ML050560347

Draft Version of Edwin I. Hatch Nuclear Power Plant - NRC Triennial Fire Protection IR 05000321-03-006 and 05000366-03-006

ML050560347
Person / Time
Site:	Hatch
Issue date:	08/31/2003
From:	Belcher S, Ogle C, Schin R, Casey Smith, Sullivan K, Wiseman G Division of Reactor Safety II
To:
References
FOIA/PA-2004-0277 IR-03-006
Download: ML050560347 (39)

See also: IR 05000321/2003006

Text

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U.S. NUCLEAR REGULATORY COMMISSION.

REGION I .

Docket Nos.: 50-321, 50-366

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• License Nos.: ..DPR-57, NPF-5

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Report No.: 05000321/200306 and 05000366/200306

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Licensee: Southern Nuclear Operating Company . . i

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Facility: E. I. Hatch Nuclear Plant

Location: P.O. Box 2010

Baxley, GA. 31513

Dates: July 7-11, 2003 (Week1) . .

July 21-25, 2003 (Week 2).

Inspectors: C. Smith, P E., Senior Reactor Inspector, (Lead Inspector)

R. Schin, Senior Reactor Inspector . .

G. Wiseman, Fire Protection Inspector

K. Sullivan, Consultant, Brookhaven National Laboratory

Accompanying S. Belcher, Nuclear Safety Intern,

Personnel:

Approved by: Charles R. Ogle, Chief

Engineering Branch 1

Division of Reactor Safety

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CONTENTS

SUMMARY OF FINDINGS ..................................................

REPORT DETAILS .............................................. ........

REACTOR SAFETY

FIRE PROTECTION

Systems Required to Achieve and Maintain Safe Shutdown .................

Fire Protection of Safe Shutdown Capability ................................

Post Fire safe Shutdown Capability.,..........................

Operational Implementation of Alternative Shutdown Capability..............

• Communications................

Emergency Lighting .....................................................

&old Shutdown Repairs ....................... I.........................

Fire Barriers and Fire Area/Zone/Room Penetration Seals .....................

Fire Protection Systems, Features, and Equipment............................

SAFETY SYSTEM DESIGN AND PERFORMANCE CAPABILITY

• • DCR 91-134, SRV Backup Actuation via Pressure Transmitter Signals .............

• OTHER ACTIVITIES.

Identification and Resolution of Problems .................................... ..

Meetings Including Exit....................................... ......

• . Supplemental Information ..............................................

List of Items Opened, Closed, and Discussed.................................

List of Documents Reviewed..............................................

2

SUMMARY OF FINDINGS.

IR 05000321/2003-006, 05000366/2003-006; Southern Nuclear Operating Company,-.

7/7-11/2003 and 7/21-25/2003; E. I. Hatch Nuclear Plant, Units'I and 2; Triennial FIre'

Ptectiidn

'The report coverdatowepridf inspection by three regional ispectors and a'

co actor from Brookhaven'National Laboratory. Three Green non-cited violations (NCVs) and

• rnresolved items with potential safety significance greater than Green were identified. The

5infcance 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-1 649, "Reactor Oversight Process," Revision 3,'dated July 2000.

A. NRC-Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

• URI. The team identified an unresolved item in that a local manual operator action, to

. prevent spurious opening of all eleven safety relief valves (SRVs) during a fire event,.

would not be performed in sufficient time to be effective. Also, licensee reliance on this

manual action for hot shutdown during a fire, instead of physically 'protecting cables from.

• fire damage, had not been approved by the NRC.

• ~This finding is unresolved pending coripletion of a significance determination. In

response to this potential issue, the licensee promptly moved the manual action step to

• the front of the Fire Procedure to enable operators to accomplish the action' much

sooner during a fire event. This finding was determined to have potential safety

• significance greater than very low significance because of the use of manual actions In

• * lieu of physical protection as required by 10 CFR 50 Appendix R, Section IIll.G.2.

(Section11R05.05.b.1)

• * URI. The team identified an unresolved item in that a fire in Fire Area 2104 could

• ~cause all eleven SRVs to open at a time when residual heat removal (RHR) system

may not be available. To mitigate this event, the licensee's safe shutdown analysis

report (SSAR) credits the use of Core Spray Loop A to provide reactor coolant makeup.

However, the licensee did not provide any objective evidence (e.g., specific calculation

or analysis) which demonstrated that, assuming worst-case fire damage In.Fire Area

• -2104, the limited set of equipment available would be capable of mitigating the event In

.a manner that satisfies the shutdown performance goals specified in Appendix R,

section L.1.e to1IOCFR 50.

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This finding is unresolved pending completion of the NRC review of a calculation of -

record which demonstrates the capability of the Core Spray system to mitigate the'

. above event. This finding was determined to have potential safety significance greater:

than very low significance because of a lack of a calculation of record and'

documentation of the limited set of equipment that would be credited for safe shutdo

under these conditions. (Section lR.05.03.b).'

' URIS The team identified an unresolved item in that the licensee's current fire '

protection licensing basis characterizes the opening of terminal board links in control

panels as a repair activity which is not permitted to achieve and maintain hot shutdown

conditions. The licensee could not provide any evidence to justify why these actions

were not characterized as a repair activity in its current SSAR. In response to this'

-inspectionfinding the licensee initiated a Condition Report (CR 2003800152, dated

7/24/03) to evaluate actions to open links, in order to determine if they are necessary to

achieve hot shutdown, and if an exemption from Appendix R is required.'

This finding is unresolved Pending completion of a significance determination. This

finding is greater than minor because it impacts the mitigating system cornerstone and

has the potential for the operator not successfully completing the action because of

adverse human factor conditions. (Section 1R.05.01.b) . '

URI: The team identified an unresolved item in connection with the implementation of

design change request (DCR)91-134, SRV Backup Actuation via Pressure Transmitter.

Signals. The installed'plant modification failed to implement the one-out-of-two taken

twice logic that was specified as design input requirements in the design change.

' package. Additionally, implementation of a two-out-of-two coincident taken twice logic,'

has introduced a potential common cause failure of all eleven SRVs because of fire

- induced damage to two instrumentation circuit cables in close proximity to each other.

This finding is unresolved pending completion of a significance determination. This

finding is greater than minor because it impacts the mitigating system cornerstone. This

finding has the potential for defeating manual control of Group 'A" SRVs that are

required for ensuring that the suppression pool temperature will not exceed the 't

capacity temperature limit (HCTL) for the suppression pool. (Section 1R21.0185 /

Green. The team identified a finding with very low safety significance in that a local

manual operator action to operate safe shutdown equipment was too difficult and was'

also unsafe. The licensee had relied on this action instead of providing physical

protection of cables from fire damage or preplanning cold shutdown repairs. However,

the team judged that some operators would not be able to perform the action.

This finding involved a violation of 10 CFR'50, Appendix R,Section III.G.1 and

Technical Specification 5.4.1. The finding is greater than minor because it affected the

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availability and reliability objectives and the equipment performance attribute of the-,.'

mitigating systems cornerstone. Since the licensee could have time to develop and'.

implement cold shutdown repairs to facilitate accomplishment of the action, this firiding

did not have potential safety significance greater than very low safety significance.'...

(Section 1R05.05.b.2) ' . ..

• . . Green. The team identified a finding with very low safety significance in that the '  :

licensee relied on some manual operator actions to operate safe shutdown equipment,

instead of providing the required physical protection of cables from fire damage- and

• without NRC approval..

This finding involved a violation of 10CFR 50, Appendix R, Section lIl.G.2. The finding

is greater than minor because it affected the availability and reliability objectives and the

equipment performance attribute of the mitigating systems cornerstone., Since the

actions could reasonably be accomplished by operators in a timely manner, this finding'

did not have potential safety significance greater than very low safety significance., .. . ' i

(Section 1R05.05.b.3) .

• Green. The team identified a finding with very low safety significance in that emergency

supportlpost-iretoperationofasafe shutdwntequipment

lighting was not adequate for some manual operator actions that were needed to .

' ,, t ~~support post-fire operation of safe shutdown equipment. - ,

This finding involved a violation of 10 CFR 50, Appendix R,Section lll.J; The finding is

greater than minor because it affected the reliability objective and the equipment - .

performance attribute of the mitigating systems cornerstone.. Since operators would be

able to accomplish the actions with the use of flashlights, this finding did not have

potential safety significance greater than very low safety significance. (Section

1R05.07.b)

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 '

The purpose of this inspection was to review the Hatch Nuclear Plant fire protection program

(FPP) for selected risk-significant fire areas. Emphasis was placed on verification that the post-

fire safe shutdown (SSD) capability and the fire protection features provided for ensuring that at

least one redundant train of safe shutdown systems is maintained free of fire damage.: The

inspection was performed in accordance with the Nuclear Regulatory Commission (NRC)

Reactor Oversight Program using a risk-inf6rmed approach for selecting' the fire areas and

attributes to be inspected. The team used the licensee's Individual Plant Examination for

External Events and in-plant tours to choose four risk-significant fire areas for detailed

inspection and review. The fire areas chosen for review during this inspection were:

Fire Area 2016, West 600 V Switchgear Room, Control Building, Elevation 130 feet.

Fire Area 2104, East Cableway, Turbine Building, Elevation 130 feet.

Fire Area 2404, Switchgear Room 2E, Diesel Generator Building, Elevation 130 feet.

' . Fire Area 2408, Switchgear Room 2F, Diesel Generator Building, Elevation 130 feet.:-

The team evaluated the licensee's FPP against applicable requirements, including Operating -

License Condition 2.D, Fire Protection; Title 10 of the Code of Federal Regulations, Part 50 (10

CFR 50), Appendix R; 10 CFR 50.48; Appendix A of Branch Technical Position (BTP) Auxiliary

and Power Conversion Systems Branch (APCSB) 9.5-1; related NRC Safety Evaluation

Reports (SERs); the Hatch Nuclear Plant Updated Final Safety Analysis Report (UFSAR); and.

plant Technical Specifications (TS). The team evaluated all areas of this inspection, as

documented below, against these requirements.

Documents reviewed by the team are listed in the attachment.

01 Systems Required to Achieve and Maintain Post-Fire Safe Shutdown

a. Inspection Scope

The licensee's Safe Shutdown Analysis Report (SSAR) was reviewed to determine the

components and systems necessary to achieve and maintain safe shutdown conditions

in the event of fire in each of the selected fire areas. The objectives of this evaluation '-

were as follows:

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* :: (a) Verify that the licensee's shutdown methodology has correctly identified

the components and systems necessary to achieve and maintain a safe

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(b) Confirm the adequacy of the systems selected for reactivity control,

reactor coolant makeup, reactor heat removal, process monitoring and

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support system functions.'

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(c) Verify that a safe shutdown can be achieved and maintained without off-

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site power, when it can be confirrrn "nvnf the

selected fire areas could cause tl4

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(d) Verify that local manual operatorl , It

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fire protection licensing basis.  :

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.. bz.lg:*nFindings i-t- I

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Licensing Basis for Repair Activities(Opening/ r 2,

Safe Shutdown Condition.

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Introduction: The team iden fied a potential concern wi ice...

- actions to disconnect terminal board sliding linkstin order to isolate two 4-20 ma

instrumentation control loop cirbuits in order to prevent the spurious actuation of eleven

• SRVs.

-: rescition: The licensee has ide tified the systems required to perform the shutdown

functions of reactor shutdown, overT ressure protection, maintenance of coolant

inventory, and decay heat removal fo4 SSD paths 1 and 2. The reactor shutdown

function is provided by the reactor prof ction system (RPS) for all paths.

Path 1 utilizes reactor core isolation dooling (RCIC), two group MA"SRVs, and the

residual heat removal (RHR) systeri in the alternate shutdown cooling mode of'

operation to provide inventory ma~keup, decay heat removal, and depressurization..

RCIC would be used until approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into the event, at which time the

reactor pressure will be within the low-pressure coolant injection (LPCI) operability range

(approximately 135 psig). To mitigate the impact of a spurious actuation of the automatic

- depressurization system (ADS) at a time when hHR system may not be available due to

• fire damage, the licensee has assured that Core Spray (CS) would be available.

• ' Path 2 utilizes the High Pressure Coolant Injection (HPCI), two group MA"SRVs, and the

RHR system in the alternate shutdown cooling mode of operation. The HPCI system

and one SRV are utilizes during the first 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of a~fire event to maintain the reactor

water level and pressure within acceptable limits. Afteirapproximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the RHR

system is started in the alternate shutdown cooling mode of operation.

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For the fire areas evaluated, the licensee identified the structures, systems and

components needed to achieve and maintain safe shutdown conditions in the event of

fire. The team evaluated required manual operator actions in order to verify that they

were consistent with the plant's fire protection licensing basis.NBased on this evaluation

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• the team determined that the licensee r operator actions to open

terminal board links as a means of prev n sired actuation of all eleven

SRVs. (see section 1R2 1.01b).

Analysis: This finding is greater-t minor because it affected the availability.and

. . reliability objectives and the.equipment performance attribute of the. mitigating systems

• cornerstone. Additionally,-the licensee's current licensing basis documents (Reference:

'Georgia Power requestfor exemption dated May 16, 1986 and a subsequent Safety

Evaluation Report (SER) dated January 2, 1987) characterized the opening of links as a..

repair activity that is not permitted as a means of complying 'with Section' III.G of,

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. :. - .- Enforcement: The licensee's current licensing basis documents((Referen:e:eGeorgia:

. . Power request for exemption dated May 16, 1986. and a subsequent Safety Evaluation'...

- Report (SER) dated January 2,1987) characterized the opening of links asa repair.,

. . activity that is not permitted as a means of complying with Section II.G of Appendix R.

Based on these documents the opening of links was considered a repair by both the

licensee and the NRC staff in 1987. The licensee could not provide any evidence to -

. . justify why these actions are not characterized as a repair activity in its current SSAR: .

In response to this inspection finding the licensee initiated a Condition Report (CR'.

2003800152, dated 7/24/03) to evaluate actions to open links, in order to determine If. /

necessary

they are hot shutdown, and if an exemption fromAppendix R is

to achieve

required. This issue is combined w Untimely and Unapproved :

• . Operator Action for Post Fire 'af6 Shutdown discussed in section

-Manual

• :1R05.05.b.1of the report. b;elicensii basis concerns will be disposition upon review.

. and acceptance of additihnali sing which demonstrates that

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. actions necessary to op not be considered a repair necessary to achieve

and maintain hot sbutdo n conditions.

• .02. _ Fire Protection of Safe Shutdown Capabilitv . -..

a.. Inspectiori Scope

For the selected fire areas, the team evaluated the frequency of fires or the potential for

fires, the combustible fire load characteristics and potential fire severity, the separation

of systems necessary to achieve safe shutdown (SSD), and the separation of electrical

components and circuits located within the same fire area to ensure that at least one

SSD path was free of fire damage. The team also inspected the fire protection features II

to confirm they were installed in accordance with the codes of record to satisfy the i

applicable separation and design requirements of 10 CAR 50, Appendix R,Section III.G, ... II

and Appendix A of BTP APCSB 9.5-1. The team reviewed the following documents, I

which established the controls and practices to prevent fires and to control combustible

fire loads and ignition sources, to verify that the objectives established bythe I

NRC-approved fire protection program (FPP) were satisfied:

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l\: l *: Administrative Procedure 40AC-ENG-008-OS, Fire Protection Program.  : .

Administrative Procedure 42FP-FPX-01 8-OS, Use, Control, and Storage of

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Flammable/Combustible Materials

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pressure within acceptable limits. After approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the RHR system is

started in the alternate shutdown cooling mode of operation.

For the fire areas evaluated, the licensee identified the structures, systems and

components needed to achieve and maintain safe shutdown conditions in the event of

fire. The team evaluated required manual operator actions in order to verify that they..

were consistent with the plant's fire protection licensing basis. Based on this evaluation

the team determined that the licensee relies on manual operator actions to open

terminal board links as a means of preventing an undesired actuation of all eleven

SRVs. (see section 1R21.01).

Analysis: This finding is greater than minor because it affected the availability and

reliability objectives and the equipment performance attribute of the mitigating systems

cornerstone. Additionally, human factors problems created the potential for the operator

to not successfully complete the task. The above concerns along with the fact that the

opening of terminal board links are considered "repairs" causes this finding to have

potential safety significance greater than low safety significance.

Enforcement: The licensee's current licensing basis documents (Reference: Georgia

Power request for exemption dated May 16, 1986 and a subsequent Safety Evaluation

Report (SER) dated January 2, 1987) characterized the opening of links as a repair

activity that is not permitted as a means of complying with Section III.G of Appendix R.

Based on these documents the opening of links was considered a repair by both the

licensee and the NRC staff in 1987. The licensee could not provide any evidence to

justify why these actions are not characterized as a repair activity in its current SSAR.

In response to this inspection finding , the licensee initiated a Condition Report (CR

2003800152, dated 7/24/03) to evaluate actions to open links, in order to determine if

they are necessary to achieve hot shutdown, and if an exemption from Appendix R is

required. This issue is identified as URI 50-366/03-06-01, Licensing Basis for Repair

Activities (Opening/Closing of Links) to Achieve Safe Shutdown Conditions. This item

remains open pending review and acceptance of additional licensing basis

documentation which demonstrates that actions necessary to open links should not be

considered a repair necessary to achieve and maintain hot shutdown conditions.

.02 Fire Protection of Safe Shutdown Canabilitv

a. Inspection Scone

For the selected fire areas, the team evaluated the frequency of fires r the potential for

fires, the combustible fire load characteristics and potential fire se rity, the separation

of systems necessary to achieve safe shutdown (SSD), and separation of electrical

components and circuits located within the same fire area ensure that at least one*

SSD path was free of fire damage. The team also inspected the fire protection features

to confirm they were installed in accordance with the co es of record to satisfy the

applicable separation and design requirements of 10 CAR 50, Appendix R,Section III.G,

and Appendix A of BTP APCSB 9.5-1. The team reviewed the following documents,

which established the controls and practices to prevent fires and to control combustible

fire loads and ignition sources, to verify that the objectives established by the

NRC-approved fire protection program (FPP) were satisfied:

lUpdated Final Safety Analysis Report (UFSAR) Section 9.1-A, Fire Protection

Plan

• Administrative Procedure 40AC-ENG-008-OS, Fire Protection Program

• Administrative Procedure 42FP-FPX-01 8-OS, Use, Control, and Storage of

Flammable/Combustible Materials

• Preventive Maintenance Procedure 52PM-MEL-012-0, Low Voltage Switchgear

Preventive Maintenance

The team toured the selected plant fire'areas to observe whether the licensee had

properly evaluated in-iufr lad and limited transient fire hazards in a manner

consistent with the fire prevention and combustible hazards control procedures. In

addition, the team reviewed the licensee's fire safety inspection reports and corrective

action program (CAP) condition reports (CRs) resulting from fire, smoke, sparks, arcing,

and overheating incidents for the years'2000-2002 to assess the effectiveness of the fire

prevention program and to identify any maintenance or material condition problems

related to fire incidents.

The team reviewed fire brigade response, fire brigade qualification training, and drill

program procedures; fire brigade drill critiques; and drill records for the operating shifts

from January 1999 - December 2002. The reviews we're performed to determine

whether fire brigade drills had been conducted in high fire risk plant areas and whether

fire brigade personnel qualifications, drill response, and performance met the

requirements of the licensee's approved FPP.

The team walked down the fire brigade equipment storage areas and dress-out locker

areas in the fire equipment building and the turbine building to assess the condition of

fire fighting and smoke control equipment. Fire brigade personal protective equipment

located at both of the fire brigade dress-out areas and fire fighting equipment storage

area in the turbine building were reviewed to evaluate equipment accessibility and

functionality. Additionally, the team observed whether emergency exit lighting was

provided for personnel evacuation pathways to the outside exits as identified inthe

National Fire Protection Association (NFPA) 101, Life Safety Code, and the

Occupational Safety and Health Administration (OSHA) Part 1910, Occupational Safety

and Health Standards. This review also included examination of whether backup

emergency lighting was provided for access pathways to and within the fire brigade

equipment storage areas and dress-out locker areas in support of fire brigade

operations should power fail during A fire emergency. The fire brigade self-contained

breathing apparatuses (SCBAs) were reviewed for adequacy as well as the availability

of su plemental breathing air tanks and their refill capability.

The team reviewed fire fighting pre-fire plans for the selected areas to determine if

appropriate information was provided to fire brigade members and plant operators to

facilitate suppression of a fire that could impact SSD. Team members also walked down

the selected fire areas to compare the associated pre-feire plans and drawings with as-

built plant conditions. This was done toverify that fire fighting prefire plans and

drawings were consistent with the fire protection features and potential fire conditions

described in the Fire Hazards Analysis (FHA).

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The team reviewed the adequacy of the design, installation, and operation of the manual

suppression standpipe and fire hose system for the control building. This was

accomplished by reviewing the FHA, pre-fire plans and drawings, engineering

mechanical equipment drawings, design flow and pressure calculations and NFPA 14

for hose station 03.~~~

location, ~water

~ ~ flow~requirements

Pot S. and effective reach capability. Team -t2&~pbit

members also walked down the selected fire areas in the control building to ensure that

hose stations were not blocked and to verify that the required fire hose lengths to reach

the safe shutdown equipment in each of the selected areas were available. Additionally,

the team observed placement of the fire hoses and extinguishers to assess consistency

with the fire fighting pre-fire plans and drawings.

b. Findings

No findings of significance were identified.

a. Inspection Scope

10 CFR 50.48, "Fire Protection," and Appendix R to 10 CFR 50, "Fire Protection

Program for Nuclear Power Facilities Operating Prior to January 1, 1979" establish

specific fire protection features required to satisfy General Design Criterion 3, AFire

Protection" (GDC 3, Appendix Atol10CFR 50). Section lll.G of Appendix Rrequires

fire protection features be provided for equipment important to safe shutdown. An

acceptable level of fire protection may be achieved by various combinations of fire

protection features (barriers, fire suppression systems, fire detectors, and spatial

separation of safety trains) delineated in Section lII.G.2. For areas of the plant where

compliance with the technical requirements of Section IlI.G.2 can not be achieved,

licensees must either seek an exemption from the specific requirement~s) or provide an

alternative shutdown capability in accordance with Sections Ill.G.3 and lll1L of the

regulation.

For each selected fire area, the results of the licensee's analysis for compliance with

Section IIlLG of Appendix R is documented in a SSAR . The overall approach of these

evaluations was t9 determine the fire-induced losses for a fire in each fire area and then

assess the plant' pat given those loses.

Ona asaple s an evaluation was performed to verify that systems and equipment

identified in e iensee's SSAR as being required to achieve and maintain hot

shutdown oditions would remain free of fire damage in the event of fire in the selected

fire areas . The evaluation included a review of cable routing data depicting the location

of power and control cables associated with SSD Path 1 and Path 2 components 'of the

RCIC and HPCI systems. Additionally, on a sample basis, the team reviewed the

licensee's analysis of electrical protective device (e.g., circuit breaker, fuse, relay)

coordination.

b. Findings

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Capability of Equipment Credited in SSAR to MitiqaW6 the Sdr5Urius An.ti.='4- -' -

SRVs  ;.

Introduction: The licensee's evaluation of tll 2

2.100 of the SSAR) states that a fire in this aIi -

automatic depressurization system (ADS) atl bl

ADS includes seven of the eleven SRVs. To

use of Core Spray Loop A for a fire in this are ___;

(which involves seven SRVs), the SSAR also' 7

cause all eleven SRVs to spuriously actuate £ J6 ,,*

are located in close proximity in this area. II /, -f

Description: The SSAR states that a fire in Fire Area 2104 could cause all eleven SRVs

to spuriously actuate as a result of fire damage to two cables that are located in close

proximity in this area. The specific circuits that could cause this event have been -

identified by the licensee (circuit nos.: ABE019C08 and ABE019C09). Each of these

two circuits provides a 4 to 20 milliamp instrumentation signal from SRV high-press

actuation transmitters (28211-N1271 and 2B21-N127D) to master trip units 2B21-N

and 2B21-N697D, respectively. The purpose of this circuitry ipto provide an eleical

backup to the mechanical trip capability of the individual SRys.,l the event igh

reactor pressure, the circuits would provide a signal to th ejflunits which would cause

all eleven SRVs to actuate (open). The pressure signal fSr each transmitter is

conveyed to its respective trip unit via a two-conductor, Instrument cable that is routed

through this fire area (two separate cables). Each cable consists of a single twisted pair

of insulated conductors, an uninsulated drain wire that is wound around the twisted pair

of conductors, and a foil shield. In Fire Area 2104 the two cables are located in close

proximity, in the same cable tray. Actuation of the SRV electrical backup is completely

"blind" to the operators. Unlike ADS, it does not provide any pre-actuation indication

(e.g., actuation of the ADS timer) or an inhibit capability (e.g., ADS inhibit switch). Since

the operators typically would not initiate a manual scram until fire damage significantly

interfered with control of the plant, its possible that all eleven SRVs could open at 100%

power, prior to scramming the reactor. This scenario could place the plant in an

unanalyzed condition.

Unlike a typical control circuit, a direct short or "hot short" between conductors of a 4 to

20 milliamp instrument circuit may not be necessary to initiate an undesired (false high)

signal. For cables that transmit low-level instrument signals, any degradation of the

insulation of the individual twisted conductors due to fire damage may be sufficient to

cause leakage currents to be generated between the two conductors. Such leakage

current would appear as a false high pressure signal to the trip units. If both cables

were damaged as a result of fire, false signals gen---'"-' -

in each cable, would actuate the SRV electrical ban wR

eleven of the SRVs to open. The conductor inJat

is cross-linked polyethylene (XLPE). Sin oth ccCo

exposed to the same eating ateh thrOis a reaso 4/

shorting inhtwpKables at aira[mately the san 1X

The licensee's SSAR recognizes the potential safe-

describes methods that have been developed to p/

I>--'

7

its impact on the plant's post-fire safe shutdown capability should it occur. To prevent

this scenario, the licensee has developed procedural guidance which directs operators

to open link BB-10 in panel 2H11-P927 and link BB-10 in panel 2H11-P928. Opening of

these links would prevent actuation of the SRV trip units by removing the 4 to 20

milliamp signal fed by the pressure transmitters. In the event the SRVs were to open

prior to operators completing this action, the SSAR credits Core Spray loop A to mitigate

the event. However, the inspection team had several concerns regarding the

effectiveness of the licensee's approach. Specific concerns identified by the team

included:

1. The timing of operator actions necessary to prevent the event (the time

from fire detection to the time the two links would be opened);

/ 2. Whether the operator actions (opening of links) were consista+ '

plants current fire protection licensing " W 11

needed to achieve and maintain hot st .

/-1 I .*-' u .

V 3. The capability of the limited set of systiA

SSAR for accomplishing post-fire safe x. /

of fire in Fire Area 2104 to mitigate thel

the shutdown performance goals specif,

,) With regard to t timing of rtrt ctions to prevent

SRVs to open, during the inspectic ,t e licensee perfoi

estimated that approximately thi futes would pass_

the time an operator would implement procedural actionr. to prevent its occurrence

(opening of links). The licensee concurred with the insp '*--- ' - en^nrrn that this

time (30 minutes) may be too long to provide an effectiv SiA

actuation. To improve the effectiveness of this action th e F

its existing procedures so that the action would be tak6

A confirmation of fire in areas where the spurious actuatic __ -

ill -\

. .,

I , Analysis: This findin is greater than minor becausei

reliability objectives nd the equipment performance at

0

cornerstone. In o er to achieve safe shutdown conditl

areas chosen fo review, manual control of two SRVs i vv

2121 -F013B a id 2B21-F013F are required to remain mariumy M

for Path 2 s221-FO13D

BS and 2B21-FO13G are required to remain manually

operable. T ese actions are necessary to ensure that the suppression pool temperature

will not exc ed the heat capacity temperature limit (HCT' a-- Vn c-mrnnr=Qqinn nool.

One SRV ( per Path) is opened to manually control dep!

and a half hours after event inedn order to maintai,

HCTL. The second SRV is opened approximately four I

alternate shutdown cooling mode of operation. The ind , In 3

U manually control the listed SRVs because of spurious a

have potential safety significance greater than low safe

s

00

..iVI- kti- -

8

1.,

.-.

- .V& 5

4

Enforcement: 10 CFR 5Appendix RsectionL..e states thatlduringthepostfire'

shutdown, the reactor coolant system process variables shall be maintained within

those predicted for a loss of normal ACpwi ease there Is a -potential for-all SRs

UsiUS~ acuate as a result of tir inFr rea 2104 at a;'--

available, the SSAR credits the use of Core Spray Loop A to, '

coolant makeup function. During the inspection, on 7/24103,l .

performed a simulator exercise of an event which caused all/

this exercise, simulator RPV level instruments indicated that! , 4tu

capable of maintaining level above the top of active fuel. How i 17

/

provide any objective evidence (e.g., specific calculation or X :k7 -

demonstrated that, assuming worst-case fire damage in FirE

of equipment available would be capable of mitigating the e e -

satisfies the shutdown performance goals specified in Appe

1OUFR 50. Pending review and acceptance of objective ey

his capability, this issue is identified as URI 50-:366/03-06Q)2)Capability of Equlpment

creited i-n the`SSAR-to lMitigate -the Spurious Actuation of EBeevKSRVs./

.o< C p erational Implementation of Alternative Shutdown Capabilit\

a. Ins.ectionSconeeA/7 ~

The selected fire areas that were the focus of this inspection all involved reactor

shutdown from the control room. None involved abandoning the control room and

alternative safe shutdown from outside of the control room. However, the licensee's

plans for SSD following a fire in the selected areas involved many local manual operator

actions that would be performed outside of the control area of the control room. This

section of the inspection focused on those local manual operator actions.

The team reviewed the operational implementation of the SSD capability for a fire in the

selected fire areas to determine if: (1)the procedures were consistent with the

Appendix R safe shutdown analysis (SSA); (2)the procedures were written so that the

operator actions could be correctly performed within the times that were necessary for

the actions to be effective; (3)the training program for operators included SSD

capability; (4) personnel required to achieve and maintain the plant in hot standby could

be provide from the normal onsite staff, exclusive of the fire brigade; and (5)the

licensee periodically performed operability testing of the SSD equipment.

The team walked down SSD manual operator actions that were to be performed outside

of the control area of the main control room for a fire in the selected fire areas and

discussed them with operators. These actions were documented in abnormal operating

procedure (AOP) 34AB-X43-001-2, Version 10.8, dated May 28, 2003. The team

evaluated whether the local manual operator actions could reasonably be performed,

using the criteria outlined in NRC Inspection Procedure (IP)71111.05, Enclosure 2. The

team also reviewed applicable operator training lesson plans and job performance

measures (JPMs) and discussed them with operators. In addition, the team reviewed

records of actual operator staffing on selected days.

b. Findings

9

Untimely and Unapproved Manual Operator Action for Fire Safe Shutdown

Introduction: The team found that a local manual operator action to prevent spurious

opening of all eleven SRVs would not be performed in sufficient time to be effective.

Licensee reliance on this manual action for hot shutdown during a fire, instead of

physically protecting cables from fire damage, had not been approved by the NRC.

Description: The team noted that step 9.3.2.1 of AOP 34AB-X43-001-2, Fire Procedure,

Version 10.8, dated May 28,2003, stated: 'To prevent all eleven SRVs from opening

simultaneously, open links 13B-1 0 in Panel 21-1 1-P927 and 13B-1 0 in Panel 2H1 1-P928."

The team noted that spurious opening of all eleven SRVs would be considered a large

loss of coolant accident (LOCA), and that a LOCA must be prevented from occurring

during a fire event. Additionally, the team observed that this step was sufficiently far

back in the procedure that it may not be completed in time to prevent potential fire

damage to cables from causing all eleven SRVs to spuriously open.

The licensee had no preplanned estimate of how Ion it would take operators to

complete this step during a fire event. There was no event time line or operator training

JPM on this step. The team noted that, d i g a revent, operators could be using

many other procedures concurrent with th ire jedure. For example, they could be

using other procedures to communicate w ire brigade about the fire, respond to a

reactor trip, deal with a loss of offsite pow rndprovide emergency classifications and

offsite notifications of the fire event. During the inspection, licensee operators estimated

that, during a fire event, it could take about 30 minutes before operators would

• accomplish step 9.3.2.1. The team concurred with that time estimate. However, NRC

fire models indicated that fires could potentially cause damage to cables in as little as

--- abouit five to ten -minu-te-s. Consequently, the team concluded that during afire event the

licensee's procedures would not ensure that step 9.3.2.1 would be accomplished in time

to prevent potential spurious opening of all eleven SRVs.

The team also identified other issues with step 9.3.2.1. There was no emergency

lighting inside the panels, so th~at if tije fire caused a loss of normal lighting (e.g., by

causing aloss of offsite power), ope~aors would need to use flashlights to perform the

• actions inside the panels. Consequi ntly, the team considered the emergency lighting

for step 9.3.2.1 to be m ade e3 section 1R05.07.b). In addition, labeling of the

links inside the panels wa o ~t at operators stated that they would not fully rely on

the labeling. Also, the tool thafp q-btors would use to loosen and slide the links inside

the energized panels was mad of steel and was not professionally electrically

insulated. Further, licensee reliance on this operator action, instead of physically

protecting the cables as required by 1.0 CFR 50, Appendix R, Section lll.G.2, had not

been approved by the NRC.

The licensee stated that cable damage to two instrument cables, for reactor pressure

signals, would be needed to spuriously open all eleven SRVs. Since the licensee stated

that the two cables were in the same cable tray in fire area 2104, the Unit 2 east

cableway, the team considered that a fire in that area could potentially cause all eleven

SRVs to spuriously open (see section 1R21 .01).

I "t

10

In response to this potential issue, the licensee initiated CR 2003008203 and promptly

revised the Fire Procedure before the end of the inspection, moving the actions of step

9.3.2.1 to the beginning of the procedure. The procedure change enabled the actions to

be accomplished much sooner during a fire in the Unit 2 east cableway or in other fire

areas that were vulnerable to the potential for spuriously opening all eleven SRVs.

Analysis: The team determined that this potential issue is related to associated circuits.

As described in NRC Inspection Procedure (IP) 71111.05, Fire Protection, inspection of

associated circuits is temporarily limited. Consequently, the team did not pursue the

cable routing or circuit analysis that would be necessary to evaluate the possibility, risk,

or potential safety significance of Group "B and C" SRVs spuriously opening due to fire

damage to the instrument cables. The team did, however, perform a circuit analysis of

Group"A" SRVs for which the licensee takes credit for a fire in fire area 2104. (see

section 1R21.01).

Enforcement: 10 CFR 50, Appendix R, Section 1Il.G.2 requires that where cables or

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

cause mal-operation 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 the primary containment, one of the

following means of ensuring that one or the redundant trains is free of fire damage shall

be provided: 1) a fire barrier with a 3-hour rating; 2) separation of cables by a

horizontal distance of more than 20 feet with no intervening combustibles and with fire

detectors and automatic fire suppression; or 3) a fire barrier with a 1-hour rating with fire

detectors and automatic suppression.

The licensee had not provided physical protection against fire damage for the two

instrument cables by one of the prescribed methods. Instead, the licensee had relied on

manual operator actions to prevent the spurious opening of all eleven SRVs. Licensee

personnel contended that fire damage to two cables was outside of the Hatch licAncinn

basis and consequently that there was no requ!

However, the licensee could provide no evident

This potential issue will remain unresolved penrc

associatede of a significes V

identified as URI 50-366/03-06 PO )Untimely aeIr

for Fire Safe Shutdown. cpFs" 0

2. Local Manual Operator Action was Too Difficult and Unsafe

Introduction: A finding of very low safety significance was identified in that a local

manual operator action to operate SSD equipment was too difficult and was also unsafe.

The team judged that some operators would not be able to perform the action. This

finding involved a violation of NRC requirements.

• 11

Description: The team observed that steps 4.15.8.1.1 and 9.3.5.1 of the Fire Procedure

were relied on instead of providing physical protection for cables or providing a

procedure for cold shutdown repairs. Both steps required the same local manual

operator action: "Manually OPEN 2E1 1-F01 5A, Inboard LPCI Injection Valve, as

required." This action was to be taken in the Unit 2 drywell access, which was a locked

high radiation, contaminated, and hot area with temperatures over 100 degrees F.

Valve 2E1 I -F01 5A was a large (24-inch diameter) motor-operated gate valve with a

three-foot diameter handwheel. The main difficulty with manually opening this valve was

lack of an adequate place to stand. An operator showed the team that to perform the

action he would have to climb up to and stand on a small section of pipe lagging (a

curved area about four inches wide by 12 inches long), and then reach back and to his

right side, to hold the handwheel with his right hand, while reaching forward and to his

right to hold the clutch lever for the motor operator with his left hand. He would not have

good balance while performing the action. The foothold, which was large enough to

support only one foot, was well flattened and appeared to have been used in the past to

manually operate this valve. The foothold was ab6ut six to seven feet above a steel

grating, and the team observed that space available for potential use of a ladder to

better access the 2E1 1-F01 5A valve handwheel was not good.

Other difficulties with manually opening the valve included the heat; required wearing of

full anti-contamination clothing, a hardhat, and safety glasses; and inadequate -

emergency lighting (see section 1R05.07). Also, there was no note or step in the

procedure to ensure that the RHR pumps were not running before attempting to '

manually open the 2E1 1-F01 5A valve. If an RHR pump were running, it could create a

differential pressure across the valve which could make manually opening it much more'

difficult.- If the operator-did not have-sufficient agility or strength or stamina, he would be

unable to complete the action. Also, the team judged that inability to remove sweat from

his eyes, due to wearing gloves that could be contaminated, would be a limiting factor

for the operator. In addition, if the operator slipped or lost his balance, he could fall and

become injured. Considering all of the difficulties, the team judged that this action was

unsafe and that some operators would not be able to perform it.

The licensee had no operator training job performance measure (JPM) for performing

this action and could not demonstrate that all operators could perform the action. One

experienced operator, who appeared to be in much better physical condition that an

average nuclear plant operator, stated that he had manually operated the valve in the

past, but that it had been very difficult for him.

The team judged that, since this action was not required to maintain hot shutdown and

was required for cold shutdown following a fire in one of the four selected fire areas,

licensee personnel could have time to improve the working conditions after a fire. They

could have time to install scaffolding or temporary ventilation; improve the lighting; and

assign multiple operators to manually open the valve. They could have time to perform

a 'cold shutdown repair.' However, the licensee had not preplanned any cold shutdown

repairs for opening this valve.

Analysis: This finding is greater than minor because it affected the availability and

reliability objectives and the equipment performance attribute of the mitigating systems

1 .- I

. S: , . , . ,

12

cornerstone. Since the licensee could have time t

shutdown repairs to facilitate accomplishment of t!

potential safety significance greater than very low /'-'? tIL ~ -k- '.

Enforcement: 10 CFR 50, Appendix R, Section Il

features shall be provided for systems important

of limiting fire damage so that systems necessarn

shutdown from either the control room or emergency control stations can uV

within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In addition, TS 5.4.1 requires that written procedures shall be

established, implemented, and maintained covering activities including Fire Protection

Program implementation and including the applicable procedures recommended in

Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33

recommends procedures for combating emergencies including plant fires and

procedures for operation and shutdown of safety-related BWR systems. The Fire

Protection Program includes the SSAR which requires that valve 2E11 -FOl 5A be -

opened for SSD following a fire in fire area 2104, the Unit 2 east cableway. AOP 34AB-

X43-001-2, Fire Procedure, Version 10.8, dated May 28, 2003, implements these

requirements in that it provides information and actions necessary to mitigate the

consequences of fires and to maintain an operable shutdown train following fire damage

to specific fire areas. Also, AOP 34AB-X43-001-2 provides steps 4.15.8.1.1 and 9.3.5.1

for manually opening valve 2E1 1-F01 5A following a'

&'"4ee

Contrary to the above, the licensee 1admontprovii

damage, for electrical operation valve 2E1 1-F01 5A 1

--e

procedure for repairing any related fire damage with

74" '

relied on local manual operator actions, as describe( f

Add Ago .,

AOP 34AB-X43-001-2. However, those procedure /

operators would not be able to perform them becaus

difficult and also were unsafe. In response to this is C AL

203008202. Because the identified inadequate opel '

I

significance and the issue has been entered into the

program, this violation is being treated as an NCV consistent with Section VI.A.1 of the

NRC's Enforcement Policy: NCV 50-366/03-0tfocal Manual Operator Action for

Post-Fire Safe Shutdown Equipment was Too fic t andU

Unapproved Manual Operator Actions for Post-Fire Saf hutdown

Introduction: A finding of very low safety significance was identified in that the licensee

relied on some manual operator actions to operate SSD equipment, instead of providing

the required physical protection of cables from fire damage, and without NRC approval.

This finding involved a violation of NRC requirements.

Description: The team observed that AOP 34AB-X43-001-2, Fire Procedure, included

some local manual operator actions to achieve and maintain hot shutdown that had not

been approved by the NRC. Examples included:

)/ S)p 4.15.2.2; ...if a loss of offsite power occurs and emergency busses energize

I. Place Station Service battery chargers 2R42-S026 (2R42-S029), 2R42-S027

(2R42-S030) AND 2R42-S028 (2R42-S031) in service per 34SO-R42-001-2."

13

• >Step 4.15.4.5; ...lf HPCI fails to automatically trip on high RPV level... "OPEN the

following links to energize 2E41 -F1 24, Trip Solenoid Valve, AND to fail 2E41-

F3025 HPCI Governor Valve, in the CLOSED position:

• TT-75 in panel 2H11-P601

• TT-76 in panel 2H1 1-P601"

/* Up 4.15.4.6; ...lf HPCI fails to automatically trip on high RPV level.... OPEN

breaker 25 in panel 2R25-S002 to fail 2E41 - - ,

CLOSED position." r Xv

accomplished by operators in a timely manner. In

these operator actions were being used instead of I

damage that could cause a loss of station service b

runaway. /

Analysis: The finding is greater than minor because

reliability objectives and the equipment performance

cornerstone. Since the actions could reasonably be accomplished by operators in a

timely manner, this finding did not have potential safety significance greater than very

low safety significance.

Enforcement: 10 CFR 50, Appendix R, Section lll.G.2 requires 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 the primary containment,-one of the

following means of ensuring that one or the redundant trains is free of fire damage shall

be provided: 1) a fire barrier with a 3-hour rating; 2) separation of cables by a

horizontal distance of more than 20 feet with no intervening combustibles and with fire

detectors and automatic fire suppression; or 3) a fire barrier with a 1-hour rating with

fire detectors and automatic suppression.

Contrary to the above, the licensee had not provided the required physical protection

against fire damage for power to the station service battery chargers or for HPCI

electrical control cables. Instead, the licensee relied on local manual operator actions,

without NRC approval. In response to this issue, the licensee initiated CR2003800166

dated 7/25/2003. Because the issue had very low safety significance arid has been

entered into the licensee's corrective action program, this violation is being treated as an

NCV consient wi~At 4$ectron VI.A.1 of the NRC's Enforcement Policy: NCV 50-366/03-

06 U d Manual Operator Actions for Post-Fire Safe Shutdown.

.06 Comm tions

a. Inspection Scope

The team reviewed the plant communications systems that would be relied upon to

support fire brigade and safe shutdown activities. The team walked down portions of

the safe shutdown procedures to verify that adequate communications equipment would

14

be available for personnel performing local manual operator actions. Inaddition, the

team reviewed the adequacy of the radio communication system used by the fire

brigade to communicate with the main control room.

b. - Findings

No findings of significance were identified.

.07 Emergency Lighting

a. Inspection Scope

The team inspected the licensee's emergency lighting systems to verify that 8-hour

emergency lighting coverage was provided as required by 10 CFR 50, Appendix R,

Section III.J., to support local manual operator actions that were needed for post-fire

operation of SSD equipment. During walkdowns of the post-fire SSD operator actions

for fires in the selected fire areas, the team checked if emergency lighting units were

installed and if lamp heads were aimed to adequately illuminate the SSD equipment, the

equipment identification tags, and the access and egress routes thereto, so that

operators would be able to perform the actions without needing to use flashlights.

b. Findings

Inadequate Emergency Lighting for Operation of Safe Shutdown Eguipment

Introduction: A finding with very low safety significance was identified in that emergency

lighting was not adequate for some manual operator actions that were needed to

support post-fire operation of SSD equipment. This finding involved a violation of NRC

requirements.

Description: The team observed that emergency lighting was not adequate for some

manual operator actions that were needed to support post-fire operation of SSD

equipment. Examples included the following operator actions in procedure 34AB-X43-

001 -2, Fire Procedure, Version 10.8, dated May 28, 2003:

Step 4.15.2.2; ...if a loss of offsite power occurs and emergency busses energize

"...PlaceStation Service battery chargers 2R42-S026 (2R42-S029), 2R42-S027

(2R42-S030) AND 2R42-S028 (2R42-S031) in service per 34SO-R42-001-2."

• S 4p 4.15.4.5; ...IfHPCI fails to automatically trip on high RPV level... UOPEN the

lowing links to energize 2E41-F124, Trip Solenoid Valve, AND to fail 2E41-

3025 HPCI Governor Valve, in the CLOSED position:

• TT-75 in panel 2H11-P601

• TT-76 in panel 2H11-P601"

• Step 4.15.5; "IF 2R25-S065, Instrument Bus 2B, is DE-ENERGIZED perform the

following manual actions to maintain 2C32-R655, Reactor Water Level

Instrument, operable:

• 4.15.5.1; At panel 2H1 1-P612, OPEN links AAA-1 1 and AAA-1 2.

15

• 4.15.5.2; At panel 2H11-P601, CLOSE links HH-48 and HH-49. -

• Steps 4.15.8.1.1 and 9.3.5.1; "Manually OPEN 2E11-F015A, Inboard LPCI

Injection Valve, as required."

• Steps 4.15.8.1.2 and 9.3.5.2; 'Manually CLOSE 2E11 -F018A, RHR Pump A

Minimum Flow Isolation Valve, as required."

• Step 9.3.2.1; 'To prevent all 11 SRVs from opening simultaneously, open links

BB-10in Panel 2H11-P927 and BB-10 in Panel 2H11-P928."

• Step 9.3.3; UAt Panel 2H11-P627, open links M-19, AA-20, AA-21, and AA-22,

to prevent spurious actuation of SRVs 2B21-F013D AND 2B21-F013G."

• Step 9.3.6; "OPEN link TB9-21 in Panel 2H11-P700 to open Drywell Pneumatic

System Inboard Inlet Isolation, 2P70-F005."

• Step 9.3.7; "OPEN link TB1-12 in Panel 2H11-P700 to open Drywell Pneumatic

System Outboard Inlet Isolation, 2P70-F005."

• Step 9.3.9.1; "Confirm OR manually CLOSE RHR Shutdown Cooling Valve

2E11-FO06D."

• Step 9.3.9.2; "Manually OPEN Shutdown Cooling Suction Valve 2E11-F008, IF

required..."

The team verified that flashlights were readily available arid judged that operators would

be able to use the flashlights to accomplish the actions, with two exceptions. One

exception was the action to open terminal board links in two panels to prevent all eleven

SRVs from spuriously opening, which was judged to be untimely (see section

1R05.05.b.1). The other exception was the action to open 2E1 -FO15A, which was

judged to be too difficult (see section 1R05.05.b.2). For all of these actions, the lack of

adequate emergency lighting could make the actions more difficult to complete in a

timely manner and increase the chance of operator error.

Analysis: This finding is greater than minor because it affected the reliability objective

and the equipment performance attribute of the mitigating systems cornerstone. Since

operators would be able to accomplish the actions with the use of flashlights, this finding

did not have potential safety significance greater than very low safety significance.

Enforcement: 10 CFR 50, Appendix R, Section III.J. requires that emergency lighting

units with at least an 8-hour battery power supply shall be provided in all areas needed

for operation of safe shutdown equipment and in access and egress routes thereto.

Contrary to the above, emergency lighting units were not adequately provided in all

areas needed for operation of safe shutdown equipment. In response this issue, the

licensee initiated CRs 2003008237 and 2003008179. Because the identified lack of

emergency lighting is of very low safety significance and has been entered into the

licensee's corrective action program, this violation is being treated as an NCV,

-

.

16

consistent with Section VI.A.1 of the NRC's Enforcement Policy: NCV 50-366/03-0

Inadequate Emergency Lighting for Operation of Safe Shutdown Equipment.

.08 Cold Shutdown Repairs

The licensee had identified no needed cold shutdown repairs. Also, with the exception

of the potential need for a cold shutdown repair to open valve 2E1 1-FO1SA (see section

1R05.05.b.2), the team identified no other need for cold shutdown repairs.

Consequently, this section of IP 71111.05 was not performed.

.09 re Barriers and Fire Area/Zone/Room Penetration Seals

a.

The team reviewed the selected fire areas to evaluate the adequacy of the fire

resistance of fire area barrier enclosure walls, ceilings, floors, fire barrier mechanical

and electrical penetration seals, fire doors, and fire dampers. The team selected

several fire barrier features for detailed evaluation and inspection to verify proper

installation and qualification. This was accomplished by observing the material condition

and configuration of the installed fire barrier features, as well as construction details and

supporting fire endurance tests for the installed fire barrier features, to verify the as-built

configurations were qualified by appropriate fire endurance tests. The team also

reviewed the FHA to verify the fire loading used by the licensee to determine the fire

resistance rating of the fire barrier enclosures. The team also reviewed the installation

instructions for sliding fire doors, the design details for mechanical and electrical

penetrations, the penetration seal database, Generic Letter (GL) 86-10 evaluations, and

the fire protection penetration seal deviation analysis for the technical basis of fire

barrier penetration seals to verify that the fire barrier installations met design

requirements and license commitments. In addition, the team revie ed completed

surveillance and maintenance procedures for selected fire f atures to verify the

fire barriers were being adequately maintained.

The team evaluated the adequacy of the fire resta e of fi rrier electrical raceway

fire barrier system (ERFBS) enclosures for ca e pr tection to satisfy the applicable

separation and design requirements of 10 50, Appendix R,Section III.G.2.

Specifically, the team examined the design drawings, construction details, installation

records, and supporting fire endurance tests for the ERFBS enclosures installed in fire

area 2104, the Unit 2 East Cableway. Visual inspections of the enclosures were

performed to confirm that the ERFBS installations were consistent with the design

drawings and tested configurations.

The team reviewed abnormal operating fire procedures, selected fire fighting pre-plans,

fire damper location and detail drawings, and heating ventilation and air conditioning

(HVAC) system drawings to verify that access to shutdown equipment and selected

operator manual actions would not be inhibited by smoke migration from one area to

adjacent plant areas used to accomplish SSD.

b. Findings

- I ; , ..

,r:. ; .1 I

, ;.'I . 1. -I - , -,

17

No findings of significance were identified.

m--

.10 Fire Protection Systems. Features, and Equipment

Inspection Scope

The team reviewed flow diagrams, cable routing information, and operational valve

lineup procedures associated with the fire pumps and fire protection water supply

system. The review evaluated whether the common fire protection water delivery and

supply components could be damaged or inhibited by fire-induced failures of electrical

power supplies or control circuits. Using operating and test procedures, the team toured

the fire pump house and diesel driven fire pump fuel storage tanks to observe the

system material condition, consistency of as-built configurations with engineering

drawings, and determine correct system controls and valve lineups. Additionally, the

team reviewed periodic test procedures for the fire pumps to assess whether the

surveillance test program was sufficient to verify proper operation of the fire protection

water supply system in accordance with the program operating requirements specified

in Appendix B of theFHA.

The team reviewed the adequacy of the fire detection systems in the selected plant fire

areas in accordance with the design requirements in Appendix R, III.G.1 and III.G. 2.

The team walked down accessible portions of the fire detection systems in the selected

fire areas to evaluate the engineering design and operation of the installed

configurations. The team also reviewed engineering drawings for fire detector types,

spacing, locations and the licensee's technical evaluation-of the detector locations for

the detection systems for consistency with the li6ensee's FHA, engineering evaluations

for NFPA code deviations, and NFPA 72E. Inaddition, the team reviewed surveillance

procedures and the detection system operating requirements specified in Appendix B of

the FHA to determine the adequacy of fire detection component testing and to ensure

that the detection systems could function when needed.

The team performed in-plant walk-downs of the Unit 2 East Cableway automatic wet

pipe sprinkler suppression system to verify the proper type, placement and spacing of

the sprinkler heads as well as the lack of obstructions for effective functioning. The

team examined vendor information, engineering evaluations for NFPA code deviations,

and design calculations to verify that the required suppression system water density for

the protected area was available. Additionally, the team reviewed the physical

configuration of electrical raceways and safe shutdown components in the fire area to

determine whether water from a pipe rupture, actuation of the automatic suppression

system, or manual fire suppression activities in this area could cause damage that could

inhibit the plant's ability to safely shutdown.

The team reviewed the adequacy of the design and installation of the manual C02 hose

reel suppression system for the diesel generator building switchgear rooms 2E and 2F

(fire areas 2404 and 2408). The team performed in-plant walk-downs of the diesel

generator building C02 fire suppression system to determine correct system controls

and valve lineups to assure accessibility and functionality of the system, as well as

18

associated ventilation system fire dampers. The team also reviewed the licensee's

actions to address the potential for C02 migration to ensure that fire suppression and

post-fire safe shutdown actions would not be impacted. This was accomplished by the

review of engineering drawings, schematics, flow diagrams, and evaluations associated

with the diesel generator building floor drain system to determine whether systems and

Operator actions required for SSD would be inhibited by C02 migration through the floor

drain system.

b. Findings

No findings of significance were identified.

.11 Compensatorv Measures

a. Inspection Scope

The team reviewed Appendix B of the FHA and applicable sections of the fire protection

program administrative procedure regarding administrative controls to identify the need

for and to implement compensatory measures for out-of-service, degraded, or

inoperable fire protection or post-fire safe shutdown equipment, features, and Warns.

The team reviewed licensee reports for the fire protection status of Unit 1, U 2a of

shared structures, systems, and components. The review was performed to ihat

the risk associated with removing fire protection and/or post-fire systems or

components, was properly assessed and implemented in accordance with the approved

fire protection program. The team also reviewed Corrective Action Program Condition

Reports generated over the last 18 months for fire protection features that were out of

servie for long periods of time. The review was conducted to assess the licensee's

effl tiveness in returning equipment to service in a reasonable period of time.

@ dinas

No findings of significance were identified.

1R21 SAFETY SYSTEM DESIGN AND PERFORMANCE CAPABILITY

01. DCR 91-134, SRV BackuD Actuation via Pressure Transmitter Sianals

(.\ Inspection ScoDe

The team performed an inde pendent design review of plant modification DCR 91-134 in

( order to evaluate the techrial adequacy of the design change package and its

associated 10 CFR 50.59lrvaluation. The scope of the review and circuit analysis

performed by the team wM limited to the group 'A' SRVs -or which the licensee takes

credit in mitigating a fire irj the fire areas selected for t inspection.

Findinat i

Ina 7e uatelent:Mo ic iJon Re ults i o se Falu ~of Sa~fe

t,.

19

14

oduction / 1A

Design Change Request (DCR)91-134 was impl /

raised in General Electric Report NEDC-3200P, A

January-February 1991 Turbine Trip Events for PlI

ensure that individual SRV(s) will actuate at or neat I./

allowable limits, a backup mode of operation was ir'

design was intended to mitigate the effects of corrc

k _ 5.'z

arget Rock SRVs.

Automatically controlled two stage SRVs are installed on the main steam lines inside

containment for the purpose of relieving nuclear boiler pressure either by normal

mechanical action or by automatic action of an electro-pneumatic control system. Each

SRV can be manually controlled by use of a two position switch located in the main

control room. When placed in the uOpen' position the sw'4 ' 4 -. .t."k

4k ^f

the individual SRV and causes it to go open. When the X

position the SRV is opened upon receipt of either an Autc

(ADS), or Low-Low Set (LLS) control logic signal. Either A

the valve. DCR 91-134 provided a backup mode for initie

valve sole old, which was independent of ADS or LLS loi

no opprabr action to initiate opening of the SRVs and wE

77X0,al_4

A- 41/1110/117

- . -loa he operators. I> .*

The pe of the plant modification involved the installati

transmitters ( Model No. 1154GP9RJ), 0-3000 psig, in th

instrument racks at EL. 158 of the Reactor Building. Ea6ai, ,,ooui,

part of v 4-20 ma current loop and provided the analog trip signal for SRV actuation

within the following set point groups.

SRV Identification Taps SRV Set Point

A 2B21-FO13B, D, F, and G 1120 psig

B 2B21-F013A, C, K, and M 1130 psig

C 2B21-F013E, H, and D 1140 psig

Pressure transmitters 2B21-N127A and 2B21-N1P1 ^71 '*-';- to ATTQ nnhinate

2H11-P927. Pressure transmitter 2B21-N127A ir 6

of a trip unit master relay K308C and trip unit sla) ' ,/u:/c. C

loop components for pressure transmitter 2B21-h '

relay K335C in addition to trip unit slave relays K'

instrument loops constituted a "Division" pressurc / 1/

intended to provide the one-out of two logic signsWX

backup actuation. The design objective of having'<

assure compliance with the single failure criterior

20

Additionally, pressure transmitters 2B21-N127B and 2B21-N127D were wired to ATTS'

cabinet 2H11-P928. Pressure transmitter 2B21-N127B instrument loop components

consited of a trip unit master relay K31OD and trip unit slave relays KK312D and

K33 D. The loop components for pressure transmitter 2B21-N127D consisted of a trip

un/ master relay K335D in addition to trip unit slave relays K336D and K363D. These.

yo instrument loops constituted a separate "Division" pressure monitoring channels and

,4vas intended to provide the one-out of two logic signal from this Division for initiating

SRV backup actuation. The design objective of having two instrument channels was to,

assure compliance with the single failure criterion of 10 CFR 50 Appendix A.

The Group "A"SRVs were provided logic input signals from the trip unit master relays.

The Group UB and C"SRVs were provided logic input signals from the trip unit slave

relays. The total of 12 relays described above, (6 in ATTS cabinet 2H1 1-P927 and 6 in

ATTS cabinet 2H1 1-P928), were intended to be wired to provide "one-out-of-two taken

twice logic" for actuation of the SRVs. The design objective was to assure that a single

relay failure in either Division would not cause an inadvertent SRV actuation.

Coincident logic input is required from both Division instrument loops in order to initiat-

a SRV backup actuation via the pressure tran*

Analysis: The licensee in their SSAR takes c-

in order to achieve and maintain safe shutdow

SRVs are required for a fire in the fire areas seY

The team performed a circuit analysis of SRVi

F01 3G (Path 2) in order to verify that the desig

of-two taken twice logic had been achieved. B

- that the design objective of implementing a on(

been installed for the SRVs. The logic installet

opincident taken twice logic in addition to a one-oui-ot-two coincident taken twice logic.A -/

I he1tam also determined that the two-out-of-two coincident logic inrom trp unit

master relays K31 OD and K335D represented a common cause failure for both SRVs'ef

a fire in fire area 2104. Specifically, cable ABEOl9C08 associated with pressure

transmitter 2B21-N127B current loop, and cable ABE019CO9 associated with pressure

transmitter 2B21-N127D current loop, are both routed in the same cable tray in'fire area

2104. Both shielded twisted pair instrument cables are unprotected from the effects of a

'fire in this fire area. Fire induced insulation damage to both cables could result in

leakage currents which causes the instrument loops to fail high. This failure mode

simulates a high nuclear boiler pressure condition which would initiate SRV backuD

>riousactuation of both SRVs for a

-lire in fire area 2104 defeats the capability to manually control these SRVs as is required

per the SSAR.

Enforcement 10 CFR 50, Appendix B, -; -

measures shall provide for verifying or ch C . 11 ted

industrystandard, ANSI N45.2.11 -1974

for relating the final design back to the so _

The logic implemented by the licensee for. d

design input requirements. The plant instE -

.Ai

21

out-of-two taken twice logic that was specified for the SRV backup actuation via

pressure transmitter signals design change package. This failure has created a

condition where fire induced failures of two instrument circuit cables, (within close

proximity to each other), could result in spurious actuation of all eleven SRVs based on

the logic input from trip master unit relays K31 OD, and K335D and their associated trip

unit slave relays. The 10 CFR 50.59 Evaluation performed for the plant modification

failed to identify this failure mode. Additionally, the 10 CFR 50.59 Evaluation was

inadequate in that it did not provide an adequate technical basis that an Unreviewed

Safety Question (USQ) had not been created by implementation of the plant

modificatiorLending additional review by the NRC, this item is identified as URI 50-

366/03-06 U7Imleentation of DCR 91-1 34 Results in Spurious Actuation of Eleven

SRVs; bec ofpFire Induced Faults.

This inspection ding may be a"Potentially Generic Issue" by having implications for

other licensees whave implemented a plant modification similar to DCR 91-1 34 for a

BWR having a Mar ontainment.

4. OTHER ACTIVITIES

40A2 Identification and Resoluti p rblems.

a. Inspection Scope

The team reviewed a sample of licensee audits, self-assessments, and condition reports

(CRs) to verify that items related to fire protection and to SSD were appropriately

entered into the licensee's CAP in accordance with the Hatch quality assurance program

and procedural requirements. The items selected were reviewed for classification and

appropriateness of the corrective actions taken or initredtoriesolve ithisse

addition, the team reviewed the licensee's applicability evaluations and corrective

actions for selected industry experience issues related to fire protection. The operating

experience (QE) reports were reviewed to verify that the licensee's review and actions

were appropriate.

The team reviewed licensee audits and self-assessments of fire protection and safe

shutdown to assess the types of findings that were generated and to verify that the

findings were appropriately entered into the licensee's corrective action program.

b. Findings

No findings of significance were identified.

i.(A6 Meetinas.lIncluding Exit

e team presented the inspection results to Mr. R. Dedrickson, Assistant

neral Manager, and other members of your staff at the conclusion of the

oon July 25, 2003 The licensee acknowledged the findings presented.

oprietary information is not included in the inspection report.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel:

M. Beard Acting Engineering Support Supervisor

V. Coleman Quality Assurance Supervisor

M. Dean Nuclear Specialist, Fire Protection

B. Duval Chemistry Superintendent

R. Dedrickson Assistant General Manager for Plant hatch

M. Googe Maintenance Manager

J. Hammonds Operations Manager

D. Javorka Administrative Assistant, Senior

R. King Acting Engineering Support Manager

1.Luker Senior Engineer, Licensing

T. Metzer Acting Nuclear safety and Compliance Manager

A. Owens Senior Engineer, Fire Protection

J. Payne Senior Engineer, Corrective Action Program

D. Parker Senior Engineer, Electrical

J. Rathod Bechtel Engineering Group Supervisor

K. Rosanski Oglethorpe Power Corporation Resident Manager

M. Raybon Summer Intern

J. Vance Senior Engineer, Mechanical & Civil

R. Varnadore Outages and Modifications Manager

NRC personnel:

N. Garret, Senior Resident Inspector

C. Payne Fire Protection Team Leader

Attachment

. 4

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

50-366/03-066 URI Capability of Equipment Credited iphe SSAR to Mitigate the

Spurious Actuation of Eleven SRV, (Section 1R05.03.b)

50-366/03-06 URI Untimely and Una roved Manual Operator Action for Post-Fire

Safe Shutdow (Section 1R05.05.b.1)

0 k

50-366/03-06,-V URI Implementation of DCR 91-134 Results in Spuyi'ous Actuation of

Eleven SRVs because of Fire Induced Fault (Section 1R21.01.b)

Opened and Closed

50-366/03-06> NCV Local Manual Operator Action for Post-Fire Safe Shutdown

Equipment was Too Difficult and Unsafl( (Section 1R05.05.b.2)

50-366/03-06@ NCV Unapproved Manual Operator Actions for Post-Fire Safe

Shutdow ection 1R05.05.b.3)

50-366/03-060 NCV Inadeq a e Emergency~lghting for Operation of Post-Fire Safe

Shutdown Equipmen (Section 1R05.07.b)

Discussed

None

Attachment

LIST OF DOCUMENTS REVIEWED

Procedures

Administrative Procedure 40AC-ENG-008-OS, Fire Protection Program, Rev. 9.2

Administrative Procedure 42FP-FPX-018-OS, Use, Control, and Storage of

Flammable/Combustible Materials, Rev. 1.0

Department Instruction DI-FPX-02-0693N, Fire Fighting Equipment Inspection, Rev. 5

Fire Protection Procedure 42FP-FPX-005-OS, Drill Planning, Critiques and Drill

Documentation Rev. 1 ED1

Fire Protection Procedure 42FP-FPX-007-OS, Hot Work, Rev. 1.2

Preventive Maintenance Procedure 52PM-MEL-01 2-0, Low Voltage Switchgear Preventive

Maintenance, Rev. 25.0

Preventive Maintenance Procedure 52PM-MEL-01 4-0, Transformer Maintenance, Rev. 10.1

Surveillance Procedure 42SV-FPX-002-OS, Low Pressure CO2 System Surveillance, Rev. 7.1

Surveillance Procedure 42SV-FPX-004-OS, Fire Pump Test, Rev. 8.6

Surveillance Procedure 42SV-FPX-006-OS, Fire Damper Surveillance, Rev. 1 ED 1

Surveillance Procedure 42SV-FPX-021-OS, Surveillance of Swinging Fire Doors, Rev. 1.6

Surveillance Procedure 42S V-FPX-024-OS, Fire Hose Stations 31 Day Surveillance, Rev. 1

Surveillance Procedure 42SV-FPX-030-OS, Fire Emergency Self Contained Breathing

Apparatus Inspection and Test, Rev. 1

Surveillance Procedure 42SV-FPX-032-OS, Automatic Sliding Fire Door Visual Inspection, Rev.

.3.3

Surveillance Procedure 42SV-FPX-036-OS, Annual Fire Pump Capacity Test, Rev. 8.6

Surveillance Procedure 42SV-FPX-037-OS, Fire Detection Instrumentation Surveillance, Rev.

5.1

System Operating Procedure 34SO-X43-001-1, Fire Pumps Operating Procedure, Rev. 4.3

Training Procedure 73TR-TRN-003-OS, Fire Training Program, Rev.4

AOP 34AB-C1I-001-2, Loss of CRD System, Version 2.3

AOP 34AB-C71-001-2, Scram Procedure, Version 9.9

AOP 34AB-C71-002-2, Loss of RPS, Version 4.3

AOP 34AB-N61-002-2S, Main Condenser Vacuum Low, Version 0.4

AOP 34AB-P41-001-2, Loss of Plant Service Water, Version 8.1

AOP 34AB-P42-001-2S, Loss of Reactor Building Closed Cooling Water, Version 1.4

AOP 34AB-P51-001-2, Loss of Instrument and Service Air System or Water Intrusion into the

Service Air System, Version 3.0

AOP 34AB-R22-001-2, Loss of DC Busses, Version 2.4

AOP 34AB-R22-002-2, Loss of 4160V Emergency Bus, Version 1.4

AOP 34AB-R22-003-2, Station Blackout, Version 2.3

AOP 34AB-R22-004-02, Loss of 4160V Bus 2A, 2B, 2C, or 2D, Version 1.3

AOP 34AB-R23-001-2S, Loss of 600V Emergency Bus, Version 0.4

AOP 34AB-R24-001-2, Loss of Essential AC Distribution Buses, Version 1.3

AOP 34AB-R25-002-02, Loss of Instrument Buses, Version 5.4

AOP 34AB-T47-001-2, Complete Loss of Drywell Cooling, Version 1.8

AOP 34AB-X43-001-2, Fire Procedure, Version 10.8

AOP 34AB-X43-002-0, Fire Protection System Failures, Version 1.3

SOP 34SO-C71-001-2, 120VAC RPS Supply System, Version 10.2

SOP 34SO-N40-001-2, Main Generator Operation, Version 10.8

Attachment

- I t -

i: .'.. 'a.

2

SOP 34SO-R42-001-2S, 125V DC and 125/250 VDC System, Version 7.1

SOP 34SO-S22-001-2, 500 KV Substation Switching, Version 5.2

31 EO-EOP-010-2S, RC RPV Control (Non-ATWS), Rev. 8, Attachment 1 . . .

31 EO-EOP-012-2S, PC-1 Primary Containment Control, Rev. 4, Attachment 1 . . .

.... .

31 EO-EOP-013-2S, PC-2 Primary Containment Control, Rev. 4, Attachment 1 - . .  :

31 EO-EOP-014-2S, SC - Secondary Containment Control, Rev. 6, Attachment 1 ..

31 EO-EOP-01 6-2S, CP-2 RPV Flooding, Rev. 8, Attachment 1

.. .-  :.-.

Procedure 34AB-X43-001-2S, Rev.1 OED3, "Fire Procedure," dated 5/28/03.

Calibration Procedure 57CP-CAL-097-2, Rosemount 1153 and 1154 transmitters, Revision No

19.9.

Drawings

H-1 1814, Fire Hazards Analysis, Control Bldg. El. 130'-0", Rev. 5

H-11821, Fire Hazards Analysis, Turbine Bldg. El. 130'-0", Rev. 0

H-1 1846, Fire Hazards Analysis, Diesel Generator Bldg., Rev. 2

H-26014, R.H.R. System P&ID Sheet 1, Rev. 49

H-26015, R.H.R. System P&ID Sheet 2, Rev. 46

H-26018, Core Spray System P&ID, Rev. 29

B-1 0-1326, Rectangular Fire Damper Schedule, Rev. 2

B-1 0-1 329, Rectangular Fire Damper, Rev. 1

H-1 1033, Fire Protection Pump House Layout, Rev. 47

H-11035, Fire Protection Piping and Instrumentation Diagram, Rev. 22

H-11226, Piping-Diesel Generator Building Drainage, Rev. 6

H-11814, Fire Hazards Analysis Drawing, Control Building, Rev. 5

H-1 1821, Fire Hazards Analysis Drawing, Turbine Building, Rev. 11

H-1 1846, Fire Hazards Analysis Drawing, Diesel Generator Building, Rev. 2

H-11894, Fire Detection Equipment Layout-Diesel Generator Building, Rev. 2

H-11915, Fire Detection Equipment Layout-Control Building, Rev. 2

H-13008, Conduit and Grounding, Fire Pump House, Rev. 9

H-13615, Wiring Diagram, Fire Pump House, Rev. 13

H-1 6054, Control Building HVAC System, Rev. 19

H-41509, Diesel Generator Building CO2 System-P&ID, Rev. 5

H-43757, Penetration Seals-Type, Number, and as-Built Location, Rev. 3

Calculations. Analyses, and Evaluations

E. l. Hatch Nuclear Plant Units 1 and 2 Safe Shutdown Analysis Report, Rev. 20.

Edwin I. Hatch Nuclear Plant Fire Hazards Analysis and Fire Protection Program, Rev. 20

Calculation SMFP88-001, Hydraulic Analysis of Sprinkler Systems in Control Building East

Cableway, dated 03/11/1988

Calculation SMNH94-046, FCF-F1 OB-006, Fire Resistance of Concrete Block at HNP, dated

09/30/1994

Calculation SMNH94-048, FCF-F1 OB-006, Cable Tray Combustible Loading Calculation, dated

09/30/1994

Calculation SMNH98-023, HT-98617, Fire Protection Penetration Seal Deviation Analysis,

dated 10/28/1998

Calculation SMNHOO-01 1, HT-00606, Hose Nozzle Pressure Drop Analysis, dated 09/08/2000

Evaluation HT-91722, Fire Protection Code Deviation Resolution, dated 04/22/1992

Attachment

3

Hatch Response to NRC IN 1999-005, dated 05/04/1 999

Hatch Response to NRC IN 2002-024, dated 09/20/2002

Calculation SENH 98-003, Rev. 0, plot K, protective relay settings 4kV bus 2E

Calculation 85082MP, Plot 29, 600V Switchgear 2C

Calculation SENH 94-004, Attachment A, Sheets 7&8, 600/208 Reactor Building MCC 2C

Calculation SENH 91-01 1, Attachment P, Sheet 6, Reactor Building DC MCC 2A

Calculation SENH 94-01 3, Sheets 28 and 29, 600V Reactor Building MCC 2E-B3

Calculation SENH 91 -01 1, Attachment P, Sheet 16, Reactor Building 25OVDC MCC 2B3

Audits and Self-Assessments

Audit No. 01-FP-1, Audit of the Fire Protection Program, dated April 12, 2001

Audit No. 02-FP-1, Audit of the Fire Protection Program, dated February 28, 2002

Audit No' 03-FP-1, Audit of Fire Protection, dated April 21, 2003

1999-001106, Lighting in Fire Equipment Building

2002-000629, Inordinate Number of Buried Piping Leaks

2002-002127, Inadequate Bunker Gear

2002-002129, Health Physics Support and Participation for Fire Brigade

2003-000735, Impact on Cold Weather on Operating Units

Audit Report 01 -FP-1, Audit of Fire Protection Program, dated 04/12/2001

Audit Report 02-FP-1, Audit of Fire Protection Program, dated 02/28/2002

Audit Report 03-FP-1, Audit of Fire Protection Program, dated 04/21/2003

CRs Reviewed

CR 2000007119, Fire Procedure 34AB-X43-001 -1pS Needs to be Enhanced

. CR 2001002032, Fire Procedure 34AB-X43-001-2S Needs Actions for Diesel Fuel Oil Pumps

CR 2003004377, Fire Procedure 34AB-X43-001-1 Enhancements

CR 2003004379, Fire Procedure 34AB-X43-001P-2 Enhancements

CR 2003004382, SSAR Discrepancies

CRs Generated During this Inspection

CR 2003007129, No Fire Procedure Actions for a Fire in the 2C Switchgear Room

CR 2003007719, Use of Link Wrench

CR 2003007978, Fire Damper Corrective Action

CR 2003008141, Breaker Maintenance Handle

CR 2003008165, SSAR Section 2.100

CR 2003008179, Drywell Access Emergency Lights

CR 2003008181, Link Labeling

CR 2003008202, Manually Opening MOV 2E1 1-F0l 5A

CR 2003008203, SRV Manual Action Steps in Fire Procedure

CR 2003008237, Emergency Lights and Component Labeling for Manual Actions

CR 2003008238, c02 Migration Through Floor Drains

CR 2003800132, SSAR Error for Position of 2E1 1P-FO04A

CR 2003800151, Instruments for Manual Actions

CR 2003800152, Sliding Links in SSAR

Attachment

4

CR 2003800153, Promat Test Report

CR 2003008250, Communications for Post-Fire SSD

CR 2003800166, Review Fire Procedure Step 34AB-X43-001 -2 Steps to Verify Compliance

with Appendix R.

Design Criteria and Standards

Design Philosophy for Fire Detectors at E. I. Hatch Nuclear Plants, Rev. 2

Completed Surveillance Procedures and Test Records

42SV-FPX-021 -OS, Surveillance of Swinging Fire Doors, Task # 1-3367-1 (completed on

01/09/2003)

42SV-FPX-024-OS, Fire Hose Stations, Task # 1-3359-1 (completed on 06/27/2003)

42SV-FPX-030-OS, Fire Emergency Self Contained Breathing Apparatus Inspection and Test,

Task # 1-4200-3 (completed on 07/07/2003)

42SV-FPX-032-OS, Automatic Sliding Fire Door Surveillance, Task # 1-3361-2 (completed on

08/13/2002

Promatec Technologies Installation Inspection Report for Fire Area 2104, MWO 2-98-00881,

Record 09367-2289, dated 09/03/1998

Technical ManualsNendor Information

Dow Corning Fire Endurance Test on Penetration Seal Systems in Precast Concrete F

Using Silicone Elastomers, dated 10/28/1975

Dow Corning 561 Silicone Transformer Fluid Technical Manual,1 0-453-97, dated 1997

S-80393, Mesker Instructions for Installing d&H "Pyromatic" Automatic Sliding Fire Door Closer

S-27874B, General Electric Instruction Book GEK-26501, Liquid-Filled Secondary Unit

Substation Transformers, Rev. 2

S-52429A, Bisco, Fire Rated Penetration Seal Qualification Data, dated 08/16/1990

S-52480, Factory Mutual, Fire Rated Penetration Seal Qualification Data-Chemtrol Design FC-

225, dated 08/31/1990

S-54875B, Promatec, Fire Barriers-Unit 2 East Cableway, Rev. 2

Omega Point Laboratories, SR90-005, Three Hour Wall Test, dated 06/06/1990

Attachment

5

Promatec Technologies Inc., PSI-001, Issue 1,General Construction Details, dated 07/21/1998

Promatec Technologies Inc., IP-2031, Installation Inspection for Promnat's Three Hour Solid

Wall/Ceiling Protection System, Issue C, dated 06/16/1998

System Information Document No. SI-L-P-01401-03, Main Steam and Low Low Set System,

dated 4/3/2000

Applicable Codes and Standards

ANSI N45.2.1l-1974, Quality Assurance Requirements for the Design of Nuclear Power Plants

NFPA 12, Standard for Carbon Dioxide Systems; 1973 Edition.

NFPA 13, Standard for the Installation of Sprinkler Systems, 1976 Edition.

NFPA 14, Standard for the Installation of Standpipe and Hose Systems, 1974 Edition.

NFPA 20, Standard for the Installation of Centrifugal Fire Pumps, 1973 Edition.

NFPA 72D, Standard for the Installation, Maintenance, and Use of Proprietary Protection

Signaling Systems, 1975 Edition.

NFPA 72E, Standard on Automatic Fire Detectors, 1974 Edition

NFPA 80, Standard on Fire Doors and Windows, 1975 Edition.

NUREG-1 552, Supplement 1, Fire Barrier Penetration Seals in Nuclear Power Plants, dated

January 1999

OSHA Standard 29 CFR 1910, Occupational Safety and Health Standards,

Underwriters Laboratory, Fire Resistance Directory, January 1998

Other Documents

Design Change Package 91-009, Retrof ill Dielectric Fluid on Unit 2 Transformers, Rev. 1

Fire Protection Inspection Reports for the period 2001-2002

Fire Service Qualification Training, FP-LP-1 0003, Fire Fighter Safety, dated 01/1 4/2002

Fire Service Qualification Training, FP-LP-1 0004, Fire Fighter Personal Protective Equipment,

dated 01/14/2002

Fire Service Qualification Training, FP-LP-1 0014, Fire Streams, dated 01/22/2002

Attachment

. r. .

6

Fire Service Qualification Training, FP-LP-10018, Fire Fighting Principles and Practices, dated

01/22/2002

Hatch Response to NRC Information Notice 1999-05, Inadvertent Discharge of Carbon Dioxide

Fire Protection System and Gas Migration, dated 05/04/1999

Hatch Response to NRC Information Notice 2002-24, Potential Problems with Heat Collectors

on Fire Protection Sprinklers, dated 09/20/2002

1OCFR21 -001, ELECTRAK Corporation, Software Error within TRAK2000 Cable Management

and Appendix R Analysis System, dated 03/07/2003

U. S. Consumer Product Safety Commission, Invensys Building Systems Announce Recall of

Siebe Actuators in Building Fire/Smoke Dampers, dated 10/02/2002

Pre-f ire Plan A-43965, Power-Block Areas Methodology, Rev. 0

Pre-fire Plan A-43966, Fire Area 2404, Diesel Generator Building Switchgear Room 2E, Rev. 2

Pre-fire Plan A-43966, Fire Area 2408, Diesel Generator Building Switchgear Room 2F, Rev. 2

Pre-fire Plan A-43965, Fire Area 2016, DW 600V Switchgear Room 2C, Rev. 4

• Pre-f ire Plan A-43965, Power-Block Areas Methodology, Rev. 0

Pre-fire Plan A-43965, Power-Block Areas Methodology, Rev. 0

Pre-fire Plan A-43965, Power-Block Areas Methodology, Rev. 0

License Basis Documents

Hatch UFSAR Section 3.4, Water Level Flood Design, Rev. 20

Hatch UFSAR Section 9.1-A, Fire Protection Plan, Rev. 18C

Hatch UFSAR Section 17.2, Quality Assurance During the Operations Phase, Rev. 20B

Hatch Fire Hazards Analysis, Appendix B, Fire Protection Equipment Operating and

Surveillance Requirements, Rev. 12B

Hatch Fire Hazards Analysis, Appendix H, Application of National Fire Protection Association

Codes, Rev. 12B

Hatch SER dated April 18,1994

Safe Shutdown Analysis Report for E.l. Hatch Nuclear Plant Units 1 and 2, Rev. 26

Attachment

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Fire Hazards Analysis for E. l. Hatch Nuclear Plant Units 1 and 2, Rev.18C, dated 7/00.

NRC Safety Evaluation Report dated 01/02/1987; Re: Exemption from the requirements of

Appendix B to 10 CFR Part50 for Hatch Units 1 and2 (response to letter dated May 16,,

1986).

Letter dated 05/16/86, From L. T. Guewa (Georgia Power) to D. Muller, NRC/NRR; Re: Edwin I

Hatch Nuclear Plant Units 1 and 2 10 CFR 50.48 and Appendix R Exemption Requests

Design Change Request Documents

DCR No.91-134, SRV Backup Actuation via Pressure Transmitter Signals, Revision 0.

Drawing No. H-26000, Nuclear Boiler System P&ID, Sheet 1, Revision 39

Drawing No. H-27403, Automatic Depressurization System 2B21 C Elementary Diagram, Sheet

6 of 6, Revision 2

Drawing No. H-27472, Automatic Depressurization System 2B21 C Elementary Diagram, Sheet

3 of 6, Revision 2

Drawing No. H-27473, Automatic Depressurization System 2B21 C Elementary Diagram, Sheet

4 of 6, Revision 2

Drawing No. H-24427, Elementary Diagram, ATTS System 2A70 Sheet 27 of 35, Revision 3

Drawing No. H-24428, Elementary Diagram, ATTS System 2A70 Sheet 28 of 35, Revision 3

Drawing No. H-24429, Elementary Diagram, ATTS System 2A70 Sheet 29 of 35, Revision 5

Drawing No. H-24430, Elementary Diagram, ATTS System 2A70 Sheet 30 of 35, Revision 3

Drawing No. H-24431, Elementary Diagram, ATTS System 2A70 Sheet 31 of 35, Revision 3

Drawing No. H-24432, Elementary Diagram, ATTS System 2A70 Sheet 32 of 35, Revision 6

Attachment

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Attachment