ML20137V727

Insp Rept 50-302/97-04 on 970127-0321.No Violations Noted. Major Areas Inspected:Operations & Engineering

ML20137V727
Person / Time
Site:	Crystal River
Issue date:	04/11/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20137V706	List: ML20137V702 ML20137V727
References
50-302-97-04, 50-302-97-4, NUDOCS 9704170383
Download: ML20137V727 (15)

See also: IR 05000302/1997004

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U.S. NUCLEAR REGULATORY COMMISSION

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1 REGION 2

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! Docket No: 50-302

License No: DPR-72

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Report No: 50-302/97-04

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Licensee: Florida Power Corporation

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Facility

Crystal River 3 Nuclear Station

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i Location: 15760 West Power Line Street

j Crystal River. FL 34428-6708

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Dates: January 27 through March 21, 1997

Inspectors: S. Cahill. Senior Resident Inspector, paragraph.01.1

T. Cooper. Resident Inspector paragraph 01.1 '

P. Fillion. Reactor Inspector, paragraph E8.2

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R; Schin. Reactor Inspector. paragraph E8.1

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Approved by: H. Christensen. Chief. Engineering Branch

} Division of Reactor Safety

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Enclosure

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9704170393 970411

PDR ADOCK 05000302

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EXECUTIVE SUMMARY

Crystal River 3 Nuclear Station

NRC Inspection Report 50-302/97-04

This special inspection included aspects of licensee operations and

engineering functional areas. The purpose of the inspection was to follow up

on the licensee not reporting the emergency feedwater net positive suction

head issue and to follow up on other licensee problems in reporting conditions

to the NRC as required.

Doerations

- A weakness was identified regarding an Emergency Action Level classification

i that was not made in a timely manner following a transformer explosion at an

aajacent fossil power plant. An apparent Violation (EEI 50-302/97-04-01) was  ;

identified for failure to make an emergency phone report within the time

requirements of 10 CFR 73.71. Another apparent Violation (EEI 50-302/97-04-

02) was identified for failure to hand carry a suspected reportable issue to

the Shift Manager for a reportability review as required by the licensee's I

procedures (Section 01.1).

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Enoineerina

An apparent Violation (EEI 50-302/97-04-03) was identified for failure to l

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' report to the NRC the outside design basis condition, involving-insufficient l

emergency feedwater pump net positive suction head, that was identified in

April 1996. This was a failure to report a condition that resulted in )

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l escalated enforcement, and the failure to report the condition contributed to

, a lack of timely NRC awareness and review of the condition. As'a result, the

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NRC missed an opportunity to ensure that appropriate corrective actions were

taken to address an outside design basis condition. This failure to report

was also a repeat of 3revious Violations 50-302/94-27-02, 50-302/94-27-03. and )

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50-302/96-06-06. whic1 involved failures to report outside design basis '

conditions to the NRC as required by 10 CFR 50.72 and 50.73. (Section E8.1).

A second example of ap)arent Violation EEI 50-302/97-04-03 was identified for

failure to report to tie NRC in a timely manner the outside design basis

condition, involving a non-safety-related transfer switch installed in safety-

related emergency safeguards status indicating light circuitry, that was

identified in December 1995. This example also involved a concern with

inaccurate information in LER 96-19 regarding the date on which the engineer

discovered the nonconforming condition and With the related failure of the LER

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to address, or include corrective action for untimely engineering review of

l- the nonconforming condition. (Section E8.2) '

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The inspectors assessed the licensee's performance in the five areas of

continuing NRC concern.in the following paragraphs: the assessment is limited

to the specific issue addressed in the respective paragraph.

l NRC AREA 0F CONCERN ASSESSMENT PARAGRAPH

01.1 E8.1 E8.2

Hanagement Oversight 1 I- 1

Engineering Effectiveness A I I

Knowledge of Design Basis I A

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Compliance With Regulations I

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Operator Performance I I A

5 - Superior G = Good A = Adequate / Acceptable I = Inadequate

Blar.k - Not Evaluated / Insufficient Information

01.1: Timeliness of Recent Licensee Reporting to the NRC

E8.1: Reporting of Emergency Feedwater Net Positive Suction Head Condition

E8.2: Reporting of Non-Safety-Related Transfer Switch Used in Safety-Related

Engineered Safeguards Status Indicating Light Circuitry

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Report Details

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L. Ooerations

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01 Conduct of Operations

] 01.1 Timeliness of Recent Licensee Reoortina to the NRC

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a. Insoection Scooe (71707)

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The inspectors followed up on three observed examples of re)orting l

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deficiencies. The licensee had several potentially reporta)le events i

that were not thoroughly evaluated within the required time to ensure  !

NRC notification time requirements could be fulfilled.

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4 b. Observations and Findinas

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On January 30, 1997, at approximately 1:17 a.m.. a main step-up

transformer at the adjacent coal electric generation plant. Crystal '

l River Unit-4 (CR4), exploded and caught fire. The force of the

' explosion lifted the transformer off of its base and toppled it onto its

side. Although the impact on the nuclear plant. Crystal River Unit 3

(CR3), was only limited to-a switchgear perturbation due to CR4  ;

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separating from the grid, the licensee's Radiological Emergency Response !

Plan. Revision 16. requires declaration of a Notice of Unusual Event 1

(NOUE) classification for a " severe explosion near or within the 0.83 '

,' Site Boundary but not affecting plant operations". CR4 is approximately

0.7 miles away from the nuclear plant. The control room operators

i originally believed the event was a fire that did not involve a severe

3 explosion, although the Shift Manager (SM) log referred to the event as

an explosion and plant management discussed the event as an explosion at

the Plan of the Day meeting at 8:00'a.m. After being questioned by the

licensee's Emergency Preparedness Manager about the lack of a

declaration and upon receiving further information that indicated the

transformer failure was an explosion. the Shift Supervisor on Duty

(SSOD) administratively entered and immediately exited a NOUE, at

a) proximately 1:45 a.m.. over 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the event. The SS00 made

tie subsequent 10 C R 50.72 report to the NRC Operations Center within

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one hour of the event classification as required. The licensee

initiated corrective action program precursor cards (PC) 97-0680 and 97-

0724 to investigate and correct the cause of the delay. The inspectors

concluded the event classification and subsequent notification were not

. timely in that sufficient information was available and well known

i shortly after the event for the SS00 to make the classification and

i notifications.Section IV of Appendix E of 10 CFR 50 requires licensees

to have the capability to notify offsite authorities within 15 minutes

of the declaration of an emergency. . 10 CFR 50.72 requires that the

licensee notify the NRC not later than one hour after the time the

licensee declares one of the emergency classes. The 15 minute and the

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one-hour periods are measured from the time of declaration of an

emergency class. Although the regulations do not specify any time

requirement for the classification process itself. they do im)ly that

classification should be made without delay. The SS00 did ma<e a

preliminary and timely evaluation of the event against the

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classification requirements shortly after the explosion and did refer to  !

them again 16ter in the event. However, the licensee's investigation

determined that the SS00 did not adequately pursue final resolution of

the classification determination by investigating and gathering the  ;

available information. The inspectors concluded the' delay was i

indicative of a weakness in the licensee's process for promptly

assessing and reporting events.

The second example of reporting deficiencies also occurred on January

30, 1997. At 6:45 p.m. a potential breach in the Protected Area as a

result of maintenance work on a main condenser circulating waterbox was

discovered by a security officer. A Protected Area breach is a one-hour

reportable event per 10 CFR 73.71 but it was not reported until 1:18 .

a m. on January 31, 1997. Although the security force needed some time

after initial discovery to assess the o)ening to determine if it was

above the allowable security plan breac1 size of 96 square inches. these

efforts were not expedited sufficiently to make a timely verification.

Proper priority was not placed on the investigation by shift management

considering it was a suspected reportable problem so the necessary

coordination of several plant groups was limited. Efforts were

suspended during Operations shift turnover, and delays were encountered

due to confined space entry permit requirements. Consequently, the

inspector determined from interviews with licensee personnel that the

licensee did not determine that the breach was reportable until

approximately 10:30 p.m. Then the problem was not officially screened

for reportability by the SM until 12:20 a.m. on January 31 while

paperwork documenting the breach was prepared by the security staff. 4

Some members of the licensee's staff were not aware that the one-hour l

reportability requirement starts at the time of recognition or

10:30 p.m., and not the time of re)orting the event officially to the )

SM at 12:20 a.m. on January 31. T1e inspectors concluded this did not ,

meet the requirements of 10 CFR 73.71 to report the event to the NRC

within one hour from the time of discovery of the event. A report was

required to have been made by 11:30 p.m. Consequently this delay was  :

identified as ap)arent Violation EEI 50-302/97-04-01. Failure to Hake an '

Emergency Phone Report Within the Time Requirements of 10 CFR 73.71.

The third example of reporting deficiencies occurred on February 6.

1997, when corrective action document PC 97-055 was rece Ned by the SM

for review. This PC documented a situation identified during NRC

Generic Letter 96-06 reviews where reactor building system components

were potentially outside their design basis because they were not

designed to withstand post-accident conditions. Precursor Card 97-055

was generated on January 31 but was not received by the SM for

reportability screening until February 6. Although part of the delay

was due to verifying the scope and; extent of the issue prior to

submitting it for review, which is%cceptable. a portion of the delay

was due to the PC originator mailing it to the SM. This was contrary to

Compliance Procedure (CP) 111. Processing of Precursor Cards for

Corrective Action Program. Revision 55. which recuires all PCs that are

suspected reportable to be hand carried to the SF for immediate

evaluation. Precusor Card 97-055 was annotated as potentially

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reportable due to the suspected design basis problem and was therefore

required to be hand carried to the SM for immediate review for

reportability requirements. Although the SM determined that only a

written 30 day Licensee Event Report (LER) per 10 CFR 50.73 was

required, the problems were significant enough that a 4-hour phone

report per 10 CFR 50.72 could potentially have been required. It would

not have been made in time due to the several day delay from mailing the

PC. The inspector determined the SM's reportability evaluation was not

timely relative to the recognition of the design basis problem and would

not have met the 4-hour reporting requirement if it had been applicable.

The inspectors identified this as apparent Violation (EEI 50-302/97-04-

02). Failure to Hand Carry a Suspected Reportable Issue to the Shift

Manager for Reportability Review.

The licensee initiated PC 97-0841 to evaluate if the above three

problems had similar root causes. This effort was not yet finalized at

the end of the report period and was being incorporated into the

corrective actions for Item OP-4. Upgrade the Operability /Reportability

(CP-150/151) Program, on the licensee's and NRC's restart restraint

list. The inspector observed that the licensee's Quality Assurance

group responded to these problems and performed two surveillance

inspections on the licensee's reporting process that found similar

deficiencies.

c. Conclusions

The inspectors concluded these examples were indicative of deficiencies

in the licensee's reportability screening process. All screening was

done via PCs reviewed by the SM which can result in delays while

paperwork to complete a PC is generated. The existing prccess was not

always followed. The inspectors also concluded that some licensee

personnel did not understand that the reporting time requirements were

from time of discovery versus submittal of a PC for review, which

created further delays. Additionally, proper priority was not placed on

determining the correct status of the event expeditiously in order to

make a timely reportability determination.

The inspector assessed the licensee *s performance, with respect to this >

issue. in the five areas of continuing NRC concern:

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. Management Oversight - Inadequate

. Engineering Effectiveness - Adequate

+ Knowledge of the Design Basis - Not Applicable

Compliance with Regulations - Inadequate

. Operator Performance - Inadequate

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II. Enaineering

E8 Miscellaneous Engineering Issues ,

E8.1 (Ooen) EEI 50-302/96-19-03. EFW NPSH US0 due to Inadeauate 10 CFR 50.59

Safety Evaluation for a Modification  !

(Closed) VIO 50-302/94-27-02 (dated January 26. 1995). Failure to Make

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Two 10 CFR 50.73 Reoorts to the NRC Within the Reauired Time (olus one

subseauent additional examole in IR 95-02)

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(Closed) VIO 50-302/94-27-03 (dated January 26. 1995). Failure to Make a I

10 CFR 50.72 Reoort to the NRC Within the Reauired Time (olus one

abseauent additional examole in IR 95-08)

(Closed) VIO 50-302/96-06-06 (dated July 27. 1996). Failure to Notify

the NRC of a Condition Outside the Accendix R Licensina Desian Basis in

a Timely Manner

a. Insoection Scoce (92903)

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The inspector noted that EEI 96-19-03 involved a condition apparently

outside the design basis of the plant that the licensee had identified

in April 1996 and that. a'., of January 27. 1997, the licensee had not l

reported to the NRC. As described in Inspection Report (IR) 96-19. the i

condition had existed from 1987 through April 1996. Licensee PC 96-2196

dated April 20. 1996, had identified the condition and engineering .

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analysis had confirmed it that same month. The condition involved

insufficient net positive suction head (NPSH) for the turbine-driven

emergency feedwater (EFW) pump in a certain accident scenario [ loss of

coolant accident (LOCA) and loss of offsite power (LOOP) with loss of

the B battery, which would fail the B Emergency Diesel Generator (EDG) )

and also fail open the discharge flow control valves for the turbine-

l driven EFW pump]. In that scenario, the turbine-driven EFW pump would

l automatically start and go to runout, with insufficient NPSH. Also, as

! described in the Final Safety Evaluation Report (FSAR). in that scenario

l the A EDG would rely on the operation of the B train turbine-driven EFW-

Sump to share the EFW flow requirements with the A train motor-driven

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EFW pump in order to maintain the A EDG within its electrical loading

limits.

In this ins ection. the inspector followed up on the above reportability

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issue and a so followed up on the licensee's corrective actions for

three previous violations that involved inadequate reportin9 of outside

design basfs conditions. .

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b. Observations and Findinas, t '

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In response to the inspector's questions regarding' reportability of the

EFW NPSH issue, the licensee initiated PC 97-0052 on January 28. 1997.

In reviewing the PC. the licensee concluded the same day that the

condition was outside the design basis of the plant, that the condition

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was reportable in accordance with 10 CFR 50.73, and that the condition

had not been re)orted. The licensee subsequently reported the EFW NPSH

condition in LE1 97-001. Ineffective Change Management Results -in ,

Unrecognized NPSH Issue Affecting Emergency Feedwater Availability. '

dated February 27, 1997. i

The inspector noted that the EFW NPSH condition represented a  ;

significant safety concern that had warranted NRC escalated enforcement

action (it was addressed at an enforcement conference on January 24

1997). The inspector also noted that the plant had been shut down in

April and May 1996 when this condition was identified, had operated i

between May 1996 and September 1996, and had been shut down since then.

The licensee's failure to report .the condition had contributed to a lack '

of timely NRC awareness and review of the condition. As a result. the

NRC had missed an opportunity to ensure that appropriate corrective

actions were taken to address an outside design basis condition. The

related condition resulted from the licensee's modification to ASV-204

to correct the EFW NPSH condition, and was also a subject of escalated

enforcement discussed at the January 24. 1997, enforcement conference.

The inspector reviewed the EFW NPSH issue with engineering and licensing

3ersonnel who had been involved with it, and concluded that the licensee

lad many opportunities to recognize that the condition was outside the

design basis of the plant and to report it. PC 96-2196 was reviewed in

A)ril 1996 for reportability by the Nuclear SM and by the plant managers j

w1o were present at the daily PC review meeting. The PC did not result

in a Problem Report and did not receive a formal documented operability

or reportability review. Engineering management and the Plant Review

Committee (PRC) reviewed the condition in April 1996 when they approved

the ASV-204 modification to correct the condition. Licensing reviewed

the condition and mentioned the EFW NPSH concern in LER 96-20 on EDG

loading but did not identify the EFW NPSH concern as a condition outside

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the design basis. The licensee's ASV-204 root cause team reviewed the

condition but did not identify that it was reportable. The licensee's

senior management, in ) reparation for the January 24, 1997, enforcement

conference, reviewed tie condition (it was the subject of an apparent

violation) but did not identify that it was reportable and was not

reported.

' The inspector reviewed the licensee's corrective actions in response to

three previous violations for failing to report conditions outside the

design basis as required by 10 CFR 50.72 and 10 CFR 50,73, to assess

whether those actions should have prevented the failure to report the

EFW NPSH condition. The inspector verified that the licensee had

accomplished the corrective actions.for Violations 94-27-02, 94-27-03,

and 96-06-06. as stated in their responses to the Notices of Violation.

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Reporting each condition to the NRC.

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Revising CP-111. Initiation and Processing of Precursor Cards and

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Problem Reports, to include steps that direct the originator to

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1mmediately notify the Nuclear SM if the issue is believed to

involve safety, reportability, or operability.

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Daily review of all new Precursor Cards where the Director.

' Nuclear Plant Operations (DNPO) and other line managers can assist

in the determination of reportability.

Issuing a letter from the DNP0 to all nuclear operations personnel

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about the reportability of exceeding the breach allowance of the

,. control complex habitability envelope (CCHE).

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Submitting a Technical Specification (TS) change request to the :

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NRC to address the CCHE. including applicable completion times and

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surveillance requirements.

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Issuing a new procedure. CP-150. Identifying and Processing

, Operability Concerns, in October, 1995.

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The licensee's corrective actions for the three violations were

completed in 1995. In addition, the inspector noted that the licensee

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had issued another new procedure, CP-151. External Reporting

Requirements, in November 1996. Also, the licensee had recently revised

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the Corrective Action Program. including a more extensive review process

for Precursor Cards. Violations 50-302/94-27-02. 50-302/94-27-03, and

50-302/96-06-06 are closed.

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c. Conclusions

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The inspector concluded that the failure to report to the NRC the

" emergency feedwater net positive suction head outside design basis

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condition that was identified in April 1996 was an apparent Violation of

10 CFR 50.73. The licensee's failure to report the condition ,

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contributed to a lack of timely NRC awareness and review of the

. condition. As a result, the NRC missed an opportunity to ensure that

appropriate corrective actions were taken to address an outside design

basis condition. This failure to report was a repeat of previous

Violations 50-302/94-27-02, 50-302/94-27-03, and 50-302/96-06-06 which

involved failures to report outside design basis conditions to the NRC

' as required by 10 CFR 50.72 and 50.73. This apparent Violation is '

identified as EEI 50-302/97-04-01. Repeat Failure to Report Outside

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Design Basis Conditions to the NRC.

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The inspector assessed the licensee's performance. with respect-to this

issue, in the five areas of continuing NRC concern:

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. ManagementOversight-Inadequite

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Engineering Effectiveness - Inadequate

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Knowledge of the Design Basis - Inadequate

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Compliance with Regulations - Inadequate

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. Operator Performance - Inadequate

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E8.2 (Closed) Unresolved Item 96-06-03. Non-Safety-Related Transfer Switch

Used in ES Statgs Indicatina Licht Circuitry

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(Ocen) LER 96-19. Classification of Transfer Switch Causes Potential for

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Loss of Power to ES Status Lichts

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In PC 95-2770. dated December 4.1995, the licensee had identified that

manual transfer switch ESCP-1 did not meet the requirements with regard

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' to qualification of the equipment. This transfer switch, which is

< locatedbreakers,

circuit on a wallbus

in the

bar,main

and control room. consists of four molded-case

an enclosure. ESCP-1 is in the power

supply circuit for the equipment status monitoring panel on the main

control board. The transfer switch is original plant equipment. and it

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provides a means to energize the equipment status monitoring panel from

' either vital bus 3A (train A) or 3B (train B). The switch is arranged

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such that, in the normal alignment, the train A power source energizes

status lamps for train A equipment, and similar for train B. The

equipment status monitoring panel falls under the requirements of

Regulatory Guide (RG) 1.97. Instrumentation for Light Water Cooled  !

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Nuclear Reactors to Assess the Plant and Environs Conditions During and

Following an Accident. RG 1.97 requires that the monitoring

i 3anel meet

several design criteria including seismic events and that it ]e treated

as safet

however,y-related. There is no requirement to have the transfer switch:  ;

since it is installed in the circuit. it must also meet all the i

' requirements that apply to the monitoring panel and its power supply.

The problem identified by the license was that ESCP-1 was not purchased i

safety-related and was not necessarily seismically qualified. '

As stated above. the non-conforming condition was discovered in December

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1995 and documented in PC 95-2770 on December 4. 1995. According to

proceJure. PC 95-2770 was reviewed by the Nuclear SM. His instructions.

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dated December 6, 1995, were to evaluate and respond, which were in

accord with the recommendations of the originating engineer. This meant

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the engineers should determine whether the equipment can be accepted as

is by perfcrming an upgrade evaluation.

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Evidence indicates that this evaluation was not performed in a timely

manner commensurate with the importance to safety. On June 13, 1996,

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- the licensee determined that the issue warranted a Problem Report and

PR 96-195 was initiated. Also, an operability evaluation was performed

(OCR-96-ESCP-1) which concluded that the equipment was non-conforming

but OPERABLE. On June 14, 1996. the licensee reported this non-

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conforming condition to the NRC by telephone pursuant to 10 CFR

50.72(b)(1)(11)(B) which requires a one-hour report. In addition. '

compensatory measures were initiated. i.e. red tag of alternate position

circuit breakers. LER 96-19 was submitted on July.15. 1996, pursuant to

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.10 CFR 50.73.

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The inspector determined that a six-month delay in performing the

operability evaluation was not commensurate with the safety significance

of the issue. Precursor Card 95-277 initiated the proper evaluation.

However, no time limit for completion was specified, and the controlling

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procedure. CP-111. Initiation and Processing of Precursor Cards and

Problem Reports, Rev 54. did not specify any time limit for this type

evaluation.

The inspector noted that Revision 55 of Compliance Procedure CP-111.

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dated November 22. 1996. requires in section 4.3.3.5.1. that Nuclear

Operations Engineering is to validate suspected Design Basis Issues or

unanalyzed conditions within 10 working days.

The LER states that the corrective action will be to either remove or

replace transfer switch ESCP-1. Since that time the licensee has

prepared modification 97-01-03-01 to install a fully qualified switch in

place of the existing one. The inspector did not review this

modification package, because it was not officially issued. The

inspector noted that replacement of ESCP-1 was tracked by the licensee

as their Restart Item D-21. to be completed prior to plant restart.

The six-month delay in reporting the non-conforming transfer switch to

the NRC constitutes a violation of 10 CFR 50.72 and 50.73, which require

that reports be made within one hour of the occurrence of the event and

within thirty days of the discovery of the event. respectively. The

root cause for the late report was the same as the cause for the

untimely corrective action mentioned above; i.e., since the evaluation

was delayed, recognition of the need for reporting was delayed. The

matter is identified as a second example of apparent Violation EEI 50-

302/97-04-01. Repeat Failure to Report Outside Design Basis Conditions

to the NRC.

Inspection activity related to this issue included the following:

e The inspector examined the equipment status monitoring panel on

the main control board, and noted that it included indications for

the operator to confirm the position of certain containment

isolation valves. The inspector then reviewed the licensee's

submittal made pursuant RG 1.97 on FMrch 21, 1988, and the Design

Basis Document for Post-Accident Monitoring Instrumentation. Both

these documents indicated that automatic containment isolation

valve position was a Type B. Category 1. variable as defined by RG

1.97. Therefore, the transfer switch in question was required to

be fully qualified. The fact that it was not fully qualified

created the potential (i.e. assuming failure of the switch) that

control room indications needed to mitigate the consequences of an

accident would not be available,

o The inspector verified that switch ESCP-1 was red tagged to ensure

that it remained in the normal alignment as stated under

corrective action in the LER) The tag number was Eco No. 96-07-

05-6. dated July 10. 1996.

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e The inspector verified that other components in the power supply

circuit to the monitoring panel, namely 115 - 25 V transformers OK-

and OL, were purchased safety-related through review of

documentation. It appears these were originally safety-related.

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e The inspector reviewed the operability evaluation, and found that

it met the guidance of Generic Letter 91-18. Resolution of

Degraded and Nonconforming Conditions and on Operability,

e The inspector noted that LER 96-19 states that the engineer's

discovery of the nonconformance occurred on June 13, 1996. 1his

date apparently represents the date that the need for a report to

the NRC was recognized as opposed to the initial discovery date.

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PC 95-2770, dated December 4.1995. clearly documented the

engineer's discovery of the nonconformance at that time. In this

regard, the LER is_ inaccurate. As written, the LER implies that

the corrective action program was effective. In fact the

corrective actions were not timel

Also. LER 96-19 did not address,ory,include

as described in thisaction

corrective section.

for.

the untimely engineering review of the nonconforming condition.

In summary: a second example of apparent Violation EEI 590-302/97-04-

03. Repeat Failure to Report Conditions to the NRC, was identified for

failure to report to the NRC in a timely manner the outside design basis

condition, with a non-safety-related transfer switch installed in

safety-related emergency safeguards status indicating light circuitry,

that was identified in Decenber 1995. This example also involved an

incorrect date for the discevery of the nonconformance. The LER also

did not recognize or include corrective action for the untimely l

engineering review of the nonconforming condition. Unresolved Item l

96-06-03 is closed. LER 96-19 Classification of Transfer Switch Causes '

Potential for Loss of Power to ES Status Lights, remains o)en for NRC i

verification that the licensee's modification to correct tie problem is i

completed. l

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-With regard to Unresolved Item 96-06-03, the inspector assessed the

licensee's performance, for the time period of June 1996 to present. in

the five NRC continuing ares of concern as follows

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• Management Oversight - Inadequate

• Engineering Effectiveness - Inadecuate

• Knowledge of the Design Basis - Acequate

*

Compliance with Regulations - Inadequate

* Operator Performance - Adequate

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E Manaaement Meetinas

X1 Exit Meeting Summary

The inspection scope and findings were summarized in exit meetings held on

January 31, February 27, and March 21, 1997. Proprietary information is not

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

licensee.

PARTIAL LIST OF PERSONS CONTACTED

Licensees

t

R. Anderson, Senior Vice President, Nuclear Operations

J. Baumstark, Director. Quality Programs

J. Cam) bell, Assistant Plant Director, Maintenance

W. Conclin, Jr.. Director, Nuclear Operations Materials and Controls

J. Cowan, Vice President, Nuclear Production  ;

R. Davis, Assistant Plant Director. Operations

B. Gutherman, Manager. Nuclear Licensing

G. Halnon Assistant Director. Nuclear Operations Site Support

8. Hickle, Director, Nuclear Plant Operations

J. Holden. Director, Nuclear Engineering and Projects

D. Kunsemiller, Director, Nuclear Operations Site Support

NRC

R. Schin, Reactor Inspector, Region II (January 27 through 31: February 10

through 14: March 3 through 7: and March 19 through 21, 1997)

P. Fillion. Reactor Inspector, Region II (March 17 through 21 1997)

INSPECTION PROCEDURES USED

IP 71707: . Plant Operations

IP.92903: Followup - Engineering

,

ITEMS OPENED. CLOSED, AND DISCUSSED

Opened

Typ3 Item Number Status Descriotion and Reference

V

EEI 50-302/97-04-01 'Open Failure to Make an Emergency Phone Re3 ort

Within the Time Requirements of 10 CFR

73.71. (paragraph 01.1)

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EEI 50-302/97-04-02 Open Failure to Hand Carry a Suspected

Reportable Issue to the Shift Manager for

Reportability Review. (paragraph 01.1)

EEI 50-302/97-04-03 Open

Repeat Failure to Report Outside Design

Basis Conditions. (paragraphs E8.1. E8.2)

Closed

Iyp3 Item Number Status Descriotion and Reference

VIO 50-302/94-27-02 Closed Failure to Make Two 10 CFR 50.73 Reports

to the NRC Within the Required Time.

(paragraph E8.1)

VIO 50-302/94-27-03 Closed Failure to Make a 10 CFR 50.72 Report to

the NRC Within the Required Time.

(paragraph E8.1)

VIO 50-302/96-06-06 Closed Failure to Notify the NRC of a Condition

Outside the Appendix R Licensing Design

Basis in a Timely Manner. (paragraph E8.1)

URI 50-302/96-06-03 Closed Non-Safety-Related Transfer Switch Used in

ES Status Indicating Light Circuitry.

(paragraph E8.2)

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Discussed

Typ_g Item Number Status Descriotion and Reference

EEI 50-302/96-19-03 Open EFW NPSH US0 due to Inadequate 10 CFR

50.59 Safety Evaluation for a l

Modification. (paragraph E8.1)

LER 96-019-00 Open Classification of Transfer Switch Causes

Potential for Loss of Power to ES Status

Lights. (paragraph E8.2)

LIST OF ACRONYMS USED

CCHE - Control Complex Habitability Envelope

CFR - Code of Federal Regulations

CP - Compliance Procedure d

CR3 - Crystal River Unit 3 s

CR4 - Crystal River Unit 4

DNP0 - Director. Nuclear Plant Operations

EDG - Emergency Diesel Generator

EEI - Escalation Enforcement Item

EFW Emergency Feedwater

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ES - Engineered Safeguards

FSAR - Final Safety Evaluation Report

IP - (NRC) Inspection Procedure

IR - Inspection Report

LER - Licensee Event Report

LOCA - Loss of Coolant Accident

LOOP - Loss of Offsite Power

NOUE - Notification of Unusual Event

NPSH - Net Positive Suctior. Head

OP - Operating Procedure i

PC - Precursor Card

PRC - Plant Review Committee

RG - (NRC) Regulatory Guide

SM - Shift Manager

SS00 - Shift Supervisor on Duty

TS - Technical Specification

URI - Unresolved Item

US0 - Unreviewed Safety Question

VIO - Violation

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