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IR 05000269-04-012, 05000270-04-012, 05000287-04-012; 02/17 - 07/13, 2004; Oconee Nuclear Station, Units 1, 2, and 3; Significance Determination of Unresolved Item from Triennial Fire Protection Inspection. Proprietary Information Removed
	ML042170356
Person / Time
Site:	Oconee
Issue date:	07/20/2004
From:	Casto C; Division of Reactor Safety II
To:	Rosalyn Jones; Duke Energy Corp
References
	EA-04-115 IR-04-012
	Download: ML042170356 (6)
	v • d • e

ATTACHMENT 2 CONTAINS PROPRIETARY INFORMATION

July 20, 2004

EA-04-115

Duke Energy Corporation

ATTN: Mr. Ronald A. Jones

Vice President

Oconee Nuclear Station PROPRIETARY INFORMATION REMOVED

7800 Rochester Highway

Seneca, SC 29672

SUBJECT: OCONEE NUCLEAR STATION - NRC TRIENNIAL FIRE PROTECTION

INSPECTION (FOLLOW UP) REPORT 05000269/2004012, 05000270/2004012

AND 05000287/2004012 PRELIMINARY GREATER THAN GREEN FINDING

Dear Mr. Jones:

On February 18, 2004, the U.S. Nuclear Regulatory Commission (NRC) completed the on-site

portion of an open item inspection for your Oconee Nuclear Station Units 1, 2, and 3. Following

additional in-office review, the inspection was completed on July 13, 2004. The enclosed

inspection report documents the inspection findings, which were discussed on July 13, 2004,

with Mr. Noel Clarkson and other members of your staff.

This inspection was an on-site and in-office examination of an Unresolved Item (URI)

05000269, 270, 287/2002003-001, Failure to Meet License Basis Commitment For Staffing the

Standby Shutdown Facility (SSF) for a Confirmed Fire. This issue was unresolved pending a

review by the NRC to determine if a loss of function occurs and a safety significance

determination.

PROPRIETARY INFORMATION

Exempt from public release under the Freedom of

Information Act (5 U.S.C. 552). Ref: 2.390(d)(1)

Nuclear Regulatory Commission review required

before public release.

Determination made by:

Charles R. Ogle, Chief, EB1, DRS, RII

Signature /RA/ Date 7/20/4

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENT 2, THIS DOCUMENT IS DECONTROLLED

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ATTACHMENT 2 CONTAINS PROPRIETARY INFORMATION

Based on the results of this inspection, the NRC determined that the Oconee fire response

procedures were not consistent with the licensing basis in regards to the criteria for manning of

the Standby Shutdown Facility (SSF). In some scenarios this could result in a delay of transfer

of control to the SSF which could challenge the capability of the installed SSF makeup pump.

As a result, in some scenarios, this could result in pressurizer level failing to be maintained

within the indicating range as required by Appendix R. This apparent violation is identified as

AV 05000269, 270, 287/2002004-12, Failure to Meet Licensing Basis for Staffing the SSF in the

Event of a Confirmed Plant Fire.

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

assumptions, using the applicable Significance Determination Process (SDP) and was

preliminarily determined to be a Greater than Green finding. The NRCs calculated change in

core damage frequency for this finding is provided in Attachment 2. However, due to

uncertainties in developing the plant response to fires that leave the SSF as the exclusive core

damage mitigation strategy, this finding has been preliminarily characterized as Greater than

Green. Additional information in this regard from Duke Energy Corporation would allow a more

refined risk analysis.

The finding has a greater than very low safety significance because it could affect fire protection

defense in depth. The finding does not represent a current safety concern because you have

modified your procedures to ensure that the SSF is manned upon confirmation of a fire.

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

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 "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement

Policy), NUREG-1600. The current Enforcement Policy is included on the NRCs Web site at

http://www.nrc.gov/reading-rm/adams.html

Please contact Mr. Charles R. Ogle at (404) 562-4605 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.

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENT 2, THIS DOCUMENT IS DECONTROLLED

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ATTACHMENT 2 CONTAINS PROPRIETARY INFORMATION

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

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

change as a result of further NRC review.

In accordance with 10 CFR 2.390 of the NRCs "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 3 Evaluation (Attachment 2) within our records

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

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. 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/

Charles A. Casto, Director

Division of Reactor Safety

Docket Nos.: 50-269, 50-270, 50-287

License Nos.: DPR-38, DPR-47, DPR-55

Enclosure: Inspection Report 05000280,281/2004012

w/Attachments: 1. Supplemental Information

2. Phase 3 SDP Evaluation (Proprietary Information)

cc w/encl: (See page 4)

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENT 2, THIS DOCUMENT IS DECONTROLLED

4

ATTACHMENT 2 CONTAINS PROPRIETARY INFORMATION

cc w/encl:

B. G. Davenport

Compliance Manager (ONS)

Duke Energy Corporation

ON03RC

7800 Rochester Highway

Seneca, SC 29672

Lisa Vaughn

Legal Department (PB05E)

Duke Energy Corporation

422 South Church Street

P. O. Box 1244

Charlotte, NC 28201-1244

R. L. Gill, Jr., Manager

Nuclear Regulatory Issues

and Industry Affairs

Duke Energy Corporation

526 S. Church Street

Charlotte, NC 28201-0006

Distribution:

L. Olshan, NRR

L. Slack, RII, EICS

C. Ogle, RII, DRS

B. Westreich, NSIR (hard copy w/encl)

E. McNiel, NSIR (hard copy w/encl)

RIDSNRRDIPMLIPB

OEMAIL

OFFICE RII:DRS RII:DRS RII:DRS RII:DRP RII:EICS RII:DRS

SIGNATURE RA RA RA RA RA RA

NAME ODONOHUE ROGERS OGLE RHAAG EVANS CASTO

DATE 7/19/2004 7/15/2004 7/15/2004 7/16/2004 7/16/2004 7/20/2004

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

OFFICIAL RECORD COPY DOCUMENT NAME: C:\ORPCheckout\FileNET\ML042170356.WPD

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENT 2, THIS DOCUMENT IS DECONTROLLED

ATTACHMENT 2 CONTAINS PROPRIETARY INFORMATION

U. S. NUCLEAR REGULATORY COMMISSION

REGION II

Docket Nos.: 50-269, 50-270, 50-287

License Nos.: DPR-38, DPR-47, DPR-55

Report Nos.: 05000269/2004012, 05000270/2004012, 05000287/2004012

Licensee: Duke Energy Corporation

Facility: Oconee Nuclear Station

Location: 7800 Rochester Highway

Seneca, SC 29672

Dates: February 17 - July 13, 2004

Inspectors: K. ODonohue, Fire Protection Team Leader (Lead Inspector)

W. Rogers, Senior Reactor Analyst

Approved by: C. Casto, Director

Division of Reactor Safety

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENT 2, THIS DOCUMENT IS DECONTROLLED

Enclosure

ATTACHMENT 2 CONTAINS PROPRIETARY INFORMATION

SUMMARY OF FINDINGS

IR 05000269/2004012, 05000270/2004012, 05000287/2004012; 02/17 - 07/13, 2004; Oconee

Nuclear Station, Units 1, 2, and 3; Significance Determination of Unresolved Item from Triennial

Fire Protection Inspection.

This in-office and on-site review were conducted by a regional inspector 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, Sections III.L.2.b and III.L.3 was identified, in that, for a severe fire in areas

requiring the manning of the Standby Shutdown Facility (SSF) and activation of

the SSF makeup pump, the licensees safe shutdown strategy and related

response procedures delayed the manning of the SSF until there was a loss of

function of high pressure injection and component cooling or feedwater. In some

scenarios, this would delay transfer of control to the SSF, thereby challenging

the operability of the installed SSF makeup pump. This could result in

pressurizer level failing to be maintained within the indicating range. The

licensee has revised the affected procedures and is evaluating the need for

additional corrective action.

This finding is greater than minor because it was associated with protection

against external factors and procedure quality cornerstone attributes. It

affected the objective of the Mitigating Systems cornerstone to ensure the

availability, reliability and capability of systems that respond to initiating events.

This degraded condition increased plant risk because, if a severe fire occurred in

areas requiring the manning of the SSF and activation of the SSF makeup pump,

the licensee's procedures may not preclude loss of reactor coolant beyond the

capability of the SSF makeup pump. (Section 4OA5.01)

B. Licensee-identified Violations:

None

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENT 2, THIS DOCUMENT IS DECONTROLLED

Enclosure

ATTACHMENT 2 CONTAINS PROPRIETARY INFORMATION

Report Details

4. OTHER ACTIVITIES

4OA5 OTHER

.1 (Closed) URI 05000269, 270, 287/2002003-001: Failure to Meet License Basis

Commitment for Staffing the Standby Shutdown Facility in the Event of a Confirmed

Plant Fire.

Introduction: An apparent violation (AV) was identified for failure to comply with 10 CFR

50, Appendix R, Sections III.L.2.b and III.L.3, in that, for a severe fire in areas requiring

the manning of the SSF and activation of the SSF makeup pump, the licensees method

for implementing their alternative shutdown capability did not ensure that the reactor

coolant makeup function would be capable of maintaining the reactor coolant level within

the level indication of the pressurizer. 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 the timeliness associated with manning the SSF, having

potential safety significance greater than very low significance. Specifically, per

procedure, the manning of the SSF was not to be initiated until the fire damage caused

a loss of both the high pressure injection (HPI) and component cooling (CC) systems or

a loss of all feedwater. The inspectors were concerned that waiting until loss of function

occurred to man the SSF could result in a loss of RCS inventory due to additional

challenges to and subsequent failures of a power operated relief valve (PORV) or

pressure safety valve (PSV). (Abnormal Procedure, AP/0/A/1700/025, Standby

Shutdown Facility Emergency Operating Procedure, Revision 20, implemented the

Auxiliary Shutdown design capability from the SSF in the event of a fire in areas

requiring the manning of the SSF and activation of the SSF makeup pump.)

By letter dated September 20, 1982, the licensee responded to a staff request for

additional information (letter dated July 17, 1982). In response to staff concerns

regarding the potential for spurious operation of reactor coolant system (RCS) isolation

valves (Question No. 4) the licensee stated:

"Upon confirmation of a fire in the plant, operating personnel will be dispatched to the

SSF where they will establish communication with the control room...If vital control and

monitoring functions (eg., reactor coolant pressure boundary, reactor coolant makeup

capability) become unacceptably degraded or unavailable from the control room, a

prompt transfer can be made and control established from the SSF....As stated above,

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENT 2, THIS DOCUMENT IS DECONTROLLED

Enclosure

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ATTACHMENT 2 CONTAINS PROPRIETARY INFORMATION

spurious operation is extremely unlikely within the first 10 minutes. To preclude

unacceptable consequences of spurious operation in the longer term, circuits are

designed to either preclude spurious operation or retain operability of the systems

necessary to mitigate such operation."

By letter dated April 28, 1983, the staff issued its SER of the Oconee Nuclear Station

SSF. This SER was based, in part, on the above referenced Duke response. Based on

review of these documents, the inspectors determined that upon a confirmed fire in the

Cable Spread Room or other areas where shutdown from the SSF may be required, the

SSF should be manned immediately upon the confirmation of a fire and communication

established with the MCR. However, as described above, the licensees procedures did

not require the SSF to be manned (or activated) until significant fire damage occurred.

This delay in manning the SSF could result in additional challenges to the PORVs and

PSVs and result in loss of RCS inventory beyond the SSF makeup capability. In some

scenarios this could result in a failure to maintain the RCS PZR level within the

indicating range.

The licensee captured this issue in its corrective action program in Problem

Investigation Process (PIP) O-02-00609.

Analysis: This finding affects the protection against external factors and procedure

quality cornerstone attributes. It affected the objective of the Mitigating Systems

cornerstone to ensure the availability, reliability and capability of systems that respond to

initiating events because existing procedural guidance may result in a fire damage

scenario that could impact the ability of the unit (s) to achieve and maintain safe

shutdown conditions. 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). The Phase 1

screening conclusion was that a Phase 2 screening should be performed. However,

due to the unique failure mechanism associated with the performance deficiency, the

senior risk analyst determined that a Phase 2 SDP would not be performed and the

finding was screened to a Phase 3 analysis.

Summary of Phase 3 SDP Analysis

This evaluation was performed by Region II SRAs. The Oconee Phase 3 SDP Analysis

is included in this inspection report as Attachment 2.

The Phase 3 analysis discusses the approach, site visit observations, assumptions,

screening analysis, fire ignition frequencies, fire scenario analysis, contributors to fire

risk, integrated assessment of fire-induced core damage frequency, and conclusions

developed from this analysis. The report also contains several appendices documenting

supplemental information used in the Phase 3 analysis.

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENT 2, THIS DOCUMENT IS DECONTROLLED

Enclosure

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ATTACHMENT 2 CONTAINS PROPRIETARY INFORMATION

The Phase 3 considers the conforming case where operation personnel are dispatched

to the SSF at the confirmation of a fire in the fire areas of concern. The Phase 3 then

considers the non-conforming case where operation personnel are dispatched only after

loss of critical plant functions. This time delay increases the challenges to the PORVs

or if they are not available, challenges to the PSVs. The additional risk is quantified by

the CDF. The core damage frequency (CDF) for fires which require the SSF to

prevent core damage was calculated in each case and summed for each unit.

The risk analysts concluded that the CDF [the difference between the conforming

case CDF and the non-conforming case CDF] was 3E-6 (low to moderate importance to

safety.)

SDP/Enforcement Review Panel (SERP) Evaluation

The total change in CDF due to the performance deficiency was found to be 3 E-6 / yr

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

developed fires that require manning of the SSF. 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.

Enforcement: Oconee Unit 1 Operating License DPR-38, Oconee Unit 2 Operating

License DPR-47, and Oconee Unit 3 Operating License DPR-55 Condition D provide, in

part, that the licensee implement and maintain in effect all provisions of the approved

fire protection program as described in the UFSAR and as approved in the SER dated

April 28, 1983 and subsequent supplements.

The licensees UFSAR commits to 10 CFR 50, Appendix R, Sections III.G and III.L.

Section III.G.3 states that alternative shutdown capability should be provided where the

protection of systems whose function is required for hot shutdown, does not satisfy the

requirements of III.G.2.Section III.L of Appendix R provides requirements to be met by

alternative shutdown methods.Section III.L.2.b states, in part, that The reactor coolant

makeup function shall be capable of maintaining the reactor coolant level. . . within the

level indication in the pressurizer in PWRs."Section III.L.3 specifies that procedures

shall be in effect to implement this capability.

Contrary to the above, on February 8, 2002, the inspectors determined that the

procedures specified for a fire requiring SSF manning and activation would not assure

that reactor cooling makeup function would be capable of maintaining reactor coolant

level within the indicated range of the pressurizer. Specifically, delaying the manning of

the SSF until after the occurrence of a loss of function of the high pressure injection and

component cooling or feedwater rather than manning the SSF immediately upon

confirmation of a fire in the areas of concern may not preclude an extended loss of

reactor coolant system inventory. This could result in pressurizer level failing to be

maintained within the indicating range. This apparent violation is identified as AV

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENT 2, THIS DOCUMENT IS DECONTROLLED

Enclosure

4

ATTACHMENT 2 CONTAINS PROPRIETARY INFORMATION

05000269, 270, 287/2004012-01, Failure to Meet Licensing Basis for Staffing the SSF in

the Event of a Confirmed Plant Fire.

4OA6 Meetings, Including Exit

On July 13, 2004, the inspectors presented the inspection results by telephone to Mr.

Noel Clarkson and other members of your staff, who acknowledged the findings. The

inspectors confirmed that proprietary information was not provided or examined during

the inspection.

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENT 2, THIS DOCUMENT IS DECONTROLLED

Enclosure

ATTACHMENT 2 CONTAINS PROPRIETARY INFORMATION

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

G. Davenport, Compliance Manager (ONS)

D. Garland, Sr. Engineer

J. Smith, Regulatory Compliance

J. Weast, Regulatory Compliance

H. Barrett, Sr. Engineer (Design Basis Engineering)

N. Constance, Operations Training

D. Henneke, Sr. Engineer (PRA Engineering)

NRC personnel

M. Shannon

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000269,270,287/2004012-01 AV Failure to Meet Licensing Basis for Staffing the

SSF in the Event of a Confirmed Plant Fire

(Section 4OA5.1)

Closed

05000269,270,287/2002003-01 URI Failure to Meet License Basis Commitment for

Staffing the SSF in the Event of a Confirmed Plant

Fire (Section 4OA5.1)

LIST OF DOCUMENTS REVIEWED

Procedures:

AP/0/A/1700/025, Standby Shutdown Facility Emergency Operating Procedure, Revision 20

AP/0/A/1700/025, Standby Shutdown Facility Emergency Operating Procedure, Revision 25

AP/1/A/1200/008, Loss of Control Room, Revision 8

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENT 2, THIS DOCUMENT IS DECONTROLLED

Attachment 1

2

ATTACHMENT 2 CONTAINS PROPRIETARY INFORMATION

Plant Issue Reports Reviewed:

PIP-O2-00609, Questions of Procedural Guidance for Spurious Actuation of EFW and

Acceptability of the Start of the 10 minute Time for Spurious Actuations

ML042170356

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