ML20137V727
| 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: |
| 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.
including:
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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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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
'
- Management Oversight - Inadequate
- Engineering Effectiveness - Inadecuate
- Knowledge of the Design Basis - Acequate
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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
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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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