ML20134C313

Insp Rept 50-302/96-08 on 960714-0810.Violations Noted.Major Areas Inspected:Operations,Maintenance,Engineering & Plant Support

ML20134C313
Person / Time
Site:	Crystal River
Issue date:	09/12/1996
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20134C308	List: ML20134C304 ML20134C309 ML20134C313
References
50-302-96-08, 50-302-96-8, NUDOCS 9609270052
Download: ML20134C313 (30)

See also: IR 05000302/1996008

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l U.S. NUCLEAR REGULATORY COMMISSION l

REGION 2

Docket No: 50-302

License No:- DPR-72

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Report No: 50-302/96-08

Licensee: Florida Power Corporation

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Facility: Crystal River 3 Nuclear Station

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Location: 15760 West Power Line Street-

Crystal River, FL 34428-6708

l Dates: July 14 through August 10, 1996

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L Inspectors: R. Butcher, Senior Resident Inspector ,

T. Cooper, Resident Inspector i

J. Bartley, Resident Inspector, Farley Nuclear Plant, j

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paragraphs, 01.1, 08.1, 08.2 '

W. Bearden, Reactor Inspector, paragraphs M8.2, M8.3,

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M8.4

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G. Hopper, Reactor Engineer, paragraphs 01.1, 08.2 i

F. Wright, Senior Radiation Specialist, paragraphs  !

R2.1, R3.1, R3.2, R3.3, and R5.1 )

Approved by: K. Landis, Chief, Projects Branch 3

Division of Reactor Projects

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PDR ADOCK 05000302

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EXECUTIVE SUMMARY

Crystal River 3 Nuclear Station

NRC Inspection Report 50-302/96-08

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This integrated inspection included aspects of licensee operations,

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l engineering, maintenance, and plant support. The report covers a four week I

period of resident inspection; in addition, it included the results of

announced inspections by a reactor inspector, a reactor engineer, a visiting

resident inspector, and a senior radiation specialist.

Operations

The plant staff's response to a main condenser tube rupture event and a plant

transient due to a governor valve component failure were considered

appropriate. However, there was a concern regarding an operating crew that

did not implement a valid abnormal procedure because they were not comfortable

with it. (paragraph 01.1)

An Operations self-assessment appeared to be of insufficient depth. There

were no findings identified, although recent NRC inspections have identified

problem areas. (paragraph 07.1 and 07.2)

Maintenance

The vital battery charger change out work was accomplished in a professional

manner, with good project manager oversight. (paragraph M1.1)

A Weakness was identified in maintenance personnel communications with

operators during the conduct of a surveillance. (paragraph M3.1)

An Inspector Followup Item (50-302/96-08-02) was identified for the followup

of a permanent fix for thermal relief protection for reactor building cavity

cooling piping. (paragraph M8.2)

Interim corrective actions for a deficiency regarding the use of non-safety

related positioners on safety related valves (identified by the Integrated

Performance Assessment Process team as Unresolved Item 50-302/96-201-04) were

inspected and found to be acceptable. (paragraph M8.3)

Enaineerino

A Weakness was identified in the Problem Report tracking system in that

incomplete corrective actions for make-up system audit findings were closed

out as complete. (paragraph E8.1)

A Violation (50-302/96-08-01) was identified for failure to take timely

corrective actions for make-up system audit findings and excessive vibration

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on a spent fuel pump cooling fan motor. (paragraphs E8.1 & E2.1)

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FPC 2

Plant Support

Inspectors observed that calibration of Reactor Building Radiation Monitor

process was adequate. (paragraph R2.1)

The 1995 Annual Radiological Environmental Monitoring and Annual Radiological

Effluent Monitoring Reports met Technical Specification requirements, and did

not report any adverse radiological trends. The licensee appeared to be

adequately managing radiological effluents to maintain offsite doses as low as i

reasonably achievable. (paragraph R3.1 and R3.2) '

Shipping papers for the transportation of radioactive material, and

radioactive waste met regulatory requirements. (paragraph R3.3)

A weakness was identified in that a Quality Assurance audit failed to properly

identify that two PCS issued during the audit should have been characterized 1

as prs. (paragraph F7.1)

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Report Details

Summary of Plant Status

l The unit began the inspection period with the output breakers closed, I

and the unit at 100 % power. No major evolutions occurred during this

inspection period.

l h Goerations

01 Conduct of Operations

01.1 Review of Previous Operational Events

l a. Inspection Scope (92700)

The inspectors reviewed a Main Condenser Tube Rupture Event, and a

l Governor Valve transient to determine if the licensee responses to the

events were adequate and met regulatory requirements, license

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conditions, and commitments and to verify that the licensed operator's

l performance was adequate to ensure safety.

b. Observations and Findings

The inspectors reviewed the Main Condenser tube rupture event which

occurred on January 9, 1996, in detail. This event was previously

described in NRC Inspection Report (IR) 50-302/96-03, paragraph 2.4.

l The inspectors interviewed some of the operators involved, and reviewed

the plant's procedures and the licensee's investigation. The inspectors

found that the licensee responded to the event in accordance with the

existing procedures with the exception of implementing abnormal

procedure (AP) AP-510, Rapid Power Reduction, Revision 00, dated .

December 14, 1995.

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During the event the operators had discussed alternatives such as

tripping the reactor early and initiating Emergency Feedwater (EFW), but

l did not feel the event warranted placing the plant in such a transient,

and deviating from plant procedures. The operators also stated that the

increased attention to procedural compliance affected the decision not

to initiate EFW. The licensee had a valid procedure for conducting a

rapid shutdown which was appropriate for the existing situation.

Procedure AP-510, Rapid Power Reduction, Revision 00, was issued on

December 14, 1995. If the licensee staff had implemented procedure AP-

510, the plant could have been in Mode 3 within one hour of commencing

the shutdown, instead of taking four hours and 40 minutes. This would

! have significantly reduced the amount of salt water that entered the

Once Through Steam Generators (OTSGs). The operators stated that they

did not implement procedure AP-510 because they had not yet received

simulator training on the new procedure, and that one of the reactor

operators (R0s) was not a normal member of the crew. The operators on

this crew had only received on shift training for procedure AP-510, the

crew's simulator training was scheduled for their next training

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4 rotation. The inspectors were concerned that a crew would not implement

a valid AP because they were not comfortable with using a newly issued

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procedure. The inspectors found that the delay in feeding the OTSGs

with clean water after the shutdown was due to: 1) the licensee staff

prioritization of minimizing the transient on the plant, 2) developing a ;

procedure for feeding the OTSGs with the EFW system, and 3) a lack of i

management direction in emphasizing the importance of initiating clean

feedwater to the OTSGs.

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The inspectors concluded that the plant staff responded to the plant i

transient adequately using the existing plant procedures and equipment.

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NRC Inspection Report 50-302/95-21 described the " Conservative and i

thorough management of the shutdown" as a strength. However, use of

i procedure AP-510 to shut the plant down in a more expeditious manner

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could have been more appropriate to reduce chloride intrusion into the

OTSGs. In addition, the licensee could have been more prompt in

developing and implementing the procedure to feed the OTSGs using EFW.

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The inspectors found that no safety limit or procedural violations

occurred during this event. The inspectors also noted that the licensee

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issued procedure AP-610, Waterbox Tube Failure, on April 8, 1996. This

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procedure was developed from the lessons learned in this event, and

provided explicit guidance to the operators for condenser tube failures. ,

The inspectors also reviewed the governor valve event which occurred on l

May 20, 1996. This event was previously described in NRC Inspection '

Report 50-302/96-05. The event caused three power transients within a l

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30 minute period. The inspectors interviewed the operators involved and

the plant's investigation report. The inspectors verified that the

plant never met the conditions to initiate a manual reactor trip as  ;

specified in procedure AI-0505, Conduct of Operations During Operational '

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Events and Emergency Events.

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The inspectors determined that the operators adequately controlled the l

plant, and that their actions were appropriate. The inspectors also '

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determined that while a reactor trip may have been considered a more  !

conservative response, it would have initiated a larger transient on the )

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plant than the power swings caused by the governor valve. The

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inspectors asked the operators if any other actions were considered in

response to the event. The operators all stated that the option of

using the test feature to slow close the governor valve was considered, i

but no one mentioned it until the post-event critique. Further l

discussions revealed that the senior reactor operator (SR0) did not  !

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pursue the slow close option because it was contained in a performance

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test (PT), and operators could not use the PT in response to the event.

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The operators reduced power to close the governor valve, and then

isolated electro-hydraulic control (EHC) oil to the valve. The event

, was witnessed by the resident inspector who was in the control room at i

j the time. He considered the crew's response to the event demonstrated

technically sound judgement.

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c. Conclusions

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The inspectors recognized that the Emergency and Abnormal procedure I

network was not intended to be the optimal recovery strategy for any )

single event, but rather an enveloping strategy for a family of events.

While the inspectors found a procedure whose use may have been more

appropriate, the licensee appropriately followed an existing procedure.

The licensee has taken the lessons learned from this event and developed

a more appropriate strategy for combatting this specific type of event.

The inspectors noted that any potential consequences of the chloride

intrusion were longterm and could not be evaluated during the

inspection. The inspectors concluded that the licensee's response to

both events was adequate.

06 Operations Organization and Administration

06.1 Operator Shift Schedulina

a. Inspection Scoce (71707)

Technical Specification 5.2.2.e requires that the amount of overtime

worked by unit staff members performing safety related functions shall

be limited and controlled in accordance with approved administrative

procedures.

b. Observations and Findinas

Licensee procedure 01-11, Operations Schedules, section 4.0, Overtime

Policy, provides guidance for the operations staff for the control of

overtime. The requirements in 01-11 are in accordance with Genetic

Letter (GL) 82-12, Nuclear Power Plant Staff Working Hours, as required

by the NRC order transmitted on May 14, 1983, confirming the commitments

to implement NUREG-0737 post-TMI related requirements contained in a

letter from the licensee responding to GL 82-10.

01-11 requirements address the issue of controlling overtime to avoid

fatigue in operators and a degradation in their performance. 01-11

limits include:

e An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />

straight.

e An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />

in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />

period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven day period.

• A break of at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> should be allowed between work

periods.

Licensee procedure, AI-100, Administrative Policies, Section 4.10.2,

i Scheduled Work Implementation, places certain administrative limitations

on the scheduling of overtime

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e No work will normally be scheduled for more than 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> per

week.

• If it becomes necessary to schedule or work more than 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> but'

'less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, department management approval is required.

• If more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per week needs to be scheduled-or worked, it

must be approved by the Director, Nuclear Plant Operations (DNPO). ,

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The licensee's practice is to have the Manager of Nuclear Plant

Operations Support review shift schedules for compliance and approval

prior to issuance. The inspectors reviewed the shift schedules for the

operations department since start up from the recent refueling outage.  ;

! During the refueling outage, large amounts of overtime were routinely i

scheduled. Since the outage, it has been determined that time in excess I

of 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> in any seven day period has been routinely scheduled, for

the licensed R0s and SR0s. During this period, it has been a common

practice to schedule shift periods from five to nine days long. No 4

times'were found where more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> were worked in a.seven day '

L period, but a large number of times where a licensed operator worked

! between 60 and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> were found. A review of the NLO schedule

! revealed that the NL0s rarely worked more than 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> in a seven day

period.

l c. Conclusions

A PC has been issued to address concerns with the scheduling practices

in the operations department, including overtime control practices.

Actions have been taken to address the concern expressed in the PC,

including maintaining support personnel with active licenses to provide

-additional' coverage and starting a new license training class to provide

additional personnel for shift coverage. The inspectors have concluded

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that the licensee did not violate any overtime restrictions but the

continued use of routine overtime provided the potential for degradation

of operator performance.

07 Quality Assurance in Operations

07.1 96-02-REFL. Audit Report of 1996 Refuel Outaae

a. Inspection Scope (40500)

The inspectors reviewed the-licensee quality assurance (QA) audit for

the 1996-refueling outage, 96-02-REFL. The audit included, but was not

limited to, field observations, review of the plant shutdown, control of.

contractors, nuclear quality control activities, ISI activities, fuel

movement, adherence to technical specifications, corrective actions, and

maintenance activities,

b. Observations and Findinas

~ As a result of the audit, five findings (characterized by the Licensee

as violations), fifteen weaknesses, numerous negative comments, and

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fifteen strengths were identified. Several conclusions were identified

by the licensee in the report:

1. Ineffective training and understanding of procedures was evidenced

by the problems with foreign material exclusion.

2. Lack of attention to detail in failures to complete signoffs as

actions occurred.

3. Management expectations in the areas of communications, self

checking and the utilization of available resources were not

consistently met.

4. Management oversight is lacking as evidenced by improper reviews

of procedures.

c. Conclusions

The inspectors discussed the audit with the audit team leader and

reviewed the audit report. There were many individual findings,

however, conclusions drawn tended to be specific to the identified items

(a list), without any programmatic findings discussed. Plant management

is aware and is currently assessing these findings, therefore no NRC

follow-up will be taken.

07.2 Operations Self-Assessment

a. Inspection Scope (405001

The inspectors reviewed the operations self-assessment performed at the

site by current and formerly SR0 licensed individuals on loan from other

licensees. The areas assessed by the team included: professionalism and

skills of plant operators, self checking and questioning attitude,

interface and communications, conservative decision making, shift

turnover, control room traffic, and procedural adherence and usage. l

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b. Observations and Findinas

The assessment team provided a number of recommendations, areas for

improvement, and strengths. No findings or problem reports were

identified as a result of this assessment. A number of recommendations

were provided, which have the potential to improve the effectiveness of

operations. However, no programmatic assessments were evident, although

a number of areas for improvement were identified.

c. Conclusions

The operations self-assessment, while providing a number of observation;

characterized as strengths, weaknesses, areas in need of improvement, or

recommendations; did not provide an assessment of operational programs.

There were no problem reports issued as a result of the assessment.

Three PCs were issued as a result of the assessment, two on recurring

recorders problems in the main control room. No conclusions were

reached on the overall effectiveness of the licensee's operations

department, by the self-assessment.

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Recent inspections by the NRC have identified multiple problems,

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practices. Even though recommendations exist in these areas, no

assessment of the current effectiveness, nor any examples of specific

problems were identified.

l The performance of the self-assessment was a good initiative, but the

implementation was not of sufficient depth.

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08 Miscellaneous Operations Issues

08.1 (Closed) Licensee Event Report (LER) 50-302/96-017-01. Reactor Trio on

Hiah Reactor Coolant Pressure Durina Turbine Testina Caused by Debris in

Manual Isolation Valve

a. Inspection Scope (92901) i

This event was discussed in IR 50-302/96-05. The inspector performed

additional followup interviews with the operators to clarify their-

response when the main steam safety valve (MSSVs) did not operate at the

proper setpoints.

b. Observations and Findinas

The inspector determined that the operators followed plant procedures

and that their actions were adequate.

c. Conclusions

This LER is closed.

08.2 (00en) Violation (VIO) 50-302/93-16-07. Inadeauate Emeraency Operatina I

Procedure (EOP) and AP Procedures

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a. Inspection Scope (92901)

This item concerned multiple examples of a violation of 10 CFR 50,

Appendix B, Criterion V. The inspector reviewed the remaining

procedures cited in the violation, AP-581, Loss of (Non-Nuclear

Instrumentation power supplies) NNI-X, and AP-582, Loss of NNI-Y.

b. Observations and Findinas

The inspector noted that the procedural discrepancies noted in the i

Notice of Violation had been corrected for these procedures. However,

an outstanding Request for Engineering Assistance (REA) 95-0406 was

initiated in April 1995, and has yet to be completed. This item

requested that an engineering review be conducted on the above

procedures to ensure the lists of reliable and unreliable

instrumentation contained in the above procedures were correct.

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c. Conclusions

This item will remain open pending completion of REA 95-0406, and final

review of the procedures for accuracy.

II. Maintenance

M1 Conduct of Maintenance

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M1.1 Vital Battery Charaer IF Chanae Out

a. Inspection Scope (62707) '

The inspectors witnessed portions, on various days, of the performance

of WR NU 0331781, Vital battery charger IF change out, per Modification

Approval Record 93-05-07-01.

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b. Observations and Findinas

The inspectors reviewed the work package and verified that all reviews l

and approvals had been received prior to the beginning of work.

The licensee utilized SMC support personnel to perform this task. The

inspectors verified that the personnel were qualified by the licensee's

program and were the same personnel used to change out battery chargers ,

lA, 18, 1C, and 10 during the recent refueling outage,

c. Conclusions

The vital battery charger change out work was accomplished in a

! professional manner, with good project manager oversight.

i No problems were identified.

M1.2 Shoot and Clean Nuclear Services Closed Cycle Coolina Water System Heat

Exchanger (SWHE) 10

a. Inspection Scope (62707)

The inspectors witnessed the performance of portions of WR NU 0336700,

Shoot and Clean SWHE-1D.

b. Observations and Findinas

A review of the work package revealed that all reviews and approvals had

been received prior to beginning the task. The inspectors discussed the

task with the technicians.

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c. Conclusions

The inspectors determined, that the technicians were knowledgeable of

the requirements and performance of the work. No problems were

identified.

M3 Maintenance Procedures and Documentation

M3.1 Surveillance Observations

a. Inspection ScoDe (61726)

The inspectors witnessed the performance of surveillance procedure, SP-

130, Engineered Safeguards Monthly Functional Test. This procedure was

performed to verify operability of the Engineered Safeguards Actuation

System (ESAS) instrumentation, as required by TS 3.3.5, ESAS

Instrumentation.

b. Observations and Findinas ,

The inspectors witnessed the performance of SP-130, Engineered

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Safeguards Monthly Functional Test. The inspectors noted some r

weaknesses with the instrument technicians communications. When the  !

technician asked the R0 to verify the reactor building pressure, he did_

not specify the instrument to be used until the R0 asked if one was l

.specified in the procedure. -While performing steps of the procedures  :

which would cause an alarm, the technician would just state that .there i

would be noise coming,-and did not specify which alarms to expect, until ;

the R0 requested that the technician be more specific. Even though j

there were no problems with the completion of the surveillance, the.R0 i

had to prompt the technician for more accurate communications.

c. Conclusions

Weaknesses were identified with the instrument technician communications

skills, but the R0 prompted better communications during the performance

of the test.

M8 Miscellaneous Maintenance Issues

M8.1 Main Steam Safety Valve

a. Inspection Scope (92902)

A reactor trip followed by one steam generator dryout due to the failure I

of one main steam safety valve to close occurred at Arkansas Nuclear I

on May 19, 1996. -The main steam safety valve had apparently been

maintained improperly, allowing a stem nut (release nut) to move, i

preventing valve closure. An Industry Report had previously been issued i

due to the same failure mechanism having occurred at Crystal River. I

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b. Observations and findinas

In IR 50-302/96-06, Paragraph M8.1, the residents discussed their review

of applicable licensee documents, and concluded that the corrective

actions taken for the main steam line safety valves release nut

restraint problem were adequate to prevent recurrence. Subsequently,

the residents walked down the main steam line safety valves and verified

that the release nuts were properly restrained by the release nut cotter

pin.

c. Conclusions

No further follow-up of this issue is required.

M8.2 Cavity Coolina Pipina Thermal Relief Protection

a. Inspection Scope (62703)

The inspector reviewed the licensee's interim corrective actions

associated with a potential concern associated with containment

integrity. During the recent NRC IPAP inspection an NRC inspector had

questioned the adequacy of the existing cavity cooling (CI) piping

configuration associated with AHHE-14A and AHHE-14B within the Reactor

Building. This system provides cavity cooling during power operation

and prnvides no post-accident safety related function. The specific

concern was that the CI piping or cavity coolers could fail following a

loss of cooling accident (LOCA) due to over pressurization after

containment isolation. CI piping within the Reactor Building was not

provided with thermal relief protection.

b. Observation and Findinas

The inspector reviewed Problem Report (PR) 96-0261 which was issued by

the licensee to address this issue along with the licensee's interim and

proposed long term corrective actions and determined that they were

adequate. The licensee evaluated this issue and determined that

containment integrity would not have been affected by a failure of the

CI piping located within the reactor building (RB). However, the

licensee also issued temporary modification, (T-MAR) T96-07-16-01, to

provide interim thermal relief protection for the CI piping in the

containment. The inspector reviewed this T-MAR along with maintenance

work request WR 0336793 which implemented this temporary modification.

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This WR added CI thermal relief valves, CIV-279, 280, at AHHE-14A and

148. The relief valves were added downstream of piping vent valves,

CIV-90 and 91, and the vent valves administratively controlled under the

licensee's equipment clearance program. Installation of this T-MAR

required the licensee to modify their system venting instructions to

allow removal of a relief valve for venting purposes. The inspector was

further informed that a permanent plant modification would be issued at

a later date to replace this T-MAR.

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c. Conclusions

The inspector determined that the licensee's interim corrective actions

were acceptable. IFI 50-302/96-08-02, Reactor building cavity cooling

piping thermal relief protection, will be issued to track permanent

resolution of this concern.

M8.3 Non Safety-Related Positioners on Safety-Related Valves

a. Inspection Scope (62703) I

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The inspector reviewed the licensee's interim corrective actions j

associated with a concern associated with the safety-related and non-  !

safety-related interface on Decay Heat Closed Cycle Cooling valve

positioners and whether a failure of the valve positioner could

potentially result in a failure of the associated valve to remain in the

correct position during a LOCA. During the recent NRC IPAP inspection,

an inspector had questioned the adequacy of the configuration associated

with the non-safety related positioners installed on the DC outlet

valves, DCV-177 and DCV-178, and bypass valves, DCV-17 and DCV-18, for

the Decay Heat Removal Heat Exchangers. During a review of the subject

valves it was discovered that although control air to the pneumatic

controls were isolated, the supply air to these non safety-related

positioners was not normally isolated.

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b. Observation and Findinas

The inspector determined that PR 96-0220 had been subsequently issued by

the licensee to address this concern. The licensee's interim resolution

of this issue was to temporarily close IAV-228 to isolate the supply air

to the pneumatic controls to the valves and place IAV-228 under

administrative control by use of an equipment clearance.

These DC valves are required to be in their safety related ES positions

during a LOCA to provide maximum cooling flow to the DH heat exchangers

(DCV-17 and 18 full closed with DCV-177 and 178 full open). However,

these valves also have a non-safety related function in that they are

throttled to control the rate of decay heat removal from the reactor

core during plant shutdown. During power operation these valves were

previously failed to their safety related position by isolating control

air to the positioners by Procedure OP-404, Decay Heat Removal System,

while supply air was not isolated. The actual valve position on loss of

air would be determined by the actuator spring. Although the

positioners were not considered safety related, the valve actuators were

classified by the licensee as safety-related.

The inspector reviewed plant modification MAR 94-09-02-01, DC Cooling

Instrumentation Enhancement. The inspector determined that this MAR,

when implemented, will simplify the control logic of DC cooling to the

heat exchangers by permanently removing the existing pneumatic

temperature controller and piping the emergency preparedness (E/P)

output directly to the valve positioners. This modification will remove

the temperature control function and allow the operator to command a

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valve position demand signal directly to the valve positioners. This

modification would also result in a change in the method of failing the

valves to their safety-related ES positions by isolating DCV-194 and

DCV-196 and venting of supply air by opening DCV-195 and DCV-197 rather

than only isolating the input control air signal to the positioners.

This would isolate all air to the valve positioners and no longer

require closure of IAV-228.

c. Conclusions

The inspector determined that the licensee's interim corrective actions

were acceptable. Additionally, the inspector determined that completion

of this MAR was listed as part of the corrective action plan (CAP) for

PR 96-0220 with a completion date of September 30, 1996. This item was

identified by the IPAP team, and will be tracked as part of IPAP Open

Issues. (See IPAP URI 50-302/96-201-04.)

M8.4 Safety-Related Battery Chargers

a. Inspection Scope (62703)

The inspector reviewed the status of licensee actions to address an

industry issue which could potentially affect reliability of the safety-

related battery chargers. During installation testing of replacement

safety-related 125 VDC battery chargers at Salem, it was found that

three wire lugs were landed on a terminal with insufficient thread

engagement to the terminals. This deficiency could have resulted in

failure of those battery chargers during power operation. All six

safety-related chargers at Crystal River were recently replaced with new

battery chargers from the same manufacturer as those at Salem.

b. Observation and Findinas

The inspector was informed that licensee management had contacted Salem

management, and had determined that the actual problem was that up to

three wire lugs were landed on common screw terminal connections on the

battery charger high voltage shutdown board. This board is located

within the alternating current (AC) side of the battery chargers and

provides high AC voltage input protection for the battery and charging

circuits. The terminal screws were not of sufficient length to provide

adequate thread engagement with three wire lugs landed on the associated

terminal.

Each of Crystal River's two safety-related batteries is supported by two

125 VDC chargers and a single 125 VDC spare charger (total of six

chargers). The safety-related chargers at Crystal River are from the

same manufacturer but are a different size and model (Charter Power

Systems Model ARR200F) than those used at Salem (Charter Power Systems

Model ARR300F). However, the non-safety related chargers in use at

Crystal River are the same type chargers as those being questioned at

Salem. The inspector was informed that the non-safety related chargers

had also been inspected and no loose connections were identified.

I

i

)

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

d

12

1

However, the engineer noted that there were three wire lugs landed on

i

common screw terminal connections on the high voltage shutdown boards on

these chargers. A precursor card had been written to address this

i potential problem with the non-safety related chargers at Crystal River.

'

The inspector was informed that a licensee engineer had inspected all

six safety-related chargers and verified the adequacy of all screw

.

terminals located within the safety-related chargers. The inspector

interviewed the engineer that conducted these inspections. The engineer

informed the inspector that all screw terminals located on the high

voltage shutdown boards had less than three wire lugs installed. The

inspector was also informed that each of these chargers did have three

wire lugs landed on a common terminal on the power terminal block. The

lug connections on these terminals had been closely examined by the

licensee and the licensee determined that the installations were

acceptable. The power terminal blocks and connections appeared sound

, and the type and length of the screw used to secure the wire lugs was

verified to be of sufficient length to allow full thread engagement with

three wire lugs attached by a common terminal screw. Additionally, the

engineer verified all wire terminal points on the six safety-related

i

chargers which utilized lug nuts had full thread engagement.

The inspector selected one safety-related charger, DPBC-1E, which had

been previously inspected by the licensee and visually inspected all

internal wire lug connections. The inspector verified that all screw

terminals and lug nut installations were acceptable.

,

c. Conclusions

J

The results of the inspector's observations were consistent with the

results from the inspection performed by the licensee engineer. Based

,

on this review the inspector determined that this potential industry

1 issue had been adequately dispositioned by the licensee. No further

review of this issue is required.

l III. Engineerina

"

E2 Engineering Support of Facilities and Equipment

4

E2.1 Deqraded Safety Related Eauipment

,

a. Inspection Scope (37551)

On July 17, 1996, the inspectors reviewed the licensee's Plant Equipment

Condition Monitoring Program report dated July 12, 1996. This report is

a quarterly report summarizing the plant equipment condition monitoring

program.

b. Observations and Findinas

The inspectors noted that AHF-8A, Spent Fuel Pump Motor Cooling Fan 8A,

continued to be operated with an increasing trend of excessive vibration

__

13

levels on the motor and with a high priority action level. The report

referenced REA 94-0026, which was previously written to improve the fan

motor base attachment. The fan housing is not stiff enough to dampen

the belt vibrations, and a contributing factor may be a mismatch between

the adjustable dual groove sheave and the non-adjustable double wide

powerband belt, which drives the fan. The inspector reviewed the Plant

Equipment Condition Monitoring Program report dated October 15, 1995.

That report also indicated that AHF-8A had excessive vibration levels

but with a low priority action level.

On July 17, 1996, the inspector questioned the licensee regarding the

operable status of AHF-8A. This safety related component has been

operating with known excessive vibration levels since 1994 and the

monitoring data shows an increasing trend of vibration levels.

Licensee procedure CP-150, Identifying and Processing Operability

Concerns, Revision 1, dated May 7, 1996, paragraph 4.1.3 states in part

that the discovery of degraded conditions of components, where

performance is called into question, requires an operability

determination. CP-150, paragraph 4.2, Phase 2: Evaluation, states in

part that the shift supervisor on duty (SS00) evaluates the degraded  ;

condition for immediate disposition. If the component is important to

safety the SS0D makes an immediate disposition of either Operable,

Inoperable, or Complex Requiring Further Review. If the component

requires further review, the SSOD is to initiate an Operability Concern

Resolution (OCR). The 0CR is, among other requirements, to contain the

applicable Problem Report number and an immediate disposition.

10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, requires that

measures be established to assure that conditions adverse to quality,

such as failures, malfunctions, deficiencies, deviations, defective

material and equipment, and nonconformances are promptly identified and

corrected.

The failure of the licensee to initiate prompt corrective actions to

address the July 12, 1996 identification of an increasing trend in

excessive vibration on the spent fuel pump cooling fan motor, which was

originally identified in 1994, is first of two examples of a violation.

This violation will be tracked as the first example of VIO 50-302/96-08-

01, Failure to take timely corrective actions.

The system engineer put together an action plan for AHF-8A as follows:

Description of Problem

Vibration levels on the fan motor were first identified as high in 1992

through the Plant Equipment Condition Monitoring Program, and again

early in 1994 through REA 94-0026. The motor subsequently burned up in

May 1994 and was replaced. Vibrations have since increased to the

present level of about 2"/sec., which is extremely rough. Contributing

factors may include a mismatch between the adjustable dual groove sheave

and the non-adjustable double wide powerband belt which drives the fan.

14

Another factor seems to be insufficient stiffness of the sheet metal

housing upon which the motor base is mounted.

Proposed Resolution

1. Implement WR318701 to troubleshoot high vibration problem via MP-

531. Include correct size fixed motor sheave (pitch diameter must

provide desired fan RPM of approximately 1111) to replace original

adjustable sheave; determine whether existing dual powerband belt

is acceptable or whether two single belts should be used. This

may require a review of FIMIS documentation to determine how this

mismatch (if it is) may have occurred.

2. If Step 1 improves the overall vibrational characteristics, the

Rapid Response Team will take steps to make a permanent sheave /

belt change via PEERE, CGWR or MAR. It must be remembered that

i

this cooling unit is Safety Related. 1

3. If, upon installation of a correct sheave / belt configuration,

vibration continues to be amplified by flexing of the mounting

surface, stiffening of the housing will be pursued. REA 96-0727

is in place to begin this modification process.

On July 19, 1996, PR 96-0239, AHF-8A High Vibration, was initiated. The

SS0D dispositioned the AHF-8A as Conditionally Operable /Potentially l

Inoperable with a due date for 0CR AH-96-AHF-8A of September 19, 1996.

No corrective actions had been initiated at the time this report period

ended.

c. Conclusions

The Plant Equipment Condition Monitoring report is very comprehensive,

and is an excellent summary of the status of plant equipment. One

violation was identified for a failure of the licensee to initiate

prompt corrective actions.

E8 Miscellaneous Engineering Issues

E8.1 IFI 50-302/95-08-02, Corrective Actions for Makeup System Audit Findinas

a. Inspection Scope (92901)

The inspectors reviewed corrective action plan developed in response to

quality assessments audit 95-02-MAKP, 1995 Audit Report for Make-Up

System, February, 1995.

I

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_ )

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

.

I

d

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15

-

b. Observations and Findinas

4

In February 1995 an audit of the make up system disclosed numerous

, problems with the system (see IR 95-08 and IR 95-18). A Corrective

!

Action Plan (CAP) was documented in PR 95-0041. In June 1996 a follow-

up review, performed by quality assessments, was performed to verify the

status of the corrective action plan.

i The auditor reviewed the corrective actions for CAP item 1 and found

that the item was partially completed with a request from engineering to

open a CAP item to track additional engineering review of discrepancies

identified during subsequent piping walkdowns. CAP. item 3 was completed

with a request by. engineering to add two additional CAP items to develop-

, a DCN to modify the support drawings for MUH-518 and MUH-519 and to t

l' track completion of the work requests written to verify torque on-anchor

l bolts to MUH-807 and MUH-819. To address CAP item 5, systems engineers

-

initiated 19 work requests to correct deficient conditions identified in

. the original audit. This CAP item was closed, based on.the issuance of

j. the WRs.

. .

The auditor's review revealed that the tracking system for prs,

PRSTATUS, showed all CAP items completed for PR 95-0041, with no

. additional CAP items added. One of the 19 WRs written had been -

! completed. The other 18 WRs had not been planned. The failure to

, create additional steps in PRSTATUS to document the additional CAP items

'

or track the implementation of the CAP items, such as planning and

performing work activities, resulted in failure to complete.these

actions in a timely manner. Since engineering relied on PRSTATUS to

drive the activities committed to in the additional CAP items, these

additional corrective actions were not completed.

10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires that

measures be established to assure that conditions adverse to quality,

such as failures, malfunctions, deficiencies, deviations, defective

material and equipment, and nonconformances are promptly identified and

corrected. The failure of the licensee to take prompt and adequate

corrective actions to correct deficiencies identified in February 1995

during a make-up system audit is the second example of a violation, VIO

50-302/96-08-01, Failure to take timely corrective actions. This also

indicates a weakness in the PR tracking system. '

c. Conclusions

IFI 50-302/95-08-02 is closed. One Violation and one Weakness were

identified.

___ _

16

IV. Plant Support

R2 Status of Radiation Protection and Chemistry Facilities and Equipment

R2.1 Calibration of Reactor Buildina Air Sample Line Monitor RM-A6

a. Inspection Scope (84750, 83750)

The inspectors observed the calibration of the gaseous detector channel

on the Reactor Building Air Sample Line Monitor RM-A6 to evaluate the

adequacy of the licensee's process and procedures,

b. Observations and Findings

The licensee utilized CH-232, " Atmospheric Radiation Monitoring System

Calibration Procedure, Revision 30, dated August 6, 1996, for the

calibration of plant atmospheric monitors. The procedure was utilized

to calibrate all of the atmospheric monitors and was somewhat cumbersome

to use, in that it required the user to skip portions of the procedure

not applicable to the monitor being calibrated. Additionally, some of

the steps were not specific and could result in inconsistent actions.

For example, the step for connecting the radioactive calibration gas

loop in the observed calibration did not specify the specific connecting

points to the radiation monitor. The calibration gas loop connecting

points were not identified on the radiation monitors. The technicians

performing the calibration spent considerable time looking for the most ,

appropriate connecting points and selected the most direct path to the

detector chamber. The inspectors noted the radioactive gas could have .

been connected at several locations, which could have resulted in some

'

calibration problems if the equipment was not placed in an appropriate

valve alignment. The procedure also did not include sign-offs for valve

alignment verifications. These issues were discussed with licensee

personnel, and the inspectors learned that the licensee was developing

specific procedures for each monitor to make the calibration procedure

more user friendly. However, the inspectors noted that the draft

procedures also did not include valve position. verification sign-offs.

The inspectors discussed the value of the valve position verification

activities with licensee management. The licensee did not indicate

whether valve position verification sign-offs would be included in the

new procedures.

During the observation of calibration activities in the control room,

the inspectors noted numerous (approximately 10) deficiency and work

order tags on the radiation monitoring panel. Most of the radiation

monitoring system was fully operational and most of the monitors with a

deficiency tag remained operable to some extent. Many of the tags were

less than a month old, however, three were four to eight months old. An

assessment of the adequacy of the maintenance activities on the

radiation monitoring system was not made during this inspection.

However, the inspectors discussed the maintenance of the radiation

monitoring equipment in general with licensee management. The licensee

personnel reported that other maintenance activities having a higher

,

17

priority were resulting in the maintenance delays on the radiation

monitoring system. Tb" inspectors acknowledged that there were  :

maintenance activities having a higher safety significance, but reported :'

the operability of the plant radiation monitoring equipment was an

important component in detecting adverse plant and radiological

conditions of fission product barriers, which could warrant a higher

.

'

work priority assignment. The licensee management noted the inspector's

concerns, but did not make any specific commitments to revise l

l maintenance priorities for the radiation monitoring system.  ;

c. Conclusions

The calibration process was adequate. The technicians performing the  !

csH bration appeared to understand the characteristics of the radiation .

monitor being calibrated and the calibration procedure arJ process. The l

calibraticn procedure was properly utilized throughout th calibration

process. Instrument settings were proper considering the instrument's

response during the calibration. The quality controls in the

calibration procedure could be improved by identifying calibration

connecting points for gaseous source loop and inclusion of valve

position verification sign-offs.

The operability of plant radiation monitoring system could be improved

with more timely maintenance of system equipment.

R.3 Radiological Protection and Chemistry Procedures and Documentation

R3.1 Annual Radioloaical Environmental Operatina Report

a. Inspection Scope (84750)

The Annual Radiological Environmental Operating Report for 1995 was

reviewed to identify any adverse trends and to verify that the ,

requirements of Technical Specifications (TS) were met. j

i b. Observations and Findinas

The report indicated that plant operations in 1995 had not resulted in

any significant impact on the environment resulting from radiological

effluents. The inspectors compared the reported radiation measurements

in the 1995 report with those of previous years and did not identify any

adverse trends. The report was complete and met TS requirements.

c. Conclusions

No concerns with the 1995 Annual Environmental Monitoring Report were

i identified.

!

.

i

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,

l

18

R3.2 Annual Radioloaical Effluent Release Report

l a. Inspection Scope (84750)

The Annual Radiological Effluent Release Report for 1995 was reviewed to

identify any adverse trends and to verify that the requirements of TS

were met.

b. Observations and Findinas

The inspectors compared the reported measurements.in the 1995 report

with those of previous years and did not identify any adverse trends.

The quantity of radioactive liquid and gaseous effluents were generally '

l down from the previous year (1994) with the exception of liquid tritium,

l which was slightly higher than reported in the last few years. The

quantity of radioactive iodine released in gaseous effluents was

l reported as zero. The licensee did not have a refueling outage in 1995

l and as a result did not need to generate and release the additional

l water needed for refueling activities. The licensee also stored i

l additional water to take advantage of radioactive decay prior to its

release. Radioactive gases were also stored for radioactive decay prior

to their release. The licensee appeared to be effectively managing

l radiological effluents to maintain offsite doses as low as reasonably

i

achievable.

c. Conclusions j

l l

l The 1995 Effluent Report was complete and met TS requirements. No

concerns with the 1995 report were identified.

R3.3 Transportation of Radioactive Waste and Material l

l a. Inspection Scope (86750)

Shipping papers were reviewed to determine whether they met applicable I

regulatory requirements.

4

b. Observations and Findinas

The inspectors reviewed the licensee's documentation for selected

radioactive material and radioactive waste shipments to verify that the

( licensee was properly documenting radioactive material transportation

activities.

c. Conclusions

Reviewed shipping documentation appeared appropriate and in compliance

with applicable transportation regulations. No concerns with the

-

licensee's transportation records were identified.

i

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l

! 19

l R5 Staff Training and Qualification in Radiological Protection and  :

l Chemistry l

l

l R5.1 Review of Trainina Lesson Plans

a. Inspection Scope (84750)

The inspectors reviewed selected training lesson plans for staff

l performing functions related to the program areas reviewed during the

'

inspection. The selected training plans addressed activities associated

l with transportation of radioactive waste and material, calibration of

l radiation monitoring systems, and radioactive effluent monitoring and

l releases. The reviews were made to verify lesson plans included

appropriate information,

b. Observations and Findinas

In general, the lesson plans reviewed by the inspectors clearly

, addressed lesson objectives, addressed regulatory requirements and

referenced other related guidance and procedures. Basic physical

concepts, analytical techniques, equipment operation and system

descriptions were included in the lesson plans, as appropriate. Lesson

l plans also addressed proper documentation and response requirements and

l included appropriate laboratory exercises.

l

C. Conclusions

1

( The inspectors found the reviewed lesson plans addressed appropriate

topics in the program area. No concerns with the reviewed lesson plans

were identified.

F7 Quality Assurance in Fire Protection Activities

F7.1 96-04-FPEP. Audit Report of Fire Protection /Emeroency Plannina

l

a. Inspection Scope (40500)

l The inspectors reviewed the quality assessments audit, 96-04-FPEP, in

the areas of fire protection and emergency planning.

l b. Observations and Findinas

l

l No direct audit findings were identified. Seventeen weaknesses were

!

identified, which resulted in the issuance of PCs. Two of the PCs have

resulted in three problem reports being issued, after review by licensee ,

management. These problem reports addressed various problems with fire

protection surveillance procedures, inadequate corrective actions to

address concerns identified in LER 90-002, the failure to transfer fire

'

i protection technical specifications to the fire protection plan per GL 88-12, and failure to take corrective actions for questions on the fire

service tank water volume. The audit team did not consider that any of

these issues met the criteria for issuing a PR. However, the review of

- . _ . . _ . - .- = - - - . . ,

l

l

l

l 20

1

the PCs by line management did identify the issues as requiring a PR and

upgraded the concern.

l A total of 17 PCs were identified, which were evaluated by the audit

! team leader for trends or common concerns. The team leader reported

that there were four basic types of issues:

1. There were indications of a weakness at the first and second level

of defense of quality. These represent a deficiency that

originated at the worker level and was not detected during the

i

,

review and approval process.

2. There was the need for more attention to detail when documenting

l or otherwise describing activities.

3. There appeared to be an indication of a weakness in management

oversight function.

l 4. A weakness was identified in the timely distribution of industry

information specific to emergency planning issues.

!

c. Conclusions

l

) The review of the audit noted that programmatic assessments were made,

j however the assessments were restricted to the specific areas being

l audited. The licensee appears to be attempting to improve their audits

to include programmatic assessments, but the assessments produced to

date are limited in scope.

A weakness was identified in that a Quality Assurance audit failed to

l properly identify that two PCs issued during the audit should have been

characterized as prs.

l

1

'

V. Management Meetinos

l

X1 Exit Meeting Summary

The inspection scope and findings were summarized on August 9, 1996.

i The inspectors described the areas inspected and discussed in detail the

l inspection results listed below. Proprietary information is not

l contained in this report. Dissenting comments were not received from the

l licensee.

X3 Management Meeting Summary

On July 25, 1996 the IPAP team held a pre-exit meeting for the findings

identified during their inspection. A special public exit meeting was ,

held on August 6, 1996 at the CR-3 site. The meeting was attended by

NRC staff from Region II and from HQ. The results of this inspection

were issued on August 23, 1996 as Final Assessment Report 50-302/96-201.

!

!

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l 21 j

PARTIAL LIST OF PERSONS CONTACTED

Licensees

l

l

!

K. Baker, Manager, Nuclear Configuration Management  :

P. Beard, Senior Vice President Nuclear Operations )

G. Becker, Operations

C. Bergstrom, Jr., Manager, Nuclear Plant Operations Support

l G. Boldt, Vice President Nuclear Production

J. Campbell, Manager, Nuclear Security

'

l

J. Campbell, Assistant Plant Director, Maintenance and Radiation Protection

l W. Conklin, Jr., Director, Nuclear Operations Materials and Controls

i

R. Davis, Assistant Plant Director, Operations and Chemistry

D. DeMontfort, Manager, Nuclear Plant Operations  !

R. Enfinger, Manager, Safety Assessment Team i

R. Fuller, Manager, Nuclear Chemistry i

! B. Gutherman, Manager, Nuclear Licensing

i

G. Halnon, Manager, Nuclear Licensing

B. Hickle, Director, Nuclear Plant Operations

L. Kelley, Director, Nuclear Operations Site Support

, -

H. Koon, Manager, Nuclear Outages

l K. Lancaster, Manager, Nuclear Projects

l J. Maseda, Manager, Engineering Programs l

l P. McKee, Director, Quality Programs j

R. McLaughlin, Nuclear Regulatory Specialist '

i

'

B. Moore, Manager, Nuclear Integrated Scheduling

W. Rossfeld, Manager, Site Nuclear Services

J. Stephenson, Manager, Radiological Emergency Planning

l F. Sullivan, Manager, Nuclear Engineering Design

i

R. Widell, Director, Nuclear Operations Training

,

D. Wilder, Manager, Radiation Protection

<

l NRC

l

) J. Bartley, Resident Inspector, Farley NP (August 5 through 9,1996)

l W. Bearden, Reactor Inspector, Region II (August 6 through 9, 1996)

J. Cummins, IPAP Contractor (July 14 through 25,1996)

T. Foley, IPAP Team Member (July 14 through 25,1996)

R. Gallo, Chief, Special Inspection Branch, NRR (August 6,1996)

,

!

A. Gibson, Director, Div. of Reactor Safety, Region II (July 25 and August 6,

1996)

G. Hopper, Reactor Engineer, Region II (August 5 through 9, 1996)

J. Isom, IPAP Team Member (July 14 through 25,1996)

,

J. Jacobson, IPAP Team Leader (July 14 through 25, 1996 and August 6, 1996)

'

S. Klementowicz, IPAP Team Member (July 14 through 25,1996)

K. Landis, Branch Chief, Region II (August I through 2, 1996)

R. Mathew, IPAP Team Member (July 14 through 25,1996)

0. Mazzoni, IPAP Contractor (July 14 through 25,1996)

D. Norkin, Section Chief, Special Inspection Branch, NRR (July 25,1996)

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22

L. Raghavan, Project Manager, NRR (August 6, 1996)

M. Shylamberg, IPAP Contractor (July 14 through 25,1996)

D. Solorio, IPAP Team Member (July 14 through 25,1996)

L. Stratton, Physical Security Specialist, Region II (July 29 through August

2,1996)

J. Williams, Project Manager, NRR (August 6, 1996)

F. Wright, Senior Radiation Specialist, Region II (August 5 through 9, 1996)

INSPECTION PROCEDURES USED

IP 37551: Onsite Engineering

IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving and

Preventing Problems

IP 61726: Surveillance Observations

IP 62703: Maintenance Observation

IP 62707: Conduct of Maintenance

IP 71707: Plant Operations

IP 83750: Ocupational Radiation Exposure

IP 84750: Radioactive Waste Treatment and Effluent and Environmental

Monitoring

IP 86750: Solid Radioactive Waste Management and Transportation of

Radioactive Materials

IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power

Reactor Facilities

IP 92901: Followup - Operations

IP 92902: Followup - Maintenance

IP 92903: Followup - Engineering

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

Type Item Number Status Description and Reference

VIO 50-302/96-08-01 Open (Two Examples) Failure to take

timely corrective actions for make-

up system audit findings and

excessive vibration on a spent fuel

pump cooling fan motor. (paragraphs

E8.1 & E2.1)

IFI 50-302/96-08-02 Open Reactor building cavity cooling

piping thermal relief protection.

(paragraph M8.2)

l

!

23

Closed

l Type Item Number Status Description and Reference

!

IFI 50-302/95-08-02 Closed Corrective actions for make-up

system audit findings. (paragraph

E8.1)

LER 50-302/96-017-01 Closed Reactor trip on high reactor coolant

pressure during turbine testing.

(paragraph 08.1)

Discussed

,

TYDe Item Number Status Description and Reference

l

l VIO 50-302/93-16-07 Open Inadequate E0P and AP Procedures.

(paragraph 08.2)

l

URI 50-302/96-201-04 Open Non-safety related positioners used

on safety related decay heat cooling

system air operated valves.

(paragraph M8.3)

LIST OF ACRONYMS USED ,

ac - Alternating Current

ADI - Absolute Drift Indications

AHD - Air Handling Vent and Cooling Damper

AHV - Air Handling Vent and Cooling Valve

AI - Administrative Instruction

ALARA - As Low as Reasonably Achievable

ANSI - American National Standards Institute

ANSS - Assistant Nuclear Shift Supervisor

APC - Alternate Plugging Criteria

ASME - American Society of Mechanical Engineers -

ASV - Auxiliary Steam Valve

B&PV - Boiler and Pressure Vessel

B&W - Babcock & Wilcox

BS - Building Spray

BSP - Building Spray Pump

BVT - Below Voltage Threshold

BWST - Borated Water Storage Tank

CAL - Confirmatory Action Letter

CAP - Corrective Action Plan

l CCTV - Closed Circuit Television

! CFR - Code of Federal Regulations

CFT - Core Flood Tank

CFV - Core Flood Valve

CGWR - Commercial Grade Work Request

CI - Cavity Cooling

CP - Compliance Procedure

t

-. - -. . _-. .

l

24 l

l

l CREVS - Control Room Emergency Ventilation System

!

CR3 - Crystal River Unit 3

CST - Condensate Storage Tank

dc - Direct Current

DC - Decay Heat Closed Cycle Cooling

l DCHE - DC Heat Exchanger

DCN - Design Change Notice

DEV - Deviation

DFP - Diesel Fuel Pump

DH - Decay Heat

' DHHE - Decay Heat Heat Exchanger  !

l

DHP - Decay Heat Pump

j DHR - Decay Heat Removal

l DHV - Decay Heat Valve

!

DNP0 - Director, Nuclear Plant Operations j

dp - Differential Pressure '

'

EA - Enforcement Action

ECCS - Emergency Core Cooling System (s)

EDBD - Enhanced Design Basis Document  ;

EEI - Escalation Enforcement Item i

'

EFIC - Emergency Feedwater Initiation and Control

EFP - Emergency Feedwater Pump ,

EFT - Emergency Feedwater Tank i

EFW - Emergency Feedwater l

,

EFV - Emergency Feedwater Valve

- Emergency Diesel Generators

'

EGDG

EM - Emergency Plan Implementing Procedure

E0P - Emergency Operating Procedure 1

.

EP - Emergency Preparedness

l ES - Engineered Safeguards

ESF - Engineered Safeguards Feature

'

ESAS - Engineered Safety Actuation System

ET - Eddy Current Test

EVS - Emergency Ventilation System

F - Fahrenheit ,

! FIMIS - Fully Integrated Materials Information System

l FPC - Florida Power Corporation

FSAR - Final Safety Analysis Report

FWP - Feedwater Pump

FWV - Feedwater Valve

GL - Generic Letter

gpm - Gallons Per Minute

HELB - High Energy Line Break

HP - Health Physics

l HPI - High Pressure Injection

!- in. Hg - Inches of Mercury

l I&C - Instrumentation and Control

ICC - Inadequate Core Cooling

ICS - Integrated Control System

IEEE - Institute of Electrical and Electronics Engineers

IFI - Inspection Followup Item

INP0 - Institute of Nuclear Power Operations

l

i

L

!

l

25

IP -

Inspection _ Procedure

IR - Inspection Report

'

ISA- -

Instrument Society of America

ISI -

Inservice Inspection

ISO - Isometric Drawing

IST' -

Inservice Test

ITS- -

Improved Technical Specification

JC0 -

Justification for Continued Operation

JPM - Job Performance Measure

Kv' -

Kilovolt

i Kw -- Kilowatt

LC0 - Limiting-Condition for Operation

LER -

Licensee Event Report

LOCA - Loss of Coolant Accident

LOOP - Loss of Offsite Power

LTE -

Lower Tube End

! LTS -

Lower Tube Sheet

l MAR -

Modification Approval Record

MCB -

- Main Control Board

l MCC - Motor Control Center

! MFW -

Main Feedwater

l MOV -

Motor Operated Valve

i M0 VATS - Motor Operated Valve Analysis and Test System

MP -

Maintenance Procedure

MRP -

Management Review Panel

MSSV -

Main Steam Safety Valve

MSV - Main Steam Valve

j' MT . - Magnetic Particle Testing

l MU -- Make Up

,

MVP - Make-up Pump l

l

'

MUT -

Make-up Tank

MUV - Make-up Valve

MW . -

Megawatt

! NCV -

Non-cited Violation-

f. NDE -

Nondestructive Examination

i NEP -

Nuclear Engineering Procedure

l N0D -

Nuclear Operations Department

NOV -

Notice of Violation ,

NPSH -

Net Positive Suction Head l

NQI - Non-Quantifiable Indication l

NRC -

Nuclear Regulatory Commission

NRR -

Office of Nuclear Reactor Regulation

l NSM -

Nuclear Shift Manager

NSSS -

Nuclear Steam System Supplier  !

NUREG -

NRC technical report designation i

OCR - Operability Concerns Resolution )

OP - Operating Procedure l

, OSB -

Operations Study Book

i.' OTSG - Once Through Steam Generator j

PEERE -

Plant Equipment Equivalency Replacement Evaluation '

4 -PM -

Preventive Maintenance l

j' PORV - Power Operatsd Relief Valve  !

!  ;

I

!

!

1

26

ppb - Parts Per Billion  :

PR - Problem Report l

PRC - Plant Review Committee l

PSI - Preservice Inspection

psig - pounds per square inch gauge i

PT - Liquid Penetrant  ;

PTLR - Pressure and Temperature Limits Report l

QC - Quality Control l

QA - Quality Assurance

QAP - Quality Assurance Procedure

RB - Reactor Building

RC - Reactor Coolant

RCA - Radiation Control Area

RCP - Reactor Coolant Pump

RCPPM - Reactor Coolant Pump Power Monitor

RCS - Reactor Coolant-System

REA - Request for Engineering Assistance

RF0 - Refueling Outage

RG - Regulatory Guide

RO - Reactor Operator

RPC - Rotating Pancake Coil

RP&C - Radiological Protection and Chemistry

RT - Radiographic Inspection

RW - Nuclear Services and Decay Heat Seawater

RWP - Nuclear Services and Decay Heat Seawater Pump

RWV - Nuclear Services and Decay Heat Seawater Valve

SALP - Systematic Assessment of Licensee Performance

SAT - Systems Approach to Training

SDT - Station Drain Tank

SER - Safety Evaluation Report *

SFPD - Safety Function Determination Program '

SG - Steam Generator

SOER - Significant Operating Event Report

SP - Surveillance Procedure

SR - Surveillance Requirement

SR0 - Senior Reactor Operator

SS0D - Shift Supervisor on Duty

STI - Short Term Instruction

SW - Nuclear Services Closed Cycle Cooling System

SWHE - SW Heat Exchanger

SWP - SW System Pump

SWV - SW System Valve

T, - Cold Leg Temperature -

TI - Temporary Instruction

TMAR - Temporary Modification Approval Record

TMI - Three Mile Island

TS - Technical Specification ,

TSC - Technical Support Center

TSCR - Technical Specification Change Request

TW - Through Wall

UAf - A measure of heat exchanger effectiveness

4

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27

UHS - Ultimate Heat Sink

URI - Unresolved Item

USAS - United States of America Standards

UT - Ultrasonic Test

VIO - Violation

V0TES - Valve Operation Test and Evaluation System

Vpp - Volts point-to-point

WR - Work Request

i