ML20195J046
ML20195J046 | |
Person / Time | |
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Site: | Comanche Peak |
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
l/h/E
'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
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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
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
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
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
. 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
.
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 :
l
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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.
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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. ,
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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.
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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
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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- -
FD-Chemical & Volume Control System
'
2323-M1-0253 -
" " "
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 -
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
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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
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- fiumber Title Sheet. Rev.
2323-El-0033 460 V Switchgear Bus lEB3 03 CP-1
Supply Breaker AEB3-1 .
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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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