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IR 05000302-05-007, on 01/28/2005 - 06/16/2005, Crystal River Nuclear Plant, Unit 3; Significance Determination of Unresolved Item from Triennial Fire Protection Inspection, (Followup) Report, Preliminary Greater than Green Finding
	ML051670575
Person / Time
Site:	Crystal River
Issue date:	06/16/2005
From:	Mccree V; Division of Reactor Safety II
To:	Young D; Florida Power Corp
References
	EA-05-114, IR-04-009 IR-05-007
	Download: ML051670575 (13)
	v • d • e

June 16, 2005

EA-05-114

Mr. Dale E. Young, Vice President

Crystal River Nuclear Plant (NA1B)

ATTN: Supervisor, Licensing &

Regulatory Programs

15760 West Power Line Street

Crystal River, FL 34428-6708

SUBJECT: CRYSTAL RIVER UNIT 3 - NRC TRIENNIAL FIRE PROTECTION INSPECTION

(FOLLOW UP) REPORT NO. 05000302/2005007; PRELIMINARY GREATER

THAN GREEN FINDING

Dear Mr. Young:

On June 8, 2005, the U.S. Nuclear Regulatory Commission (NRC) completed an in-office

examination of an unresolved item (URI) associated with the Crystal River 3 facility which was

identified in NRC Inspection Report 05000302/2004009 (ADAMS Accession Number

ML050740113) forwarded to you on March 14, 2005. Specifically, URI 05000302/2004009-01,

Unprotected Post-Fire Safe Shutdown Cables and Related Non-feasible Local Manual Operator

Action, was identified as unresolved pending completion of a significance determination. The

finding, as indicated by the enclosed supporting documentation, appears to be of greater than

very low safety significance (GREEN). The finding and the preliminary significance

determination were discussed on June 16, 2005, with you and other members of your staff.

As discussed in NRC Inspection Report 05000302/2004009, a fire in the 3A 4160V engineered

safeguards (ES) compartment could result in a total loss of offsite power. In addition fire

damage to the metering and protection electrical cables located in or just above the 3A 4160V

ES switchgear could trip and lock out all feeder breakers to both 4160V ES busses causing a

loss of all safety-related A.C. power. These protection and metering circuits were not physically

separated or protected from fire damage as required by 10 CFR 50, Appendix R, Section

III.G.2. Instead an unapproved local manual operator action was used to restore A.C. power.

However, this local manual operator action to reset the 3B emergency diesel generator breaker

lockout on the 3B 4160V ES switchgear was not feasible.

This finding was assessed based on the best available information, including influential

assumptions, using the applicable Significance Determination Process (SDP). The finding was

preliminarily determined to be Greater than Green. The finding has a greater than very low

safety significance because it could affect fire protection defense in depth. In addition,

uncertainties associated with identifying all the fire scenarios that can impact the metering and

protection circuitry, the primary plant response to a delay in secondary side heat removal, the

effectiveness of secondary side heat removal systems following an overcooling event, and

FPC 2

operator response to these fires contribute to the preliminarily characterization of the finding as

greater than green. This issue was also determined to be an apparent violation of NRC

requirements, as discussed in the Enclosure. However, the finding does not represent a

current safety concern because you have modified and corrected the nonconforming condition

before the inspection team left the site.

Our SDP Phase 2 evaluation of this finding is provided in Attachment 2. Additional information

from Progress Energy Florida (Florida Power Corporation) that addresses the assumptions in

our evaluation will permit a more refined risk analysis.

Before we make a final decision on this matter, we are providing you an opportunity to (1)

present to the NRC your perspectives on the facts and assumptions, used by the NRC to arrive

at the finding and its significance, at a Regulatory Conference or (2) submit your position on the

finding to the NRC in writing. If you request a Regulatory Conference, it should be held within

30 days of the receipt of this letter and we encourage you to submit supporting documentation

at least one week prior to the conference in an effort to make the conference more efficient and

effective. If a Regulatory Conference is held, it will be open for public observation. If you decide

to submit only a written response, such submittal should be sent to the NRC within 30 days of

the receipt of this letter.

This apparent violation is being considered for escalated enforcement action in accordance with

the Enforcement Policy, because it is associated with a greater than Green finding. The current

Enforcement Policy is included on the NRCs Web site at http://www.nrc.gov/reading-

rm/adams.html

Please contact Mr. D. Charles Payne at (404) 562-4669 within seven days of the date of this

letter to notify the NRC of your intentions regarding the regulatory conference for the

preliminary Greater than Green finding. If we have not heard from you within 10 days, we will

continue with our significance determination and associated enforcement processes on this

finding, and you will be advised by separate correspondence of the results of our deliberations

on this matter.

Since the NRC has not made a final determination in this matter, no Notice of Violation is being

issued for the inspection finding at this time. In addition, please be advised that the number

and characterization of the apparent violation described in the referenced inspection report may

change as a result of further NRC review. For administrative purposes, this letter is issued as a

separate NRC Inspection Report, No. 05000302/2005007, and the above apparent violation is

identified as AV 0500302/2005007-01, Unprotected Post-Fire Safe Shutdown Cables and

Related Non-feasible Local Manual Operator Action. Accordingly, URI 05000302/2004009-01

is closed.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter,

portions of its enclosure and your response (if any) will be available electronically for public

inspection in the NRC Public Document Room or from the Publicly Available Records (PARS)

component of NRCs document system (ADAMS). However, the NRC is continuing to review

the appropriate classification of the SDP Phase 2 Evaluation (Attachment 2) within our records

management program, considering changes in our practices following the events of September

FPC 3

11, 2001. Using our interim guidance, the attached analyses have been marked as Proprietary

Information or Sensitive Information in accordance with Section 2.390(d) of Title 10 of the Code

of Federal Regulations and will not be placed in the PDR. Please control the document

accordingly (i.e., treat the document as if you had determined that it contained trade secrets

and commercial or financial information that you considered privileged or confidential). We will

inform you if the classification of these documents change as a result of our ongoing

assessments. ADAMS is accessible from the NRC web site at http://www.nrc.gov/reading-

rm/adams.html (the Public Electronic Reading Room).

If you have any questions regarding this letter, please contact me at 404-562-4600.

Sincerely,

\\RA\\

Victor M. McCree, Director

Division of Reactor Safety

Docket No.: 50-302

License No.: DPR-72

Enclosure: Inspection Report 05000302/2005007

w/Attachments: 1. Supplemental Information

cc w/encl:

Daniel L. Roderick

Director Site Operations

Crystal River Nuclear Plant (NA2C)

Electronic Mail Distribution

Jon A. Franke

Plant General Manager

Crystal River Nuclear Plant (NA2C)

Electronic Mail Distribution

Richard L. Warden

Manager Nuclear Assessment

Crystal River Nuclear Plant (NA2C)

Electronic Mail Distribution

cc w/encl contd - (See page 4)

FPC 4

cc w/encl:

Michael J. Annacone

Engineering Manager

Crystal River Nuclear Plant (NA2C)

Electronic Mail Distribution

R. Alexander Glenn

Associate General Counsel (MAC - BT15A)

Florida Power Corporation

Electronic Mail Distribution

Steven R. Carr

Associate General Counsel - Legal Dept.

Progress Energy Service Company, LLC

Electronic Mail Distribution

Attorney General

Department of Legal Affairs

The Capitol

Tallahassee, FL 32304

William A. Passetti

Bureau of Radiation Control

Department of Health

Electronic Mail Distribution

Craig Fugate, Director

Division of Emergency Preparedness

Department of Community Affairs

Electronic Mail Distribution

Chairman

Board of County Commissioners

Citrus County

110 N. Apopka Avenue

Inverness, FL 36250

Jim Mallay

Framatome Technologies

Electronic Mail Distribution

_________________________

OFFICE RII:DRS RII:DRS RII:DRP* RII:OI

SIGNATURE WGR1 DCP JTM CFE

NAME WRogers DPayne JMunday CEvans

DATE 6/14/2005 6/14/2005 6/13/2005 6/16/2005 6/ /2005 6/ /2005 6/ /2005

E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO

U. S. NUCLEAR REGULATORY COMMISSION

REGION II

Docket No.: 50-302

License No.: DPR-72

Report No.: 05000302/2005007

Licensee: Progress Energy Florida (Florida Power Corporation)

Facility: Crystal River Unit 3

Location: 15760 West Power Line Street

Crystal River, FL 34428-6708

Dates: January 28, 2005 - June 16, 2005

Inspectors: R. Rodriguez, Reactor Inspector

R. Schin, Senior Reactor Inspector (Lead Inspector)

W. Rogers, Senior Reactor Analyst

Approved by: D. Charles Payne, Chief, Engineering Branch 2

Division of Reactor Safety

SUMMARY OF FINDINGS

IR 05000302/2005007; Crystal River Nuclear Plant, Unit 3; Significance Determination of

Unresolved Item from Triennial Fire Protection Inspection.

This in-office review was conducted by two regional inspectors and a senior reactor analyst.

One preliminary Greater than Green finding with an apparent violation was identified. The

significance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 0609 Significance Determination Process (SDP). Findings for which the SDP does not apply

may be Green or be assigned a severity level after NRC management review. The NRC's

program for overseeing the safe operation of commercial nuclear power reactors is described in

NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.

A. NRC-Identified and Self-Revealing Findings

Cornerstone: Initiating Events and Mitigating Systems

Preliminary Greater than Green. An apparent violation of 10 CFR 50, Appendix

R,Section III.G.2, for failure to physically protect or separate cables from fire

damage and instead relying on an unapproved local manual operator action.

The unprotected cables were associated with a common electrical protection and

metering circuit which was installed such that fire damage to a cable in or just

above the 3A 4160V engineered safeguards (ES) switchgear could result in

tripping and locking out all feeder breakers to both 4160V ES busses, resulting in

a loss of all safety-related alternating current power.

In addition, the local manual operator action to reset the 3B emergency diesel

generator breaker lockout on the 3B 4160V ES switchgear was determined to be

non-feasible. During a severe fire in the adjacent 3A 4160V Switchgear Room

the fire response activities would cause the location for the operator action (the

3B 4160V Switchgear Room) to be exposed to hot smoke, water mist, and water

on the floor.

This finding is greater than minor because it degraded the defense in depth for

fire protection and also because it is associated with the protection against

external factors attribute and degraded the reactor safety mitigating systems

cornerstone objective. The finding adversely affected the reliability and

capability of equipment required to achieve and maintain a safe shutdown

condition following a severe fire in the 3A 4160V ES Switchgear Room.

(Section 4OA5.01)

B. Licensee-identified Violations:

None

Report Details

4. OTHER ACTIVITIES

4OA5 OTHER

.1 (Closed) URI 05000302/2004009-01. Unprotected Post-Fire Safe Shutdown Cables and

Related Non-feasible Local Manual Operator Action.

Introduction. An apparent violation (AV) of 10 CFR 50, Appendix R, Section III.G.2, for

failure to physically protect or separate cables from fire damage and instead relying on a

local manual operator action that was not approved by the NRC. The unprotected

cables were in common electrical protection and metering circuits which were installed

such that fire damage to a cable in or just above the 3A 4160V ES switchgear could trip

and lock out all feeder breakers to both 4160V ES busses, resulting in a loss of all

safety-related alternating current (a.c.) power.

In addition, the team found that the licensees local manual operator action to mitigate

this condition was not feasible. The action was to reset the 3B emergency diesel

generator (EDG) breaker lockout on the 3B 4160V ES switchgear. However, that action

was not feasible because the fire in the 3A 4160V Switchgear Room and fire fighting

activities through the adjacent 3B 4160V Switchgear Room would cause the location for

the operator action (in the 3B 4160V Switchgear Room) to be exposed to hot smoke,

water mist, and water on the floor. This inspection finding was assessed using the SDP

and preliminarily determined to be Greater than Green (i.e., an issue with low to

moderate increased importance to safety, which may require additional NRC

inspections.)

Description. During the baseline triennial fire protection inspection, the inspectors

identified a finding involving cables for the electrical protection and metering circuit

located in the 3A 4160V ES Switchgear Room were vulnerable to fire damage that could

disable both the 3A 4160V ES switchgear and the redundant train 3B 4160V ES

switchgear, having potential safety significance greater than very low significance.

Specifically, the licensees 10 CFR 50 Appendix R Fire Study and post-fire safe

shutdown (SSD) procedure OP-880A, Appendix R Post-Fire Safe Shutdown

Information, Rev. 5, Step 9-6, included a local manual operator action that was not

approved by the NRC and also was not feasible. The action was to reset the 3B EDG

breaker lockout on the 3B 4160V ES switchgear during a fire in the 3A 4160V ES

Switchgear Room. This action was time critical and required to be completed within 30

minutes of entering OP-880A. Operators were to trip the reactor and enter OP-880A if a

fire in the 3A 4160V Switchgear Room impacted safe operation of the plant. The

licensee had considered that the action was needed because a fire in the 3A 4160V ES

switchgear could affect cables for the electrical protection and metering circuit and could

lock out all feeder breakers to both the 3A and the 3B 4160V ES switchgear. However,

the licensees post-fire SSD methodology relied upon equipment powered from the 3B

4160V ES switchgear. Specifically, the licensees analysis determined that power to the

3B 4160V switchgear was needed within 30 minutes to enable operators to restore

2

ventilation cooling to the Emergency Feedwater Initiation and Control (EFIC) rooms.

The EFIC system was needed for automatic EFW flow control.

The team found that cables for the electrical protection and metering circuit were

located within and directly above the 3A 4160V switchgear, where a fire originating in

certain sections of the switchgear could immediately damage them. The team noted

that these cables were four-conductor, #8 American Wire Gage (AWG), Institute of

Institute of Electrical and Electronic Engineers (IEEE) 383 qualified, thermoset-type

cables with no protective fire wrap. Damage to one of these cables could result in

immediate loss of both the 3A and the 3B 4160V ES switchgear, and a loss of all safety-

related a.c. power. Plant operators would conclude this had an impact on safe

operation of the plant, would trip the reactor, and immediately enter OP-880.

During this fire condition, the primary plant operator (PPO) had a number of

proceduralized time-critical local manual operator actions to perform in a prescribed

sequence. Based on Pre-fire Plans and fire brigade drill results, the fire brigade would

attack a fire in the 3A 4160V Switchgear Room through the 3B 4160V Switchgear Room

about 15 minutes after confirmation of the fire. Based on licensee time validations and

NRC team walkdowns of the actions, the team determined that the PPO would arrive at

the 3B 4160V Switchgear Room about 25 minutes into the fire event. When the PPO

arrived, the fire brigade would have the door between the two switchgear rooms open

and would have sprayed fire water into the 3A 4160V Switchgear Room. Hot smoke

from the fire would have filled both the 3A and 3B 4160V ES switchgear rooms and the

hallway leading to those rooms because the fire brigade would have all doorways

between the two rooms and the hallway blocked open with their fire hose. In addition,

water from the fire hose would have created mist in the air and water on the floors of

both switchgear rooms (the switchgear rooms had no floor drains). At 25 minutes into

the fire event, the fire brigade would not yet have had time to evacuate the smoke with

portable fans. In addition, the portable fans would not have electrical power available

because the outlets for the fans were powered from the 3A and 3B 4160V switchgear,

which would potentially be de-energized. While the fire brigade could obtain a portable

generator to power the fans, this would take too long to allow the operator to complete

the action within the time-critical 30 minutes. In view of all of these conditions, the team

concluded that the operator action was not feasible.

This finding was an immediate safety concern and the licensee made modifications to

correct the nonconforming condition before the inspection team left the site.

Analysis: This finding degraded the defense in depth for fire protection and also it is

associated with the protection against external factors attribute and degraded the

reactor safety mitigating systems cornerstone objective. The finding adversely affected

the reliability and capability of equipment required to achieve and maintain a SSD

condition following a severe fire. The finding is applicable to post-fire SSD from the

control room during a fire in the 3A 4160V ES Switchgear Room. Because the finding

affects fire protection, it was assessed in accordance with the NRC Reactor Oversight

Processs SDP as described in NRC Inspection Manual Chapter 0609, Appendix F (MC 0609, App. F). In the Phase 1, the finding was associated with post-fire safe shutdown,

3

it was assigned a high degradation rating and it existed for more than 30 days. As a

result, a Phase 2 Risk Evaluation was required.

Summary of Phase 2 SDP Analysis

This evaluation was performed by Region II inspectors with the assistance of the

regional SRA. The Crystal River Phase 2 SDP Analysis is included in this inspection

report as Attachment 2.

The Phase 2 analysis involves a quantitative assessment of CDF increase given a

finding. There are nine analysis steps and five screening checks. This assessment

includes quantification of a Fire Frequency, Fire Damage State, Non-Suppression

Probability and Conditional Core Damage Probability (CCDP). The report also contains

several appendices documenting supplemental information used in the Phase 2

analysis.

Effects from a fire in the 3A 4160V switchgear room were postulated and evaluated. Not

all ignition sources were counted in the fire area, only the cabinets where the CT

circuitry is located were considered. The room has no automatic suppression and no

manual suppression credit is applied with the target damaged within 1 minute. This

yielded an Fire Frequency of 2.4E-04.

A loss of offsite power and a plant trip were postulated and the appropriate plant

initiating event worksheet from the plant risk-informed inspection notebook was used to

account for the plant SSD response and required human recovery actions in order to

quantify the factor CCDP for each fire scenario of interest.

The Phase 2 analysis concluded that the change in Core Damage Frequency ( CDF)

[the difference between the conforming case CDF and the non-conforming case CDF]

was 2.4E-05 (substantial importance to safety).

SDP/Enforcement Review Panel (SERP) Evaluation

The total change in CDF due to the performance deficiency was found to be 2.4 E-05/yr

for the unit. The dominant accident sequences that cause the largest CDF are fully

developed fires that require Emergency AC Power and Emergency Feedwater. The

color associated with this magnitude of change in CDF is Greater than Green.

Therefore, the SERP has preliminarily determined this issue to be a Greater than Green

finding.

4

Enforcement: 10 CFR 50.48(b)(1) requires, in part, that all nuclear power plants

licensed to operate prior to January 1, 1979, must satisfy the applicable requirements of

Appendix R,Section III.G.Section III.G.2 applies to the ability to achieve and maintain

hot SSD from the control room during a fire. It states, in part, that where cables or

equipment, including associated non-safety circuits that could prevent operation or

cause maloperation due to hot shorts, open circuits, or shorts to ground, of redundant

trains of systems necessary to achieve and maintain hot shutdown conditions are

located within the same fire area outside of primary containment, one of three means of

protecting cables to ensure that one of the redundant trains is free of fire damage shall

be provided. The three means include, among others, the physical protection or

separation of cables to preclude fire damage.

Contrary to the above, on January 26, 2005, cables for the electrical protection and

metering circuit located in the 3A 4160V ES Switchgear Room were vulnerable to fire

damage that could disable both the 3A 4160V ES switchgear and the redundant train 3B

4160V ES switchgear. Specifically, these protection and metering circuits were not

physically separated or protected (as discussed above) from fire damage as required by

10 CFR 50, Appendix R, Section III.G.2. This apparent violation is identified as AV

05000302 /2005007-01, Unprotected Post-Fire Safe Shutdown Cables and Related

Non-feasible Local Manual Operator Action. Accordingly, URI 05000302/2004009-01 is

closed.

4OA6 Meetings, Including Exit

On June 16, 2005, the inspectors presented the inspection results by telephone to

Mr. Dale E. Young and other members of your staff, who acknowledged the findings.

The inspectors confirmed that proprietary information was not provided or examined

during the inspection.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

NRC personnel

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000302/2005007-01 AV Unprotected Post-Fire Safe Shutdown Cables and

Related Non-feasible Local Manual Operator Action

(Section 4OA5.1)

Closed

05000302/2004009-01 URI Unprotected Post-Fire Safe Shutdown Cables and

Related Non-feasible Local Manual Operator Action

(Section 4OA5.1)

LIST OF DOCUMENTS REVIEWED

Procedures:

OP-880A, Appendix R Post-Fire Safe Shutdown Information, Rev. 5

Drawings:

E-215-031, Electrical Conduit Layout Control Complex, Rev. 56

Attachment 1

2

ML051670575

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