Text

IR 05000280-03-008, IR 05000281-03-008; on 04/21/2003 - 01/07/2004; for Surry Power Station Units 1 and 2; Significance Determination of Unresolved Items from Triennial Fire Protection Inspection
	ML040490131
Person / Time
Site:	Surry
Issue date:	02/02/2004
From:	Casto C; Division of Reactor Safety II
To:	Christian D; Virginia Electric & Power Co (VEPCO)
References
	EA-04-005 IR-03-008
	Download: ML040490131 (25)
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ATTACHMENTS 2 AND 3 CONTAIN PROPRIETARY INFORMATION

EA-04-005 February 2, 2004

Virginia Electric and Power Company

ATTN: Mr. David A. Christian

Senior Vice President and

Chief Nuclear Officer

Innsbrook Technical Center

5000 Dominion Boulevard

Glen Allen, VA 23060

SUBJECT: SURRY POWER STATION - NRC INSPECTION REPORT 05000280/2003008

AND 05000281/2003008; PRELIMINARY WHITE FINDING

Dear Mr. Christian:

On January 7, 2004, the U.S. Nuclear Regulatory Commission (NRC) completed an open item

inspection for your Surry Power Station, Units 1 and 2. The enclosed inspection report

documents the inspection findings, which were discussed on February 2, 2004, with

Mr. R. Blount and other members of your staff.

This inspection was an in-office examination of three unresolved items (URIs) which were

identified in NRC Inspection Report 05000280/2003007 and 05000281/2003007 (ADAMS

Accession Number ML030930560) forwarded to you on March 31, 2003. The three URIs were:

URI 05000280/2003007-001, Fire Response Procedures 1-FCA-4.00 And 0-FCA-14.00 Not

Adequate To Assure Safe Shutdown Of Unit 1; URI 05000280/2003007-002, Fire Response

Procedures 1-FCA-3.00 And 1-FCA-14.00 Not Adequate To Assure Safe Shutdown Of Unit 1;

and URI 05000280,281/2003007-003, Alternate Shutdown Panel Ventilation System Not

Independent From Impacts Of A Main Control Room Fire. These issues were unresolved

pending a safety significance determination.

Based on the results of this inspection, the inspectors identified that the Surry fire response

procedures were not effective in ensuring a safe shutdown of Unit 1 during a severe fire in

Emergency Switchgear and Relay Room (ESGR) Number (No.) 1 (URI 05000280/2003007-

001). Specifically, these procedures may not preclude an extended loss of reactor coolant

pump seal injection flow and may initiate a reactor coolant pump seal loss of coolant accident

which could result in pressurizer level failing to be maintained within the indicating range as

required by 10 CFR 50, Appendix R. This inspection finding was assessed using the applicable

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENTS 2 AND 3, THIS DOCUMENT IS DECONTROLLED

VEPCO 2

ATTACHMENTS 2 AND 3 CONTAIN PROPRIETARY INFORMATION

significance determination process (SDP) and preliminarily determined to be White (i.e., an

issue with low to moderate increased importance to safety, which may require additional NRC

inspections.) This issue was also determined to be an apparent violation of NRC requirements.

In accordance with the "General Statement of Policy and Procedure for NRC Enforcement

Actions - May 1, 2000" (Enforcement Policy), NUREG-1600, this apparent violation is being

considered for escalated enforcement action because it is associated with a White finding.

Additionally, because relative equipment and cable locations in Unit 2s ESGR No. 2 are similar

to those in Unit 1s ESGR No. 1, the fire ignition frequencies for both areas are similar, and the

procedures used to respond to a severe fire are similar, the inspectors determined that the safe

shutdown of Unit 2 during a severe fire in ESGR No. 2 may be similarly impacted. This issue is

documented in this report as a new finding on Unit 2 (URI 05000281/2003008-001) that has

potential safety significance greater than very low significance. This finding did not present an

immediate safety concern. To date, no SDP of this new finding on Unit 2 has been completed

by the NRC staff. Accordingly, you are requested to provide any information as to why the

outcome of the SDP for ESGR No. 1 should not be applied to Unit 2s ESGR No. 2.

Before the NRC makes a final decision on these matters, we are providing you an opportunity

to request a regulatory conference where you would be able to provide your perspectives on the

significance of the findings, the bases for your position, and whether you agree with the

apparent violation. If you choose to request a regulatory conference, we encourage you to

submit your evaluation and any differences with the NRCs evaluation at least one week prior to

the conference in an effort to make the conference more efficient and effective. Should you

request a conference, the NRC requests that you provide information in your written evaluation

and at the conference on the design and performance of the reactor coolant pump breakdown

bushings, and their effect on limiting reactor coolant system leakage in the event of a seal

failure. If a regulatory conference is held, it will be open for public observation. The NRC will

also issue a press release to announce the regulatory conference.

In addition, the report documents two NRC-identified findings of very low safety significance

(Green). These findings which resulted from URI 05000280/2003007-002 and URI 05000280,

281/2003007-003 were determined to involve violations of NRC requirements. However,

because of the very low safety significance and because they are entered into your corrective

action program, the NRC is treating these two findings as non-cited violations (NCVs)

consistent with Section VI.A of the NRC Enforcement Policy. If you contest any NCV in this

report, you should provide a response within 30 days of the date of this inspection report, with

the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control

Desk, Washington, DC 20555-0001; with copies to the Regional Administrator Region II; the

Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington,

DC 20555-0001; and the NRC Resident Inspector at Surry Power Station.

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 White finding. If we have not heard from you within 10 days, we will continue with

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ATTACHMENTS 2 AND 3 CONTAIN PROPRIETARY INFORMATION

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.

In accordance with 10 CFR 2.790 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 Summary of Phase 2 SDP Risk Analysis and Phase 3 SDP

analysis (Attachments 2 and 3) 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 2.790(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.: 50-280, 50-281

License Nos.: DPR-32, DPR-37

Enclosure: Inspection Report 05000280,281/2003008

w/Attachments: 1. Supplemental Information

2. Summary of Phase 2 SDP Risk Analysis

3. Phase 3 SDP Analysis

cc w/encl and Attachments: See page 4

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENTS 2 AND 3, THIS DOCUMENT IS DECONTROLLED

VEPCO 4

ATTACHMENTS 2 AND 3 CONTAIN PROPRIETARY INFORMATION

cc w/encl and Attachments:

Chris L. Funderburk, Director

Nuclear Licensing and

Operations Support

Virginia Electric & Power Company

6000 Dominion Boulevard

Glen Allen, VA 23060

Richard H. Blount, II

Site Vice President

Surry Power Station

Virginia Electric & Power Company

5570 Hog Island Road

Surry, VA 23883

cc w/encl and Attachment 1:

Virginia State Corporation Commission

Division of Energy Regulation

P. O. Box 1197

Richmond, VA 23209

Lillian M. Cuoco, Esq.

Senior Counsel

Dominion Resources Services, Inc.

Millstone Power Station

Building 475, 5th Floor

Rope Ferry Road

Rt. 156

Waterford, Connecticut 06385

Attorney General

Supreme Court Building

900 East Main Street

Richmond, VA 23219

Distribution w/encl and Attachments:

F. Congel, OE

N. Kalyanam, NRR

M. Sykes, NRR

R. Laufer, NRR

D. Nelson, OE

C. Ogle, RII

Distribution w/encl and Attachments contd - See page 5:

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Distribution w/encl and Attachments cont;d)

L. Garner, RII

K. Landis, RII

C. Evans, RII

G. McCoy, RII

G. Wiseman, RII

W. Rogers, RII

S. Sparks, RII

OEMAIL

Distribution w/encl and Attachment 1:

RIDSNRRDIPMLIPB

PUBLIC

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

SIGNATURE * * * *

NAME PAYNE ODONOHUE ROGERS OGLE LANDIS EVANS CASTO

DATE 1/ /2004 1/ /2004 1/ /2004 1/ /2004 1/ /2004 1/ /2004 1/ /2004

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U. S. NUCLEAR REGULATORY COMMISSION

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REGION II

Docket Nos.: 50-280, 50-281

License Nos.: DPR-32, DPR-37

Report Nos.: 05000280/2003008 and 05000281/2003008

Licensee: Virginia Electric and Power Company (VEPCO)

Facility: Surry Power Station

Location: 5850 Hog Island Road

Surry, VA 23883

Dates: April 21, 2003 - January 7, 2004

Inspectors: C. Payne, Senior Reactor Inspector (Lead Inspector)

W. Rogers, Senior Reactor Analyst

Approved by: C. Casto, Director

Division of Reactor Safety

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SUMMARY OF FINDINGS

IR 05000280/2003-008, 05000281/2003-008; 04/21/2003 -01/7/2004; Surry Power Station,

Units 1 and 2; Significance Determination of Unresolved Items from Triennial Fire Protection

Inspection.

This in-office review was conducted by a regional inspector and a senior reactor analyst. One

preliminary White finding with an apparent violation, one unresolved item with potential safety

significance greater than Green and two Green non-cited violations (NCVs) were 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 White. 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 the Emergency

Switchgear and Relay Room Number 1 (Fire Area 3), the licensees fire

response procedures were not effective in assuring a safe shutdown of the

Unit 1 reactor. The licensee has revised the affected fire response procedures

and is evaluating the need for additional corrective action.

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

against one of the external factors attribute. It affected the objective of the

Initiating Events cornerstone to limit the likelihood events that challenge critical

safety functions as well as 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 Fire Area 3, these procedures may not

preclude an extended loss of reactor coolant pump seal injection flow and may

initiate a reactor coolant pump seal loss of coolant accident which could result in

pressurizer level failing to be maintained within the indicating range as required.

(Section 4OA5.01)

TBD. The inspectors identified a violation having potential safety significance

greater than very low significance because the licensees safe shutdown strategy

and related fire response procedures may be inadequate to ensure a safe

shutdown of the Unit 2 reactor for a severe fire in Emergency Switchgear and

Relay Room (ESGR) Number (No.) 2. This finding is similar to one for ESGR

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No. 1. The licensee has revised the affected fire response procedures and is

evaluating the need for additional corrective action.

This finding is unresolved pending completion of a significance determination.

The finding is greater than minor because it is associated with the ability to

achieve a safe shutdown of the Unit 2 reactor following a fire in ESGR No. 2 and

affects the Initiating Events and Mitigating Systems cornerstone objectives. The

finding has potential safety significance greater than very low, safety significance

because RCP seal package failure could cause a seal loss-of-coolant accident

and failure of the specified alternative shutdown strategy. (Section 1R05)

Green. A Green non-cited 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 the Unit 1 Cable Vault and

Tunnel (Fire Area 1), the licensees alternative shutdown capability may not

ensure that the reactor coolant makeup function would be capable of maintaining

the reactor coolant level within the level indication of the pressurizer. The

licensee has entered this finding into its corrective action program.

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

against one of the external factors attribute. It affected the objective of the

Initiating Events cornerstone to limit the likelihood events that challenge critical

safety functions as well as affected the objective of the Mitigating Systems

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

respond to initiating events. This finding was determined to be of very low safety

significance because the likelihood of a severe fire in the service building cable

vault (SBCV) or the cable tunnel that could cause a loss of all three Unit 1

charging pumps is very low and a 3-hour rated fire door would prevent a severe

fire in the remaining sections of Fire Area 1 from spreading through the cable

tunnel to the SBCV. (Section 4OA5.02)

Green. A Green non-cited violation was identified for failure to comply with

10 CFR 50, Appendix R, Sections III.G.3.a and III.L.3. Specifically, the shared

ventilation system between the main control room (MCR) and the Unit 1 and Unit

2 emergency switchgear and relay rooms (ESGRs), did not have adequate

separation, isolation, or barriers to preclude smoke and toxic gases from being

transported to the ESGRs during a fire in the MCR. The alternative shutdown

capability for an MCR fire is located in each units ESGR, respectively.

Consequently, operators may not have the environmental conditions or visibility

to safely man and accomplish a successful shutdown of either Unit 1 or Unit 2

from the Auxiliary Shutdown Panels. The licensee has entered this finding into

its corrective action program.

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

against external factors attribute and affected the objective of the Mitigating

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Systems cornerstone to ensure the availability, reliability, and capability of

systems that respond to initiating events. This finding was determined to be of

very low safety significance because heat from a fire, and the natural buoyancy

of smoke, will cause the smoke gas layer to accumulate near the ceiling of the

MCR (away from the ESGRs), the likelihood of a severe fire in the MCR is low,

and the prompt response and actions of the MCR operators and the fire brigade

would prevent any fires that start from becoming severe. (Section 4OA5.03)

B. Licensee-identified Violations:

None

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Report Details

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems and Barrier Integrity

1R05 FIRE PROTECTION (71111.05T/SDP)

Section 4OA5.01 of this inspection report documents the resolution of unresolved item

(URI)05000280/2003007-001, Fire Response Procedures 1-FCA-4.00 And 0-FCA-

14.00 Not Adequate To Assure Safe Shutdown Of Unit 1 (NRC Inspection Report

05000280, 281/2003007, dated March 31, 2003, ADAMS Accession Number

ML030930560). Following completion of the significance determination process (SDP)

for this Emergency Switchgear and Relay Room (ESGR) Number (No.) 1 finding, the

inspectors recognized that an evaluation of the finding for applicability to Unit 2s ESGR

No. 2 would be appropriate. [ESGR No. 2 was not included in the original scope of the

baseline triennial fire protection inspection. Thus, it was not evaluated as part of the

ESGR No. 1 SDP analysis.]

a. Inspection Scope

The inspectors conducted an in-office review of plant drawings 11448-FE-27C and

11548-FE-27A (arrangement drawings for ESGR No. 1 and ESGR No. 2, respectively)

to compare the location and design features of safe shutdown equipment and fire

protection features. [The inspectors had walked down both these areas during the

triennial fire protection inspection.] The inspectors also compared Fire Contingency

Action (FCA) procedure 2-FCA-4.00, Limiting ESGR Number 2 Fire, Revision 14, with 1-

FCA-4.00, Limiting ESGR Number 1 Fire, Revision 13, to identify any differences in

operator implementation of the alternative safe shutdown strategies for Unit 1 and

Unit 2. In addition, the Surry Non-Seismic Individual Plant Examination of External

Events (IPEEE) was reviewed for plant fires in ESGR No. 1 and ESGR No. 2. This

included a review of the applicable fire area/compartment Ignition Source Data Sheets.

b. Findings

Because relative equipment and cable locations in Unit 2s ESGR No. 2 are similar to

those in Unit 1s ESGR No. 1, the fire ignition frequency for both areas is similar, and

the alternative safe shutdown procedures used to respond to a severe fire are similar,

the inspectors determined that the finding associated with URI 05000280/2003007-001

was applicable to Unit 2. Specifically, for a severe fire in ESGR No. 2 the procedural

guidance in 2-FCA-4.00, Limiting ESGR Number 2 Fire, may not prevent loss of seal

injection cooling to the Unit 2 reactor coolant pump (RCP) seal packages nor be timely

in restoring seal injection flow, via the charging system cross-connect line with Unit 1, to

prevent damage to the RCP seal packages. Additionally, procedural guidance in 0-

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FCA-14.00, Charging and Seal Injection Flow Paths, to restore seal injection flow

following cross-connect with Unit 1 may aggravate potential damage to the seal

packages and, consequently, increase the severity of leakage from the RCP seals.

Thus, safe shutdown of Unit 2 during a severe fire in ESGR No. 2 would not be ensured.

The licensee captured this issue in its corrective action program under Plant Issue (PI)

S-2003-0637 [during review of this similar issue in ESGR No. 1]. This finding has

potential safety significance greater than very low significance and is a URI pending

completion of the SDP.

10 CFR 50.48 states, in part, Each operating nuclear power plant must have a fire

protection program that satisfies Criterion 3 of Appendix A to this part. Surry Unit 1

Operating License DPR-32, and Surry Unit 2 Operating License DPR-37 Condition 3.I,

specifies, 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

Safety Evaluation Report (SER) dated September 19, 1979, 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 14, 2003, the alternative shutdown capability and

response procedures specified for a fire in ESGR No. 2, an Appendix R,Section III.G.3

area, were not effective and did not meet this requirement. Specifically, the licensees

procedures may not preclude an extended loss of reactor coolant pump seal injection

flow and may initiate a reactor coolant pump seal loss of coolant accident which could

result in pressurizer level failing to be maintained within the indicating range. Pending

determination of the safety significance, this finding is identified as URI

05000281/2003008-001, Fire Response Procedures 2-FCA-4.00 And 0-FCA-14.00 Not

Adequate To Ensure Safe Shutdown Of Unit 2.

4. OTHER ACTIVITIES

4OA5 OTHER

.01 (Closed) URI 05000280/2003007-001: Fire Response Procedures 1-FCA-4.00 And 0-

FCA-14.00 Not Adequate To Assure Safe Shutdown Of Unit 1

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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 the ESGR

No. 1, the licensees 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 White (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 having potential safety significance greater than very low

significance, involving the strategy, and related fire response procedures, for assuring a

safe shutdown of the Unit 1 reactor during a severe fire in ESGR No. 1. Specifically, the

procedural guidance in 1-FCA-4.00, Limiting ESGR Number 1 Fire, may not prevent loss

of seal injection cooling to the Unit 1 RCP seal packages nor be timely in restoring seal

injection flow, via the charging system cross-connect line with Unit 2, to prevent damage

to the RCP seal packages. Additionally, procedural guidance in 0-FCA-14.00, Charging

and Seal Injection Flow Paths, to restore seal injection flow following cross-connect with

Unit 2 may aggravate potential damage to the seal packages and, consequently,

increase the severity of leakage from the RCP seals. The licensee captured this issue

in its corrective action program under PI S-2003-0637. Subsequent licensee

investigation of this issue generated two additional PIs (S-2003-1490 and S-2003-

5254). Pending determination of the safety significance, this finding was documented

as a URI in the triennial fire protection inspection report.

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

quality cornerstone attributes. It affects the objective of the Initiating Events

Cornerstone to limit the likelihood of events that challenge critical safety functions

because existing procedural guidance may result in RCP seal package damage and

increase the likelihood of an RCP seal loss of coolant accident (LOCA). Additionally, the

finding affects the Mitigating Systems Cornerstone to ensure the availability, reliability,

and capability of systems that respond to initiating events [fire] because continuous RCP

seal injection flow is not maintained nor, once seal injection is lost, is it restored quickly

enough to preclude RCP seal damage so that pressurizer level can be maintained in the

indicating range. 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). However, the MC 0609, App. F, Phase 1 screening criteria did not apply to the deficiencies related to

Surrys safe shutdown strategy or fire response procedures. As a result, a Phase 2 risk

analysis was performed. A Summary of the Phase 2 analysis is provided as

Attachment 2.

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Summary of Phase 3 SDP Analysis

This evaluation was performed by contractors supporting the Office of Nuclear Reactor

Regulation with the assistance of senior reactor analysts from headquarters and Region II.

The Surry Phase 3 SDP Analysis is included in this inspection report as Attachment 3.

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 four appendices documenting supplemental

information used in the Phase 3 analysis (circuit analysis, fire propagation, accidental water

spray on both switchgears and event tree analysis).

Five fire scenarios were developed and considered during the Phase 3 analysis.

1. A severe fire in emergency bus room 1H damaging equipment or cables in emergency

bus room 1J. In this scenario, both emergency buses fail without a possibility of

recovery.

2. A severe fire in emergency bus room 1J damaging equipment or cables in emergency

bus room 1H. (Similar to Scenario 1 above.) In this scenario, both emergency buses

fail without a possibility of recovery.

3. A relatively severe fire occurs in emergency bus room 1H. The fire brigade uses water

to extinguish the fire. Water is accidentally sprayed on equipment in emergency bus

room 1J. It is assumed that as a result, both emergency buses fail without a possibility

of recovery.

4. A relatively severe fire occurs in emergency bus room 1J. The fire brigade uses water

to extinguish the fire. Water is accidentally sprayed on equipment in emergency bus

room 1H. (Similar to Scenario 3 above.) It is assumed that as a result, both emergency

buses fail without a possibility of recovery.

5. A severe fire in emergency bus room 1J leads to complete loss of emergency bus 1J,

loss of some of the cable above the electrical cabinets, and recoverable loss of

emergency bus 1H.

Based on an analysis (Appendix B of the Phase 3 SDP Analysis), it was determined that

multi-compartment fires were very unlikely. Thus, Scenarios 1 and 2 were not analyzed

further for risk significance. As a result, only Scenarios 3, 4 and 5 were analyzed in detail

with probabilistic modeling.

The core damage frequency (CDF) for each scenario was calculated by multiplying the

scenario frequency and associated conditional core damage probability (CCDP). The table

below presents the set of scenarios, their associated occurrence frequencies, CCDPs and

CDFs.

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Scenario CCDPs and CDFs

Scenario Frequency CCDP CDF %

3. Water Spray on Both Buses - Room 1H 1.1E-06 3.3E-01 3.6E-07 16.0

4. Water Spray on Both Buses - Room 1J 1.9E-07 3.3E-01 6.1E-08 2.7

5. Severe switchgear fire in Room 1J 5.8E-05 3.2E-02 1.8E-06 81.3

Total 5.9E-05 2.3E-06

As the table above indicates, the total CDF for the entire set of scenarios (the non-

conforming case) was calculated to be 2.3E-06 per reactor year. Due to the low base case

CDF, the risk analysts concluded that the delta CDF [the difference between the base case

CDF and the non-conforming case CDF] was essentially the same as the non-conforming

case CDF (i.e., 2.3E-06). This result indicates the risk significance of the finding is of 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 2.3E-06. The

key factors in the risk determination which most influenced this result were the CDF

associated with a severe fire in the 1J 4160V switchgear and the lack of an automatic fire

suppression system in the fire area. The color associated with this magnitude of change in

CDF is White. Therefore, the SERP has preliminarily determined this issue to be a White

finding.

Enforcement: 10 CFR 50.48 states, in part, Each operating nuclear power plant must have

a fire protection program that satisfies Criterion 3 of Appendix A to this part. Surry Unit 1

Operating License DPR-32, and Surry Unit 2 Operating License DPR-37 Condition 3.I,

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

approved fire protection program as described in the Updated Final Safety Analysis Report

(UFSAR) and as approved in the SER dated September 19, 1979, 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.

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENTS 2 AND 3, THIS DOCUMENT IS DECONTROLLED

Enclosure

6

ATTACHMENTS 2 AND 3 CONTAIN PROPRIETARY INFORMATION

Contrary to the above, on February 14, 2003, the alternative shutdown capability and

response procedures specified for a fire in ESGR No. 1, an Appendix R,Section III.G.3

area, were not effective and did not meet this requirement. Specifically, the licensees

procedures may not preclude an extended loss of reactor coolant pump seal injection flow

and may initiate a reactor coolant pump seal loss of coolant accident which could result in

pressurizer level failing to be maintained within the indicating range. This apparent violation

is identified as AV 05000280/2003008-002, Alternative Shutdown Capability and Response

Procedures Not Adequate to Ensure Safe Shutdown of Unit 1.

.02 (Closed) URI 05000280/2003007-002: Fire Response Procedures 1-FCA-3.00 And 0-FCA-

14.00 Not Adequate To Assure Safe Shutdown Of Unit 1

Introduction: A Green non-cited violation (NCV) was identified for failure to comply with 10

CFR 50, Appendix R, Criterion III.L.2.b and III.L.3, in that, for a severe fire in the Unit 1

Cable Vault and Tunnel (CV&T), the licensees 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.

Description: During the triennial fire protection inspection, the inspectors identified a finding

having potential safety significance greater than very low significance, involving the

strategy, and related fire response procedures, for assuring a safe shutdown of the Unit 1

reactor during a severe fire in Unit CV&T (Fire Area 1). Specifically, the procedural

guidance in 1-FCA-3.00, Limiting Cable Vault and Cable Tunnel Fire, may not prevent loss

of seal injection cooling to the Unit 1 RCP seal packages nor be timely in restoring seal

injection flow, via the charging system cross-connect line with Unit 2, to prevent damage to

the RCP seal packages. In addition, procedural guidance in 0-FCA-14.00, Charging and

Seal Injection Flow Paths, to restore seal injection flow following cross-connect with Unit 2

may aggravate potential damage to the seal packages and, consequently, increase the

severity of leakage from the RCP seals. The licensee captured this issue in their corrective

action program under PI S-2003-0637. Subsequent licensee investigation of this issue

generated two additional PIs (S-2003-1490 and S-2003-5254). Pending determination of

the safety significance, this finding was documented as a URI in the triennial fire protection

inspection report.

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

against external factors and procedure quality cornerstone attributes. It affects the

objective of the Initiating Events Cornerstone to limit the likelihood of events that challenge

critical safety functions because existing procedural guidance may result in RCP seal

package damage and increase the likelihood of an RCP seal LOCA. Additionally, the

finding affects the Mitigating Systems Cornerstone to ensure the availability, reliability, and

capability of systems that respond to initiating events [fire] because continuous RCP seal

injection flow is not maintained nor, once seal injection is lost, is it restored quickly enough

to preclude RCP seal damage so that pressurizer level can be maintained in the indicating

range.

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENTS 2 AND 3, THIS DOCUMENT IS DECONTROLLED

Enclosure

7

ATTACHMENTS 2 AND 3 CONTAIN PROPRIETARY INFORMATION

During the significance determination period, the inspectors determined that the issue was

of very low safety significance (Green). Some of the factors (including assumptions used in

the SDP, MC 0609, App. F) causing the issue to be of very low safety significance were:

The Unit 1 CV&T is comprised of four interconnected rooms - the service building cable

vault (SBCV), the cable tunnel, the cable penetration vault (CPV) and the motor control

center (MCC) room. A normally open 3-hour rated fire door (and wall) separates the

CPV and the MCC room from the cable tunnel and the SBCV. The operators and fire

brigade can enter the Unit 1 CV&T through several different doors: from the outside

yard to the MCC room, from the auxiliary building (15'-0" elevation) to the CPV, or from

ESGR No.1 to the SBCV (two separate doors).

The CV&T is protected by a total flooding, automatic carbon dioxide (CO2) suppression

system. The CO2 system is divisional such that a fire in one section of the fire area will

only dump CO2 in that section. The cable tunnel fire door automatically shuts upon

actuation of the CO2 suppression system within the CV&T. This essentially creates two

separate fire zones: 1) the MCC room and the CPV and 2) the cable tunnel and the

SBCV. No findings were associated with this rated fire door.

The alternative safe shutdown procedure for the Unit 1 CV&T (i.e., cross-connecting the

Unit 1 charging system with Unit 2's) would only be implemented if all Unit 1 charging

flow is lost. Control and power cables for all three Unit 1 charging pumps pass from

ESGR No. 1 into the SBCV. The cables for the 1A and 1C charging pumps then pass

from the SBCV directly into the auxiliary building. In contrast, the cables for the 1B

charging pump first pass from the SBCV, down through the cable tunnel into the CPV,

and then into the auxiliary building. Assuming the cable tunnel fire door functions

correctly, only a severe fire in the SBCV and/or cable tunnel could cause a loss of all

Unit 1 charging flow.

The cable tunnel contains no ignition sources. The only ignition sources in the SBCV

are three relay panels associated with the cooling water canal level system. All three

panels contain only relays and cables that are energized by low voltage power.

Combined with their low fire ignition frequency, the likelihood of these relay panels

causing a severe fire in the SBCV is very low.

Implementation of the alternative safe shutdown procedures directs the performance of

twenty-nine manual operator actions in the Unit 1 CV&T. All twenty-nine steps would be

performed in the CPV. Assuming the cable tunnel fire door and the divisional CO2 fire

suppression system function as designed, the effects of a severe fire in the SBCV would

not prevent the operators from entering the CPV to perform their required actions. [To

avoid the fire, the operators could enter the CPV from either the MCC room or the

auxiliary building.]

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENTS 2 AND 3, THIS DOCUMENT IS DECONTROLLED

Enclosure

8

ATTACHMENTS 2 AND 3 CONTAIN PROPRIETARY INFORMATION

Enforcement: Because this failure to comply with 10 CFR 50, Appendix R, Sections III.L.2.b

and III.L.3, is of very low safety significance and has been entered into the corrective action

program (PI S-2003-0637), this violation is being treated as an NCV, consistent with Section

VI.A of the NRC Enforcement Policy: NCV 05000280/2003008-003, Fire Response

Procedures 1-FCA-3.00 And 0-FCA-14.00 Not Adequate To Ensure Safe Shutdown Of Unit

1.

.03 (Closed) URI 05000280,281/2003007-003: Alternate Shutdown Panel Ventilation System

Not Independent From Impacts Of A Main Control Room Fire

Introduction: A Green NCV was identified for failure to comply with 10 CFR 50, Appendix R,

Sections III.G.3 and III.L.3. Specifically, the shared ventilation system between the MCR

(Fire Area 5) and ESGR No. 1 and ESGR No. 2 (Fire Areas 3 and 4, respectively), did not

have adequate separation, isolation, or barriers to preclude smoke and toxic gases from

being transported to the ESGRs during a fire in the MCR. The alternative shutdown

capability for an MCR fire is located in each units ESGR, respectively. Consequently,

operators may not have the environmental conditions or visibility to safely man and

accomplish a successful shutdown of either Unit 1 or Unit 2 from the Auxiliary Shutdown

Panels (ASP).

Description: During the triennial fire protection inspection, the inspectors identified a

finding of having potential safety significance greater than very low significance,

involving the lack of adequate separation, isolation, or barriers to preclude smoke and

toxic gases from being transported to the ESGRs during a fire in the MCR. The Surry

Appendix R Report identified the MCR fire area as an alternative shutdown area. During

a severe fire in the MCR, the operators would abandon the MCR and utilize the Unit 1

and Unit 2 ASPs, located in the Unit 1 and Unit 2 ESGRs respectively, to achieve a safe

shutdown of the units. The ESGRs share a common ventilation system with the MCR.

Fire dampers, located in the ventilation system ducts, were designed to isolate the

ESGR area to contain the Halon within the ESGRs, and to prevent smoke and toxic

gases from spreading from the ESGRS to the MCR. Although an ESGR fire alarm

signal or manual actuation of the Halon system (in response to an ESGR fire) would

signal these dampers to close, the inspectors found that there were no smoke or fire

actuation devices to signal them to shut during a fire in the MCR. Additionally these

dampers do not have the capability of being manually actuated from the MCR. During a

severe fire in the MCR, large amounts of heavy black smoke and toxic gases could be

generated. The open dampers could permit smoke and toxic gases to spread from the

MCR to the ESGR. This situation could present a habitability concern for the operators

attempting to achieve shutdown at the respective units ASP.

Fire procedure 0-FCA-1.00, Limiting MCR Fire, Revision 29, does not require the

operators to bring self-contained breathing apparatus (SCBA) gear to the ESGR nor are

any SCBAs readily available at the ESGRs. The Surry Appendix R Report did not

include an evaluation of potential maloperation of the ventilation system, its

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENTS 2 AND 3, THIS DOCUMENT IS DECONTROLLED

Enclosure

9

ATTACHMENTS 2 AND 3 CONTAIN PROPRIETARY INFORMATION

components, or its effect on habitability at the ASP. As a result, the alternative

shutdown capability was not physically independent of the fire area as required by

Sections III.G.3 and III.L of Appendix R. The licensee initiated PI S-2003-0643 to

evaluate the independence and operability of the ESGR ventilation system during an

MCR fire.

Analysis: The inspectors determined that the finding was associated with the protection

against external factors attribute and affected the objective of the Mitigating Systems

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

to initiating events. Therefore, the finding is greater than minor.

During the significance determination period, the inspectors determined that the issue

was of very low safety significance (Green). Some of the factors (assumptions used in

the SDP, MC 0609, App. F) causing the issue to be of very low safety significance were:

Heat from a fire and the natural buoyancy of smoke will cause the smoke gas

layer to accumulate near the ceiling of the MCR. Because the ESGR is located

below the MCR, smoke and toxic gases must nearly fill the MCR envelope in

order to drive the smoke gas layer down through ventilation ducts to the room

below.

Due to the large volume in the MCR, more than two bench boards would need to

be involved in a fire to generate sufficient smoke to fill the MCR. The likelihood

of fire spreading to more than two bench boards is very low due to their low fire

ignition frequency and due to their construction (self-contained cabinets).

Additionally, the MCR is a normally manned station so the MCR operators would

attempt to fight the fire in its early stages.

The fire brigade nominally responds in 10-15 minutes (based on fire drills over

the last 18 months) of fire notification. At that time, an MCR door(s) would be

opened to allow fire brigade access to fight the fire. This action would serve to

vent smoke out of the MCR to the turbine building and reduce the likelihood of

smoke migration down to the ESGRs. In addition, the fire brigade would set up

portable ventilation equipment to enhance smoke removal from the area.

Enforcement: This finding was considered a failure to comply with 10 CFR 50, Appendix

R, Sections III.G.3 and III.L.3, which specify that the alternative shutdown capability

shall be independent of the affected fire area(s). Contrary to the above, the shared

ventilation system between the MCR and the ESGRs did not have adequate separation,

isolation, or barriers to preclude smoke and toxic gases from being transported to the

ESGRs during a fire in the MCR. Because this finding is of very low safety significance

and has been entered into the licensees corrective action program (PI S-2003-0643),

this violation is been treated as an NCV, consistent with section VI.A of the NRC

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENTS 2 AND 3, THIS DOCUMENT IS DECONTROLLED

Enclosure

10

ATTACHMENTS 2 AND 3 CONTAIN PROPRIETARY INFORMATION

Enforcement Policy: NCV 05000280,281/2003008-004, Alternate Shutdown Panel

Ventilation System Not Independent From Impacts Of A Main Control Room Fire.

4OA6 Meetings, Including Exit

On February 2, 2004, the inspectors presented the inspection results by telephone to

Mr., 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 ATTACHMENTS 2 AND 3, THIS DOCUMENT IS DECONTROLLED

Enclosure

11

ATTACHMENTS 2 AND 3 CONTAIN PROPRIETARY INFORMATION

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

M. Adams, Site Engineering Manager

R. Blount, Site Vice President

T. Carlisle, Nuclear Engineering

B. Garber, Licensing

T. Gunning, Fire Protection Engineer

J. Kloecker, Mechanical Engineer

H. Le, Supervisor Engineering

M. Smith, Systems Engineering Manager

T. Sowers, Director Operations and Maintenance

B. Staley, Maintenance Manager

J. Swientoniewski, Operations Manager

M. Thomas, Electrical

NRC personnel

G. McCoy, Senior Resident Inspector, Surry

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000281/2003008-001 URI Fire Response Procedures 2-FCA-4.00 And 0-FCA-14.00

Not Adequate To Ensure Safe Shutdown Of Unit 2

(Section 1R05)05000280/2003008-002 AV Alternative Shutdown Capability and Response

Procedures Not Adequate to Ensure Safe Shutdown of

Unit 1 (Section 4OA5.01)

Opened and Closed

05000280/2003008-003 NCV Fire Response Procedures 1-FCA-3.00 And 0-FCA-14.00

Not Adequate To Ensure Safe Shutdown Of Unit 1

(Section 4OA5.02)

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENTS 2 AND 3, THIS DOCUMENT IS DECONTROLLED

Attachment 1

12

ATTACHMENTS 2 AND 3 CONTAIN PROPRIETARY INFORMATION

05000280,281/2003008-004 NCV Alternate Shutdown Panel Ventilation System Not

Independent From Impacts Of A Main Control Room Fire

(Section 4OA5.03)

Closed

05000280/2003007-001 URI Fire Response Procedures 1-FCA-4.00 And 0-FCA-14.00

Not Adequate To Assure Safe Shutdown Of Unit 1

(Section 4OA5.01)05000280/2003007-002 URI Fire Response Procedures 1-FCA-3.00 And 0-FCA-14.00

Not Adequate To Assure Safe Shutdown Of Unit 1

(Section 4OA5.02)

05000280,281/2003007-003 URI Alternate Shutdown Panel Ventilation System Not

Independent From Impacts Of A Main Control Room Fire

(Section 4OA5.03)

LIST OF DOCUMENTS REVIEWED

Procedures:

0-FCA-14.00, Charging and Seal Injection Flow Paths, Rev. 2

1-FCA-3.00, Limiting Cable Vault and Cable Tunnel Fire, Rev. 12

1-FCA-4.00, Limiting ESGR Number 1 Fire, Rev. 13

2-FCA-4.00, Limiting ESGR Number 2 Fire, Rev. 14

Drawings:

(Note: 11448 indicates Unit 1, 11548 indicates Unit 2)

11448-FAR-205, Equipment Location - Appendix R Auxiliary Building Plan El 13-0", sh. 2,

Rev. 16

11448-FB-25A, Ventilation & Air Conditioning Service Building, sh. 1, Rev. 9

11448-FB-25B, Ventilation & Air Conditioning Service Building, sh. 2, Rev. 9

11448-FB-25C, Ventilation & Air Conditioning Service Building, sh. 1, Rev. 17

11448-FB-25D, Ventilation & Air Conditioning Service Building - El. 9-6", sh. 1, Rev. 16

11448-FB-25E, Ventilation & Air Conditioning Service Building - El. 9-6", sh. 1, Rev. 22

11448-FB-25F, Ventilation - Service Building Floor El. 42'-0" and 47'-0" Columns 21/4 to 6, sh. 1,

Rev. 13

11448-FB-25G, Ventilation - Service Building Floor El. 42'-0" Columns 10 to 131/2, sh. 1, Rev.

12

11448-FB-25H, Ventilation - Service Building Floor El. 27'-0" Columns 21/4 to 5, Rev. 7

11448-FB-25J, Ventilation & Air Conditioning Service Building, sh. 9, Rev. 9

DOCUMENT TRANSMITTED HEREWITH CONTAINS SENSITIVE UNCLASSIFIED INFORMATION

WHEN SEPARATED FROM ATTACHMENTS 2 AND 3, THIS DOCUMENT IS DECONTROLLED

Attachment 1

13

ATTACHMENTS 2 AND 3 CONTAIN PROPRIETARY INFORMATION

11448-FB-25K, Ventilation - Service Building Floor El. 27-0" Columns 10 to 14, sh. 1, Rev. 10

11448-FB-25L, Ventilation & Air Conditioning Service Building, sh. 1, Rev. 10

11448-FB-25M, Ventilation & Air Conditioning Service Building, sh. 12, Rev. 4

11448-FB-25N, Ventilation & Air Conditioning Service Building, sh. 13, Rev. 3

11448-FB-25K, Ventilation - Service Building Roof El. 60'-0" and 75'-0" Columns 21/2 to 14,

Rev. 6

11448-FB-25R, Ventilation - Service Building Floor El. 27'-0" Columns F to 21/4, Rev. 4

11448-FB-25S, Ventilation - Service Building Floor El. 42'-0" Columns F to 21/4, Rev. 2

11448-FB-25T, Ventilation - Service Building Roof/Floor El. 56'-0" and 70'-0" Columns F to 21/4,

Rev. 3

11448-FB-25U, Ventilation - Service Building Floor El. 27'-0" Part Plans, Sections and Detail,

Rev. 2

11448-FB-25V, Ventilation - Service Building Part Plans El. 42'-0" Columns 11 to 121/2, Rev. 4

11448-FE-3FH, Wiring Diagram Control Cabinet 1-CW-PNL-1A & 1-CW-PNL-1B, sh. 1, Rev. 2

11448-FE-3FJ, Wiring Diagram Control Logic Cabinet 1-CW-PNL-2, sh. 1, Rev. 0

11448-FE-27C, Arrangement Emergency Switchgear and Relay Rooms El. 9'-6", sh. 1, Rev. 31

11448-FE-42T, Conduit Plan Emergency Swgr Rm El. 9'-6", Rev. 18

11448-FE-45A, Conduit & Cable Tray Plan Cable Tunnel & Vaults, sh. 1, Rev. 19

11448-FE-48C, Conduit Plan Auxiliary Building - El. 13'-0", sh. 1, Rev. 19

11448-FE-48F, Cable Terminations & Conduit Sleeve Loading Tables Auxiliary Building, sh. 1,

Rev. 31

11448-FE-90BA, Appendix R Block Diagram Charging Pump System, sh. 1, Rev. 2

11448-FE-90BB, Appendix R Block Diagram Charging Pump System, sh. 2, Rev. 2

11448-FM-5B, Arrangement Auxiliary Building, sh. 1, Rev. 13

11548-FE-27A, Arrangement Emergency Switchgear and Relay Rooms El 9'-6", sh. 1, Rev. 25

Plant Issue Reports Reviewed:

S-2003-1490, Review FCA procedures to determine the need for additional guidance on

establishment of charging flow to both units via the charging pump cross-tie.

S-2003-5254, The design data used to support Dominions methodology for maintaining

pressurizer level following an Appendix R fire in the Unit 1 emergency switchgear room

appears to be inadequate and non-conservative.

ML040490131

Text

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