ML20195J046

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Insp Rept 50-445/87-22 on 871019-23.No Violations Noted. Major Areas Inspected:Implementation of Fire Protection Program & Compliance W/Branch Technical Position CMEB-9.5-1, Fire Protection for Nuclear Power Plants
ML20195J046
Person / Time
Site: Comanche Peak Luminant icon.png
Issue date: 01/11/1988
From: Kelley D, Mckee P, Singh A
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV), NRC OFFICE OF SPECIAL PROJECTS
To:
Shared Package
ML20195J034 List:
References
50-445-87-22, NUDOCS 8801200459
Download: ML20195J046 (31)


See also: IR 05000445/1987022

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-U.S. NUCLEAR REGULATORY COMMISSION

OFFICE OF SPECIAL PROJECTS

NRC Inspection Report: 50-445/87-22 Construction Permit: CPR-126

Docket No: 50-445

Applicant: TU Electric

Skyway Tower

-400 North Olive Street

Lock Box 81

Dallas, Texas 75201

Facility Name: Comanche Peak Steam Electric Station (CPSES),

Unit 1

Inspection At: Comanche Peak Site, Glen Rose, Texas

Inspection Conducted: October 19-23, 1987

Inspectors: W"I Y '4 b\

Amarjit Singhi, Reactor Operation Engineer

//G/N

Date

0ffice of Spec'al Projects.

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Den'nis Kelley, Senior' Resider}t / inspector

Comanche Peak Steam Electric Mation

Also participating and contributing to the report were:

Harvey Thomas, Brookhaven National Laboratory (BNL)

Anthony Fresco, BNL

Thomas Storey, Science Application International

Reviewed by:  !-- .

Phillip F.6McKee, Deputy Director

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'Da te

Comanche Peak Project Division

Office of Special Projects

Inspection Summary

Inspection Conducted October 19-23, 1987 (Report 50-445/87-22)

Areas inspected: Special announced inspection of the implementation of

fire protection program and compliance with Branch Technical Position (BTP)

CMEB 9.5-1, Fire Protection for Nuclear Power Plants," (formerly Appendix A

to BTP APCSB 9.5-1); per FSAR commitments and SER evaluation.

Results: Within the areas inspected, no violations were identified.

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8801200459 880111

PDR ADOCK 05000445

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DETAILS

1.0 Persons Contacted

TV Electric

R. Bab, Fire Protection Engineer

J. Barker,10 Electric

H. Beck, CPE/FP

C. Becket, CPE/FP

M. Blevins, TU Electric

J. Boothroyd, TU OPS F.P

B. Browning, Startup

F. Cobb, Prof.

C. Creamer, Project ISE Engineer

P. Desar, CPE/ISC

J. Disewwright, TV Electric

T. Evans, CPE/EE

D. Fuller, TU Electric

W. Grace, TV Electric (Nuc Ops) '

R. Howe, EPM /FP

J. Jamer, CPE/ MECH

J. Kelly, TV Electric

J. LaMarca, CPE/EE

B. Lancaster, TV-Electric

0. Lowe, TV Electric

R. Laytun, Fire Protection Coordinator

F. Madden, CPE-MECH

S. Popek, CPE/FP

J. Reywerson, TV Electric

W. Rowe, CPE/ civil

C. E. Scott, TV Electric

J. Smith, TU Electric

N. Terrel, TV Electric

D. hoodlen, TV Electric

IMPELL

John Echternacht

Steven Einbinder

Kevin C. Warapius

John Wawreeniak

SWEC

J. T. Conly

Thomas G. Persurer

D

Enrique Margalejo

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2.0 Background and Inspection Approach

This report documents findings during an inspection conoucted by Mr. A.

Singh and Mr. D. Kelley of the Office of Special Projects (OSP), Mr. T. A.

Storey of Science Applications Internatinnal Corporation (SAIC) and Messrs.

H. Thomas and A. Fresco of Brookhaven National Laboratory during the period

October 19-23, 1987.

The fire protection program for Comanche Peak Steam Electric Station

(CPSES) is described in the applicant's Fire Protection Report (Ref. A.1)

and the FSAR. The applicant is committed to the Fire Prctection Program of

Appendix A to APCSB 9.5-1, as modified by applicant correspondence to the

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NRC that docunents additional concitments and deviations from FSAR

censni tments . Supplement 12 to the Safety Evaluation Report (NUREG-0797)

issued in October 1985 presents the staff review of the CPSES Fire

Protection Program. In Supplen,ent 12 the staff reviewed the applicant's

program against branch Technical Position (BTP) CMEB 9.5.1, which

superseded Appendix A to BTP APCSB 9.5.1. Among other changes, the

criteria of Appendix R to 10 CFR Part 50 were factored into GTP CMEB 9.5.1.

TUEC letter dated October 9, 1987 provided the staff with an advance copy

of a change to the FSAR sections relative to the fire protection program.,

TUEC letter dated October 2, 1987 provided the staff with revised deviations

to BTP APCSB 9.5-1 Appendix A and 10 CFR 50, Appendix R.

A site inspection of the CPSES fire protection program was conducted during

October 29 thrcuch November 2, 1984. The inspection was documented in

Inspection Report (IR) 50-445/84-44. This inspection (hereafter referred

to as 84-44 inspection) included personnel from the Office of Nuclear

Reactor Regulation, Regico IV and the Of fice of Inspection and Enforcement

and resulted in a number of open items.

Areas examined during the 84-44 inspection included establishment and

implementation of the fire protection program and compliance with the

requirements of BTP "Fire Protection for Nuclear Power Plants," per FSAR

ccnmitments and SER evaluation. Within these areas, the inspection

consisted of selective examination of procedures and representative

records, interviews with personnel, ar.d observations by the inspectors.

During this inspection, open items resulting from previous NPC audits and

inspections were reviewed. The results of these reviews are included

within this report.

0.0 Fire Protection Program Requirements

3.1 Fire Protection Program

In SSER 12, the staff stated that the fire protection progran meets the

guidelines of BTP CMEB 9.5-1 and is therefore, acceptable. During the

84-44 inspection, the inspectors found that the applicant's procram did

not specifically designate responsibility for fire brigade training

and maintenance of training records. In addition, the inspectorc found

that the prcgram dio not identify that a QA program was established for

the fire protection program (Unresolved item 445/8444-0-01,1st item).

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During this inspecticn the applicant prasented procedure FIR-101, "Fire

Protection Program" which had been revised to address the staff concerns

stated above. The revisions were found to adequately address the

assignment of fire brigade training and records maintenance

responsibilities and clearly established that a QA program would te

provided for fire protection. Open Item 445/8444-0-01, 1st item, is

therefore closed.

3.2 Fire Hazards Analysis

In SSER 12, the staff concluded that the fire hazards analysis (FHA)

met the guidelines of BTP CMEB 9.5-1. The applicant has since revised

the FHA and has incluced it in the Fire Protection Report dated

September 22, 1987. Revisions to the FHA reflect changes in p bnt

design or changes in the Fire Safe Shutdown Analysis report. As a

result of this revision, a new deviatien relating to the RHR isolation

valves was identified. Also, a number of changes to previous deviations

were n.ade. Where these changes may have affected previous staff evalua-

tions, they are discussed in this inspection report. The new deviation

is discussed in Section 4.2 of this report.

3.3 Administrative Ccr,trols

The staff concluded in SSER 12 that the administrative controls

identified by the applicant met the guidelines of BTF CMEB 9.5-1.

During the 84-44 inspection, four items were identified where

ac'ministrative procedures were inaaequate. The items were as follows:

Failure to designate who is respcnsible for obtaining a fire permit

for controlling ignition sources. (0 pen Item 445/0444-0-01,

4th itea)

Failure to delete a temporary instruction for protection of the

new fuel area af ter the permanent procedure was in place. (0 pen

Itera 445/8444-0-01, 5th item)

Discrepancies between the proposed Technical Specifications and the

fire protection surveillance procedures. (0 pen Item 445/8444-0-02)

Failure to include a fire pump performance curve in the preoperational

test procedure. (0 pen Item 445/8444-0-03)

L;uring this inspection the applicant demonstrated that all of the above

aentioned discrepancies had been addressed in revisions to prc:edures.

These procedures weta, reviewed during the inspection ard found acceptable.

The above listed open items are therefore closed.

3.4 Fire Brigade and Fire Brioade Training

In SSER 12, the staff stated that the fire brigade and fire brigade

training program meet the guidelines of BTP CMEB 9.5-1. During the

84-44 inspection, the definition of the fire brigade compositicn was

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found to be in conflict with several plant procedures (0 pen Item

445/84-44-0-01, 3rd item). Also, the applicant's fire protection

training procedure did not adequately address the tracking of the

continuing qualification-of fire brigade ren.bers.

During this inspection, the team reveiwed the fire brigrade training

records and the revised fire protection training procedures and found

them acceptable. Therefore, these issues are considered resolved and

Open Item 445/844-0-01, 3rd item, is closed.

3.5 Reactor Coolant Pump (RCP) Oil Collecticn System

An inspector reviewed the installation of the RCP oil collection system.

The inspector luoked at two of the four RCPs and verified that all

external potential leakage areas were adequately covered and would drain

oil into a separate collection tank. The design drawings were reviewed

and the inspector confirmed that each collection tank was desigred to

hold all of the oil inventory from its associated pump. During the

inspection the applicant stated that seismic analysis for the RCPs had

not been conpleted to verify that the system was seismically qualified.

This item is considered open pending completion of the analysis by TV .

Electric (445/8722-0-01).

4.0 General Plant Guidelines

4.1 Building Design

Section D.1.j of Appendix A to BTP APCSB 9.5-1 states that floors, walls

and ceilings enclosing separate fire areas should have a minimum fire

rating of three hours, including penetration seals, fire coors and dampers.

The staff stated in SSER 12 that all fire rated assemblies are tested

for three hours in accordance with American Society for Test t.g and

Materials (ASTM) E 119, are designed in accordance with three-hour-

rated fire barrier designs obtaineo f rom the fire Resistance Directory

published by Underwriters Laboratories (UL), or are constructed of

8-inch-thick reinforced concrete in accordance with the "Uniform

Building Code" (International Conference of Building Code Officials)

for a minimum fire resistance rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The staff concluded

in SSER 12 that the fire-rated walls and floor / ceiling assemblies are

provided in accordance with the guidelines of BTP CMEB 9.5-1 fection

C.5.a and are therefore acceptable.

During this inspection several barriers separating redundant trair.s of

safe shutdown equipment were identified by the inspector as not being

three-hcur-rated. Specifically, unrated steel hatches were located in

fire area bouncaries. The applicant presented an analysis which stated

that cue to low combustible loacing on either side of the hatches,

automatic suppression on at least one side of the hatch and a one hour

fire resistive coating on both sides cf the hatch, it was not likely

that a fire would propagate through the hatch. The inspector reviewed the

analysis and found it acceptable. However, it was identified that this

was a deviation from Section 0.1.j of Appendix A to BTP APCSB 9.5-1 and

must be identified as such in the FSAR. The applicant comitted to

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identify these unrated steel hatches in a future FSAR amendment. This

item is considered open penaing submittal by the applicant of an FSAR

amendment addressing this oeviation (445/8722-0-02).

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Section D.4 (f) of Appendix A to BTP APCSB 9.5-1 states that "Stairwells,

elevators and chutes should be enclosed in casonry towers with minitrum

fire rating of three hours...." In Acenament 65 to the FSAR, the

applicant identified as a ceviation that stairwells providing access and

egress rcutes to areas containing safe shutdown equipment were provided with

two hour rated barriers. Due to the negligible combustible leading inside

stairwells and the lack of safe shutdown equipment being separated by the

stairwell walls, the inspector founo nc major issues with applicant's

stairwell boundaries. Acceptance of the deviatio1 frcm Section D.4(f) of

Appendix A to BTP APCSB 9.5-1 will be addressed ,y the staff in their

review of Amendment 65 to the FSAR.

A number of stairwell walls were identified during the inspection

where the inspector considered the justification was not adequate to

support two hcur rated construction. The applicant presented an evaluation

which was cenducted to determine the rating of fire area and stairwell

tcundaries. This evaluation was used to justify the fire rating of those,

boundaries which were not built specifically to the specifications cf an

indepencent testing organization. Where specific installation criteria

of a recognized approval E.gency was not followed, the evaluation was used

to determine if criteria were cet or exceeded in such items as wall

thickness and material type. The inspector identified six stairwell

walls that could not be directly related to the installation criteria

established by a recognized approval agency. The applicant has comitted

to take actions to resolve this issue. Pending actions taken by the

applicant to resolve this issue and NRC review ct those actions, this

item is considered unresolved (445/8722-u-01).

Appendix A to APCSB 9.5-1 Section D.1.(j) states that "Penetrations in

fire barriers, including conduits and piping, should be sealed or

closed to provide a fire resistive rating at least equal to that of the

fire barrier itself. Door openings should be protected with equivalent

ratea coor frcmes and hardware that have been tested ano approved by a

nationally recognized laboratory." During the inspection, the inspector

expressed concern that the method of sealing conduits four inches in

diameter and smaller was not in accordar.ce with rated configurations and

had not been identified as a deviation from staff guidance. The

applicant stated that conduits with either suppression or aetection on

both sides of the penetration would only be sealed on one siae while

conduits with no detection or suppression en at least one side would be

sealed on both s1ces at the first opening. The inspector was ccncerned

that this plan would allow for only one seal outside of the barrier in

locations where their was only detection en both sides of the barrier

with no suppression on either side. The applicant agreed to revise their

! position and committed to seal conduits four inches and smaller on both

i sides at the first opening regardless of the presence of detection or

l suppression. This item is considered open pending the completion of the

seal installaticn (445/8722-0-03).

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In NRC Inspection Report 50-445/85-16; 446/85-13 concerns were raised that

certain BISCO seals used at the plant may not have adequate documentattun

to justify the rating of tha seal. Specifically, American Nuclear Insurers

(ANI) had identified a seal being used by BISCO which had failed a fire

test. During this inspection the inspector reviewed dccumentation

presented by the applicant which demonstrated that the BISCO seals being

installed at the plant were acccmpanied by documentation which demonstrated

that the seals had passed fire tests. The inspector found the

uocumentation acceptable and therefore Unresolved Items 445/8516-U-06,

446/8513-U-06 and 445/8516-U-07, 446/8513-U-07 are therefore closed.

During this inspection a nunber of modifications to fire doors,

primarily for security hardware, were observed. Although the doors and

frames contained labels which demonstrated compliance with testing

criteria of Underwriter's Laboratory, the inspector was concerned that

these modifications would degrade the perfurmance of the door under

fire conditions. The applicant presanted documentation from Underwriter's

Laboratory concerning how security modifications could be made without

jeopardizing the ratina of the door. Hcwever, these guidelines may not have

been implemented during modification of the plant fire doors. The applicant

committed to review all fire donrs presently installed to determine if .

modifications comply with guidance provided by Underwriter's Laboratory.

Where compliance cannot be established, the applicant committed to bring the

dcor into compliance or replace the door with one that conforms to the

guidelines. The applicant also committed to ensure that all future

modifications will conform to the guidance established by Underwriter's

Labora tory. This item is considered open pending the completicn of

applicant's review of this issue (445/8722-0-04).

SSER 12 addressed a number of ceviaticns dealinn with heating, ventilation

and air conditioning (HVAC) penetrations of fire rated barriers. Due to

demonstrated difficulties in the operation of these dampers under air

flow concitions, the applicant has instituted a prograra to completely

change out the campers with the exception of those dampers remaining in

stairwells. The previously approved deviation associated with the

remaining dampers still applies since they cannot be mounted completely

inside the barrier due to interference with tornado pressure relief

dampers. The fire dampers protrude approximately two inches and are

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covered with a one hour rated fire resistive material. Combustible

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icading on both sides of the stairwell dampers is low. The inspector

l confirmed there is reasonable assurance that these dampers would prevent

i the propagatien of fire from cne side of the barrier to the other since

the dampers are essentially in the barrier anc would function normally.

4.2 Fire Protection of the Safe Shutdown Capability

l During the 84-44 inspection, the redundant pressurizer transfctmers

l located in the Safeguards Building were found not to be in compliance

! with the separation criteria of Section III.G.2 of Appenoix R to

l to 10 CFR 50. The applicant stated durina this inspection that, based

l on Fire Separation Calculation 152, Rev. 3, and Westinghouse's Thermal

Hydraulic Analysis (WCAP #11331), the pressurizer transformers are no

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longer required to achieve safe plant shutdown. The inspector reviewed

the analysis ano found it acceptable. Therefore, open item 445/3444-0-05

is considered closed.

By letter of October 2. 1987 the applicant identified an additional

deviation to Section III.G.2.d of Appendix R for the Residual Heat

Removal inlet isolation valves because the redundant valves are within

the same fire area and are not protected with automatic suppression.

One set of redundant valves are within 20 feet of each other. Valves

1-8701A and 1-87018 are located in the corridor outside of the

steam g2nerator compartment, fire zone 1018. Valves 1-8702A and

1-87028 are located within the steam generator compartment, fire zone

101C. The valves in the corridor are separated by approximately 40

fee t. Irtervening combustibles consist of three cable trays which do

not run directly between the valves. The valves inside the compartment

are separated by approximately six feet; however, a partial height

concrete wall extends from just below the valve bonnet up several

elevaticns. Thermistor strip heat detection is provided in both

zones containing the valves. Combustible loading inside the

containment is 34,200 BTU / square feet, comprised mainly of reactor

coolant pump lubrication oil. All four pumps are provided with oil ,

collection systems.

The inspector was concerned that a tire in containment could spread

between redundant RHR inlet isolation valves and effect the ability of

the plant to safely shutdown. However, the combustible loading inside

the containment is low. Due to the large volurte, any fires that were to

occur, would develop slowly and dissipate its heat due to the large air

volume. In addition, detection is provided in both zones containing

the redundant valves. The detection which alarms in the control room

would elert the operators to a fire in the area of the valves who in

turn could have the plant fire brigade respond. Also, since access to

the containment is restricted during plant operaticn, it is unlikely

that transient combustibles or ignition scurces would be introduced

into the area. Based on the above, the inspector determineo it would be

unlikely that a fire cculd occur in the containtrent that would disable the

redundant valves in both sets of RHR inlet isolation valves. Acceptance of

the deviation frcm Section Ill.G.2.d of Appendix R to 10 CFR 50 will be

addressed by the staff in their review of the applicant's October 2,1987

letter.

During the inspection, two adjacent manholes were found which provided

eccess to service water purrp power and control ccbles. At the time of

the inspection, both manhole covers were removea for maintenance

reasons. The inspector was concerned that a flamable liquia spill

and subsequent fire et the same time both covers were retcoved could

jeopardize redundant trains of safe shutdown cables. The concern was

heightened when it was observed that the manholes were approximately 40

feet from the unloading area for erwrgency diesel fuel oil and could be

cirectly adjacent to the path that tanker trucks would travel to the

unloading station. It was also observed that a minimal grarie existed

that would direct the flow of flammable liquios away from the manholes.

The manhole covers were of substantial steel construction and when in

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place, provided an environmentally tight cover. The applicant had

performed an evaluation to demonstrate that the manhole covers would

provide a barrier equivalent to three hours. Hcwever, the applicant did

not address the flamable liquids issue. During the inspection, the

applicant committaa to administratively control the manhole covers to

ensure that only one cover is removed at any time during plant cperation.

In addition, a procecure change was presented to the inspection team

which called from the operations department to ensure that the manhole

covers were in place during diesel fuel unloading operations, lhis

resulution is found to be satisfactory to ensure the integrity of both

trains of service water pump cables.

In SSER 12 the staff approved a deviation from Section III.G.2 of

Appendix R to 10 CFR 50 for lack of one hour separation between

redundant service water pumps. By letter dated October 2, 1967 the

applicant requested that this deviation request be expanded to include

the service water isolation valves, service water recirculation

valves, branch circuits, exhaust fans and branch circuit MCCs. The

previous deviation was granted based on negligible combustible loading,

and the presence of early warning smoke detection and area wide

auton.atic suppression. Based on inspection of the area in question, the .

inspector determined that previous ccnclusions for granting the deviation

appear to remain valid. Acceptance of the deviation from Section III.G.2

of Appendix R to 10 CFR 50 will be addressed by the staff in their review

of the applicant's October 2, 1987 letter.

4.3 Lightino and Communication

SSER 12 stated that "emergency lighting will be installed in all areas

of the plant that may have to be manned for safe shutdown operations

and at access and egress routes to and from all areas." During the

84-44 inspection, a number of lights, were found misaligned and some

areas requiring safe shutdown operations were found not to have emergency

lights (445/8444-0-04). During the inspection, the applicant presented

procedures that were designed to ensure the proper alignment of emergency

lights. While a number of lights were observed to be niisaligned, the

applicant stated that due to the present ccnstruction status of the plant,

it was difficult to maintain the lights in alignment. However, the

applicant stated that a complete alignment of lights would be performed

prior to operation and then routinely thereafter. The applicant also

presented a procedure for identifying locations requiring emergency lights.

The areas icentified in the 84-44 inspection as lacking lights had been

provided with lights and therefore open item 4A5/8444-6 44 is considered

closed. New areas requiring lights haa been identified by the applicant

resulting from changes in the safe shutdown analyses. As noted in

Section 6.1.2 of this report, areas were identified by inspectors where

additional emergency lights may be required. Pending completion of TV

Electric's evalu6 tion identifying locations reauiring additional lights,

l including resolution of the emergency lighting issues discussed in

Section 6.1.2 of this report, this item is considered open (455/8722-0-05).

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At the time of the 84-44 inspection, plant procedures identified the

"Gaitronics" page system 6s the method for notifying fire bricade and

other emergency respcnse personnel. The inspection teen was concerned

that a control room fire would disable the page thereby leaving no

emergency communications system (0 pen Item 445/8444-0-01, item 2). During

this inspection the applicant provided details of a recently installed

raoio system that would provide communications independent of the control

room. Therefore, the concerns raised during the 84-44 inspection have

been resolved and Open Item 8444-0-01, item 2, is considered closed.

During a review of the raoio system, it was noted that the radio system

may be disabled by a fire in certain plant areas. Additienally, a fire

in the same area may require nianual operator actions in the ficid;

therefore, leaving the plant page as the only method for operator-control

roon connunications. The inspector was concerned that since some of these

manual operatiens involved regulating flows, the proximity of plant pages

did not lend this system for adequate communications for this type of

operation. In order to adcress the inspector's ccncerns, the applicant

simulated these manual operations utilizing the page as the method of

communications from the control recm to the operator in the field. Even

with the assumption that the pages nearest the valves were inoperable,

the applicant den'onstrated that the page would provide an adequate means

of conynunication fcr these manual operations in the event the radio system

was disabled.

4.4 Fire Detection and Suppression

4.4.1 Fire Detection

Section E.1 of Appendix A to APCSB 9.5-1 provides the minimum require-

ments for fire detection systems. Detection systems should comply with

flFPA 72D, "Standard for the Installation, Maintenance and Use of

Proprietary Protective Signaling Systems." tiFPA 72D requires that fire

alarm control panels be listed or approved for the purpose for which they

are intended. During the 84-44 inspection, it was observed that the fire

alarm panels used in the plant were not listed or approved in accordance

with NFPA 720 (0 pen Item 445/8444-0-06). To address this issue, an alarm

panel, originally designated for training, was provided by the applicant

to Factory Mutual for testing. Factory Mutual performed the same series

of tests cn this panel that are used to approve coninercial systems.

During this inspection the applicant presented a report from Factory

Mutual to the inspectiun team which documented approval of the plant

fire alarm panels. The report was reviewed and fcund acceptable.

lherefore, Open Iten 445/8444-0-06 is considered closed.

NFPA 72D indicates that detector placerrent should be in accordance with

NFPA 72E which provides guidance on the location and spacing of

detectors. During the inspection the inspector was concerned that early

warning smoke detectors may not be located in accordance with flFPA 72E.

lhe applicant presented an evaluation in which each plant area was

reviewed for compliance with t1FPA 72E. As a result of this review,

a number of plant Ueas had been identified where additieral detectors

were required. Although many of these areas had not yet had the new

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detectors installed, the applicant prest.nted documentation which was

established to track the new installations. Some areas were identified

by the applicant that were not in strict compliance to NFPA 72E. For these

areas, TV Electric presented evaluations allowing for deviations from NFPA

due to low combustible loading and the lack of safe shutdcwn requirements.

The inspector reviewed these evaluations and found no issues.

4.4.2 Fire Protection Water Supply System

As a result of problems with microbiological induced corrosion (MIC) fr

the fire water piping, the applicant is planning to replace the current

lake fire water supply with dedicated fire water tanks. This

n.odificatien will include adding redundant 504,000 gallons storage

tanks and three 50 percent capacity fire pumps (2000 gpm, 160 psi).

Two of the pumps will.be diesel driven and the third will be electric.

The new design was reviewed during the inspection and found to comply

with the guidance as outlined in Section E.2 of Appendix A to BTP APCSB

o.5-1 "Fire Protection Vater Supply Systems."

4.4.3 jp_rinkler and Standpipe Systems

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Section E.3.(c) of Appendix A to BTP APCSB 9.5-1 states that "Automatic

sprinkler systems should as a minimum conform to requirements of

appropriate standards such as NFPA 13 Standard for the Installation of

Sprinkler Systems." During the 84-44 inspection, a number of sprinkler

systems in the plant were found that did not conform to the requirements

of NFPA 13 (0 pen Item 445/8444-0-07). Specifically, sprinkler spacina

exceeded the maximum requirements for distance from the ceiling. As a

result of this open item, the applicant perforu d a review of all of the

installed sprinkler systems against the requirements of NFPA 13. This

review identified a numt'er of areas where sprinkler installation was in

conflict with the code. These areas were then tddressed by a major

retrofit program to bring all sprinkler systems in compliance with NFPA

13. During this inspection the sprinkler installations were reviewed

for compliance with NFPA 13. All areas reviewed were found to be in

compliance with NFPA 13. Therefore, Open Item 445/8444-0-07 is considered

closed.

NRC IE Information Netice 83-41 discusses cases in which inadvertent

actuations of fire suppression systems had adversely affected the

operability of safety related equipment. The inspector was concerned

during the inspection whether the applicant had adequately addressed

this issue. The applicant presented an evaluation in which safety

related equipment had been walked down to ensure that the placerent

of fire suppression systems would not effect the operation of the

safety systems in the event the fire protection systems were to

operate. The inspector reviewed the evaluation and determined that it

adequately addressed the issue of fire protection systems adversely

affecting safety related systems,

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4.4.4 Halon Suppression Systems

Section E.4 of Appendix A to BTP APCSB 9.5-1 states that "The use of

Halon fire extinauishing agents should as a minimum con. ply with the

requirements of NFPA 12A and 128, Halogenated Fire Extinguishing Agent

Systems - Halen 1301 ano halen 1211." During this inspection, the

inspector was concerned that the Palon system provided in the Cable

Spreading Room may not be in compliance with NFPA 12A. It'e applicant

indicated that the review of the system against the requirements of

NFPA 12A had not been performed. Therefore, the applicant needs to

perform a review of the Cable Spreading Room Halon system against the

requiren,ents of NFPA 12A. Any deviations identified in this review will

be required to be submitted to the stafi for evaluetion. The NRC

considers this item open pending applicant completion of the eveluation

and NRC review of the results (445/8722-0-06).

5.0 POST FIRE SAFE SHUTC0WN CAPABILITY

During the 84-44 inspection, numerous apparent inconsistencies were noted

in the applicant's ar,alysis and assumptions concerning the protection of

fire safe shutdown equipn.ent for areas outside of the control room and ,

cable spreading room where alternative safe shutdown is not required.

Since the 84-44 inspection, the applicant has provided a more ccmprehensive

methcdology and analysis in two docurrents, the Fire Safe Shutdewn Design

Basis Document (DBD), DBD-ME-020, and the Fire Protection Report (FPR).

The Fire Hazards Arelysis Report (FHAR) [Ref. Appendix A, A.1(b)] which

is contained within the FPR, describes each fire area and its associated

fire protection features. The fire safe shutdown equipment lccated within

an area is listed in the Fire Safe Shutdown Analysis Repcrt (FSSAR) [Ref.

Appendix A, A.1(c)] also contained within the FPR. For each fire area

which contains safe shutdown components, the reference to the components

protected to achieve safe shutdown is typically a ceneral statement:

"One train of the required redundant equiptrent and components within the

4rea is protected by one of the means provided in Section II.4.5."

Section II.4.5 contains only a listing of all of the potend al means

of complying with CMEB 9.5.1 C.S.b separation requirements. Therefore,

the FHAR does not identify specifically what components are protected for

a postulated fire in that area, except in certain circumstances such as

for Fire Area AA where the protection of CCW isolation valves 1HV4512,

1HV4513,1HV4514, and 1HV4515 and their associated circuits is described.

The listing of protected components for each fire area is provided in

three volume docurrent collectively referred to as Calculation No.152,

Revision 3 [Ref. Appendix A. A.3]. Calculation No. 152 is predoninently

a computer printout for each fire area of the raceways, the safe shutdewn

cables, the cables which mu;t be thermolagged in the area, the

corresponoing safe shutdown o? vices and associated equipnent locaticn

(fire zones of the devices), tea electrical nodes (junction boxes) and

the raceaay length. A discussion of protection of associated circuits is

provided in Section 7 of this repcrt.

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From the perspective of mechanical systems operability, Calculation

No. 152 provides two tables in Attachment 16 of Volume 3: Table 1

"Fire Area Compliance Table" and Table 2 "Operator Actions for Fire

Areas." Table 1 summarizes the compliance trethod for separation for

each fire arca, but in the inspectors opinion does not provice a clear

path for determining equipment to be protected. Table ? is a listing of

safe shutdown devices and location by fire zone which require certain

operator actions including repairs, the location of the action, and the

affected fire areas where a fire in those areas tray create a requirement

for the manual action. Also the actions are classified acccrding to

whether they are required for hot shutdown (hot standby) or cold shutdown.

The inspection team noted that Table 2 is a key document in the applicant's

justification for compliance with separation requirements for those areas

nct requiring alternative shutdown. The basis of the applicant's analysis

ano protection methodology for these areas is a combination of protecting

certain components in a given fire area, in n.any instances of either

redundant train, plus reliance on the local operator actions describeo

in Table 2.

The following procedures [Refs. App. A, B1 to 8] in addition to ,

Procedure No. ABh-803A, "Response to a Fire in the Control Room or

Cable Spredding Room," have been prepared by the applicant to address

manual actions:

ABN-804A "Response to Fire in the Safeguards Building"

ABN-805A "Response to Fire in the Auxiliary Building or the

Fuel Building"

ABN-806A "Response to Fire in the Electrical and Cont J

Building"

ABN-807A "Response to Fire in the Containment Building"

AbH-808A "Respense to Fire in Service Water Intake Structure"

ABN-809A "Response to Fire in the Turbine Building"

In view of the tranual actioris required to ensure compliance with

separation requiren,ents, the team considers the above procedures to be an

integral part of the applicant's fire hazards analysis and fire safe shut-

down analysis reports. The team considered it of considerable importance

that the feasibility of the manual actions be properly analyzed with

respect to the postulated fires and the protected components within each

fire area. As a minirrum, the manu61 actions should be sorted so that

those which neeo to be perfortred in the same fire area or zone in response

to a postulated fire in that area or zcne are identified and the time after

reactor trip when the action must be perfonned cerrpared to the area acces-

sibility and corrponent operability after the postulated fire.

During the inspection, the team noted that the information in Table 2

concerning the manual actions was not adequately sor"d to identify

actions which must be taken in the sarre fire area as L,2 postulated fire.

Furthermore, the teasibility of each action with respect to the

postulated fire was not presented. The applicant presented a revised

listing of the manual actions with justifications for each acticn just

prior to the exit r.ceting. The list indicated that some revisions to

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.

Table 2 were necessary and that some actions had been deleted. The new

listing of actions would be presented in a previously planned Revision 4

of Calculation No. 152.

The issue of the adequacy of manual actions which must be taken in the san;e

area as the postulated

to Calculation ho.152 andfire remains

NRC reviewunresolved pending(TV

of the document Electric's revision

445/8722-U-02).

6.0 ALTERNATE SHUTDOWN

6.1 Procedures

During the 84-44 inspection, the inspection team noted that procedures for

alternate shutdcwn were preliminary and incomplete. During this inspection,

the inspectors found that procedures for alternate shutdown had been

prepared. The inspection team's evaluation concentrated en Procedure

ABN-803A, "Response to a Fire in the Control Rocm or Cable Spreading Recm,"

Revisicn 0 dated June 16, 1987, with Procedure Change Notices ABN-603A-R0-1

dated July 30, 1987 and ABN-803A-R0-2 dated October 9, 1987. Procedure

ABN-bO3A is based primarily on the previously referenced Calculation No.

152, Revision 3, and a Westinghouse document, WCAP-11331 * Comanche Peak

Steam Electric Station Thermal / Hydraulic Analysis of Fire Safe Shutdown

Scenario" dated October 30,1986 (Ref. Appendix A, A-5) which was prepared

tc deironstrate the ability to achieve safe shutdown conditions following

a Control Room or Cable Spreading Poem fire. WCAP-11331 ccapares baseline

assumptions for the Appendix R,Section III.L conditions against the

effects of single spurious sincles on safe shutdown capability. The

results of the review and walkdown of procedure ABN-803A are as follows.

6.1.1 Procedure Review

The procedure is organized into a main text with four (4) major

attachn:cnts to achieve hot shutdcwn. The main text is implemented

primarily by the Shift Supervisor in the hot shutdown phase.

Attachment 1 is entitled, "Reactor Operator Actions to Achieve Hot

Shutcown," Attachment 2 "Relief Reacter Operator Actions to Achieve Hot

Shutdewn," Attachment 3 "Auxiliary Operator No. 1 Actions to Achieve

Hot Shutdown" and Attachment 4 "Auxiliary Operator No. 2 Actions to

Achieve Hot Shutdown." Thus, there are five (5) operating staff

members requireo to implement the hot shutdown phase. Attachment 13

"Operator Action Timeliness," provides a summary of the key operator

actions and the required completion tirres for attachments 1 thrcugh 4.

The WCAP previously referenced is intended to ensure that given any

spurious signal, the completion times are such that safe shutdown can

be acccmplished.

The following items were noteo during the procedural review. Most of

these concerns were resolved through the issuance of Procedural Change

Notice (PCN) ABN-803A-R0-3 dated October 21, 1987:

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1. There was no provision for termination of spurious pressurizer

(PZR) heater operation. PCN ABN-803A-R0-3 contained a change to

the procedure that resolved this concern.

2. Upon a spurious safety injection signal, the WCAP indicates that

the rupture disk on the PZR relief tank (PRT) would burst after 52

minutes even though all operator actions would be completed

normally as required by the procedure.

This concern is made n: ore serious considering that multiple operator

actions may be or are required inside containment during the hot

shutdown phase, i.e., manually opening 1-HV-8112, the seal water return

isolation valve, and to manually close accumulator injection isolaticn

valves 1-8808A, 1-88088, 1-8808C, and 1-88080. These actions,

steps 2.4(d) and 2.4(t), would take place after 2-hour maintenance

of hot stanoby conditions. This ccncern is further discussed in

Section 6.1.3 below ano is identified there as an unresolved item.

3. An alternative step to manually cpening 1-HV-8112 as mentioned in

(2) above (by observing that seal water return flow is available from

outside the containment) was not provided. This ccncern was resolved

by PCN ABN-803A-R0-3 which directs the operator to check that the

seal water return filter delta pressure is greater than 0 psi by

observing the difference between 1-PI-175 and 1-PI-176. The

applicant agreed to consider the use of a portable delta pressure

gauge which could be installed if required by oscillation of the

gauge needles.

4. The procedure did not specifically address restoration of offsite

power at any time during the procedure impien.entation. The applicant

indicated that this action was handled by Procedure No. ABN-601A,

"Response to a 138/345 KV System Malfunction."

5. The prccedure did r.ot detail the steps required to manually

operate the steam generator atrrespheric PORVs. By means of PCN

ABN-603A-R0-3, caution statements were added concerning the safety

actions for the operators to follow such as wearing eye and hearing

protection and donning a steam suit.

6. Step 2.3.e calls for the reactor operator to perform several

operator actions prior to evacuating the control room, one of

which is to place both RHR punips in the PULL-TO-LOCK position. All

of the actions are verified by the reactor operator in attachment

1 except for the step involving the RHR pumps. This concern was

resolved by FCN ABN-803A-R0-3.

7. There was no reference in the main test of the procedure to Attachments 7

and 8 which list the controls and instrumentation available at the

remote shutdown panel. By means of PCN ABN-803A-R0-3, such a reference

was included in step 2.4.b.

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8. There was no provision in Attachment 1 for the reactor operator to

notify the relief reactor operator, who is starting diesel

generator A in attachment 2, in case of failure of service water

flow to the diesel. PCN ABN-803A-R0-3 calls for an additicnal note in

attachment I to cover this situatien. Tne applicant also provided

an entry from the operators log book (App. A, A 5) showing thst

the same diesel had been run unloaded for over 60 minutes without

service water flow.

The:a were other items which were substantially eoitorial in rature to

reonce the probability of operator error which were also resolved by

the PCN ABN-803A-R0-3,

6.1.2 Procedure Walkdown

A procedural walkdown of ABN-803A was conducted with one NRC representative

each following an assigned applicant operating staff member. There are

five operators requirea to implement the procedure: the Shift Supervisor,

' Reactor Operator, Relief Reactor Operator, Auxiliary Operator No.1,

and Auxiliary Operdtor No. 2. The walkdchn was conoucted wit:1 the

additional condition that the fire brigade would be called out simul- .

taneously to simulate a control room fire. The walkown ended once hot

standby conditions had been achieved.

The team was generally impressed with the organization of the procedure

and the operators' ability to carry it cut. However, one minor concern

was identified by the inspectors. The procedure did not direct the shift

supervisor to assist the reactor operator in tracking the progress of the

other operators in accomplishing their tasks within the time linits shown

in the Operator Action Timeliness in attachnent 13 of ABN-603A. By means

of the previously referenced PCl, on appropriate note was added to the

procedure, so that thi: item .s considered resolved.

The portions of the precedure applying to the time after hot standby

conditions have been acnieved, which involved either manual actions

inside containment or repairs to achieve cold shutdown, were separately

walked down.

For the actions inside containment, the scenario of coincident loss of

offsite power with evacuation of the control room results in the

inability to monitor the conditions insioe the containment. Therefore,

the operators must wear full respiratory gear including Scott air

packs, 4 A:h can limit the optrator's mobility and access in certain

!

2rus.

Th. 4+v ob: > '. hat Step 2.a.d of APN-803A, which required the

ct , ,

s ..en 1-hV-81:2, the seal water return isolation

ye 1s . + ~ -

r, nted reascelably well with the airpacks mounted.

h.,ie . . " x~w, to be no 8-hour battery pack emergency lighting

i r. P -~. ,dr step inside containment, step 2.4.t. to manually

close D1 .... v alator isolation valves 1-8608A,1-ESCCB,1-8808C,

Jd 1-88080, t r.ae censuming since it may require as much as 20

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minutes per valve. It was noted that access to valve 1-880GD was

difficult, but not impossible, with the full respiratory gear on. Also

there did not appear to be any 8-hour e-~rgency lighting in area rear

accun.ulator No. 4 The need for TV Electric to complete their as!,esscent

of locations where en.ergency lighting is needed is addressed in f,ection

4.3 of this report.

Regarding the section in the procedure invc1ving repairs, attachment 6

for the emergency air supply hookup to RHR valves 1-FCV-618 and 1-HCV-606

referenced actions to close instrument air valves 101-650 and 101-651

which were difficult to focate and poorly labeled. There also did not

appear to be 8-hour emergency lights in the area. The need for TUEC to

complete their assessment of locations where emergency lighting u needed

is adoressed in Section 4.3 of this report. The issue of the poorly

labeled valves is considered an open item pending further revie~v by the

staff (445/8722-0-07).

Attachnent 5 of ABN-803A does not involve repairs but rather manual

closure of valves IPS-100, 1PS-100, 1PS-113, IPS-126, and 1PS-139 for

steam generators 1 throta;h 4 sample line isolation. These valvas were

extremely difficult to locate amongst all of the other valves in the *

safeguards 810 primary sample room. It was also difficult to locate

CVCS valves 1 CS-8453, 105-6455, 1CS-8430, and ICS-E444 in the Auxiliary

Buildino 822 Blender Room. All of these locations were acceptably

clarified by the PCN previously referenced.

6.1.3 WCAP-11331 "CPSES Thermal / Hydraulic Analysis of Fire Safe

Shutcown Scenar.o"

As previously mentioned, WCAP *1331 was prepared to analyze the plant's

ability to achieve safe shutdcwn following a cont.ol roor or cable

spreading room fire by evaluating certain spuriots operation cases as

scositivity studies to a t,aseline scenario. The '.hermal/ hydraulic

analysis described in kCAP-11331 was generateo using the TREAT (Transient

Real Tice Engineering Analysis Tool) computer code. Use of this code was

approved by the MRC staff for the South Texas Plant in NUREG-0781,

Supplement No. 3, May 1987, for small-break LOCA analp es, but not for

Comanche Peak.

The spurious operation scenarios analyzed in WCAP-11331 ware:

1 Stuck Open Presm rizer PORV

2 Stuck Open Stedm Generator PORV(s)

3 Spurious Hecd Vent Operation

4) Auxiliary Feeowster System Misalignment

5) Spuricus SI System Operation

6) Main Feedwater and Turbine Do Not Trip at Reactor Trip

7) Bac.up Heaters fail On

All of the above cases were corpared to the operator actions described

in Procedure ABN-S03A. The only concerns noted were for case (5), the

spurious SI system operation. The WCAP re#ers on page 6 to

calculaticns that were performed to datermine whether or not the PRT

,

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rupture aisk would rupture. The calculations are stated to show that

rupture woula occur approximately 52 minutes following transient

initiation, with the release of approxirrately 11400 lbm of steam priur

to initiation of normal seal injection return flow at 90 minutes in

operational guidelines. It should be noted that these calculations are

not actually provided in the WCAP.

The rupture at 52 minutes would occur well before the operator actions

could be taken inside containment to n.anually open the seal water return

isolation valve 1HV-8112 and to rr.anually close the accurculator isolation

valves 1-8808 A/8/C/D (see discussion in Section 6.1.2 of this report).

The applicant attempted to adoress-the cencern raisto by the tearr

regarding the feasibility of the ruanual actions inside containment by

preparing, during the inspection, a calculation (ref. Appendix A, A.6)

intended to show that tirne for rupture of the FRT rupture was overly

censervative ano that the rupture disk would not burst at all. The team

did not bavt time to review this calculatien as it was presented on the

esening prior to the exit meeting and because the actual Westinghouse

c61culations are not proviaed in the WCAP. This it'm remains unresolved

pending the NRC review of the calculation (445/d72<-U-03). ,

6.2 _ Alternative Shntdewn Instrunientation

10 CFR 50, Appendix R, Ill.G.3 and III.L states, that, if the licensee

elects tc ntablish alternative safe shutdown capability, provisions

need to be providea for direct readings of process variables necesf ary to

perforni and control the reactor shutdown function. NRC Information

Notice 84-09 states that instrumentation be supplied to provide the fol-

lowing information:

. Pressurizer Pressu.e and Level

. Reactor Coolant Hot Leg Temperature -T hat

.

F.eactor Ccolaat Cold Leg Temperature - Teold or T,y

. Steam Generator Pressure and Level (wide ranga)

. Source Range Flux Monitor

. Level Indication for All Tanks Used During the

Shutdown Process

. Diagnostic !rstrumentation for Shutdown Systems

TV flectric has installed a remate shutdown panel which is located in the

Electrical Eauipment Area, Fire Zone SE16, Safeguards Building on Elev.

,

831'-6". The inspector found that the panel provices the capebility to

bring the plant to cold shutdown utilizing either Train A or Train B

equipment.

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The followinc instrumentation is available on the hot shutduwn panel:

. Steam Generator 1 Wide Range Level - 1LIS01A

. Steam Generato,' 2 Wid: Range Level - 1LI502A

. Steam Generator i Pressure - ILI5148

. Ste m Generatur 2 Pressure - 1LI5245

. Pressurizer Level - lL14598

. Pressurizer Pressure (liR) - 1P!455B

. Source Range Detector - INI31F

. RCS Loop 1 Het Leg TemperMure - 11R413F

. RCS Loop 1 Cold Leg Temperature _ ITR410F

. RCS Loop 2 Hot leg Temperature - IT!423F

. RCS Loop 2 Cold Leg Temperature - 1Tla20F

. RCS Loop 3 Hot Leg Temperature - 1TI433F

. RCS Loop 3 Cold Leg Tamerperature - ITE430F

. RCS Loop 4 Hot Leg Temperature - ITR443F

. RCS Loop 4 Cold Leg Temperature - ITR440F

. Condensate Storage Tank Level - 1LI2478B

The refueling water storage tank level indicaticn will be available

locally at the tank. ,

The above instrumentation is dedi. :ed to the Train A hot shutdcwn

panel and which is installed in areas outside of the control room,

where it is not subject to damage as the result of a control room fire.

The instrun.ents are serviced by oedicated power supplies which are

located at the shutdewn transfer panel. The cables for these

instrurents do not enter the control room and consequently are not

subiect to damage due to a fire in the control room.

The inspector determined that the instrumentation av provided met the

guioance in NRC Information Notice 84-09.

6.3 Hot Shutdown Panel

The hot snutdown panel contains instrumentation and controls for both

Train A and Train B components. Train A controls are isolated from the

control rocm by switches at the shutdown transfer panel rn Elevatie

810'-6" in the electric ecuipment aren, fire zone 509. The Troin d

isolation switchee are lo;ated at *,he he t shutdown panel . A fire at

the hot shutdown viel could damage both Train A and Train B cor.trols

loc &ted on the pan 1. However, due co the rem 3te location of shut < % n

transfer panel, Train A control will be available in the control roorq.

. The major shutdown devices which are operable for alterative safe

shutoown at the hot shutdown panel are as follow :

Main Steam Isolation Valves IHV2333A

1HV2334A 1HV2.'..m

1HV2336A

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Main Sten Isolation Bypass Valves 1HV2333S

1HV23348 1HV2335B

1HV23368

Turbine Driven Auxiliary Feedwater Pump

Motor Driven Auxiliary Feedwater Pump 1

Motor Driven Auxiliary Feedwater Pump 2

Steam Generator 1 PORY IPV2325

Steam Generator 2 PORV 1PV2326

Steam Generator 3 PORV 1PV2327

Steam Generator 4 PORV 1PV2328

Service Water Pump 1

Service Water Pump 2

Diesel Genarator i

Diesel Generator 2

Centrifugal Charging Pump 1

Centrifugal Charging Pump 2

Pressurizer Level Centrol Valve 1FCF121

Letdown Isolation Yalve ILCV459

Letdown Isolation Valve ILCV460

Letdown Orifice Isolation Valve 1-0149A

Letdown Orifice Isolation Valve 1-8149B '

Letdown Orifice Isolation Volve 1-8149C

Control Room Manual Reactor Trip

Backup Hedter Group A

Backup heater Group 8

Backup Heater Group C

Pressurizer Block Valve 1-8000A

Pressurizer Block Valve 1-80008

Ccaponent Cooling kater Pcmp 1

Component Cooling Water Pump 2

PHR Pump 1

RHR Pump 2

Accumulator Isolation Valve 1-8808A

Accumulater Isolation Valve 1-88088

Accumulator Isolation Valve 1-88C8C

Accumulator Isolation Valve 1-88080

Charging Pump Isolation Valve 1-8105

Charging Pump Isolation Valve 1-8106

Pressurizer Ptmp PCV455A LPCV455A

The applicant has developed modifications which will enable lccal operation

of the diesel generators. These are the subject of Design Change

Authorization DCA 61447. DCA 61447 was initiated to resolve the

consequences of uncocrdinated 125 VDC circuits EG 104509, EG 145211, ard

EG 13C661. This DCA, when implen.ented, will recuire the installatinn of

branch circuit fuses or the installation cf thermal lag protectier.

Pending completion of the codification and review by the NRC, this item

is considerad open (a45/8722-0-08).

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7.0 PROTECTION FOR ASSOCIATED CIRCUITS

Appendix R.Section III.G, states that protection be provided for

associated circuits that could prevent cperation or cause malcperation of

reduncant trains of systems necessary for safe shutdown. The circuits of

concern are ger.erally associated with safe shutdown circuits in one of three

ways:

. common bus concern

. spurious signal concern, and

, common enclosure concern

The asscciated circuits were evaluated by the team for common bus, spurious

signal, and coramon enclosure concerns. Approximately 250 power, control,

ano instrumentation circuits were examined by the inspectar for potential

problems. This sample size, which represents about 80s sf the safe shutdown

circuits, was used in making the review since many cir:dits were involved

and a determit.ation of cable routing took co ciderable time. The samples

were selected based on the components which he licenses proposed to use

for safe shutdown.

The applicant analysis of protection o 'ssocitted circuits related to safe

shutdown was found to be substar.tielly s ..r.pl e ted . The aralysis resulted

in the need for a nember of modifications, many of which have not been

ecmpleted. One area where a significent amount of work remained to be

done was installation of then90 lag. Until the analysis is completed and

the staff reviews the results, this item is considared open (445/8722-0-09).

The following sections present tha inspectors r(view of the specific areas

of comnon bus concern, spurious rignal concern, common enclosure cr..cern,

and rultiple hign impedence faults.

7,1 Common Bus

.

The licensee had reviewea the class IE and associated circuits in the

plant to ascertain the effects of ccordination or uncoordination on the

plants capability to achieve post fire safe shutdown. It was

demcastrated to the inspector that corrective action since the 84-44

inspection had been taken to correct det'iciencies in the electrical

coordination of safe shutcown circuits. Some of the actions were as

follows:

i . Replace existing fuses with new fuses which ccordinate.

. Provide thermal lag to protect safe shutdown circuits.

. Replace existing trip units with Westinghouse AMP tester units.

. Reanalysis of circuits which compensated by taking into account ,

feeder or cable lengths located in the fire area. ,

,

The team examined, on a sampling basis, the protection fcr several

circuits including coordir,ation of fuses, circuit breakers, end relays.

The samples selected for the cocrdination review were as follows:

,

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. Diesel Generator Source Breake 1EGI for Bus 1EA1

. Component Cooling Pump #1

. Motor Driven Auxiliary Feedwater Pu p #1

~

. Centrifugal Charging Pump #1

. Service Water Pemp ill

. Compartment SF MCC XEB1-1 - 480V AC

, MCC IEB1-2 480V AC SWGR

. Circuit 4 Distribution Panel 1E01-2125V DC

, Circuit 1 Distribution Panel 1E01-2125V DC

. Circuit 2-12 Future en Switchbo:ro IE01 125V Cd

. Circuit 1-6 Spare on Switchboard 1EDI 125V DC -

. IEB.1 o 400V Switchgear

The applicant's review identified some circuits which Wre not coordinated.

Some of these circuits were:

. Panel Boards - 1EC3

1EC3-1

1EC4-1

1EC3-2

1EC4-2

.

An assessrrent of neea of these circuits for safe shutdown was made by the

appl 1 Cant and, with the exception of 1EC3-1, none of the above panels were

required for safe shutdown. The applicant proposed to protect the safe .

shutdcwn circuits on IEC3-1 with thermal leg although the thennal lag had

not yet been installed.

The 480V switchgear 181 and IB? were found by the applicant to be

uncoordinated and a design change was authorized (DCA42381) to replace

the existina trip units with AMPTECTOR Type II A cevices.

Panel IED-1 (125V DC) was determined to be coordinated by including the

added impedance of effected cable located within the limits of the

fire area.

The applicant identified circults which 1.ere not completely analyzed to

account for arr.pacity effects due to the increased operating temperature

resulting frcm the thermal lag wrap. The applicant indicatsd that

Pevision 4 of the CPSES FSSA Calculation i;o. 152 will address this issue

and could result in furthar circuit modifications or thermal las changes.

The comon bus concern cannot be satisfactorily resolved until Revision 4

of the FSSA and the final thermal lag report, ECF-K1700, have been completed.

The issue of ampacity effects due to increased operating temperatures

resulting from thennal lag wrap will remain open subject to completion of

TV Electric's analysis. This item is censiderec part of Open item

(445/8722-0-09).

l

7.2 Spurious Sicnal

The spurious signal concern is made up of 2 fteu :

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. False motor, control, snd instrument indications can occur such as

those encountered during the Brown's Ferry fire. These could be

caused by fire initiated grounds, short er open circuits,

, Spurious operatien of safety related or ocn-safety rela .ed

components

capability can cccur that

(e.g.,RHR/RCS would valves

isolation adversely)

. affect shutd' wn

7.2.1 Current Transformer Secondaries

The applicant had completed the 6.mlysis for the current transformer

secondaries concern, ano had datermined that the voltage centrol and

governor control circuits were protecteo and would be functional in

the local control mode. No adoitional current transformer applications

we e identified which could affect post fire safe shutdown. The

inspector revi e d portions of the applicarts anain s and found it

acceptable.

,

7.2.2 Hich Low Pressure Interfaces

'

The high low pressure inter faces which were identified by the

applicants' analysis are as folicws: -

. Reactor Head and Pressurizer Vent Valves 1-HV3607, 1-HV3608,

leR 3607,1HV3610

RHR/RCS Boundary Isolation Valves 1-8701A, 1-8702A, 1-8701B,

1-87028

. Pressuriter fewer-operated Relief Valves 1PCV455A, IPCV456

. hermal letdchn isolation Valve ILCV459 ano excess letdown

isolation valves 1-8153 and 1-8154

The app?icant presented the following rrethods to preclude any unwanted

spurious actions:

-

Reactor Head and Pressurizer Vent Valves. One valve in either

train of the valves in each path is disabled by disconnecting DC

power.

- RHR/RCS Scundary Isolation Valves. The applicant intends to

remove power from either of the two Path A and B valves by opening

the appropriate circuit breaker.

- Pressurizer Power-operated Pelief Valves. The applicant intends to

close the respective pressurizer block valve (1-C000A, 1-80008) or

disconnect the DC power to the air centrolling solenoid valve for

the P09V. It should be noted the control cables for the block

valvcs (1-8000A, 1-80008) are vulnerable to damage as the result of

a contrcl room fire and that they 3re not equipped with handwheels.

Block Valves 1-8000 A & 8 will be clcsed at their respective MCC's.

The control circuits for the preuurizer PORV's are electrically

isolated from the control room.

23

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,

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- Excess Letdown Valve - 1-8153, 1-8154 Controls for 1-8153

Cdn be faile(i Closed at the hot shutdown panel.

- Normal letdown isolation valve ILCV459. VA 1-81498 can be closed

from the hot shutdown panel.

7.2.3 General Fire Instigated Spurious Signals

The applicant presented analyses for a number of devices which may be

spuriously operated subject to a fire in the control' room, some of

which are as follows:

. Main Steam Isolation Valves (M51Vs) 1HV23333A&B, 1HV2334ASB,

ihV2335A&B, and 1hV2336A&B. These devices can be closed frem

the hot shutdown panel and tre electrically isolated from the

control room.

. Steam Generator PORVs IPV2325 IPV2326, 1PV2327, IPV2328. The

controls for these valves are not electrically isolated from the

control room; their manual operation of the valve will bt the

neans of operation. The appropriateness of these tranual actions

is encornpassed in the previously identified unresolved item -

discussed in Section 5.1 of this report.

. The n6rmal charging isolation valves 1-8105 and 1-8106. These

val es are electrically isolated from the control room and can

be operated from the hot shutdown panel.

. Pressurizer level control valve 1FCV121. This valve can be put

in its fail safe position by either venting through instrument

air or disconnecting the DC power to the valve.

The control circuit for the accumulator isolation valves 1-8802A, B, C and

D are not electrically isolattd from the control room. The applicant

intends to utilize jumpers to niaintain centrol or manually operate the

valves during hot shutdown. The inspection team considers this action a

hot shutdown repair which is not consistent with staff guidelines. Further

informatian is needed to resolve this concern (445/8722-U-04).

7.3 Common Enclosure

The common enclosure concern occurs when nonsafety related cables are

run from one redundart train to another and a fire can thereby endanger

bcth redundant trains.

Seven levels of cable separation are in use at Comanche Peak: U Train A,

orange cable; 2) Train B. green cable; 3) nonsafety, black; 4) protecticn

channel, red; 5) protection channel, whitte; 6) protection channel, blue;

and 7) protection channel, yellow. There is no internixture of any of the

seven levels, and whenever a cable exits a raceway or enclosure, fire stops

or seals are installed. The coccon enclosure concern was founo to be

satisfactorily addressed by the applicant. ,

24

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7.4 .Multipic High Irnpedance Faults

TUEC's analyses for the effect of moltiple high, impedance faults cn post

fire safe shutdown accounted for the surranation of the high irnpedance fault

currents for the affected cables in the fire areas in addition to the

ntaximum operating current for cables outside the fire area. The power

supplies are considered oy TU Electr'c to have failed cue to fire irduced

multiple high irrgeoence faults where the total feeder current exceeds the

long-term trip carrent of the affecttd power supply feeder breaker. The

inspector determin d that separation and protection wnre adequate to prevent

loss of redundant safe shutdown power supplies. Based on the inspector's

review TV Electric's aralysis was censidered acceptable.

8.0 OpEN ITEMS

Open items are matters which have been discussed with the applicant, which

will be reviewed further by the inspector, and which involve sorre action on

the part of the NRC or applicart or both. Open items identified curing the

irispection are discussed in Sections 3 (one item). 4 (five items), 6 (three

items), ard 7 (one item) of this report.

9.0 UNRISOLVED ITEMS -

Unresolved items are matters about which more information is required in

order to determine whether they are acceptable items, violations, or

deviations. Unresolved items identified in this ins

inSections4(cneitem),5(oneitem),6(oneitem)pectionarediscusseo

, and 7 (cne item) of

this report.

10.0 EXIT INTERVIEl (30703)

An exit inte. ,tew w6s conducted on C;tober 23, 1987, with the applicant's

representatives identified in secticn 1 of this report. During this exit

Interview, the scope eno findings of the inspection were sumarized.

!

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APPENDIX A

Documents Reviewed

A. Reports and Correspondence

1. TU Electric - Generating Division, "Comanche Peak

Steam Electric Station - Unit No.1, Fire Protecticn

Report" Nevision 0, September 22, 1987, with the following sections:

a)Section I - Introduction

b'Section II - Flre Hazards Analysis Report (FHAR)

c)Section III - Fire Safe Shutdown Analysis Report (FSSAR)

d)Section IV - Appendices

2. "Texas Utilities Electric - Comanche Peak Engineerino

- CPSES Units I and 2 - Design Basis Document - .

Fire Safe Shutdown Analysis", DB0-ME-020. Revisicn 0,

June 19, 1987.

3. "FSSA Calculation No-152 Revision 3, CPSES. Unit ho. 1 Fire Area

Separation Analysis" Volumes 1, 2 and 3 with transmittal letter 4 May

1987 to Mr. John E. Krechting, TV Electric, from Mr. Elden E. York,

Engineering Planning and Management, Inc.

4. A.F. Gasner; etc. al. 'CPSES - Thermal Hydraulic Analysis of Fire Safe

Shutcown xenario", WCAP-11331 Westinghouse Electric Corp, October 30,

1986.

5. CPSES Operator's Log Entry-Evening Shift May 2, 1984 T. Beandi, Shift

Supervisor.

6. S. Popek, "Spurious Safety Inspection Analysis" Engineering Planning

and Management Calculaticn No. EPM - P257 - 167, October 22, 1987.

7. "CPSES Fire Protection Program - Advance Submittal of FSAR Update" -

Sections ~.4 (Systems Required for Safe Shutdown), 9.5.1 (Fire

Protection Program) and miscellaneous, transmittal letter October 9,

1987 to U.S. WRC Document Control Desk from W. G. Counsil, TV Electric.

8. Thermalav Arp Schedule ECE - M1 - 1700, Revision cpl.

9. Letter from D.P. Barry to Steven D. Einbinder, "Thermalag Cable

Ampacity Study."

[

10. Design Change Authorization ho. 61441, "Coordination of Breakers on

Panels 1 ED1-2 & IED2-2", October 22, 1987.

- A1 - -

  • ' .

,

11. Letter from P.B. Stevens to S. Einbinder "CPSES High Irrpendance Fault

Study."

12. Design Change Authorization No. 42381 Regarding "Change of 400 V Swgr

(181 & 182) Solid State Trip Devices to kestinghouse Amptector Type

IIA" dated June 30, 1987. Letter from 0.W. Lowe to S.L. Starn.

13. Letter from 0. W. Lowe to S. L. Stanun, "0C. Emergency Lighting for The

Control Room," dated October 13, 1987.

14. Coordination Study - 118, 120, 120/240 & 208/120 the Non-Class IE AC

Panel Board Buses - Calculation The-EE. CA-0008-574 Dase-2/26/8 Non

Class IE Buses.

15. Coordinatioa Study - Bus IEA1 - Calculation EE - CA - 0008 - M15.

.16. Coordination Study - 6.9 Kv System - Bus 1EA1 - Calculation The -

EE-CA-0008 - 15M.7 Date 10/7/86.

17. Coordination Study - 6.9 Kv Power Distributicn System - Ground Fault -

Calculation The-EE-CA - 0008 157 Rev. O Fig. 9 Date 10/7/86,

'

18. EFM Calc. - Analysis and Resolution of FSSA Associated Circuits of

Concern by Conmon Power supplies, EPfi - P257-165-000, dated 10/22/87.

19. Coordination Study - 118; 120 & 208/120 VAC Class IE Panel Board Buses.

- The -EE-CA-0008-183 Rev. O Fig. 43 Date 10/3/86.

20. Coordination Study - 118, 120 & 208/120 VAC Class IE Panelboard Buses -

The-EE-CA-0008 - 183. Rev. O Fig. 5.

B. Procedures

Abnormal Conditions Procedures

1. ABN-803A "Response to a Fire in the Control Room or Cable Spreading

Rcom" Rev-0, June 16, 1987 with PCN ABN-803A-RO-1, July 30, 1987, PCN

ABN-803A-RO-2, October 9, 1987, PCN ABN-803A-RO-3, October 21, 1987,

2. ABh-604A "Response to Fire in the Safeguards Building," Rev-0, July 15,

1987.

3. ABN-805A "Response to Fire in the Auxiliary Buildins or the Fuel

Buildir:g or the Fuel Building", Rev-0, July 15,1987 with PCN

ABN-805A-RO-1, October 13, 1987.

4. ABN-807A "Respense to Fire in the Electrical and Control Building,"

Rev-0, July 15, 1967.

5. ABN -807A "Response to Fire in the Containnent Building," Rev-0, July

15, 1987.

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6. ABN-808A "Response to Fire in Service Water Intake Structure," Rev-0,

July 15, 1987.

7. ABN-809A "Response to Fire in the Turbine Buildin9," Pee-0, July 15,

1987.

8. ABN-301A "Instrument Air System Malfuretion" Rev. 2. September 23, 1987

with PCN ABN-301A-R2-1, October 13, 1987.

9. AEN-601A "Response to a 138/345 KY System Malfunction," Rev. 2,

September 11, 1985.

C. Drawings

Mechanical

(Note FDE Flow Diagram)

Number Title Sheet. Rev.

2323-M1-0202 FD-Main Reheat & Steam Dump System -9 CP-9

2323-M1-0206 FD-Auxiliary Feedwater-System CP-7

2323-M1-0206 FD- Type "In Dump Trains 01 CP-2

2323-H1-0202 FD-Main Reheat & Steam Dump System -9 CP-9

" "

2323-M1-0203 F" Type In 1 CP-6

2323-M1-0216 FD-Compressed Air System A CP-1

ECE-M1-0216 Instrument Air Supply 01 CP-3

Electrical & Control

2323-M1-0233 FD-Station Service Water System - CP-11

Sheet 1 of 3

2323-M1-0233 "Type in A CP-1

'

Sheet 2 of 3

2323-M1-0234 "Type" In - CP-9

Sheet 3 of 3

2323-M1-0229 FO-Component Cooling Water System - CP-6

Sheet 1 of 7 " "

ECE-M1-0229

"

"Type" In" A CP-1

Sheet 3 of 8 B CP-1

ECE-H1-0229 FD-Component Coolin9 Water System - CP-10

Sheet 4 of 8

" " " " " " " "

LCE-M1-0230 A CP-1

Sheet 5 of 8

" " "

ECE-M1-0230

" B CP-1

Sheet 6 of 8

" "

2323-M1-0231

" ' - CP-8

Sheet 7 of 8

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Number Title Sheet. Rev.

" " " "

ECE-M1-0231 A CP-1

2323-M1-0250 FD-Reactor Coolant System - CP-7

" " "

2323-M1-0251 FD- -

CP-7

FD-Chemical & Volume Control System

'

2323-M1-0253 -

CP-7

" " "

ECE-M1-0253 FD- A CP-1

" " " "

2323-M1-0254 - CP-8

" "

2323-M1-0255 FD- - CP-9

Volume Control Tank loop

ECE-M1-0255 FD-Chem. & Vol. "Control "System

Charging and Positive Displacement

Pump Trains

2323-M1-0260 FD-Residual Heat Removal System - CP-9

2323-M1-0261 FD-Safety injection System - CP-6

2323-M1-0262 FO-Safety Injection System - CP-7

Sheet 2 of 3

. 2323-M1-0264 FD-Liquid Waste Processing -

CP-5

Reacter Coolant Drain Tank

Subsystem

2323-El-0001 Plant One Line Diaoram Units CP-2

Units 1 & 2

2323-El-0C01 Plant one Line Diagram Units CP-3

! Units 1 & 2

2323-El-0C03 6.9 KV Auxiliaries-One Line Diagram CP-4

ene Line Diagram Normal Buses

2323-El-0C04 6.9 KV Auxiliaries-One Line Diaoram CP-6

Safeguard Buses

'

ELE-El-CC04 6.9 KV Auxiliaries-One Line Diaaram A CP-1

Safeguard Buses

'

2323-El-0005 480V Auxiliaries-One Line Diagram CP-2

Safeguard Buses

2323-El-0007 Safeguard & Auxiliary Buildings

Safeguard 480V NCC's One Line Diagram CP-7

- A4 -

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M b,ef, Title Sheet. h

2323-El-0009 Containment & Desser Cenerator CP-4

Safequard 480V NCC's One Line Diagram

2323-El-0010 Cora.cn Auxiliary & Control Blogs CP-5

Safeguard 480 V hCC's One Line Diagram

2323-El-0014 Service Water intake Structure CP-3

and Casser Generator Safeguard

400V NCC's - Oae Line Diagran

2323-El-0018 118V AC Instrument Bus Distribution CP-4

Cne Line Diagram

2323-El-0018 118V AC & 125V D.C. One Line Diagram 02 CP-4

2323-El-0019 24/48V & 125/2ECV DC One Line Diagram CP 4

2323-El-0020 125V D.C. One Line Diagram CP-5

2323-El-C024 118V AC & 120V AC Corron & Unit il 03 CP-3'

Instr. Distribution Panels One Line Diagram

ThE-El-0067 Diesel Generator IC G1 AC Scheoratic, CP-1

Unit 1

THE-El-006M Diesel Generator IEG1 Engine Start- 05 CP-3

Stop DC Centrol Schenictic

2323-El-0031 6.9 KV Switchgear Bus IEA1 Diesel Gen. 21 CP-3

Bkr. 1EG1 Schematic Diagram

2323-El-0031 6.9 KV. Switchgear Bus. IEA1 25 CP-2

Component, Cooling, Water PP #11

Schematic Diagran

2323-El-0031 6.9 KV Switchgear Bus IEA1 Auxiliary 31 CP-3

Feedwater Pump all Schematic Diagram

2323-El-0031 6.9 KV Switchgear Eus IEA1 Station 41 CP-5

Service Water Pump W11 Scher.atic Diagram

2323-El-0031 6.9 KV. Switchgear Bus IEA1 53 CP-1

Centrifugal Charging P.P ill

Sch watic Diagram

2323-El-0033 480 V Switchgear Bus IEB1 01 CP-1

Supply Breaker IEB1-1

Schematic Diagram

- A5 -

_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ .

,

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fiumber Title Sheet. Rev.

2323-El-0033 460 V Switchgear Bus lEB3 03 CP-1

Supply Breaker AEB3-1 .

'

Schematic Diagram

2323-El-0037 Air Operated Value - 1PV-24538 07 CP-1

2323-El-0039 Main Steam Loop 1MS!V/BYP Value 41 CP

HSP. Indication & Manual BYP. '

ISOL Value IHV-23?3 B

Scherratic Otagram

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9

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