ML050540462
| ML050540462 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/31/2003 |
| From: | Ogle C Division of Reactor Safety II |
| To: | Sumner H Southern Nuclear Operating Co |
| References | |
| FOIA/PA-2004-0277 IR-03-006 | |
| Download: ML050540462 (40) | |
See also: IR 05000321/2003006
Text
UNITED STATES
.
NUCLEAR REGULATORY COMMISSION
REGION II
/) 7X
SAM NUNN ATLANTA FEDERAL CENTER
61 FORSYTH STREET SW SUITE 23T85
ATLANTA, GEORGIA 30303.8931
/f
X,
Southern Nuclear Operating Company, Inc.
ATTN: Mr. H. L. Sumner, Jr.
Vice President
P. 0. Box 1295
Birmingham, AL 35201-1295
/ro
7
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A
SUBJECT:
EDWIN I. HATCH NUCLEAR POWER PLANT - NRC TRIENNIAL FIRE
PROTECTION INSPECTION REPORT '0 382 1 &S3-66ND J0 3860/S-0
ier:
.
50003.2//12.0ob&;A/D
5eB>oo3v/zooa,0
Depar Mr. Sumr
On July 25, 2003, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
your Hatch Nuclear Plant Units 1 and 2. The enclosed inspection report documents the
inspection findings, which were discussed on that date with Mr. R. Dedrickson and other
members of your staff.
The inspection examined activities conducted under your licenrse as they relate to safety and
compliance with the Commission's rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel
.7'
This report doc ments thee findings that have potential safety significance greater than very
low significa e, however a safety significance determination has not been completed. One
issue invol ng a procedural inadequacy did present an immediate safety concern, how ver,
your staff evised the procedure prioj to the
Iefd of the inspection. The othertA issue did not
present $ immediate safety concern{. In addition, the report documents three NRC-identified
findings of very low safety significance (Green), all of which were determined to involve
violations of NRC requirements. Howev&r, because of the very low safety significance and
because they are entered into your corrective action program, the NRC is treating these three.
findings as non-cited violations (NCVs) consistent with Section VL.A of the NRC Enforcement
Policy. If you contest any NCV intfiis report, you should provide a response within 30 days of
the date of this inspection reportrwith the basis for your denial, to the Nuclear Regulatory
Commission, ATTN.: Documeent Control Desk, Washington DC 20555-0001; with copies to the
Regional Administrator Regioh I1; the Director, Office of Enforcement, United States Nuclear
Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at -4
Hatch Nuclear Power Plart.
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SNC, Inc.
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In accordance with 10 CFR 2.790 of the NRC's NRules of Practice,' a copy of this letter and its.
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publically Available Records (PARS) component of NRC's document system
(ADAMS). ADAMS is accessible from the NRC Website at
http://www.nrc.gov/readina-rmnfadams.html (the Public Electronic Reading Room).
Sincerely,
Charles R. Ogle, Chief
Engineering Branch 1
Division of Reactor Safety
Docket Nos.: 50-321, 50-366
Enclosure:
NRC Triennial Fire Protection Inspection Report 50-321/03-06, 50-366103-06
w/Attachment: Supplemental Information
cc w/encl:
J. D. Woodard
Executive Vice President
Southern Nuclear Operating Company, Inc.
Electronic Mail Distribution
George R. Frederick
General Manager, Plant Hatch
Southern Nuclear Operating Company, Inc.
Electronic Mail Distribution
Raymond D. Baker
Manager Licensing - Hatch
Southern Nuclear Operating Company, Inc.
Electronic Mail Distribution
Arthur H. Domby, Esq.
Troutman Sanders
Electronic Mail Distribution
Laurence Bergen
Oglethorpe Power Corporation
Electronic Mail Distribution
(cc w/encl cont'd - See page 3)
.
-1
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'-
-
'SNC, Inc.
(cc w/encl cont'd)
Director
Department of Natural Resources
205 Butler Street, SE, Suite 1252
Atlanta, GA 30334
Manager, Radioactive Materials Program
Electronic Mail Distribution
Chairman
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. County Courthouse
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Baxley, GA 31513
Resident Manager
Oglethorpe Power Corporation
Edwin I. Hatch Nuclear Plant
Electronic Mail Distribution
Senior Engineer - Power Supply
Municipal Electric Authority
of Georgia
Electronic Mail Distribution
Reece McAlister
Executive Secretary
Georgia Public Service Commission
244 Washington Street, SW
Atlanta, GA 30334
Distribution w/encl:
S. Bloom, NRR
L. Slack, RII EICS
RIDSNRRDIPMLIPB
PUBLIC
OFFICE
RII:DRS
RII:DRS
CONTRCTOR
RII:DRP
SIGNATURE
NAME
CSMITH
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GWISEMAN
KSULVA
BONSER
DATE
8/
/2003
8/
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8/
/12003
8/
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/
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003
E-MAIL COPY?
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Docket Nos.:
License Nos.:
Report No.:
Licensee:
Facility:
Location:
Dates:
U. S. NUCLEAR REGULATORY COMMISSION
REGION II
50-321, 50-366
05000321/2003006 and 05000366/2003006
Southern Nuclear Operating Company
E. l. Hatch Nuclear Plant
P. O. Box 2010
Baxley, GA. 31513
July 7-11, 2003 (Week 1)
July 21-25, 2003 (Week 2)
C. Smith, P E., Senior Reactor lnspec prkgead Inspec
R. Schin, Senior Reactor Inspector
G. Wiseman, Fire Protection Inspector
K. Sullivan, Consultant, Brookhaven National Laboratory
Inspectors:
S. Belcher, Nuclear Safety Intern, Week I
A/L
J. e
0 Approved by:
Charles R. Ogle, Chief
Engineering Branch 1
Division of Reactor Safety
Enclosure
CONTENTS
- ,SUMMARY OF FINDINGS.
......
REPORT DETAILS ...............................................................................................................
.
- .
.IPE T9~
.
.~~e
Systems Required to Achieve and Maintain Safe Shutdown .
Fire Protection of Safe Shutdown Capability.
Post-FireSafe Shutdown Capability .
- -Operational Implementation of Alternative Shutdown Capability.
C
L
54Cm
'u ii S ons...................;.l
-ergency
Lighting....................................................................................................
Cold Shutdown Repairs......................
................................................................
Fire Barriers and Fire Area/Zone/Room Penetration Seals.........................................
Fire Protection Systems, Features, and Equipment.................................................
SAFETY SYSTEM DESIGN AND PERFORMANCE CAPAIlITY
Dteik4g
34, giV Backup Actuation va Pressure Transmitter Signals .......................
FTHER ACTIVITIES
Identification and Resolution of Problems.
Meetings Including Et..
.
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SUMMA
OF FINDINGS
IR 05000321/20
-006 p05000366/2003-00 , E. l. Hatch Nuclear Plant, Units 1 and 2;
003 and 712125/2003;
riennial Fire Protection
The report covered' two-wek period of inspection by three regional inspectors and a,
oontr
tf
from Brookhaven National Laboratory. Three Green non-cited violations (NCVs) and -
three unresolved items with potential safety significance greater than Green were identified..-'.,'.,.
usgnificance of most findings is indicated by their color (Green, White, Yellow, Red) using %tfr
--
_M¢,)J609, "Significance Determination Process" (SDP). Findings for which the SDP does not
apply may be Green or be assigned a severity level after NRC management review. The:
NRC's program for overseeing the safe operation of commercial nuclear power reactors Is
described in NUREG-1 649, "Reactor Oversight Process," Revision 3, dated July 2000.
<1
A.
NRC-Identified and Self-Revealinq Findings
/Srnersto
itiating Events, Mitigating Systems, and Barrier Integrity
he team identified an unresolved item in that a local manual operator action, to
event spurious opening of all eleven safety relief valves (SRVs) during a fire event,
would not be performed in sufficient time to be effective. Also, licensee reliance on this
.
. manual action for hot shutdown during a fire, instead of physically protecting cables from
fire dam
age
n approved by the NRC.
This finding is u
solved pending completion of a significance determination. Th
finding is greate
an minor because it affects the objective of the mitigating systj
cornerstone.
Iso, the findin has potential safety significance greater than very I
safety sign icance bec
ailure
revent spurious operation of the SRVs cod:
result in t
openingi certain fire sce
rios, thereby compI'ating the post-fire re c c
I'
atns. (Section 1 R05.04/.05.b.1)
1- -
TBD: he team identified an unresolved item in con
oeln with t
implementati'
design phange request (DCR) 91-1
p Actuation
assure Trans
' S
The installed plant mod ication failed to implement t e o
-out-of-two ts
ce loic that was specified as esign in
quirements in t
esign chan
pac
e. Additionally, implementation of a
-out-of-two coincident taken tw e r
-
has introduced a potential common cause
ure of all eleven SRVs
induced damage to two instrumentation circuit cables in close proximity to
h er
This finding is unresolved pending completion of a significance determination. This
finding is greater than minor because it impacts the mitigating system cornerstone. This
finding has the potential for defeating manual control of Group A SRVs that are required
for ensuring that the suppression pool temperature will not exceed the heat capacity
temperature limit (HCTL) forthe suppressiorq pool.,(Section 1 R21.01.b)
Green. The team identified
manual operator action to operate safe shutdown equipment was too difficult and was
also unsafe. The licensee had relied on this action instead of providing physical
..
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protection of cables from fire damage or preplanning cold shutdown repairs. However,
the team
some operators would not be able to perform the action.'-
Thisevlo
andsP_
CFR 50, Appendix R. Se~ction l1I.!G~.2lhafn~ding
i F
Specifcation 5~.4~.VIe finding Is greater Planerbeaa it uledfthe':- '
avaiiability
and
reliability objectivu
ent performance attribute of the
mitigating systems cornerstone.5
n
Sincethe.
-
implement cold shutdown repairs to facilishe
operatorshment of the action, this finding
did not have potential safety significance greater than very low safety significanc'e.'
(Section 1 R05.04/.05.b.2)
Green. The team identified a tinding with v
my(w
c
ao
in that the
licensee relied on some manual operator actions to operate safe shutdown equipment
instead of providing the required physical protection of cables from fire damage without
<
~~NRC approval.
--
-
-
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Thri finding involved a "iolation of10 CFR 50, A
endix R, Section lI.GJ.
The findings
is greater than minor because it affecte the
reliabilitybjcectives and the
e
n
equipment performance attribute of the mitigating systems cornerstone. Since theor
w
b
actions could reasonably be accomplished by operators in a timely manner, this finding
d
did not have potential safety significance greater than very low safety signiiicance.
(
(Section 1 R05.04/.05.b.3)
Nonee
Green. The team identified
PrF low 4nfnc
gmerg
ncy.'
A
l~ighting was not adequate for some manual operator actions that were needed to
0
tef it-
support post-fire operation of safe shutdown equipment.Ll'-
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10 CR50, Appendix R. S
o l..
efnding is.I
greater than minor because it affected te reliability objectiv-eand the e-quipment-.l
performance attribute of the mitigating systems cornerstone. Since operators would be[
'
able to accomplish the actions with the use of flashlights, this finding did not have -
'
potential safety significance greater than very low safety significance. (Section
'/
1 R05.07.b)
.
Licensee-Identified Violationsl
-None
'
11 At X,?w
/O
SUMMARY OF FINDINGS
IR 05000321/2003-006, 05000366/2003-006; E. l. Hatch Nuclear Plant, Units 1 and 2;
7/7-11/2003 and 7/21-25/2003; Triennial Fire Protection
The report covered a two-week period of inspection by three regional inspectors and a.
contractor from Brookhaven National Laboratory.- Three Green non-cited violations (NCVs) and
three unresolved items with potential safety significance greater than Green were identified.
The significance of most findings is indicated by their color (Green, White, Yellow, Red) using
IMC 0609, "Significance Determination Process" (SDP). Findings for which the SDP does not
apply may be Green or be assigned a severity level after NRC management review. The
NRC's program for overseeing the safe operation of commercial nuclear power reactors is
described in NUREG-1 649, "Reactor Oversight Process," Revision 3, dated July 2000.
- A.
NRC-identified and Self-Revealin Findings
Cornerstor: Mitigating Systems ohs
Af*tk,?t/
-- *) "b~i~rThe team identified an unresolved item in that a local manual operator action, to
prevent spurious opening of all eleven safety relief valves (SRVs) during a fire event,
would not be performed in sufficient time to be effective. Also, licensee reliance on this
, manual action for hot shutdown during a fire, instead of physically protecting cables from
fire damage, had not been approved by the NRC.
o,
This finding is unresolved pending completion of a sig
determination. Tle
finding is greater than minor because it affects the tigating system cornerstone. Also,
the finding has potential safety significance greater than very low safety significance
because failure to prevent spurious operation of the SRVs could result in the opening is
certain fire scenarios, thereby complicating the post-fire recovery actions. (Section
1 R05.04/.05.b.1)
he team identified an unresolved item in that a fire in Fire Area 2104 could
all e
SRVs to open. The inspection team was concerned that the licen action
A/
to preclu
is scenario were not consistent withtthe current licensin
s of the
p nt. In additi
o objective evidence existed 4o demonstrat
he post-fire safe
shutdown equipmenas adequate to mitigateAeleven S opening. Finally the team
noted that if the Group A
Vs were to spuriously
te as a result of fire damage,
they could not be manually co
led by th ator as part of the licensee's fire
mitigation strategy
This finding is identified
resolved
N RC review of the concerns associated
with the potential fig
of SRVs. This finding
determined to have potential
safety sign
greater than very low significance bese of the concerns
asso
with potential opening of the SRVs and the limite
t of equipment that
,d be available for safe shutdowp n
hese conditions. (Se
1 R.05.03.b)
ran The team identified an unr
d item in connection with the implementation of
design change request (DCR)
1-134, SRV Backup Actuation via Pressure Transmitter
Signals. The installed plant modification failed to implement the one-out-of-two taken
twice logic that was specified as design input requirements in the design change
2
package. Additionally, implementation of a two-out-of-two coincident taken twice logic,
has introduced a potential common cause failure of all eleven SRVs because of fire
induced damage to two instrumentation circuit cables in close proximity to each other.
C
This finding is unresolved pending completion of a significance determination. This
finding is greater than minor because it impacts the mitigating system cornerstone. This
finding has the potential for defeating manual control of GroupOA!SRVs that are
required for ensuring that the suppression pool temperature will not exceed the heat
capacity temperature limit (HCTL) for the suppression pool. (Section 1 R211.01 .b)
Green. The team identified a finding with very low safety significance in that a local
J
manual operator action to operate safe shutdown equipment was too difficult and was
also unsafe. The licensee had relied on this action instead of providing physical
protection of cables from fire damage or preplanning cold shutdown repairs. However,
the team judged that some operators would not be able to perform the action.
-
This finding involved a violation of 10 CFR 50, Appendix R, Section lIl.G.1 and
Technical Specification 5.4.1. The finding is greater than minor because it affected the
availability and reliability objectives and the equipment performance attribute of the
mitigating systems comerstone. Since the licensee could have time to develop and
implement cold shutdown repairs to facilitate accomplishment of the action, this finding
J -did
not have potential safety significance greater than very low safety significance.
-
(Section 1 R05.04/.05.b.2)
Green. The team identified a finding with very low safety significance in that the
licensee relied on some manual operator actions to operate safe shutdown equipment,
instead of providing the required physical protection of cables from fire damage
d
without NRC approval.
This finding involved a violation of 10 CFR 50, Appendix R, Section Ill.G.2. The finding
is greater than minor because it affected the availability and reliability objectives and the
equipment performance attribute of the mitigating systems comerstone. Since the
actions could reasonably be accomplished by operators in a timely manner, this finding
did not have potential safety significance greater than very low safety significance.
(Section 1 R05.04/.05.b.3)
- a
Green. The team identified a finding with very low safety significance in that emergency
lighting was not adequate for some manual operator actions that were needed to
support post-fire operation of safe shutdown equipment.
This finding involved a violation of 10 CFR 50, Appendix R, Section IIU.J. The finding is
greater than minor because it affected the reliability objective and the equipment
performance attribute of the mitigating systems cornerstone. Since operators would be
able to accomplish the actions with the use of flashlights, this finding did not have
potential safety significance greater than very low safety significance. (Section
1 R05.07.b)
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Licensee-identified Violations
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REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems and Barrier Integrity
FIRE PROTECTION
The purpose of this inspection was to review the Hatch Nuclear Plant fire protection
program (FPP) for selected risk-significant fire areas. Emphasis was placed on
verification that the post-fire safe shutdown (SSD) capability and the fire protection
features provided for ensuring that at least one fed nda# train of safe shutdown
systems is maintained free of fire damage. The inspection was performed in
accordance with the Nuclear Regulatory Commission (NRC) Reactor Oversight Program
using a risk-informed approach for selecting the fire areas and attributes to be
inspected. The team used the licensee's Individual Plant Examination for External
Events and in-plant tours to choose four risk-significant fire areas for detailed inspection
and review. The fire areas chosen for review during this inspection were:
Fire Area 2016, West 600 V Switchgear Room, Control Building, Elevation 130
feet.
Fire Area 2104, East Cableway, Turbine Building, Elevation 130 feet.
,*
Fire Area 2404, Switchgear Room 2E, Diesel Generator Building, Elevation 130
feet.
!
Fire Area 2408, Switchgear Room 2F, Diesel Generator Building, Elevation 130
feet.
I
\\.-The team evaluated the licensee's FPP against applicable requirements, including
Operating License Condition 2.C.(3)(a), Fire Protection; Title 10 of the Code of Federal
Regulations, Part 50 (10 CFR 50), Appendix R; 10 CFR 50.48; Appendix A of Branch
Technical Position (BTP) Auxiliary and Power Conversion Systems Branch (APCSB)
9.5-1; related NRC Safety Evaluation Reports (SERs); the Hatch Nuclear Plant Updated
Final Safety Analysis Report (HNP-FSAR); and plant Technical Specifications (TS). The
team evaluated all areas of this inspection, as documented below, against these
requirements.
Documents reviewed by the team are listed in the attachment.
.1J
Systems Required to Achieve and Maintain Post-Fire Safe Shutdown
Inspection Scope
The licensee's Safe Shutdown Analysis Report (SSAR) was reviewed to determine the
components and systems necessary to achieve and maintain safe shutdown conditions
in the event of fire in each of the selected fire areas. The objectives of this evaluation
were as follows:
I
2
1.
Verify that the licensee's shutdown methodology has correctly identified
the components and systems necessary to achieve and maintain a safe
-
shutdown condition.
2.
Confirm the adequacy of the systems selected for reactivity control,
-"
reactor coolant makeup, reactor heat removal, process monitoring and
support system functions.
3.
Verify that a safe shutdown can be achieved and maintained without off-
site power, when it can be confirmed that a postulated fire in any of the
selected fire areas could cause the loss of off-site power.
4.
Verify that local manual operator actions are consistent with the plant's
fire protection licensing basis.
b.
Findinas
The team identified a potential concern in that the licensee used manual actions to':
disconnect terminal board sliding links in order to isolate two 4-20 milli-amp (ma
'
instrumentation loop control circuits in order to prevent the spurious actuation f eleven
SRVs, This issue is discussed in section 1 R05.03.b of the report.
4o,
a 7
.02
Fire Protection of Safe Shutdown Caabilit-y
a.
InsDection Scope
For the selected fire areas, the team evaluated the frequency of fires or the potential for
fires, the combustible fire load characteristics and potential fire severity, the separation
of systems necessary to achieve safe shutdown (SSD), and the separation of electrical
components and circuits located within the same fire area to ensure that at least one
SSD path was free of fire damage. The team also inspected the fire protection features
to confirm they were installed in accordance with the codes of record to satisfy the
applicable separation and design requirements of 10 CFR 50, Appendix R, Section III.G,
and Appendix A of BTP APCSB 9.5-1. The team reviewed the following documents,
which established the controls and practices to prevent fires and to control combustible
fire loads and ignition sources, to verify that the objectives established by the
NRC-approved fire protection program (FPP) were satisfied:
Updated Final Safety Analysis Report (UFSAR) Section 9.1 -A, Fire Protection
Plan
Administrative Procedure 40AC-ENG-008-OS, Fire Protection Program
Administrative Procedure 42FP-FPX-01 8-OS, Use, Control, and Storage of
Flammable/Combustible Materials
Preventive Maintenance Procedure 52PM-MEL-012-0, Low Voltage Switchgear
Preventive Maintenance
The team toured the selected plant fire areas to observe whether the licensee had
properly evaluated in-situ fire loads and limited transient fire hazards in a manner
consistent with the fire prevention and combustible hazards control procedures. In
addition, the team reviewed the licensee's fire safety inspection reports and corrective
action program (CAP) condition reports (CRs) resulting from fire, smoke, sparks, arcing,
and overheating incidents for the years 2000-2002 to assess the effectiveness of the fire
3
prevention program and to identify any maintenance or material condition problems
related to fire incidents.
The team reviewed fire brigade response, fire brigade qualification training, and drill
program procedures; fire brigade drill critiques; and drill records for the operating shifts
from January 1999 - December 2002. The reviews were performed to determine
whether fire brigade drills had been conducted in high fire risk plant areas and whether
fire brigade personnel qualifications, drill response, and performance met the
requirements of the licensee's approved FPP.
The team walked down the fire brigade equipment st age areas and dress-out locker
areas in the fire equipment building and the turbin
iulding to assess the condition of
fire fighting and smoke control equipment. Fire
igade personal protective equipment
located at both of the fire brigade dress-out are s and fire fighting equipment storage
area in the turbine building were reviewed to e aluate equipment accessibility and
functionality. Additionally, the team observe whether emergency exit lighting was
provided for personnel evacuation pathways a the outside exits as identified in the
National Fire Protection Association (NFPA) 101, Life Safety Code, and the
Occupational Safety and Health Administration (OSHA) Part 1910, Occupational Safety
and Health Standards. This review also included examination of whether backup
emergency lighting was provided for access pathways to and within the fire brigade
equipment storage areas and dress-out locker areas in support of fire brigade
operations should power fail during a fire emergency. The fire brigade self-contained
breathing apparatuses (SCBAs) were reviewed for adequacy as well as the availability
of supplemental breathing air tanks and their refill capability.
The team reviewed fire fighting pre-fire plans for the selected areas to determine if
appropriate information was provided to fire brigade members and plant operators to
facilitate suppression of a fire that could impact SSD. Team members also walked down
the selected fire areas to compare the associated pre-fire plans and drawings with as-
built plant conditions. This was done to verify that fire fighting pre-fire plans and
drawings were consistent with the fire protection features and potential fire conditions
described in the Fire Hazards Analysis (FHA).
The team reviewed the adequacy of the design, installation, and operation of the manual
suppression standpipe and fire hose system for the control building. This was
accomplished by reviewing the FHA, pre-fire plans and drawings, en
ring
mechanical equipment drawings, design flow and pressure calculatio
a d NFPA 14
for hose station location, water flow requirements and effective reach
ability. Team
members also walked down the selected fire areas in the control building to ensure that
hose stations were not blocked and to verify that the required fire hose lengths to reach
the safe shutdown equipment in each of the selected areas were available. Additionally,
the team observed placement of the fire hoses and extinguishers to assess consistency
with the fire fighting pre-fire plans and drawings.
b.
Findings
No findings of significance were identified.
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Inspection Scope
On a sample basis, the inspectors paluated whether the systems an equipment
identified in the licensee's SSAR s being required to achieve and mentain hot
shutdown conditions would re
in free of fire damage in the event of ire in the
fire areas. The evaluation in uded a review of cable routing data de icting the L,
of power and control cable associated with SSD Path 1 and Path 2 component.
RCIC and HPCI systems. Additionally, on a sample basis, the team reviewed the
licensee's analysis of electrical protective device (e.g., circuit breaker, fuse, relay),
coordination. The following motor operated valves (MOVs) and other components
reviewed:
IM.
.
iI
7
11
1
LiOm1ponenit IL)
2E51 -F010
2P41 -CO01 A
-2P41 -C001 B
eUsbUrILtluI
RCIC Pump Suction from Suppressionspal
RCIC Pump Suction Valve from
e
Plant SerVice Water Pump 2A
^RH-!Veat Exchanger A Drain to Suppression Pool Valve
Plant Service Water Pump 2B
HPCI Turbine Steam Supply Valve
I-
2E41 -F002
2E41 -F006
HPCI Turbine Steam Supply Inboard Containment Isolation Valve
HPCI Pump Inboard Discharge Valve
HPCI Pump Discharge Bypass Test Valve to CST
b.
Findings
The team identified a potential concern in that the licensee used manual actions to
isolate two 4 to 20 ma instrumentation loop control circuits associated with eleven SRVs
in lieu of providing physical protection. This did not appear to be cgrjtstent with theIJo
.plant's licensing basis nor 10 CFR 50 Appendix R. Spurious acti'he tlese SRVs e
ese could impact the licensee's fire mitigation strategy. In addi ion th licensee had
no objective evidence that post-fire safe shutdown equipment co
figate this event.
The SSAR stated that a fire in Fire Area 2104 could cause all eleven SRVs to spuriously
actuate as a result of fire damage to two cables located in close proximity in this area.,N
The specific circuits that could cause this event have been identified by the licensee A&Y
4c6rcuitlern: ABE01 9C08 and ABE01 9C09P Each of these two circuits provides a 4 to
20 ma instrumentation signal from SRV high-pressure actuation transmitters4B21 -
.I
N127B and 2B21-N127Dgto master trip us 2
-N697B and 2B21-N697D,
respectively. The purpose of this circui
is to provide an electrical backup to the
mechanical trip capability of the indivj ual SRVs. In the event of high reactor pressure,
the circuits would provide a signal
the master trip units which would cause all eleven
SRVs to actuate (open). The pr sure signal from each transmitter is conveyed to its
respective master trip unit thro gh a two-conductor, instrument cable that is routed
through this fire area (two se arate cables). Each cable consists of a single twisted pair
of insulated conductors, an ninsulated drain wire that is wound around the twisted pair
of conductors, and a foil
leld. In Fire Area 2104 the two cables are located in close
proximity, in the same c
le tray. Actuation of the SRV electrical backup is completely.
"blind" to the operators. hat is, unlike ADS, it does not provide any pre-actuation
indication (e.g., actuation of the ADS timer) or an inhibit capability (e.g., ADS inhibit
switch). Since the operators typically would not initiate a manual scram until fire
damage significantly interfered with control of the plant, its possible that all eleven SRVs
could open at 100% power, prior to scramming the reactor. This scenario could place
the plant in an unanalyzed condition.
Unlike a typical control circuit, a direct short or "hot short" between conductors of a 4 to
20 ma instrument circuit may not be necessary to initiate an undesired (false high)
signal. For cables that transmit low-level instrument signals, degradation of the
insulation of the individual twisted conductors due to fire damage may be sufficient to
cause leakage currents to be generated between the two conductors. Such leakage,
current would appear as a false high pressure signal to the trip units. If both cables
were damaged as a result of fire, false signals generated as a result of leakage current
in each cable, could actuate the SRV electrical backup scheme which would cause all
eleven of the SRVs to open. The conductor insulation and jacket material of each cable
is cross-linked polyethylene (XLPE). Since both cables are in the same tray and
exposed to the same heating rate, there is a reasonable likelihood that both
instrumentation cables could suffer insulation damage at the same time and both circuits
could fail high simultaneously.
The licensee's SSAR recognizes the potential safe significance of this eve
nd
describes methods that have been developed to revent its occurrence an or mitigate
its impact on the plant's post-fire safe shutdow capability should it occur. o prevent
this scenario, the licensee has developed pro edural guidance which directs operators
to open link BB-10 in panel 2H1 1-P927 and nk BB-10 in panel 2H11-P928. These
panels are located in the main control room. Opening of these links would prevent
actuation of the SRV trip units by removing the 4 to 20 ma signal fed by the pressure
transmitters to the master trip units. In the event the SRVs were to open prior to
operators completing this action, the SSAR credits core spray loop A to mitigate the
event.
The inspection team had several concerns regarding the licensee's approach to this
potential spurious actuation of the SRVs. Specific concerns identified by the team
included:
1.
The links may not be opened in time to preclude inadvertent actuation of
the SRVs.
6
2.
The use of links to avoid inadvertent actuation of the SRVs did not
appear to be consistent with the current licensing basis.
3.
No objective evidence existed to demonstrate that the post-fire safe
shutdown equipment could adequately mitigate a fire in Fire Area 2104, if
the SRVs were to open.
4.
The operations staff is unable to manually control the group A SRVs that
are credited for mitigating a fire in Fire Area 2104 if they spuriously
actuatska a result of fire induced damage.
- .
I-
-F,
I-.
With regard to t
iming of operator actions to prevent fire damage from causing all
SRVs to open, u ng the inspection, the licensee performed an evaluation which
estimated that approximately thirty minutes would pass from the time of fire detection to
the time an operator would implement procedural actions to open the links. The
inspectors independently arrived at a similar time estimate based on their review of the
procedure. In response to inspectors concerns that this interval may be too lengthy to
preclude fire damage to the cables of interest and subsequent actuation of the SRVs,
-
the licensee agreed to enhance its existing procedures so that the action would be
taken immediately following confirmation of fire in areas where the spurious actuation
could occur. This issuep0cliscussed is Section IR.04/.05.b.1,of this report.
The team also considered opening terminal board inko be not in compliance with the
plant's licensing basis. Current licensing basis d -uments, specifically Georgia Power
request for exemption dated May 16, 1986,
a subsequent NRC Safety Evaluation
Report (SER) dated January 2, 1987, ch
cterized the opening of links as a repair
activity that is not permitted as a mean of complying with Section IIL.G of Appendix R.
The inspectors concluded that, the o
ning of links was considered a repair by both the
licensee and the NRC staff in 1987. he licensee could not provide any evidence to
justify why these actions are not characterized as a repair activity in its current SSAR.
Additionally, because there is a potential for all SRVs to spuriously actuate as a result of
fire in Fire Area 2104 at a time when RHR is not available, the SSAR credits the use of
core spray loop A to accomplish the reactor coolant makeup function. During the
inspection, the licensee performed a simulator exercise of an event which caused all 11
SRVs to open. During this exercise, simulator RPV level instruments indicated that core
spray would be capable of maintaining level above the top of active fuel. However, the
licensee did not provide any objective evidence (e.g., specific calculation or analysis)
which demonstrated that, assuming worst-case fire damage in Fire Area 2104, the
limited set of equipment available would be capable of mitigating the event in a manner
that satisfies the shutdown performance goals specified in Appendix R, Section L.. .e to
-C/un
Finally, the kli~,bee s tailure o lImiipunie
1nt Je input
-
two taken twice logic for DCfl 31-134 resulted i the followl--inga1
ir
or
the logic
that was installed by DCR 91-134 for the SRVs was a two-out-of-two coincident taken
twice logic in addition to a one-out-of-two coincident taken twice logic. The team
determined that the two-out-of-two coincident logic input from trip unit master relays
K31 OD and K335D represented a common cause failure for group "A" SRVs for a fire in
I
%
r-OY
J,
F
,
.7
/-AA ie.%te .(5
!
- 2~AR
~~2-.O~'
Fire Area 2104. Specifically, cable ABE01 9C08 associated with pressure transmitter
2B21 -Ni27B current loop, and cable ABE019CO9 associated with pressure transmitter
2B21-N127D current loop, are routed in close proximity to each other in 'the same cable'
tray in Fire Area 2104. Both shielded twisted pair instrument cables are unprotected,
from the effects of a fire in this fire area. Fire induced insulation damage to both cables
could result in leakage currents which causes the instrument loops to fail high.' This - I
failure mode simulates a high nuclear boiler pressure condition and would initiate SRV
backup actuation of all the group UA! SRVs. Whenever a SRV lifts, it will remain open
until pressure reduces to about 85% of its overpressure lift setpoint The instrument
loops having ifailed high, however, will 'ensure that the trip unit master relays and the trip
unit slave re ays continue to energize the pilot valve of the individual SRV and keep'the
SRV open.
his failure mode prevents the operators from manually controlling the'-..
s as is required per the SSAR.
In response, the licensee initiated a Condition Report (CR 2003800152, dated 7/24/03)
to evaluate actions to open links, in order to determine if they are necessary to achieve
hot shutdown, and if an exemption from Appendix R is required. Pending additional
review by the NRC, this issue is identified aSAVRI 50-366/20G030G690,
Concerns'
Associated with Potential Opening of SR Ccn.
.04/.05 Alternate Shutdown Capabilitv/O erational Implementation of Alternative Shutdown
Capability
a.
Inspection Scope
The selected fire areas that were the focus of this inspection all involved reactor
shutdown from the control room. None involved abandoning the control room and1 #J
alternative safe shutdown from outside of the control room. Thus, alternate shutlown
capability was not reviewed during this inspection. However, the licensee's plans for
SSD following a fire in the selected areas Involved many local manual operator actions
that would be performed outside of the control area of the control room. This section of
the inspection focused on those local manual operator actions.
.1'
The team reviewed the operational Implementation of the SSD capability for a 4in thy
>
selected fire areas to determine if: (1) the procedures were consistent with th
pendix
R safe shutdown analysis (SSA); (2) the procedures were written so that the operator
actions could be correctly performed within the times that were necessary for the actions
to be effective; (3) the training program for operators included SSD capability; (4)
6
personnel required to achieve and maintain the plant in hot standby could be provided
from the normal onsite staff, exclusive of the fire brigade; and (5) the licensee
periodically performed operability testing of the SSD equipment.
The team walked down SSD manual operator actions that were to be performed outside
of the control area of the main control room for a fire in the selected fire areas and
discussed them with operators. These actions were documented in Abnormal Operating
Procedure (AOP) 34AB-X43-001 -2, Version 10.8, dated May 28, 2003. The team
evaluated whether the local manual operator actions could reasonably be performed,
using the criteria outlined in NRC Inspection Procedure (IP) 71111.05, Enclosure 2. The
team also reviewed applicable operator training lesson plans and job performance
-
-
-
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4
- i J
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^ i:
-
-
F
\\
.
-
-
A:
-
8
measures (JPMs) and discussed them with operators. In addition, the team reviewed
I records of actual operator staffing on selected days.
Findings
- 1.-
Untimely and Unapproved Manual Operator Action for Fire Safe Shutdown
Introduction: The team found that a local manual operator action to prevent spurious
-
opening of all eleven SRVs would not be performed in sufficient time to be effective.
Licensee reliance on this manual action for hot shutdown during a fire, instead of
physically protecting cables from fire damage, had not been approved by the NRC.
4,1
.
.
.
I -
Description: The team noted that Step 9.3.2.1 of AOP 34AB-X43-001-2, Fire
/
Procedure, Version 10.8, dated May 28, 2003, stated: 'To prevent all eleven Srgs fro
opening simultaneously, open links BB-10 in Panel 2H1 1-P927 and BB-10 in lanel
2H1-1-P928." The team noted that spurious opening of all eleven SRVs would be
considered a large loss of coolant accident (LOCA), and that a LOCA raf be
Z prevenTe -fiom
iicurR5 g a ire venIAAdditionally, the team observed that this
step was sufficiently far back in the proceddrb that it may not be completed in time to
prevent potential fire damage to cables from causing all eleven SRVs to spuriously
open.
The licensee had no preplanned estimate of how long it would take operators to
complete this step during a fire event. There was no event time line or operator training
job performance measure (JPM) on this step. The team noted that, during a fire-eyent,
operators could be using many other procedures concurrent with the Fire Procedure.
For example, they could be using other procedures to communicate with the fire brigade
about the fire, respond to a reactor trip, deal with a loss of offsite power, and provide
emergency classifications and offsite notifications of the fire event. During the
inspection, licensee operators estimated that, during a fire event, it could take aboutt,
minutes before operators would accomplish Step 9.3.2.1. The team concurred with i
time estimate. However, NRGfire models indicated that fires could potentially cause
damage to cables inaslt66
to ten minutes. Consequently, the teams
I
concluded that during a fire eveihe
licensee's procedures would not ensure thatp7
9.3.2.1 would be accomplished Ine
to prevent potential spurious opening of all
l1even SRVs.
-A
g1
The team also identified other
s with Step 9.3.2.1. There was no emergency
2
lighting inside the panels, hen ef he fire caused a loss of normal lighting (e.g., by
causing a loss of offsite pow
e erators would need to use flashlights to plerform the
actions inside the panels. Consequently, the team considered the emergency lighting
for Step 9.3.2.1 to be inadequate (see Section 1 R05.07.b). In addition labeling of the
links inside the panels was so poor that operators stated that the
y
n fully rely on
E
the labeling. Also, the tool that operators would use to loosen a
sli
he links inside
-a
the energized panels was made of steel and was not professio allele rically
insulated. Further, licensee reliance on this operator action, insad
physically
£
protecting the cables as required by
tN0
CFR 50, Appendix R. S
n I.G.2, had notNC
bleen approved by the NRC.-
tzr -1I
0 Cr-~P.
ff
-L'- YŽ)j -C-~-
~NZ
L- -
- ,a^
M L. re Zc~ e. 1etL ttU-J"t
A- f 0i4-K#e.
9
The licensee stated that cable damage to two instrument cables, for reactor pressure.:
signals, would be needed to spuriously open all eleuenNSRVs. Since the licensee stated
that the two cables were in the same cable tray in We Krea 2104,A4heI- 4 -
--
ableay, the team considered that a fire in that area could potentially cause all eleven-
SRVs to spuriously open (see section 1 R21.01 .b).
In response to this issue, the licensee initiated CR 2003008203 and promptly revised
the Fire Procedure before the end of the inspection, moving the actions of Step 9.3.2.1-
to the beginning of the procedure. The procedure change enabled the actions to be
accomplished much sooner during a fire in the Unit 2 east cableway or in other fire
areas that were vulnerable to the potential for spuriously opening all eleven SRVs. The
team determined that this issue is related to associated circuits. As described in NRC
Inspection Procedure (IP) 71111.05, Fire Protection, inspection of associated circuits Is
temporarily limited. Consequently, the team did not pursue the cable routing or circuit
analysis that would be necessary to evaluate the possibility, risk, or potential safety
significance of Group B and C SRVs .spuriously
opening due to fjre damage to the -
1 ..
instrument cables. The team did, however, perform a circuit ana ysis of Group A SRVs
for which the licensee takes credit for a fire in fire a
2104. &'e section 1 R21.01). c#
Analysis: The team determined that this finding wssociated with the protection
against external factors attribute. It affected the 4bjective of the mitigating systemo
ensure the availability of systems that respond t nitiating events and*
ater7'-
than minor. The team determined that the finding had pot
safety significance
greater than very low safety significance because, failu to prevent spurious operation
of the SRVs could result in them opening in certair -fiscenarios, thereby complicating
the post-fire recovery a
ns. Howev r the finding remains unresolved pending
completion of the
Enforcement: 1
R 50 A
endix R, Section Ill.G.2 requires that where cables or
equipment, including ass
ated non-safety circuits that could prevent operation or
cause mal-operation d to hot shorts, open circuits, or shorts to ground, of redundant
trains of systems ne
ssary to achieve and maintain hot shutdown conditions are.
located within the me fire area outside of the primary containment, one of the'
following means f ensuring that one or the redundant trains is free of fire damage shall
be provided: 1) a fire barrier with a 3-hour rating; 2) separation of cables by a horizontal
distance of more than 20 feet with no intervening combustibles and with fire detectors
and automatic fire suppression; or 3) a fire barrier with a 1-hour rating with fire detectors
and automatic suppression.
The licensee had not provided physical protection against fire damage for the two
instrument cables by one of the prescribed methods. Instead, the licensee had relied on
manual operator actions to prevent the spurious opening of all eleven SRVs. Licensee
personnel contended that fire damage to two cables was outside of the Hatch licensing
basis and consequently that there was no requirement to protect the instrument cables.
However, the licensee could provide no evidence to support that position.
This potential issue will remain unresolved pending the completion of a significance
determination by the NRC. This issue is identified as URI 50-366/03-06-02, Untimely
and Unapproved Manual Operator Action foroFire Safe Shutdown.
.10
2.
Local Manual Operator Action was Too Difficult and Unsafe
Introduction: A finding of very low safety significance was identified in at a local
manual operator action to operate SSD equipment was too difficult
d was also unsafe.'
The team judged that some operators would not be able to perfo
the action. This
finding involved a violation of NRC requirements.
Description: The team observed that Steps 4.15.8.1.1 a
9.3.5.1 of the Fire Procedure
were relied on instead of providing'physical protection
r cables or providing a
procedure for cold shutdown repairs. Both steps re
wed the same local manual
operator action: "Manually OPEN 2E1 1 -F0l 5A, In ard LPCI Injection Valve, as
required." This action was to be taken in the Uni
drywell access, which was a locked
high radiation, contaminated, and hot area with emperatures over 100 degrees F.
Valve 2E1 I -F01 5A was a large (24-inch dia eter) motor-operated gate valve with a
three-foot diameter handwheel. The main ifficulty with manually opening this valve was
lack of an adequate place to stand. An oprator showed the team that to perform the
action he would have to climb up to and
and on a small section of pipe lagging (a
curved area about four inches wide by l inches long), and then reach back and to his
right side, to hold the handwheel with
s right hand, while reaching forward and to his
right to hold the clutch lever for the m or operator with his left hand. He would not have
good balance while performing the a ion. The foothold, which was large enough to'
support only one foot, was well flatte ed and appeared to have been used in the past to
manually operate this valve. The fo thold was about six to seven feet above a steel
grating, and the team observed tha space available for potential use of a ladder to
better access the 2E1 1-F015A valve handwheel was not good.
Other difficulties with manually opening the valve included the heat; the need to wear'
full anti-contamination clothing, a hardhat, and safety glasses; and inadequate
'
emergency lighting (see Section 1 R05.07). Also, there was no note or step in the
procedure to ensure that the RHR pumps were not running before attempting to
manually open the 2E1 1-FO15A valve. If an RHR pump were running, it could create a
differential pressure across the valve which could make manually opening it much more
difficult. If the operator did not have sufficient agility, strength or stamina, he would be
unable to complete the action. Also, the team judged that inability to remove sweat from
his eyes, due to wearing gloves that could be contaminated, would be a limiting factor
for the operator. In addition, if the operator slipped or lost his balance, he could fall and
become injured. Considering all of the difficulties, the team judged that this action was
unsafe and that some operators would not be able to perform it.
The licensee had no operator trainingJPM for performing this action and could not
demonstrate that all operators could perform the action. One experienced operator,
who appeared to be in much better physical condition that an average nuclear plant
operator, stated that he had manually operated the valve in the past, but that it had been
very difficult for him.
The team judged that, since this action was not required to maintain hot shutdown and
was required for cold shutdown following a fire in one of the four selected fire areas,
licensee personnel could have time to improve the working conditions after a fire. They
I
11
ave time to in
SC
oding or temporary ventilation; improve the lighting; and
assignImultiple oper to to
anually open the valve. They could have time to perform
afcol shutdown re
H ever, the licensee had not preplanned any cold shutdown
prs for opening t Y
e.
Analysis: This finding is greater than minor because it affected the availability and
reliability objectives and the equipment performance attribute of the mitigating systems
cornerstone. Since the licensee could have time to develop and implement cold
shutdown repairs to facilitate accomplishment of the action, this finding did not impact
the effectiveness of one or more of the defense in depth elements. Hence this finding
did not have potential safety significance greater than very low safety significance
(Green).
Enforcement: 10 CFR 50, Appendix R, Section III.G.1 requires that fire protection
features shall be provided for systems important to safe shutdown and shall be capable
of limiting fire damage so that systems necessary to achieve and maintain cold
shutdown from either the control room or emergency control stations can be repaired
within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In addition, TS 5.4.1 requires that written procedures shall be
established, implemented, and maintained covering activities including fire protection
program implementation and including the applicable procedures recommended in
Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33
recommends procedures for combating emergencies including plant fires and
procedures for operation and shutdown of safety-related BWR systems. The fire
protection program includes the SSAR which requires that valve 2E1 1-FO15A be
opened for SSD following a fire in Fire Area 2104, the Unit 2 east cableway. AOP
34AB-X43-001-2, Fire Procedure, Version 10.8, dated May 28, 2003, implements these
requirements in that it provides information and actions necessary to mitigate the
consequences of fires and to maintain an operable sh
r
in
i
a
age
to specific fire areas. Also, AOP 34AB-X43-001-
ovides
PSt 4.15.8.11 and 9. 5.1
for manually opening valve 2E11 1-F015A followi
a fire in eee
Contrary to the above, the licensee had no proceue
i
ted fire
damage within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Instead, the licensee relied on local manual operator actions,
as described in Steps 4.15.8.1.1 and 9.3.5.1 of AOP 34AB-X43-001-2. However, those
procedure steps were inadequate in that some operators would not be able to perform
them because the required actions were too difficult and also were unsafe. In response
to this issue, the licensee initiated CR 203008202. Because the identified inadequate
operator actions are
significance and the issue has been entered into
the licensee's corr
ive act o progra
this violation is being treated as an NCV,
consistent with
ction VI.A 'of the
C's Enforcement Policy: NCV 50-366/03-06-03,
Inadequate Pro edure for
[Ma
al Operator Action for Post-Fire Safe Shutdown
Equipment.
3.
Unanproved Manual Operator Actions for Post-Fire Safe Shutdown
Introduction: A finding of very low safety significance was identified in that the licensee
relied on some manual operator actions to operate SSD equipment, instead of providing
the required physical protection of cables from fire damage. This finding involved a
violation of NRC requirements.
12
Description: Th team observed that AOP 34AB-X43-001-2, Fire Procedure, included
some local m ual operator actions to achieve and maintain hot shutdown that had not -
been approv d by the N C. Examples of steps from the procedure included:
4.1 5.2.2, .
a loss of offsite power occurs and emergency busses energize
lace Station Service battery chargers 2R42-S026 (2R42-S029), 2R42-S027
(2R42-S030) AND 2R42-S028 (2R42-S031) in service per 34SO-R42-001 -2."-
-
Step 4.15.4.5; ...If HPCI fails to automatically trip on high RPV level... OPEN the
following links to energize 2E41-F124, Trip Solenoid Valve, AND to fail 2E41-
9'
rJ
i
,F3025
HPCI Governor Valve, in the CLOSED position:
TT-75 in panel 2H11-P601
TT-76 in panel 2H11-P601"
Step 4.15.4.6; ... lf HPCI fails to automatically trip on high RPV level..'. "OPEN
breaker 25 in panel 2R25-S002 to fail 2E41-F3052, HPCI Governor Valve, in the
CLOSED position."
The team walked down these actions using the guidance contained in Inspection
Procedure 71111.05T and judged that they could reasonably be accomplished by
operators in a timely manner. However, the team determined that these operator
actions were being used instead of physically protecting cables from fire damage that
could cause a loss of station service battery chargers or a HPCI pump runout.
Analysis: The finding is greater than minor because it affected the availability and
reliability objectives as well as the equipment performance attribute of the mitigating
systems cornerstone. Since the actions~could reasonably be accomplished by operators
in a timely manner, this finding did n t ave potential safety significance greater than
very low safety significance.
Enforcement: 10 CFR 50
pendixR.Section III.G.2 requires that where cables or
equipment, including a ociated non-safety circuits that could prevent operation or..
cause maloperation ue to hot shorts, open circuits, or shorts to ground, of redundant
trains of systems ecessary to achieve and maintain hot shutdown conditions are
located within t
same fire area outside of the primary containment, one of the
following me
s of ensuring that one of the redundant trains is free of fire damage shall
be provided: 1) a fire barrier with a 3-hour rating; 2) separation of cables by a
horizontal distance of more than 20 feet with no intervening combustibles and with fire
detectors and automatic fire suppression; or 3) a fire barrier with a 1-hour rating with fire
detectors and automatic suppression.
Contrary to the above, the licensee had not provided the required physical protection
against fire damage for power to the station service battery chargers or for HPCI
electrical control cables. Instead, the licensee relied on local manual operator actions,
without NRC approval. In response to this issue, the licensee initiated CR2003800166.
Because the issue had very lovw safety significance and has been entered into the
licensee's corrective action prpb ram, this violation is being treated as an NCV,
consistent with Section VI.AJ$lf the NRC's Enforcement Policy: NCV 50-366/03-06-04,
Unapproved Manual Operator jctions for Post-Fire Safe Shutdown.
.06:. Communications
- a
lnSection Scope
The team reviewe
support fire brigac
the safe shutdowr
be available for p:
team reviewed the
brigade to commu
b. -
Findings
No findings of sigr
.07
Emer6env Licghtir
13
d the plant communications systems that would be relied upon to
le and safe shutdown activities. The team walked down portions of
i procedures to verify that adequate communications equipment would
ersonnel performing local manual operator actions. In addition, the
e adequacy of the radio communication system used by the fire
Jnicate with the main control room.
iificance were identified.
..
.,
a..
Inspection Scone
The team inspected the licensee's emergency lighting systems to verify that 8-hour
emergency lighting coverage was provided as required by 10 CFR 50, Appendix R,
Section III.J., to support local manual operator actions that were needed for post-fire
operation of SSD equipment. During walkdowns of the post-fire SSD operator actions
for fires in the selected fire areas, the team checked if emergency lighting units were
installed and if lamp heads were aimed to adequately illuminate the SSD equipment, the
equipment identification tags, and the access and egress routes thereto, so that
operators would be able to perform the actions without needing to use flashlights.
- b.
Findings
Inadequate Emergency Lighting for Operation of Safe Shutdown Equipment
Introduction: A finding with very low safety significance was identified in that emergency
lighting was not adequate for some manual operator actions that were needed to
support post-fire operation of SSD equipment. This finding involved a violation of NRC
-
requirements.
Description: The team observed that emergency lighting was not adequate for some
manual operator actions that were needed to support post-fire operation of SSD
equipment. Examples included the following operator actions in procedure 34AB-X43-
001 -2, Fire Procedure, Version 10.8, dated May 28, 2003:
.*.
Step 4.15.2.2; ...if a loss of offsite power occurs and emergency busses energize
..."Place Station Service battery chargers 2R42-S026 (2R42-S029), 2R42-S027
I
(2R42-S030) AND 2R42-S028 (2R42-S031) in service per 34SO-R42-001 -2."
Step 4.15.4.5; ... lf HPCI fails to automatically trip on high RPV level... 'OPEN the
following links to energize 2E41-F1l24, Trip Solenoid Valve, AND to fail 2E41-
F3025 HPCI Governor Valve, in the CLOSED position:
-14
TT-75 in panel 2H1 Il-P601
TT-76 in 'panel 2H1-1-P601"
Step 4.15.5; "IF 2R25-S065, Instrument Bus 26, is DE-ENERGIZED perform the
following manual actions to maintain 2C32-R655, Reactor Water Level.':
Instrument, operable:
4.15.5.1; At panel 2H11-P612, OPEN links AAA-11 and AAA-12.
4.15.5.2; At panel 2H11-P601, CLOSE links HH-48 and HH49."
Steps 4.15.8.1.1 and 9.3.5.1; "Manually OPEN 2E11 1-F01 5A, Inboard LPCI
?
tInjection Valve, as required."
Steps 4.15.8.1.2 and 9.3.5.2; "Manually CLOSE 2E1 1 -F01 8A, RHR Pump A
Minimum Flow Isolation Valve, as required."
Step 9.3.2.1; 'To prevent all 1 1 SRVs from opening simultaneously, open links
6B-10 in Panel 2H11-P927 and BB-10 in Panel 2H11-P928."
i E
Step 9.3.3; "At Panel 2H1 1 -P627, open links AA-1 9, AA-20, AA-21, and AA-22,
to prevent spurious actuation of SRVs 2B21-F013D AND 2B21-F013G."
Step 9.3.6; UOPEN link TB9-21 in Panel 2H11-P700 to open Drywell Pneumatic
System Inboard Inlet Isolation, 2P70-F005."
0
Step 9.3.7; "OPEN link TB1-12 in Panel 2H11-P700 to open Drywell Pneumatic
System Outboard Inlet Isolation, 2P70-F005."
e
lStep 9.3.9.1; "Confirm OR manually CLOSE RHR Shutdown Cooling Valve
2E1 1 -F006D."
Step 9.3.9.2; "Manually OPEN Shutdown Cooling Suction Valve 2E11-F008, IF
required..."
The team verified that flashlights were readily available and judged that operators would
le to use the flashlights and accomplish the actions, with two exceptions. One
exce ion was the action to open terminal board links in two panels to prevent all eleven
SRVs rom spuriously opening, which was judged to be untimely (see Section
1 R05.t 5.b.1). The other exception was the action to open 2E11 -F015A, which was
judged to be too difficult (see Section 1 R05.P5.b.2). For all of these actions, the lack of
adequate emergency lighting could make yre actions more difficult to complete In a
timely manner and increase the cha
operator error.
Analysis: This finding is greater than minor because it affected the reliability objective
and the equipment performance attribute of the mitigating systems cornerstone. Since
operators would be able to accomplish the actions with the use of flashlights, this finding
did not impact the effectiveness of one or more of the defense in depth elements.
Hence, this finding did not have potential safety significance greater than very low safety
significance (Green).
The lead inspector presented the inspectio results to license management and othe
members of the licensee's staff at the co
lusion of the onsit inspection on Aprif.A,
Fee. Subsequent to the onsite inspec
n, the lead inspect r and Rag'-n.
management held follow up exits by t ephone with Mr
e and other members of
licensee management oncune 20anJune 3.
o update the licensee on
changes to the preliminary inspection findings. The icensee acknowledged the findings.
eam
~ked
lice e
her
of
m
lal
a
dd
he
21
4.
OTHER ACTIVITIES
40A2 Identification and Resolution of Problems
a.
Inspection Scope
The team reviewed a sample of licensee audits, self-assessments, and condition reports
(CRs) to verify that items related to fire protection and to SSD were appropriately
entered into the licensee's CAP in accordance with the Hatch quality assurance program
and procedural requirements. The items selected were reviewed for classification and
appropriateness of the corrective actions taken or initiated to resolve the issues. In
addition, the team reviewed the licensee's applicability evaluations and corrective
actions for selected industry experience Issues related to fire protection. The operating
experience (OE) reports were reviewed to verify that the licensee's review and actions
were appropriate.
The team reviewed licensee audits and self-assessments of fire protection and safe
shutdown to assess the types of findings that were generated and to verify that the
findings were appropriately entered into the licensee's corrective action program.
b.
Findings
No findings of significance were Identified.
40A6 Meetings. Including Exit
The team presented the inspection results to Mr. R. Dedrickson, Assistant General
Mana
d other members of your staff at the conclusion of the inspection on July
25,
03
licensee acknowledged the findings presented. Proprietary information is
not i clu
in the inspection report.
Id
I+41;
,*H61?
Ala-01
'117-vc-450.
Z,
-I
.
0/444
-5,
I
t
-4
- K
ro-e-
I
15
Enforcement: 10 CFR 50, Appendix R, Section III.J. requires that emergency lighting
units with at least an 8-hour battery power supply shall be provided in all areas needed
for operation of safe shutdown equipment and in access and egress routes thereto.
Contrary to the above, emergencyjighting units were not adequately provided in all
areas needed for operation of safe shutdown equipment. In response this issue, the
licensee initiated CRs 2003008237 and 2003008179. Because the identified lack of
emergency lighting is of very loW safety significance and has been entered into the
licensee's corrective action r
ram, this violation is being/reated as an NCV,
consistent with Section VI.
of the NRC's EnforcementPolicy: NCV 50-366/03-06-05,
Inadequate Emergency Lighlg for Operation of Safe Slhutdown Equipment.
.08
Cold Shutdown Repairs
f
e
The licensee had identified no needed cold shutdown repairs. Also, with the exception
of the potential need for a cold shutdown repair to open valve 2E1 1 -F01 5A (see section
1 R05.05.b.2), the team identified no other need for cold shutdown repairs.
Consequently, this section of IP 71111.05 was not performed.
.09
Fire Barriers and Fire Area/Zone/Room Penetration Seals
a.
Inspection Scope
The team reviewed the selected fire areas to evaluate the adequacy of the fire
resistance of fire area barrier enclosure walls, ceilings, floors, fire barrier mechanical
and electrical penetration seals, fire doors, and fire dampers. The team selected
several fire barrier features for detailed evaluation and inspection to verify proper
installation and qualification. This was accomplished by observing the material condition
and configuration of the installed fire barrier features, as well as construction details and
supporting fire endurance tests for the installed fire barrier features, to verify the as-built
configurations were qualified by appropriate fire endurance tests. The team also
reviewed the FHA to verify the fire loading used by the licensee to determine the fire
resistance rating of the fire barrier enclosures. The team also reviewed the installation
instructions for sliding fire doors, the design' details for mechanical and electrical
penetrations, the penetration seal database, Generic Letter (GL) 86-10 evaluations, and
the fire protection penetration seal deviation analysis for the technical basis of fire
barrier penetration seals to verify that the fire barrier installations met design
requirements and license commitments. In addition, the team reviewed completed
surveillance and maintenance procedures for selected fire barrier features to verify the
fire barriers were being adequately maintained.
The team evaluated the adequacy of the fire resistance of fire barrier electrical raceway
fire barrier system (ERFBS) enclosures for cable protection to satisfy the applicable
separation and design requirements of 10 CFR 50, Appendix R, Section III.G.2.
Specifically, the team examined the design drawings, construction details, installation
records, and supporting fire endurance tests for the ERFBS enclosures Installed in Fire
Area 2104, the Unit 2 East Cableway. Visual inspections of the enclosures were
performed to confirm that the ERFBS installations were consistent with the design
drawings and tested configurations.
16
The team reviewed abnormal operating fire procedures, selected fire fighting pre-plans,,
fire damper location and detail drawings, and heating ventilation and air conditioning .
\\
(HVAC) system drawings to verify that access to shutdown equipment and selected
operator manual actions would not be inhibited by smoke migration from one area to.
adjacent plant areas used to accomplish SSD.
b.
Findings
No findings of significance were identified.
.10
Fire Protection Systems. Features, and EquiDment
a.
Inspection ScoDe
The team reviewed flow diagrams, cable routing information, and operational valve.
lineup procedures associated with the fire pumps and fire protection water supply
system. The review evaluated whether the common fire protection water delivery and
supply components could be damaged or Inhibited by fire-induced failures of electrical
power supplies or control circuits. Using operating and test procedures, the team toured
the fire pump house and diesel driven fire pump fuel storage tanks to observe the
system material condition, consistency of as-built configurations with engineering
drawings, and determine correct system controls and valve lineups. Additionally, the
team reviewed periodic test procedures for the fire pumps to assess whether the
surveillance test program was sufficient to verify proper operation of the fire protection
water supply system in accordance with the program operating requirements specified
in Appendix B of the FHA.
The team reviewed the adequacy of the fire detection systems in the selected plant fire
areas in accordance with the design requirements in Appendix R, III.G.1 and Mll.G. 2.
The team walked down accessible portions of the fire detection systems in the selected
fire areas to evaluate the engineering design and operation of the installed
configurations. The team also reviewed engineering drawings for fire detector types,
spacing, locations and the licensee's technical evaluation of the detector locations for
the detection systems for consistency with the licensee's FHA, engineering evaluations
for NFPA code deviations, and NFPA 72E. In addition, the team reviewed surveillance
procedures and the detection system operating requirements specified in Appendix B of
the FHA to determine the adequacy of fire detection component testing and to ensure
that the detection systems could function when needed.
The team performed in-plant walk-downs of the Unit 2 East Cableway automatic wet
pipe sprinkler suppression system to verify the proper type, placement and spacing of
the sprinkler heads as well as the lack of obstructions for effective functioning. The.
team examined vendor information, engineering evaluations for NFPA code deviations,
and design calculations to verify that the required suppression system water density for
the protected area was available. Additionally, the team reviewed the physical
configuration of electrical raceways and safe shutdown components in the fire area to
determine whether water from a pipe rupture, actuation of the automatic suppression
system, or manual fire suppression activities in this area could cause damage that could
inhibit the plant's ability to safely shutdown.
,1
17
The team reviewed the adequacy of the design and installation of the manual C02 hose
reel suppression system for the diesel generator building switchgear rooms 2E and 2F
(Fire Areas 2404 and 2408). The team performed in-plant walk-downs of the diesel
generator building C02 fire suppression system to determine correct system controls
and valve lineups to assure accessibility and functionality of the system, as well as
associated ventilation system fire dampers. The team also reviewed the licensee's
actions to address the potential for C02 migration to ensure that fire suppression and
post-fire safe shutdown actions would not be impacted. This was accomplished by the
review of engineering drawings, schematics, flow diagrams, and evaluations associated
with the diesel generator building floor drain system to determine whether systems and
operator actions required for SSD would be inhibited by C02 migration through the floor
drain system.
b.
Findings
No findings of significance were Identified.
.11
Comnensatorv Measures
a.
Inspection Scope
The team reviewed Appendix B of the FHA and applicable sections of the fire protection
program administrative procedure regarding administrative controls to identify the need
for and to implement compensatory measures for out-of-service, degraded, or
inoperable fire protection or post-fire safe shutdown equipment, features, and
tS.
The team reviewed licensee reports for the fire protection status of Unit 1, U
of
shared structures, systems, and components. The review was performed to
at
the risk associated with removing fire protection and/or post-fire systems or
components, was properly assessed and implemented in accordance with the approved
fire protection program. The team also reviewed Corrective Action Program Condition
Reports generated over the last 18 months for fire protection features that were out of
service for long periods of time. The review was conducted to assess the licensee's
effectiveness in returning equipment to service in a reasonable period of time.
1j~R21
Findings
No findings of significance were identified.
Design Chanae Request (DCR)91-134. SRV Backup Actuation Usiln Pressure
Transmitter Signals
a.
Inspection Scope
,
The team performed an independent design review of plant modification D
-134
in
order to evaluate the technical adequacy of the design change packag
.he
s
e of
the review and circuit analysis performed by the team was limited to e*opA
RVs
.1
a}
I
18
for which the licensee takes credit In mitigating a fire in the fire areas selected for the
inspection.
b.
Findings
.1.
Introductij
An inadeq
plant modification, DCR 91-134, failed to implement the design input
requirements of one-out-of-two taken twice logic for the SRVs backup actuation using
pressure transmitter signals.
Descrintion^.
DCR 91-134 was implemented in response in to concerns raised in General Electric
Report NEDC-3200P, Evaluation of SRV Performance during January-February 1991
Turbine Trip Events for Plant Hatch Units 1 and 2. In order to ensure that Individual
SRV(s) will actuate at or near the appropriate set point and within allowable limits, a
backup mode of operation for the SRVs was implemented by this DCR. The design
was intended to mitigate the effects of corrosion-induced set point drift of the Target
Rock SRVs.
Automatically controlled, two stage SRVs are installed on the main steam lines Inside
containment for the purpose of relieving nuclear boiler pressure either by normal
mechanical action or by automatic action of an electro-pneumatic control system.! Each
SRV can be manually controlled by use of a two position switch located in the main
control room. When placed in the 'Open" position, the switch energizes the pilot valve
of the individual SRV and causes It to go open. When the switch is placed In the "Auto"
position the SRV is opened upon receipt of either an Auto Depressurization System
(ADS), or Low-Low Set (LLS) control logic signal. Either signal will initiate opening of
the valve. DCR 91-134 provided a backup mode for initiation of electrical trip of the pilot
valve solenoid, which was independent of ADS or LLS logic. The backup mode required
no operator action to initiate opening of the SRVs and was considered a "blind control
loop" to the operators, ie. there are no Instruments that provide the operators
information concerning the open/close status of the SRVs.
The scope of the plant modification involved the installation of four Rosemount pressure
transmitters (Model No. 11 54GP9RJ), 0-3000 psig, in the 2H21 -P404 and P405
instrument racks at Elevation 158 of the reactor building. Each pressure transmitter
formed part of a 4-20 ma current loop and provided the analog trip signal for SRV
actuation within the following set point groups:
SRV Group
SRV Identification Taas
SRV Set Point
A
2B21-F013B, D, F, and G
1120 psig
B
2B21 -F01 3A, C, K, and M
1130 psig
C
2B21-F013E, H, and D
1 140 psig
--1
19
Pressure transmitters (PTs) 2B21-N127A and 2B21-N127C were wired to ATTS
cabinets 2H11-P927. Pressure transmitter 2B21-N127A instrument loop components
consisted of a trip unit master relay K308C and trip unit slave relays K321 C and K332C.
The loop components for pressure transmitter 2B21 -N127C consisted of a trip unit
master relay K335C in addition to trip unit slave relays K336C and K363C. These two
instrument loops constituted a uDivisionn pressure monitoring channels and were
intended to provide the one-out of two logic signal from this Division for initiating SRV
backup actuation.
Additionally, pressure transmitters 2B21-N127B and 2B21-N127D were wired to ATTS
cabinet 2H11-P928. Pressure transmitter 2B21-N127B instrument loop components
consisted of a trip unit master relay K31 OD and trip unit slave relays KK312D and
K332D. The loop components for pressure transmitter 2B21-N127D consisted of a trip
unit master relay K335D in addition to trip unit slave relays K336D and K363D. These
two instrument loops constituted a separate "Division" pressure monitoring channels and
were intended to provide the one-out of two logic signal from this Division for initiating
SRV backup actuation. The design objective of having two instrument channels was to
assure compliance with HNP-2-FSAR, Section 15.1.6.1, Application of Single Failure
Criteria. This criteria requires for anticipated operational occurrences (AOOs) that the
protection sequences within mitigation systems be single component failure proof. A
failure of one instrument channel in a division will therefore not eliminate the protection
provided by either of the instrument channels.
The following table identifies the Division, pressure transmitter loops and the associated
trip unit master and slave relays:
Division
A
PT Loops
Trip Unit Master Relays
Trip Unit Slave Relays
K321 C and K332C
K336C and K363C
2B21 -N127A
2B21 -N127C
K308C
K335C
B
K31OD
K312D and K332D
/B21-N127D
K335D
K336D and K363D
The Group 7SRVsere provided logic input signals from the trip unit master relays.
The Groupp and eSRVs were provided logic input signals from the trip unit slave
relays. The otal of 12 relays described above, (6 in ATTS cabinet 2H1 1-P927 and 6 in
ATTS cabinet 2H1 1 -P928), were intended to be wired to provide "one-out-of-two taken
twice logic" for actuation of the SRVs. The design objective was to assure that a single
relay failure in either division would not cause an inadvertent SRV actuation. Coincident
logic input is required from both division instrument loops in order to initiate a SRV
backup actuation using the pressure transmitter signals. This occurs when the circuit
that is used to energize the individual SRV pilot valve to open the SRV, is enabled by
receiving simultaneous logic inputs from either instrument loop in both division.
The team performed a circuit analysis of SRV 2B21 -FO1 3F (Path 1) and SRV 2B21 -
FOI3G (Path 2) in order to verify that the design objectives of implementing a one-out-
of-two taken twice logic had been achieved. Based on this review the team determined
that the design objective of implementing a one-out-of-two taken twice logic had not
20
been installed for the SRVs. The logic installed for the SRVs was a two-out-of-two
coincident taken twice logic in addition to a one-out-of-two coincident taken twice logic.
The coincident logic implemented using trip unit master relays K31OD and K335D could:
result in spurious actuation of group "A" SRVs for a fire in Fire Area 2104. In addition,
this spurious actuation defeats the capability to manually control these SRVs.
Whenever a SRV lifts, it will remain open until nuclear boiler pressure is reduced to
about 85% of its overpressure lift setpoint However, because the instrument loops
have failed high, the trip unit master relays and the trip unit slave relays will continue to
energize the pilot valve of the individual SRV and keep the SRV open. As a result, this
failure mode prevents the operators from manually controlling the group A SRVs as is
required per the SSAR.
Analysis: This finding is greater than minor because it affected the availability and
reliability objectives and the equipment performance attribute of the mitigating system
comerstone. The team determined that the finding had potential safety significance
greater than very low safety significance because it prevented the operators from
manually controlling group A SRVs which the licensee credits with mitigating a fire in
Fire Area 2104. Manual control of the group A SRVs is required to ensure that the
suppression pool temperature will not exceed the HCTL for the suppression pool.
Failure to ensure that the suppression pool temperature will not exceed the HCTL could,
result in loss of net positive suction head for the Core Spray pumps which the licensee
credits for mitigating this event. However, the finding remains unresolved pending
completion
nificance determination.
Enforcemrn10
FR 50, Appendix B, Criterion Ill, requires that design control
measure
all pr ide for verifying or checking the adequacy of design.
DCR 91-132 pified design input requirements for the sensor initiated logic that
electrically activates the SRVs to be a one-out-of-two logic scheme. It also identified the
potential worst case failure mode of this logic modification as a short in the logic which
would results in an inadvertent opening of a SRV. It concluded that the modification is
designed so that the actuation logic will not fail to cause inadvertent opening of a SRV
nor prevent a SRV from lifting upon ADS/LLS activation. Contrary to the above the logic
implemented by the licensee for DCR 91-134 was different from the specified design
input requirements. The independent design verification performed for DCR 91-134
failed to identify this error in the logic scheme. Additionally, the Appendix R Impact
Review performed for DCR 91-134 failed to identify the potential failure mode of all
eleven SRVs because of fire induce dam
in Fire Area 2104.
The plant modification install
CR 91-134 failed to correctly implement the one-
out-of-two taken twice Ic
hat was specified in the SRV backup actuation via pressure
transmitter signals desn change package. This failure has created a condition where
fire induced failures f two instrument circuit cables, (within close proximity to each'
other), could result in spurious actuation of all eleven SRVs with the eleven SRVs
assuming a stuck open mode of operation, based on the logic input from trip unit master
unit relays K31 OD, and K335D and their associated trip unit slave relays. Pending
completion of an significance determination by the NRC, this item is identified as URI
50-366/03-06-06 ;Implementation of DCR 91-134.Refults inAur,
Artutisn of
-5 AS
!,
I. . .
..
.!~
,
. I
.
.
.
.
+b.
4
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel:
M. Beard, Acting Engineering Support Supervisor
V. Coleman, Quality Assurance Supervisor
M. Dean, Nuclear Specialist, Fire Protection
R. Dedrickson, Assistant General Manager for Plant hatch
B. Duval, Chemistry Superintendent
M. Googe, Maintenance Manager
J. Hammonds, Operations Manager
D. Javorka, Administrative Assistant, Senior
R. King, Acting Engineering Support Manager
1. Luker, Senior Engineer, Licensing
T. Metzer, Acting Nuclear safety and Compliance Manager
A. Owens, Senior Engineer, Fire Protection
D. Parker, Senior Engineer, Electrical
J. Payne, Senior Engineer, Corrective Action Program
J. Rathod, Bechtel Engineering Group Supervisor
M. Raybon, Summer Intern
K. Rosanski, Oglethorpe Power Corporation Resident Manager
J. Vance, Senior Engineer, Mechanical & Civil
R. Varnadore, Outages and Modifications Manager
NRC personnel:
N. Garret, Senior Resident Inspector
C. Payne, Fire Protection Team Leader
LIST OF ITEMS OP
ODened
50-366/03-06-0
Concern~s
), AND I
Potentie
)ISCUSSED
il Opening of SRVs (Section
URI
Opened and Closed
50-366/03-06-03
v,
Lo
V
<4q
50-366/03-06-04
50-366/03-06-05
Discussed
Unapprove Manual Operator Actions for Post-Fire Safe
Shutdown (Section 1 R05.O5.g3b
Inadequate Emergency ighting for Operation of Post-Fire Safe
Shutdown Equipment (Section 1 R05.07.b)
None
76-3iCo3-Ob
QLo
-4-
!
Attachment
3
LIST OF DOCUMENTS REVIEWED
Procedures
Administrative Procedure 40AC-ENG-008-OS, Fire Protection Program, Rev. 9.2
Administrative Procedure 42FP-FPX-018-OS, Use, Controf, and Storage of
Flammable/Combustible Materials, Rev. 1.0
Department Instruction DI-FPX-02-0693N, Fire Fighting Equipment Inspection, Rev. 5
Fire Protection Procedure 42FP-FPX-005-OS, Drill Planning, Critiques and Drill Documentation
Rev. 1 ED-
Fire Protection Procedure 42FP-FPX-007-OS, Hot Work, Rev. 1.2
Preventive Maintenance Procedure 52PM-MEL-01 2-0, Low Voltage Switchgear Preventive
Maintenance, Rev. 25.0
Preventive Maintenance Procedure 52PM-MEL-01 4-0, Transformer Maintenance, Rev. 1 0.1
Surveillance Procedure 42SV-FPX-002-OS, Low Pressure 002 System Surveillance, Rev. 7.1
Surveillance Procedure 42S;V-FPX-004-OSFire Pump Test, Rev. 8.6
Surveillance Procedure 42sV-FPX-006-OS, Fire Damper Surveillance, Rev. 1 ED 1
Surveillance Procedure 42SV-FPX-021 -OS, Surveillance of Swinging Fire Doors, Rev. 1.6
Surveillance Procedure 42SV-FPX-024-OS, Fire Hose Stations 31 Day Surveillance, Rev.
9
Surveillance Procedure 42SV-FPX-030-OS, Fire Emergency Self Contained Breathing
Apparatus Inspection and Test, Rev. 1
Surveillance Procedure 42SV-FPX-032-OS, Automatic Sliding Fire Door Visual Inspection,
Rev. 3.3
Surveillance Procedure 42SV-FPX-036-OS, Annual Fire Pump Capacity Test, Rev. 8.6
Surveillance Procedurre425V-FPX-037-MS, Fire
on Itmtation Sreveince,
Rev. 5.1
system Operating Procedure 3450-X43-001 -1, Fire Pumps Operating Procedure, Rev. 4.3
Training Procedure 73TR-TRN-003-OS, Fire Training Program, Rev.4
1-001P-2, Loss of CRD System, Version 2.3
AOP 34AB-C71-001-2, Scram Procedure, Version 9.9
AOP 34AB-C71-002-2, Loss of RPS, Version 4.3
AOP 34AB-N61-002-2S, Main Condenser Vacuum Low, Version 0.4
ASP 34AB-P41-001-2, Loss of Plant Service Water, Version 8.1
AOP 4AB-P42-001 -2S, Loss of Reactor Building Closed Cooling Water, Version 1.4
AOP 34AB-P51-001-2, Loss of Instrument and Service Air System or Water Intrusion into the
Service Air System, Version 3.0
AOP 34AB-R22-001 -2, Loss of DC Busses, Version 2.4
AOP 34AB-R22-002-2, Loss of 41 60V Emergency Bus, Version 1.4
AOP 34AB-R22-003-2, Station Blackout, Version 2.3
AOP 34AB-R22-004-02, Loss of 41 60V Bus 2A, 2B, 20, or 2D, Version 1.3
AOP 34AB-R23-001-2S, Loss of 600V Emergency Bus, Version 0.4
AOP 34AB-R24-001-2, Loss of Essential AC Distribution Buses, Version 1.3
AOP 34AB-R25-002-02, Loss of Instrument Buses, Version 5.4
AOP 34AB-T47-001-2, Complete Loss of Drywell Cooling, Version 1.8
AOP 34AB-X43-001-2, Fire Procedure, Version 10.8
AOP 34AB-X43-002-0, Fire Protection SystemoFailures, Version 1.3
SOP 34S0-C71-001-2, 1L2oVAC RPS Supply System, Version 10.2
Attachment
-4
SOP 34SO-N40-001 -2, Main Generator Operation, Version 10.8
SOP 34SO-R42-001-2S, 125V DC and 125/250 VDC System, Version 7.1
SOP 34SO-S22-001-2, 500 KV Substation Switching, Version 5.2
31 EO-EOP-01 0-2S, RC RPV Control (Non-ATWS), Rev. 8, Attachment 1
31 EO-EOP-012-2S, PC-1 Primary Containment Control, Rev. 4, Attachment 1
31 EO-EOP-01 3-2S, PC-2 Primary Containment Control, Rev. 4, Attachment I
31 EO-EOP-01 4-2S, SC - Secondary Containment Control, Rev. 6, Attachment 1
31 EO-EOP-01 6-2S, CP-2 RPV Flooding, Rev. 8, Attachment 1
Procedure 34AB-X43-001-2S, Rev.1 OED3, "Fire Procedure," dated 5/28/03.
Calibration Procedure 57CP-CAL-097-2, Rosemount 1153 and 1154 transmitters, Revision,
No. 19.9.
Drawings
H-1 1814, Fire Hazards Analysis, Control Bldg. El. 1 30'-0", Rev. 5
H-1 1821, Fire Hazards Analysis, Turbine Bldg. El. 130'-O", Rev. 0
H-1 1846, Fire Hazards Analysis, Diesel Generator Bldg., Rev. 2
H-26014, R.H.R. System P&ID Sheet 1, Rev. 49
H-26015, R.H.R. System P&ID Sheet 2, Rev. 46
H-26018, Core Spray System P&ID, Rev. 29
B-1 0-1326, Rectangular Fire Damper Schedule, Rev. 2
B-1 0-1329, Rectangular Fire Damper, Rev. 1
H-1 1033, Fire Protection Pump House Layout, Rev. 47
H-1 1035, Fire Protection Piping and Instrumentation Diagram, Rev. 22
H-1i1226, Piping-Diesel Generator Building Drainage, Rev. 6
H-1 1814, Fire Hazards Analysis Drawing, Control Building, Rev. 5
H-1 1821, Fire Hazards Analysis Drawing, Turbine Building, Rev. 11
H-1 1846, Fire Hazards Analysis Drawing, Diesel Generator Building, Rev. 2
H-1 1894, Fire Detection Equipment Layout-Diesel Generator Building, Rev. 2
H-1 1915, Fire Detection Equipment Layout-Control Building, Rev. 2
H-1 3008, Conduit and Grounding, Fire Pump House, Rev. 9
H-13615, Wiring Diagram, Fire Pump House, Rev. 13
H-1 6054, Control Building HVAC System, Rev. 19
H-41509, Diesel Generator Building CO2 System-P&ID, Rev. 5
H-43757, Penetration Seals-Type, Number, and as-Built Location, Rev. 3
Calculations. Analyses, and Evaluations
E. I. Hatch Nuclear Plant Units 1 and 2 Safe Shutdown Analysis Report, Rev. 20.
Edwin I. Hatch Nuclear Plant Fire Hazards Analysis and Fire Protection Program, Rev. 20
Calculation SMFP88-001, Hydraulic Analysis of Sprinkler Systems in Control Building East
Cableway, dated 03/11/1988
Calculation SMNH94-046, FCF-F1OB-006, Fire Resistance of Concrete Block at HNP, dated
09/30/1994
Calculation SMNH94-048, FCF-FlOB-006, Cable Tray Combustible Loading Calculation, dated
09/30/1994
Attachment
fr
5
i,
- '
Calculation SMNH98-023, HT-98617, Fire Protection Penetration Seal Deviation Analysis,
- dated 10/28/1998
Calculation SMNHOO-01 1, HT-00606, Hose Nozzle Pressure Drop Analysis, dated 09/08/2000
Evaluation HT-91722, Fire Protection Code Deviation Resolution, dated 04/22/1992
Hatch Response to NRC IN 1999-005, dated 05/04/1999
Hatch Response to NRC IN 2002-024, dated 09/20/2002
Calculation SENH 98-003, Rev. 0, plot K, protective relay settings 4kV bus 2E
Calculation 85082MP, Plot 29, 600V Switchgear 2C
Calculation SENH 94-004, Attachment A, Sheets 7&8, 600/208 Reactor Building MCC 2C
Calculation SENH 91-011, Attachment P, Sheet 6, Reactor Building DC MCC 2A
Calculation SENH 94-013, Sheets 28 and 29, 600V Reactor Building MCC 2E-B
Calculation SENH 91 -011, Attachment P, Sheet 16, Reactor Building 25OVDC MCC 2B
Audits and Self-Assessments
Audit No. 01-FP-1, Audit of the Fire Protection Program, dated April 12, 2001
Audit No. 02-FP-1, Audit of the Fire Protection Program, dated February 28, 2002
Audit No. 03-FP-1, Audit of Fire Protection, dated April 21, 2003
1999-001106, Lighting In Fire Equipment Building
2002-000629, Inordinate Number of Buried Piping Leaks
2002-002127, Inadequate Bunker Gear
2002-002129, Health Physics Support and Participation for Fire Brigade
2003-000735, Impact on Cold Weather on Operating Units
'Audit Report 01 -FP-1, Audit of Fire Protection Program, dated 04/12/2001
Audit Report 02-FP-1, Audit of Fire Protection Program, dated 02/28/2002
Audit Report 03-FP-1, Audit of Fire Protection Program, dated 04/21/2003
CRs Reviewed
CR 2000007119, Fire Procedure 34AB-X43-001 -1 S Needs to be Enhanced
CR 2001002032, Fire Procedure 34AB-X43-001 -2S Needs Actions for Diesel Fuel Oil Pumps
CR 2003004377, Fire Procedure 34AB-X43-001 -1 Enhahcements
CR 2003004379, Fire Procedure 34AB-X43-001-2 Enhancements
CR 2003004382, SSAR Discrepancies
CRs Generated During this Inspection
CR 2003007129, No Fire Procedure Actions for a Fire in the 2C Switchgear Room
CR 2003007719, Use of Link Wrench
CR 2003007978, Fire Damper Corrective Action
CR 2003008141, Breaker Maintenance Handle
CR 2003008165, SSAR Section 2.100
CR 2003008179, Drywell Access Emergency Lights
CR 2003008181, Link Labeling
CR 2003008202, Manually Opening MOV 2E1 1 -F01 SA
CR 2003008203, SRV Manual Action Steps In Fire Procedure
CR 2003008237, Emergency Lights and Component Labeling for Manual Actions
Attachment
6
CR 2003008238, C02 Migration Through Floor Drains
CR 2003800132, SSAR Error for Position of 2E1 1 -F004A
CR 2003800151, Instruments for Manual Actions'
CR 2003800152, Sliding Links in SSAR
CR 2003800153, Promat Test Report
CR 2003008250, Communications for Post-Fire SSD
CR 2003800166, Review Fire Procedure Step 34AB-X43-001-2 Steps to Verify Compliance
with Appendix R.
Design Criteria and Standards
Design Philosophy for Fire Detectors at E. l. Hatch Nuclear Plants, Rev. 2
Completed Surveillance Procedures and Test Records
42SV-FPX-021-OS, Surveillance of Swinging Fire Doors, Task # 1-3367-1 (completed on
01/09/2003)
42SV-FPX-024-OS, Fire Hose Stations, Task #1-3359-1 (completed on 06/27/2003)
42SV-FPX-030-OS, Fire Emergency Self Contained Breathing Apparatus Inspection and Test,
Task # 1-4200-3 (completed on 07/07/2003)
42SV-FPX-032-OS, Automatic Sliding Fire Door Surveillance, Task # 1-3361-2 (completed on
08/13/2002
Promatec Technologies Installation Inspection Report for Fire Area 2104, MWO 2-98-00881,
Record 09367-2289, dated 09/03/1998
Technical ManualsNendor Information
Dow Coming Fire Endurance Test on Penetration Seal Systems in Precast Concrete F Using
Silicone Elastomers, dated 10/28/1975
Dow Corning 561 Silicone Transformer Fluid Technical Manual,1 0-453-97, dated 1997
S-80393, Mesker Instructions for Installing d&H uPyromaticP Automatic Sliding Fire Door Closer
S-27874B, General Electric Instruction Book GEK-26501, Liquid-Filled Secondary Unit
Substation Transformers, Rev. 2
S-52429A, Bisco, Fire Rated Penetration Seal Qualification Data, dated 08/16/1990
S-52480, Factory Mutual, Fire Rated Penetration Seal Qualification Data-Chemtrol Design
FC-225, dated 08/31/1990
S-54875B, Promatec, Fire Barriers-Unit 2 East Cableway, Rev. 2
Omega Point Laboratories, SR90-005, Three Hour Wall Test, dated 06/06/1990
Promatec Technologies Inc., PSI-001, Issue 1, General Construction Details, dated 07/21/1998
Promatec Technologies Inc., IP-2031, Installation Inspection for Promat's Three Hour Solid
WalVCeiling Protection System, Issue C, dated 06/16/1998
System Information Document No. Sl-LP-01401-03, Main Steam and Low Low Set System,
dated 4/3/2000
Attachment
7
Applicable Codes and Standards
ANSI N45.2.11-1974, Quality Assurance Requirements for the Design of Nuclear Power Plants
NFPA 12, Standard for Carbon Dioxide Systems, 1973 Edition.
NFPA 13, Standard for the Installation of Sprinkler Systems, 1976 Edition.
NFPA 14, Standard for the Installation of Standpipe and Hose Systems, 1974 Edition.
NFPA 20, Standard for the Installation of Centrifugal Fire Pumps, 1973 Edition.
NFPA 72D, Standard for the Installation, Maintenance, and Use of Proprietary Protection
Signaling Systems, 1975 Edition.
NFPA 72E, Standard on Automatic Fire Detectors, 1974 Edition
NFPA 80, Standard on Fire Doors and Windows, 1975 Edition.
NUREG-1552, Supplement 1, Fire Barrier Penetration Seals in Nuclear Power Plants, dated
January 1999
OSHA Standard 29 CFR 1910, Occupational Safety and Health Standards,
Underwriters Laboratory, Fire Resistance Directory, January 1998
Other Documents
Design Change Package 91-009, Retrofill Dielectric Fluid on Unit 2 Transformers, Rev. 1
Fire Protection Inspection Reports for the period 2001-2002
Fire Service Qualification Training, FP-LP-1 0003, Fire Fighter Safety, dated 01/:14/2002
Fire Service Qualification Training, FP-LP-10004, Fire Fighter Personal Protective Equipment,-
dated 01/14/2002
Fire Service Qualification Training, FP-LP-1 001 4, Fire Streams, dated 01/22/2002
Fire Service Qualification Training, FP-LP-10018, Fire Fighting Principles and Practices, dated
01/22/2002
Hatch Response to NRC Information Notice 1999-05, Inadvertent Discharge of Carbon Dioxide
Fire Protection System and Gas Migration, dated 05/04/1999
Hatch Response to NRC Information Notice 2002-24, Potential Problems with Heat Collectors
on Fire Protection Sprinklers, dated 09/20/2002
1 OCFR21 -001, ELECTRAK Corporation, Software Error within TRAK2000 Cable Management
and Appendix R Analysis System, dated 03/07/2003
U. S. Consumer Product Safety Commission, Invensys Building Systems Announce Recall of
Siebe Actuators In Building Fire/Smoke Dampers, dated 10/02/2002
Pre-f ire Plan A-43965, Power-Block Areas Methodology, Rev. 0
Pre-fire Plan A-43966, Fire Area 2404, Diesel Generator Building Switchgear Room 2E, Rev. 2
Pre-fire Plan A-43966, Fire Area 2408, Diesel Generator Building Switchgear Room 2F, Rev. 2
Pre-fire Plan A-43965, Fire Area 2016, W 600V Switchgear Room 2C, Rev. 4
License Basis Documents
Hatch UFSAR Section 3.4, Water Level Flood Design, Rev. 20
Hatch UFSAR Section 9.1-A, Fire Protection Plan, Rev. 1 8C
Hatch UFSAR Section 17.2, Quality Assurance During the Operations Phase, Rev. 20B
Hatch Fire Hazards Analysis, Appendix B, Fire Protection Equipment Operating and
Surveillance Requirements, Rev. 12B
Attachment
8
Hatch Fire Hazards Analysis, Appendix H, Application of National Fire Protection Association
Codes, Rev. 1 2B
Hatch SER dated April 18, 1994
Safe Shutdown Analysis Report for E.I. Hatch Nuclear Plant Units 1 and 2, Rev. 26
Fire Hazards Analysis for E. A.
Hatch Nuclear Plant Units 1 and 2, Rev.1 BC, dated 7/00.
NRC Safety Evaluation Report dated 01/02/1987; Re: Exemption from the requirements of
- Appendix R to 10 CFR Part 50 for Hatch Units 1 and 2 (response to letter dated
May.16,1986).
Letter dated 05/16/86, From L. T. Guewa (Georgia Power) to D. Muller, NRC/NRR; Re: Edwin I
Hatch Nuclear Plant Units 1 and 2 10 CFR 50.48 and Appendix R Exemption Requests
Design Chanae Request Documents
DCR No.91-134, SRV Backup Actuation via Pressure Transmitter Signals, Revision 0.
Drawing No. H-26000, Nuclear Boiler System P&ID, Sheet 1, Revision 39
Drawing No. H-27403, Automatic Depressurization System 2B21 C Elementary Diagram, Sheet
6 of 6, Revision 2
Drawing No. H-27472, Automatic Depressurization System 2B21 C Elementary Diagram, Sheet
3 of 6, Revision 2
Drawing No. H-27473, Automatic Depressurization System 2B21 C Elementary Diagram, Sheet
4 of 6, Revision 2
Drawing No. H-24427, Elementary Diagram, ATTS System 2A70 Sheet 27 of 35, Revision 3
Drawing No. H-24428, Elementary Diagram, ATTS System 2A70 Sheet 28 of 35, Revision 3
Drawing No. H-24429, Elementary Diagram, ATTS System 2A70 Sheet 29 of 35, Revision 5
Drawing No. H-24430, Elementary Diagram, ATTS System 2A70 Sheet 30 of 35, Revision 3
Drawing No. H-24431, Elementary Diagram, ATTS System 2A70 Sheet 31 of 35, Revision 3
Drawing No. H-24432, Elementary Diagram, ATTS System 2A70 Sheet 32 of 35, Revision 6
.
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Attachment