that the use of the survey meter to screen smears prior to

counting them was a strength in that it had the potential for an improved survey,

yet the use of the instrument without an operating procedure was a programmatic

weakness.

R4 Staff Knowledge and Performance

R4.1 Radiation Work Permit Reauirements

a. inspection Scone (71750)

The inspector reviewed two examples where radiation workers failed to comp'y with

the requirements of radiation work permits,

b. Observations and Findinns

On June 26,1997, a radiation worker entered the radiologically controlled area

while logged onto Radiation Work Permit 970103. After entering the ;adiologically

controlled area, the technician recognized that the radiation work permit required,

but the worker f ailed to obtain, a neutron thermoluminescent dosimeter. After

recognizing this, the technician exited the radiologically controlled area, obtained the

required dosimetry, and reentered with the proper dorimetry. On July 17,1997,

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the licensee initiated PIR 97 2195 to address this issue. The f ailure of the

technician to obtain the dosimetry required by the radiation work permit is an

example of a violation of Technical Specification 6.11 (50 482/9711-06).

On July 7,1997, a radiation worker entercd the radiologically controlled area while

logged onto Radiation Work Permit 970003. After entering the radiologically

controlled area, the technician recognized that the radiation work permit required,

but the worker failed to obtain, an electronic dosimeter. Af ter recognizing this, the

technician exited the radiologically controlled area and reported this to the lead

health physics technician. The lead health physics technician directed the radiation

worker to speak with a health physics supervicor who restricted the radiation

worker from the radiologically controlled area until af ter the radiation worker

obtained remediation. The licensee also initiated PIR 97 2051. The failure of the

technician to obtain the dosimetry required by the radiation work permit is a second

example of a violation of Technical Specification 6.11 (50-482/9711 06),

in response to NRC Notice of Violation 50-482/9710-06, the licensee initiated

PIR 97 2389 to address a noted decline in the compliance of radiation workers to

the requirements of the radiation protection program. The licensee classified this

PIR as a Severity Level 11, requiring a thorough root cause analysis and full

development of corrective actions that would have to be approved by the corrective

action review board.

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c. Conclusions

The licensee identified two instances wherein radiatien workers entered the

radiologically controlled area without the dosimetry required by the radiation work

permit,

P1 Conduct of Emergency Preparedness Activities

P1.1 Offsite Siren Failures

a. Insoection Scooe (71750)

The inspector reviewed the circumstances surrounding a 1-hour 10 CFR 50.72

report of the loss of six emergency plan sirens that was submitted approximately

1 month late,

b. Observations and Findinns

On June 16,1997, at 1:03 a.m., a storm resulted in the loss of an offsite

substation that provided power to 6 of the 11 omergency plan notification sirens.

The initiallicensee assessment of the situation concluded that the loss of this

particular substation only resulted in the loss of two sirens and led the licensee to

conclude that this event was not reportable. Licensee Administrative

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Procedure AP 26A-001, ' Reportable Events - Evaluation and Documentation,"

Revision 3, Attachment A, defined a major loss of emergency communications

capability as a loss of 3 or more of the 11 emergency p!an sirens, a condition

reportable under 10 CFR 50.72(b)(v). After making the initial determination that the

event was not reportable, operators initiated PIR 97 1792 to address the generic

implications associated with the difficulty encountered while assessing whether this

was reportable or not.

Preparation for assessing a situation of this type had been limited to the licensee

relying on the local orbanizations providing power to these sirens to notify the

, control room and local sheriff whenever they lost power supplying the sirens.

l Assessing the impact of the loss of the substation required a review of records on a

billing computer by an employee of the local organization who was unfamiliar with

the system. No determination had ever been documented to identify the power

sources for each of the sirens.

On July 16,1997, the subsequent evaluation concluded that the initial assessment

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had been in error. The licensee then made the 10 CFR 50.72 report,

On June 16,1997, the licensee accurately know which substation actually lost

power, but then inaccurately assessed the number of affected sirens. Since the

licensee had adequato information to have assessed the impact of the storm, and

failed to properly assess the impact due to inadequate assessment and assessment

preparation, the proper notification did not occur. The f ailure of the licensec to

report the major loss of emergency communication capability within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of being

notified of a power source loss is a violation of 10 CFR 50.72(b)(v) (50-482/9711

07). Since the licensee experienced difficulties in assessing the impact of the

storm, the inspector con::luded that the subsequent assessment did not occur

within a reasonable time. Therefore, while this violation was licensee identified, it

did not meet the criteria in the NRC Enforcement Policy for enforcement discretion.

c. Conclusiorn

The licensee identified a situation wherein they failed to adequately assess

information regarding an offsite power loss which resulted in a major loss of

emergency communication without the required 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> notification of the NRC. The

subsequent investigation was not completed in a reasonable time, resulting in the

notification being approximately 1 month late.

P8 Miscellaneous Emergency Preparedness issues

P8.1 (Closed) Violation 50-482/9704-OL Emergency procedures and documentation.

The inspector verified the corrective actions described in the licensee's response

letters, dated April 23 and May 8,1997, to be reasonab!e and complete. No similar

problems were identified.

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F8 Miscellaneous Fire Protection issues

F8.1 (Closod) Violation 50-482/9623 03: Diesel fire pumo test. The inspector verified

the corrective actions described in the licensee's response letter, dated January 10,

1997, to be reasonable and complete. No similar problems were identified.

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V. ManAnement Meetinan

X1 Exit Meeting Summary

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

conclusion of the inspection on August 12,1997. 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.

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ATTACHMENT

SUPPLEMENTAL INFORMATIOE

PARTIA_LilST OF PERSONS COF TACTED

Licensee

C. W. Fowler, Manager, integrated Planning and Scheduling

O. L. Maynard, President and Chief Executive Officer

B. T. McKinney, Plant Manager

R. Muench, Vice President Engineering

W. B. Norton, Manager, Performance improvement and Assessment

P. L. Sims, Manager, System Engineering

C. C. Warren, Chief Operating Officer

INSPECTION PROCEDURES USEQ

! IP 37551 Onsite Engineering

IP 61726 Surveillance Observatioris

, IP 62707 Maintenance Observations

I IP 71707 Plant Operations ~

IP 71750 Plant Support Activities

IP 92700 Onsite Follow-Up of Written Reports of Nonroutine Events at Power

Reactor Facilities

IP 92901 Followup-Plant Operations

IP 92902 Followup-Maintenance

IP 92903 Followup-Engineering

IP 92904 Followup Plant Support

ITEMS OPENED, CLOSED, AND DISCUSSED

Ooened

50-482/9711-01 NCV Pump handswitch mispositioned during surveillance test

(Section M4.1)

50-482/9711 02 IFl Turbine-driven auxiliary feedwater pump discrepancy

lSection E1.1)

50 482/9711-03 NCV USAR response time discrepancy (Section E8.2)

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50-482/9711-04 IFl Response time discrepancy with control room vent

radioactivity monitors (Section E8.2)

50 482/9711 05 NCV Failure to comply with Technical Specification

Requirement 3.4.9.1 (Section E8.4)

50-482/9711-06 VIO Radiation work permit requirements (Section R4.1)

50-482/9711 07 VIO Failure to report offsite siren failures (Section P1.1)

Closed

50-482/EA96 470- VIO Reactor coolant pump flywheelintegrity

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01023 (Section E8.3)

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50-492/9711-01 NCV Pump handswitch mispositioned during surveillance

l test (Section M4.1)

50-482/9711-03 NCV USAR responso time discrepancy (Section E8.2)

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50 482/9711 05 NCV Failure to comply with Technical Specification

Requirement 3.4.9.1 (Section E8.4)

50-482/9623 03 VIO Diesel fire pump (Section F8.1)

50-482/9624-01 VIO Turbine-driven auxiliary feedwater pump operability

(Section 08.1)

50-482/9624-02 VIO Missing main steam safety valve spindle nut cotter

pin (Section M8,1)

50 482/9624-05 VIO Review of Updated Safety Analysis Report

commitments (Section E8.1)

50-482/9704 01 VIO Operability of Valve EF HB0034 (Section 08.2)

50-482/S700 02 VIO Turbine driven auxiliary foodwater pump retest

(Section M8.2)

50-482/9704 03 VIO Turbine driven auxiliary feedwater test (Section MB.3)

50-482/9704-04 VIO On-the spot change for pump testing (Section M8.4)

50 482/9704 07 VIO Emergency procedures and documentation

(Section P8.1)

50 482/9708-04 URI Response time discrepancy with control room

ventilation radioactivity monitors (Section E8.2)

50-482/9709-02 URI Containment cleanliness (Section 08.3)

50 482/96-011 00/01 LER Failure to comply with Technical Specification

Surveillance Requirement 4.5.3.2 (Section E8.3)

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50 482/90 012 00/01 LER Failure to comply with Technical Specification

Surveillance Requ'rement 4.5.1.1.a.1 (Section E8.3)

50 482/96-013-00/01 LER Failure to comply with Technical Specification

Surveillance Requirement 4.8.1.1.20 7 (Section E8.3)

50 482/90-014 00/01 LER Failure to comply with Technical Specification

Surveillance Requirement 4.5.2.c for visual inspection

of containment (Section E8.3)

50-482/96-015-00/01 LER Failure to comply with Technical Specification

Surveillance 3.6.1.1, " Containment integrity"

(Section E8.3)

l 50 482/96 016 00/01 LER Failure to comply with Technical Specification

Requirement 3.4.9.1 (Section E8.4)

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