ML050540511

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

ML050540511
Person / Time
Site:	Hatch
Issue date:	08/31/2003
From:	NRC/RGN-II
To:
References
FOIA/PA-2004-0277, IR-03-006
Download: ML050540511 (39)

See also: IR 05000321/2003006

Text

SUMMARY OF FINDINGS

IR 0500032112eeSee 05000366/20e~e6-.

7/7-11/2003 and 7/21-25/2003;

Hac

ula lnUisIad TinilFire

Protection

The report covered a two-week period of inspection by three regional inspectors anda

-contractor from Brookhaven National Laboratory. Three Green non-cited violations (NCVs) and

Q~c 4hiree unresolved items with potential safety significance greater than Green were identified.

The significance of most findings is indicated by their color (Green, White, Yellow, Red) using

IMC 0609, "Significance Determination Process" (SDP). Findings for which the SDP does not

apply

may

be

Green

or

be

assigned

a

severity

level

after

NRC

management

review.

The

NRC's program for overseeing the safe operation of commercial nuclear power reactors is*

described in NUREG-1 649, "Reactor Oversight Process," Revision 3, dated July 2000.

A.

NRC-Identified and Self-Revealing Findings

Cornerstone: Mitigatin. 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 completion of a significance determination.-+-

.

r~esponse o-this-ee

tial

he iene

pr~tly moved the manuai action

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ec~ecion 1R.05. 3.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 WAW

SRVs that are

required for ensuring that the suppression pool temperature will not exceed the heat

capacity temperature limit (HCTL) for the suppression pool. (Section 1 R21.01 .b)

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 IL.G.1 and

Technical Specification 5.4.1. 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 licensee could have time to develop and

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

did not have potential safety signift

ce

eater than very low safety significance.

'Section IRO

5

Green. The team identified a n g 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 10 CFR 50, Appendix R, Section III.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 accompli hed by operators in a timely manner, this finding

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

(Section 1 RO 05.b.3)

.

..i

'

' .

.

.

-*

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

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

support post-fire operation of safe shutdown equipment.

This finding involved a violation of 10 CFR 50, Appendix R, Section III.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

1 R05.07.b)

B.

Licensee-Identified Violations

None

-~

U. S. NUCLEAR REGULATORY COMMISSION

REGION II

Docket Nos.:

License Nos.:

Report No.:

50-321, 50-366

DPR-57, NPF-5

05000321/2003006 and 05000366/2003006

Licensee:

Facility:

Southern Nuclear Operating Company

E. l. Hatch Nuclear Plant

Location:

Dates: -

.

Inspectors:

Accompanying

Personnel:

Approved by:

P. o. Box 2010 -

Baxley, GA. 31513

July 7-11, 2003 (Week 1)

July 21-25, 2003 (Week 2)

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

R. Schin, Senior Reactor Inspector

G. Wiseman, Fire Protection Inspector

K. Suliivan, Consultant, Brooknaven National Laburaiury

S. Belcher, Nuclear Safety Intern, Week I

Charles R. Ogle, Chief

Engineering Branch I

Division of Reactor Safety

Enclosure

4% :91

-

I

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 .............

..

Cold Shutdown Repairs...........................................................................................

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.........................................................................................

I

V

I

1.

IR

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, mitigating Systems and Barrier Integrity

05 FIRE PROTECTION

The purpose of this inspecti

was to review the Hatch Nuclear Plant fire protection

program (FPP) for select

risk-significant fire areas. Emphasis was placed on

verification that the po fire safe shutdown (SSD) capability and the fire protection

features provided for nsuring that at least one redundant train of safe shutdown

systems is maintaied free of fire damage. The inspection was performed in

accordance with) e Nuclear Regulatory Commission (NRC) Reactor Oversight Program

using a risk-infgmed approach for selecting the fire areas and attributes to be

inspected. T1Je team used the licensee's Individual Plant Examination for External

Events and flplant 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, Elev

~j~/

feet.

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

Fire Area 2404, Switchgear Room 2E, Diesel Generator Building, Ele,

feet.

.

/

Fire Area 2408, Switchgear Room 2F, Diesel Generator Building, Ele,

feet.

L6-

'ation 130

vation 130

vation 130

The team evaluated the licensee's FPP 4 *inst applicable requirements, including

Operating License Condition 2.C.(3)(a), Fire Protection; Title 10 of the Code of Federal

Regulations, Part 50 (10 CFR 50), Appendix R; 10 CGFR 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 (HNP-FSAR); and plant Technical Specifications (TS). The

team evaluated all areas of this inspection, as documented below, against these

requirements.

a.

Documents reviewed by the team are~i

d in the attachment

Systems Required to Achieve and Post-Fire Safe Shutdovn

Insoection 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

2

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

were as follows:

(a)

Verify that the licensee's shutdown methodology has correctly identified

.\\

(a -

the components and systems necessary to achieve and maintain a safe

(b

/2shutdown

condition.

Ia

(b)

Confirm the adequacy of the systems selected for reactivity control,

.

- e~ ;reactor

coolant makeup, reactor heat removal, process monitoring and

z~~g

~' x

' support system functions.

..

.;

(c)

Verify that a safe shutdown can be achieved and maintained without off-'

d -site

power, when it can be confirmed that a postulated fire in any of the

selected fire areas could cause the loss of off-site power.

-

(d)

Verify that local manual operator actions are consistent with the plant's

fire protection licensing basis.

t

Findings '-'-

- ;

The team identified a potential concern in that the licensee used manual actions to

disconnect terminal board sliding links in order to isolate two 4-20 milli-amp (ma)

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

SRVs. This issue is discussed in section 1R05.03.b of the report.

.02

Fire Protection of Safe Shutdown Capability

Inspection Scone

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

SSSD path was free of fire damange. The team a!so inspected the ftea protecticn features

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

applicable separation and design requirements of 10 CFR 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:

Updated 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

3

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

properly evaluated in-situ fire loads 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 were 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 in the

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-u 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 simp!ementel breath'

, .,d t

, .p 'I

&

,

S

L4

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-fire plans and drawings with as-

built plant conditions. This was done to verify that fire fighting pre-fire plans and

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

described in the Fire Hazards Analysis (FHA).

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 location, water flow requirements and effective reach capability. Team

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

4

hose stations were not blocked and to verify that the requ ired 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.

Findin

~ o fidings f sigifiace were Identified.

.03

PotFr

aeSudwn Cagabiliyv

Inspection Scope

~

~

On a sample basis, in wre~icnv~s crfr.

A4

~-r'egy that systems and equipment

identified in the licensee's SSAR as being required to achieve and maintain hot

shutdown conditions 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 SSID Path I 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. The following motor operated valves (MOVs) and other components were

reviewed:

.

Findi

.

Spurious Actuation of Elevn

t he team identified a potential co

e

in that the licensee used manual

action t

ate two 4 to 20 ma instrumenta

n loop control circuits associated with

eleve

in lieu of providing the required. Tis did neot aipu e rto be

consistent with

the pnt1ensing basi~por 10 CFR 50 Appendix R.____

wit

The

fi

Rstates patae

24 could

dcauie

all eleven SRVs

to spur usy actuate as a result of fire damage to two cables Ih46located in close

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

identified by the licensee (circuit no's.: ABE019CO8 and ABE019CO9). Each of these

two circuits provides a 4 to 20 ma instrumentation signal from SRV high-pressure

W~V ',t

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.03

~os-Fir~af

Shudow Capbilty,,.

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actuation transmitter (2B21-N127B and 2B21-N127D) to masterip units 2B21-N697B

and 2B21-N697D, r spectively. The purpose of this circuitry is t provide an electrical

backup to the mec anical trip capability of the individual SRVs. n the event of high

reactor pressure, he circuits would provide a signal to the master trip units which would

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

conveyed to its ;espective master trip unit through a two-conductor, instrument cable

that is routed

rough this fire area (two separate cables). Each cable consists of a

single twistqE 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"AWKIe operators. That is, unlike ADS, it does not

provide any pre-actuation indication

.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 ma instrument circuit may not be necessary to initiate an undesired (false high)

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

insulation of the individual twisted conductors due to fire dam

e 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 generated as a result of leakage current

in each cable, could actuate the SRV electrical backup scheme which would cause all

eleven of the SRVs to open. The conductor insulation and jacket material of each cable

is cross-linked polyethylene (XLPE). Since both cables are in the same tray and

-pored to the same heatinn rate, there is a reasonaqhle likelihood that hoth

instrumentation cables

suffer insulation damage at the same time and both

circuits could fail high

ul

The licensee's SSAR recognizes the potential safety significance of this event and

describes methods that have been developed to prevent its occurrence and/or mitigate

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;P27 and link BB-10 in panel 2H11-P928. These

panels are located in the xxxxLocxxx . Opening of these links would prevent actuation

of the SRV trip units by removing the 4 to 20 ma signal fed by the pressure transmitters

to the master trip units. 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.

The inspection team had several concerns regarding the licensee's approach to this

potential spurious actuation of the SRVs. Specific concerns identified by the team

included:

- 6

• 1.

The links may not be opened in time to preclude inadvertent actuation of

the SRVs.

2.

The use of links to avoid inadvertent actuation of the SRVs did not

appear to be consistent with the current licensing basis.

3.

No objective evidence existed to demonstrate that the post-fire safe

shutdown equipment could adequately mitigate a fire in Fire Area 2104, if

the SRVs were to open. .

4.

The operations staff is unable to manually control the groupthat

are credited for mitigating a fire in Fire Area 2104 beca

f spuriousIl

actuati

ire induced damage.

With regard to the timing of operator actions to prevent fire damage from causing all

SRVs to open, during the inspection, the licensee performed an evaluation which

estimated that approximately thirty minutes would pass from the time of fire detection to

the time an operator would implement procedural actions to open the links. The

inspectors independently arrived at a similar time estimate based on their review of the

procedure. In response to inspector's concerns that this interval may be too lengthy to

preclude fire damage to the cables of interest and subsequent actuation of the SRVs,

the licensee agreed to enhance its existing procedures so that the action would be

taken immediately following confirmation6

f fire in areas where the spurious actuation

could occur.

r UsA,

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The'team also considered

opening terminal board links

K

in compliance with the plant's licensing basis; Current licensing basiwuments,

specifically Georgia Power request for exemption dated May 16, 19

d a subsequent

NRC Safety Evaluation Report (SER) dated January ,1987, charac tefzed the

ening

of links as a repair activity that is not permitted as

eans of complying withYction

!!1.G of Appendin. R. T6^ In,!ces

crncuded I

L, the

L

pening f i~ifl'swas considelred

a repair by both the licensee and the NRC staff in 1987. The licensee could not provide

any evidence to justify whv these actions ar/not characterized as a repair activity in its

rersp

~rs onse

thlis-insip-fi

fining the licensee initiated a Condition

Repor (CR 2003800152, dated 7124103) 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.

Additionally, because there is a potential for all SRVs to spuriously actuate as a result of

fire in Fire Pfenp 2104 at a time when RHR is not available, the SSAR credits the use of

core spray

op A to accomplish the reactor coolant makeup function. During the

inspection, the licensee performed a simulator exercise of an event which caused all 11

SRVs to open. During this exercise, simulator RPV level instruments indicated that core

spray would be capable of maintaining level above the top of active fuel. 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

7

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 to

1OCFR 50.

Finally, the licensee's failure to implement the design input requirements of one-out-of-

two taken twice logic for DCR 91-134 resulted in the followip plant problem. The logic

that was installed by DCR 91-134 for the SRVs was a tw -out-of-two coincident taken

twice logic in addition to a on -out-of-two coincident Ken twice logic. The team

determined that the two-out-of-cwo coincident log

nput from trip unit master relays

K31OD and K335D represented

common capee failure for group "A7 SRVs for a fire in

Fire Area 2104. Specifically, cabl ABE01I 08 associated with pressure transmitter

2B21-N127B current loop, and cabl

A

19CO9 associated with pressure transmitter

ZI-N127D current loop, ar route

close proximity to each other in the same cable

ay in Fire Area 2104. Both s ie cisted pair instrument cables are unprotected

from the effects of a fire in this

e area. Fire induced insulation damage to both cables

could result in leakage curre s which ca

es the instrument loops to fail high. This

failure mode simulates a

h nuclear boile

ressure condition and would initiate SRV

backup actuation of all

e group "A" SRVs.

henever a SRV lifts, it will remain open

until pressure reduc

to about 85% of its ove ressure lift setpoint The instrument

loops having failed igh, however, will ensure that the trip unit master relays and the trip

unit slave relavs con

to energize th

ot valve of the individual

nd keep the.

0Ccs

fassociamode prevents the operators from manually controlling toS

sopAS: s srqie

er the SSAR.

-2

Pending additional review by the Nith issue is identified as URI 50-366/2003006-

01, Concerns Associated with Pote

pening of SRVs

Anal sis: This issue has the potential to

pact availabi:and reliabil

objectiv

as

i1

al

eq ipment perfor ance attribute

the

mitigating

stem corn stone.

Ho ever, analy f this issue emains incomrteiend

e

NR.

• -*

0 Alternate Shutdown Capabilitv/Oerational ImDlementation of Alternative Shutdown

AWCaoabilitv/-M°

a.

Inspection Scone

,

5

n

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

shutdown from the control room. None involved abandoni

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

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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.

Findinas

Untimely

Introducti

opening c

Licensee

physically

and UnapDroved Manual Operator Action for Fire Safe Shutdown

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

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

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

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

Descriotion: 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 BB-10 in Panel 2H11-P927 and BB-10 in Panel

2H11-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 S

uriously

open.

The licensee had no preplanned estimate of how long ituld take operators to

complete this step during a fire event. There was n

vent time line or operator training

!P on this step. The team noted that, dun g a

event, operators could be using

many other procedures concurrent with theyr

ocedure. For example, they could be

using other procedures to communicate wit th

ire brigade about the fire, respond to a

reactor trip, deal with a loss of offsite power, and provide 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

about five to ten minutes. Consequently, th

am concluded that during a fire event the

licensee's procedures would not ensure that

ep 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, hence if the fire caused a loss of normal lighting (e.g., by

causing a loss of offsite power), operator/would need to use flashlights to perform the

actions inside the panels. Consequen( the team considered the emergency lighting.-.

for Step 9.3.2.1 to be inadequate (see rection 1 R05.07.b). In addition, labeling of the

links inside the panels was so poor that operators stated that they would not fully rely on

the labeling. Also, the tool that operators would use to loosen and slide the links inside

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

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

protecting the cables as required by 10 CFR 50, Appendix R, Section III.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 werg in the same cable tray in fire area 2104, the Unit 2 east

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

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

In response o th'

the licensee initiated CR 2003008203 and pr

ly

r rocedu e before the end of the inspection, moving the action of tep

be3..

inning > be

of the procedure. The procedure change enabled the a ons

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

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

A

s

he team determined that this pe

rrissue is, related to associated circuits.

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

associated circuits is temporarily limited. Consequentry. 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 IR21.0

Enforcement: 10 C

, Appendix R,Section III.G.2 requires that where cables or

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

cause mal-oper ion due to hot shorts, open circuits, or shorts to ground, of redundant

trains of syst es necessary to achieve and maintain hot shutdown conditions are

located withir he same fire area outside of the primary containment, one of the

following me ns 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

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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 licensing

basis and consequently that there was no requirement to protect the instrument cables.

However, the licensee could provide no evidence to support that position.

This potential issue will rem, On unreso~ved pending the6dRcompletion of a

significance determination .This

Gpete~Wal issue is identified as URI 50-366/03-06-02,'

Untimely and Unapproved Manual Operator Action for Fire Safe Shutdown.

2.

Local Manual Ogerator 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.

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-FO15A, 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 1 -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 about 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-FO15A valve handwheel was not good.

Other difficulties with manually opening the valve included the heat; the need to wear

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

11

emergency lighting (see Section 1 R05.07). 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-FOI5A 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, 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 pyrorm it.

The licensee had no operator training obpcrfii

e

J

M JP

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.

Analvsis: This 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 licensee could have time to develop and implement cold

shutdown repairs to facilitate accomplishment of the action, this finding did not impact

the effectiveness of one or more of the defense in depth elements. Hence this finding

did not have potential safety significance greater than vary low safety signifinanr.e

(Green).

Enforcement: 10 CFR 50, Appendix R, Section III.G.1 requires that fire protection

features shall be provided for systems important to safe shutdown and shall be capable

of limiting fire damage so that systems necessary to achieve and maintain cold

shutdown from either the control room or emergency control stations can be repaired

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 2E1 1 -FO1 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

T

12

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 providet teps 4.15.8.1.1 and 9.3.5:1

for manually opening valve 2E11-FO15A following a fire in fire area 2104.

Contrary to the above, the licensee had no procedure for repairing any related fire.

damage within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Instead, the licensee relied on local manual operator actions,

as described in Steps 4.15.8.1.1 and 9.3.5.1 of AOP 34AB-X43-001-2. However, those

procedure steps were inadequate in that some operators would not be able to perform

them because the required actions were tbo difficult and also were unsafe. In response'

to this issue, the licensee initiated CR 203008202. Because the identified inadequate

operator actions are of very low safety significance and the issue has been entered into

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

consistent with Section VI.A.1 of the NRC's Enforcement Policy: NCV 50-366/03-06-03,

Inadequate Procedure for Local Manual Operator Action for Post-Fire Safe Shutdown

Equipment.

3.

Unapproved Manual ODerator Actions for Post-Fire Safe Shutdown

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. 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 of steps from the procedure included:

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

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

(2R42-S030) AND 2R42-S028 (2R42-S031) in service per 34.S0-R42-0On1-2."

(g

Step 4.15.4.5; ...If HPCI fails to automatically trip on high RPV level... OPEN the

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

F3025 HPCI Governor Valve, in the CLOSED position:

TT-75 in panel 2H1 1-P601

TT-76 in panel 2HI1-P601"

Step 4.15.4.6; ... lf HPCI fails to automatically trip on high RPV level... "OPEN

breaker 25 in panel 2R25-S002 to fail 2E41-F3052, HPCI Govemor Valve, in the

CLOSED position."

The team walked downd ese actions using the guidance contained in Inspection

Procedure 71111.05T nd judged that they could reasonably be accomplished by

operators in a timely manner. However, the team determined that these operator

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actions were being used instad of physically protecting cables from fire damage that

could cause a loss of statfn service battery chargers or a HPCI pump runaway.

Analysis: The finding'

greater than minor because it affected the availability and

reliability objectives

he 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.

Enforcement: 10 CFR 50, Appendix R, Sec

.G.2 requires that where cables or

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

cause maloperation due to hot shorts, ope

ircuits, or shorts to ground, of redundant

trains of systems necessary to achieve

d maintain hot shutdown conditions are

located within the same fire area outsi e of the primary containment, one of the

following means of ensuring that one orthe 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 NRG approval. In response to this issue, the licensee initiated CR2003800166

dated 7125:2003. Because the issue had very low safety significance and has been

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

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

06-04, Unapproved Manual Operator Actions for Post-Fire Safe Shutdown.

.06-

Communications

I"17

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

be available for personnel performing local manual operator actions. In addition, 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

Emeraency Lighting

14

a.

Inspection Sco.

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

emergency lighting coverage was provided as required by 1 0 CFR 50, Appendix R,

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

operation of SSD equipment. During walkdowns of the po'st-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.

Descri-tion: 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 e~nergize

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

(2R42-S030) AN

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

Step 4.15.4.5; ...

i

f HPCI fails to automatically trip on high RPV level... "OPEN the'

following links to energizs

2E41 -Ft 24, Trip Solenoid Valve, AND to fall 2E41

F3025 HPCI Governor Valve, in the CLOSED position:

TT-75 in panel 2H11-P601

TT-76 in panel 2H-11-P601",

Step 4.15.5; "iF 2R25-65, Instrument Bus 2Bf , is DE-ENERGIZED perform the

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

Instrument, operable:

4.1 5.5.1; At panel 2H-11 -P612, OPEN links AAA-1i 1 and AAA-1 2.

4.15.5.2; At panel 2HI11~-P601, CLOSE links HH-48 and HH-49."

Steps 4.15.8.1.1 and 9.3.5.1; "Manually OPEN 2E1 1-1FO15A, Inboard LPCI

Injection Valve, as required.

Steps 4.15.8.1.2 and 9.3.5.2; t "Manually CLOSE 2E1 1 -Fo ae8A, RHR Pump A

Minimum Flow Isolation Valve, as required."

6

1.5

/

Step 9.3.2.1; "To prevent all 1 1 SRVs from opening simultaneously, open links

'

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

. /

Step 9.3.3; "At Panel 2H11-P627, open links AA-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 2H1 1-P700 to open Drywell Pneumatic

<

System Inboard Inlet Isolation, 2P70-F005."

Step 9.3.7; "OPEN link TB1-12 in Panel 2H1 1-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-F006D."

.

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

required..."

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

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

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

SRVs from spuriously opening, which was judged to be untimely (see)ection.

1R05.05.b.1). The other exception was the action to open 2E11-FO15A, which was

judged to be too difficult (see

ction 1 R05.05.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

op W

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

dad

im

impact the effectiveness of one or morc of the defCnse in depth elements.

H cthi finding did not have potential safety significance greater than very low safety

sifice

(Green).

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,

consistent with Section VI.A.1 of the NRC's Enforcement Policy: NCV 50-366/03-06-05,

Inadequate Emergency Lighting for Operation of Safe Shutdown Equipment.

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. ICold Shutdown Repairs

16

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 -F01 5A (see section

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

Consequently, this section of IP 71111.05 was not performed.

I

I

Fire Barriers and Fire Area/Zone/Room Penetration Seals

Inspection Scope

.

.

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 reviewed completed

surveillance and maintenance procedures for selected fire barrier features to verify the

fire barriers were being adequately maintained.

The team evaluated the adequacy of the fire resistance of fire barrier electrical raceway

fire barrier system (ERFBS) enclosures for cable protection to satisfy the applicahle

separation and design requirements of 10 CFR 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.

II

Ob.

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.

Findings

f;

.

T

17

No findings of significance were identified.

.10

Fire Protection Systems. Features, and Equipment

a.

lns.ection Scooe

The team reviewed flow diagrams, cable routing information, and operational valv e-,

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 the FHA.

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, lll.G.1 and Ml.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 locationsfor

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

for NFPA code deviations, and NFPA 72E. In addition, 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 ihat 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 002 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 002fire suppression system to determine correct system controls

A

'I

18

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

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

actions to address the potential for 002 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 system's and

operator actions required for SSD would be inhibited by 002 migration through the floor

drain system.

b:

.

Findings

No findings of significance were identified.

p Compensatory Measures

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 systems.

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

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

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

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

effectiveness in returning equipment to service in a reasonable period of time.

b.

Findings

No findings of significance were Identified.

1 1R21

SAFETY SYSTEM DESIGN AND PERFORMANCE CAPABILITY

1;:1.44.

S

Backup Actuation 3A Pressure Transmitter Signals

- nsgection Scope(

The team performed anindependent designreview of plant modificationo

DCt I 1-134 in

order tor evaluate the techniedl adequacy of the design change packag e a,

nies

-shar=Wed 10

rcr

sys5temsluzation. The scope of the review and circuit analysis

performed by the team was limited to the group "A' SRVs for which the licensee takes

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

19

A

-

b.

.in:nc:

An inadequate plant modification, Reine IClul- .e~u 1,)Hes*CRY9 1-134, failed to

implement the design input requirements of one-out-of-two taken twice logic for the

SRVs backup actuation using pressure transmitter signals.

Description:

DCR 91-1 34 was implemented in response in to concerns raised in General Electric

Report NEDC-3200P, Evaluation of SRV Performance during January-Febr'uairy 1991

Turbine Trip Events for Plant Hatch Units I and 2. In order to ensure that individual -

SRV(s) will actuate at or near the appropriate set point and within allowable limits, a

backup mode of operation for the SRVs was implemented by this DCR. The design'

was intended to mitigate the effects of corrosion-induced set point drift of the Target

Rock SRVs.

Automatically controll

t~ 'stage SR Vs are installed on the main steam lines inside.

containment for the p r~e 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 'Open" position, the switch energizes the pilot valve.

of the individual SRV and causes it to go open. When the switch is placed, in the "Auto"

position the SRV is opened upon rece'ipt-of either an Auto Depressurization System

(ADS), or Low-Low Set (LLS) control logic signal. Either signal will initiate opening of.

the valve. DCR 91-1 34 provided a backup mode for initiation of electrical trip of the pilot

valve solenoid, which was independent of ADS or LLS logic. The backup mode required

no operator action to initiate opening of the SRVs and was considered a "blind control

loop" to the operators, ie. there are no instruments that provide the operators

information concerning the open/close status of the SRVs.

The scope of the plant modification involved the installation of four Rosemount pressure

transmitters (Model No. 1154GP9RJ), 0-3000 psig, in the 2H21-P404 and P405

instrument racks at Elevation 158 of the reactor building. Each pressure transmitter

fomd part of a 4-20 ma current loop and p~~dteaao rpsga o R

actuation within the following set point rou'

SRV Groua

SRV Identification Tags

SRV Set Point

A

2921-FOw3B, 0, F, and G

1120 psig

B

21E21-FOl3A, C, K, andPM

n J

193psig

C

C21B21-F013E, H, and D

- 1140 psig

20

Pressure transmitters (PTs) 2B21-N127A and 2B21-N127C were wired to ATTS

cabinets 2H1 1-P927. Pressure transmitter 2B21-N127A instrument loop components

consisted of a trip unit master relay K308C and trip unit slave relays K321C and K332C.

The loop components for pressure transmitter 2B21-N127C consisted of a trip unit

master relay K335C in addition to trip unit slave relays K336C and K363C. These two

instrument loops constituted a wDivision" pressure monitoring channels and were'

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

backu atation.The desi'gn objective of having two instrument channels was to assure .

compliance with HNP-2-FSAR, Section 15.1.6.1, Application of Single Failure Criteria.

g

This criteria requires for anticipated operational occurrences (AOOs) that the protectiori2

sequences within mitigation systems be single component failure proof. A failure of one

-

_ instrument channel in a division will therefore not eliminate the protection provided by

eihe of

theinstruent channels.'.;

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

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

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

K332D. The loop components for pressure transmitter 2B21-N127D consisted of a trip

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

two instrument loops constituted a separate 'Division' pressure monitoring'channels and

'were intended to provide the one-out of two logic signal from this Division'for initiating

SRV backup actuation.

mw

-assure onmplinr~ewith

NPF-AR

ion1.

The following table identifies the Division; pressure transmitter loops and the associated

trip unit master and slave relays:

Division

PT Loops

Trip Unit Master Relays

Trig Unit Slave Relays

'A

2B21-N127A

K308C

K321C and K332C

2B21-N127C

K335C

K336C and K363C

.

B

2B21-N127B

K310D

K312D and K332D

2B21-N127D

K335D

K336D and K363D

The Group A" SRVs were provided I

nput signals from the trip unit master relays.

The Group UB and C" SRVs were

ovided logic input signals from the trip unit slave

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

ATTS cabinet 2H1 1-P928), wer intended to be wired to provide none-out-of-two taken

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

relay failure in eitheriivision would not cau e an inadvertent SRV actuation.

Coincident logic inpu is required from both Iivision instrument loops in order to initiate a

SRV backup actuation using the pressure tr nsmitter signals. This occurs when the

circuit that is used to energize the individual SRV pilot valve to open the SRV, is enabled

by receiving simultaneous logic inputs from either instrument loop in both division.

21.

The team performed a circuit analysis of SRV2B21-FO13F (Path 1) and SRV 2

1-

F01 3G (Path 2) in order to verify that the design objectives of implementing a o e-out-

of-two taken twice logic had been achieved. Based on this review the team de rmined

that the design objective of implementing a one-out-of-two taken twice logic h

not

been installed for the SRVs. The logic installed for the SRVs was a two-out-o two

coincident taken twice logic in addition to a one-out-of-two coincident taken

ice logic. :

-

-

The coincident logic implemented using trip unit master relays K310D and K35D

result in spurious actuation of group KA"

SRVs for a fire in Fire Area 2104.

defeats

--

the c pability to manually control these SRVs. Whenever a SRV lifts, it will remain open

until nuclear boiler pressure is reduced to about 85% of its overpressure lift setpoint

However, because the instrument loops have failed high, the trip unit master relays and

the trip unit slave relays willjcontinue to energize the pilot valve of the individual SRV

--

and keep the SRV open

Tis failure mode prevents the operators from manually

controlling the

SRVs as is required per the SSAR.

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

reliability objectives and the equipment performance attribute of the mitigating system

cornerstone. The team determined that the finding had potential safety significance

greater than very low safety significance because it prevented the operators from

manually controlling group A SRVs which the licensee credits with mitigating a fire in

Fire Area 2104. Manual control of the group A SRVs is required to ensure that the

suppression pool temperature will not exceed the HCTL for the suppression pool.

Failure to ensure that the suppression pool temperature will not exceed the HCTL could

result in loss of net positive suction head for the Core S ay pumps which the licenses

credits for mitigatin thivent.

h/*

I-erv

Enforcement

10 CFR 50, Appendix B, Criterion l1l, requires that design control

measures shall provide for verifying or checking the adequacy of design.

DCR 91-1 34 specified design input requirements for the sensor initiated logic that

electrically activates the SRVs to be a one-out-of-two logic scheme. It also identifie.d tile

potential worst case failure mode of this logic modification as a short in the logic which

would results in an inadvertent opening of a SRV. It concluded that the modification is

designed so that the actuation logic will not fail to cause inadvertent opening of a SRV

I

nor prevent a SRV from lifting upon ADS/LLS activation. Contrary to the above the logic

implemented by the licensee for DCR 91-134 was different from the specified design

input requirements. The independent design review performed for DCR 91-134 failed to

identify this error in the logic scheme. Additionally, the Appendix R Impact Review

performed for DCR 91-134 failed to identify the potential failure mode of all eleven SRVs

because of fire induced damage in Fire Area 2104.

The plant modification installed for DCR 91-134 failed to correctly implement the one-

out-of-two taken twice logic that was specified in 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 with the eleven SRVs

22

assuming a stuck open mode of operation, based on the logic input from trip unit master

unit relays K31 OD, and K335D and their associated trip unit slave relays. Pending

completion of an significance determination by the NRC, this item is identified as URI

50-366/03-06-06, Implementation of DCR 91-134 Results in Spurious Actuation of

Eleven SRVs because of Fire Induced Faults.

4.

OTHER ACTIVITIES

40A2 Identification and Resolution of Problems

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 initiated to resolve the issues. In

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

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

experience (OE) 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.

OA6

eetinqs. Including Exit

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

Manager, and other members of your staff at the conclusion of the inspection on July

25, 2003 The licensee acknowledged the findings presented. Proprietary information is

not included in the inspection report.

Licensee personnel:

M. Beard

V. Coleman

M. Dean

B. Duval

R. Dedrickson

M. Googe

J. Hamnmonds

D. Javorka

R. King

I. Luker

T. Metzer

A. Owens.

J. Payne

D. Parker

J. Rathod

K. Rosanski

M. Raybon

J. Vance

R. Varnadore

NRC personnel:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Acting Engineering Support Supervisor

Quality Assurance Supervisor

Nuclear Specialist, Fire Protection

Chemistry Superintendent

Assistant General Manager for Plant hatch

Maintenance Manager/

Operations Manager

Administrative Assistant, Senior

Acting Engineering Support Manager

Senior Engineer, Licensing

Acting Nuclear safety and Compliance Manager

Senior Engineer, Fire Protection

Senior Engineer, Corrective Action Program

Senior Engineer, Electrical

Bechtel Engineering Group Supervisor

Oglethorpe Power Corporation Resident Manager

Summer Intern

Senior Engineer, Mechanical & Civil

Outages and Modifications Manager

r

Attachment

I

a

r

'.

N. Garret,

Senior Resident Inspector

C. Payne

Fire Protection Team Leader

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

50-366/03-06-01,

URI

Concerns Associated with Potential Opening of SRVs (Section-

-

1 R05.03.b)

50-366/03-06-02

URI

Untimely and Unapproved Manual Operator Action for Post-Fire

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

50-366/03-06-06,

URI

Implementation of DCR 91-134 Results in Spurious Actuation of

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

Opened and Closed

50-366/03-06-03

NCV

Local Manual Operator Action for Post-Fire Safe Shutdown

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

50-366/03-06-04

NCV

Unapproved Manual Operator Actions for Post-Fire Safe

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

50-366/03-06-05

NCV

Inadequate Emergency Lighting for Operation of Post-Fire Safe

Shutdown Equipment (Section 1R05.07.b)

Discussed

None

Attachment

2

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

SOP 34S0-C71-001-2,1I20VAC RPS Supply System, Version 10.2

SOP

34S-N40-001-2, Main Generator Operation, Version 10.8

SOP 34S0-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-01O0-2S, RC RPV Control (Non-ATWS), Rev. 8, Attachment 1

31 EO-EOP-01 2-2S, PC-1 Primary Containment Control, Rev. 4, Attachment 1

31 EO-EOP-013-2S, PC-2 Primary Containment Control, Rev. 4, Attachment I

31 EO-EOP-01 4-2S, SC - Secondary Containment Control, Rev. 6, Attachment I

31E0-EOP-01 6-2, CP-2 RPV. Flooding, Rev. 8,Attachment I

Procedure 34AB-X43-001-2S, Rev.10ED3, uFire 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-1 1821, Fire Hazards Analysis, Turbine Bldg 'El. 130'-O0w, 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-2015 RH.R. System P&ID Sheet 2, Re'.4

H-605

R

v. 46:

H-26018, Core Spray System P&ID, Rev. 29

B-I 0-1 326, Rectangular Fire Damper Schedule, Rev. 2

B-I10-1329, Rectangular Fire Damper, Rev. I

H-I 1033, Fire Protection Pump House Layout, Rev. 47

H-11035, Fire Protection Piping and Instrumentation Diagram, Rev. 22

H-1 1226, Piping-Diesel Generator Building Drainage, Rev. 6

H-1 1814, Fire Hazards Analysis Drawing, Control Building, Rev. 5

H-11821, Fire Hazards Analysis Drawing Turbine Building, Rev. 11

H-1 1846, Fire Hazards Analysis Drawing, Diesel Generator Building, Rev. 2

H-1- 1894, Fire Detection Equipment Layout-Diesel Generator Building, Rev. 2

H- 1915, Fire Detection Equipment Layout-Control Building, Rev. 2

H-: 3006,'Conduit and Grounding, Fire Pump House, Rev. 9

H-I1361 5, Wiring Diagram, Fire Pump House, Rev. 13.

H-I6054, Control Building HVAC System, Rev. 19

H-41509, Diesel Generator Building C02 System-P&ID, Rev. 5

H-43757, Penetration Seals-Type, Number, and as-Built Location, Rev. 3

Calculations. Analyses. and Evaluations

E. -. Hatch Nuclear Plant Units

r

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 03/11/1 988

Calculation SMNH94-046, FCF-F-OB-006, Fire Resistance of Concrete Block at HNP, dated

09/30/1994

Calculation SMNH94-048, FCF-FiOB-006, Cable Tray Combustible Loading Calculation, dated

Attachment

3

• 09/30/1994

Calculation SMVNH-98-023, 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

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 Switcfigear 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 94-013, Sheets 28 and 29, 600V Reactor Building MCC 2E-B

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 Bu~ilding

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 1S Needs to be Enhanced

CR 2001002032,.Fire Procedure 34AB-X43-O01-2S Needs Actions for Diesel Fuel Oil Pumps

CR 2003004377, Fire Procedure 34AB-X43-001-1 Enhancements

CR 2003004379, Fire Procedure 34AB-X43-001-2 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

Attachment

Fi

S

V-t

-

44

CR2003008202, Manually Opening MOV221 l-FO15A

CR 2003008203, SRV Manual Action Steps in Fire Procedure

CR 2003008237, Emergency Lights and Component Labeling for Manual Actions

CR 2003008238, 002 Migration Through Floor Drains

CR 20038001 32, SSAR Error for Position of 2E 1 -1-F04A

CR 2003800151, Instruments for Manual Actions

CR 2003800152, Sliding Links in SSAR

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 06/2712003)j

-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,10-453-97, 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

Attachment

'1-

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

Promatec Technologies Inc., PS1-001, Issue 1, General Construction Details, dated 07/21/1998

Promatec Technologies Inc., IP-2031, Installation Inspection for Promat's Three Hour Solid

Wall/Ceiling Protection System, Issue C, dated 06/16/1998

System Information Document No. Sl-LP-01401-03, Main Steam and Low Low Set System,

dated 4/3/2000

Applicable Codes and Standards

ANSI N45.2.11-1974, 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-1552, 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

Attachment

....

6

Other Documents

Design Change Package 91-009, Retrofill Dielectric Fluid on Unit 2 Transformers, Rev. i

Fire Protection Inspection Reports for the period 2001-2002

Fire Service Qualification Training, FP-LP-10003, Fire Fighter Safety, dated 01/14/2002

Fire Service Qualification Training, FP-LP-1 0004, Fire Fighter Personal Protective Equipment,

dated 01/14/2002

Fire Service Qualification Training, FP-LP-10014, Fire Streams, dated 01/22/2002

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-fire Plan A-43965, Power-Block Areas Methodology, Rev. 0

Pre-fire Plan A-43966, Fire Area 2404, Diesel Generator Building Switchgear Rsoin 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, W 600V Switchgear Room 2C, Rev. 4

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

Attachment

7

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.I. Hatch Nuclear Plant Units I and 2, Rev. 26

Fire Hazards Analysis for E. 1. Hatch Nuclear Plan't Units

and 2, Rev.18C, dated 7/00.

NRC Safety Evaluation Report dated 0110211987; Re: Exemption from the requirements of

Appendix R to 1 0 CFR Part 50 for Hatch Units I and 2 (response to letter dated May 16, 1986).

Letter dated 05/16/86, From L. T. Guewa (Georgia Power) to D. Muller, NRCINRR; Re: Edwin I

Hatch Nuclear Plant Units I and 2 10 CFR 50.48 and Appendix R Exemption Requests

Design Change Request Documents

DCR No. 91-1 34, 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 2B21C Elementary Diagram, Sheet

6 of 6, Revision 2

Drawing No. H-27472, Automatic Depressurization System 2B21C Elementary Diagram, Sheet

3 of 6, Revision 2

Drawing No. H-27473, Automatic Depressurization System 2B21C 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

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